ML19289C791

Forwards Status Summaries of Staff Actions on Category a Generic Issues (Unresolved Safety Issues) & Table W/Related Manpower & Funding Info

ML19289C791
Person / Time
Issue date:	12/27/1978
From:	Harold Denton Office of Nuclear Reactor Regulation
To:	Kennedy R NRC COMMISSION (OCM)
Shared Package
ML19289C790	List: ML19289C789 ML19289C791
References
REF-GTECI-0000, REF-GTECI-AC NUDOCS 7901230345
Download: ML19289C791 (67)

Text

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4 UNITED STATES k

NUCLEAR REGULATORY COMMissiCN f

j WASWNGTON. D. C. 20655 1

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DEC 2 71978 MEMORANDUM FOR:

Comissioner Kennedy FROM:

Harold R. Denton, Director Office of Nuclear Reactor Regulation THRU:

Lee V. Gossick Executive Director for Operations

SUBJECT:

INFORMATION ON STATUS OF STAFF ACTIONS ON GENERIC ISSUE 5 Your December 6, 1978 memorandum to Lee V. Gossick requested certain information regarding the status of staff actions on generic issues.

Specifically, you requested a brief sumary of the status of each Category A task and a summary of resources expended on each item.

Status sumaries for thosa Category A tasks that have been designated as" Unresolved Safety Issues" have been prepared for inclusion in the Arinual Report. Copies of the proposed Annual Report sections are in-cluded in Enclosure 1.

Brief status sumaries for the remaining Category A tasks are provided in Enclosure 2. is a table that. displays the manpower, technical assistance funding, and research funding infor nation that you requested for each Category A task.

With regard to your request for quarterly reports on generic issue status, we anticioate that shortly we will finalize our plans to have status sumary information displayed in a management information book to be generated by OMPA. Several drafts have already been developed.

The latest draft was dated August 11, 1978 and w's provided to your office. We hope that this type of status sumary report will serve your needs as well as ours. My staff will be glsd to work with you to assure that the type of informatica you wish to see will be displayed in that report. Nonetheless, if we P. ave not begun to generate such reports routinely by the end of the first quarter in 1979, we will pro-vide you with a sumary report like the ene enclosed at that time.

79012303'6'

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Comissioner Kennedy As indicated in the Comission briefings on December 11 and 12,1978, those issues that have recently been approved as " Unresolved Safety Issues" will receive priority in the fiRR generic issues program.

In this regard, additional resources will be reprogramed to those tasks that address the " Unresolved Safety Issues" as necessary to assure their timely completion.

If I can be of any further assistance in this regard, please let me know.

Af Harnld R. Denton, Director Office of fluclear Reactor Regulation

Enclosures:

As Stated cc: Chairman Hendrie Comissioner Gilinsky Comissionur Bradford Comissioner Ahearne K. Pedersen, OPE S. Chilk, SECY

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ENCLOSURE 1 AflilUAL REPORT SECTIONS CESCRIBING THOSF CATECORY A TASKS THAT HAVE BEE!! IDE?lTIFIED AS "UtlRESCLVED SAFETY ISSUES"

s 5

WATER DAMP 8E2 MENEMC TASK A-11 Water namer events, that is, intense pressure pulses in fluid systems, sucn as comonly excerienced unen rapidly closing a water faucet, often occur in nuclear power plant fluid systems. Since 1971, about one nundred incidents l

involving water Namer in nucletr power reactors have been reported. These water harmer incicents have involved many types of fluid systems including steam generator feed-rings, feedwater and steam supply piping, residual heat removal systems, emergency core cooling systems, containment spray systems, and service water systems. The incidents have been attributed to such

.auses as *he racid concensation of steam poc1tets, steam-driven slugs of water, pump start-up with :artially empty lines, and racid valve motions, i

tst of the damage has been relatively minor, however, there have been several cases of failure or partial failure of system piping, i

4 No water hamer incident has resulted in the release of radioactivity out-T side of a plant. However, the principal concerns are that water harmer could result in the fatture of a pipe in the reactor coolant system or disable s system required to cool the plant after a reactor shutdown.

Means to prevent one particular type of water harmer caused by the rapid condensation of steam in the steam generator feed-rings of some pressurized water reactors are being instituted. Applicants with new staam generator designs are being required to demonstrate through test or analysis that water hamer will, not occur in these designs. Plants with steam generators of the top feeding type that are subject to water harmer, are being required to modify the feed-rings and/or test the systems to a?sure water hasser will not occur. Other actions to correct the specific causes of enter hasser identified to-date are also being required.

The MRC staff's review of this safety issue has been incorporated in the 'fRC

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Program for Resolutton of Generic Issues as Generic Tast A-i as described in a nport (NURE3 0410) to Congress submitted in,anuary 1978. The potentia!

for water h r in various systems is being evaluated and appropriate re-quirements and systematic review procedures are being developed to ensure that water hamer is given appropriate consideration in all areas of licensing re-views. The task also includes a study of potential water hammer phenomena to -

6 aid in the development of design and review procedures. A technical report providing the results of a staff review of water harmer events in nuclear power plants is scheduled for pubitcation in Decenter 1978. Issuaxe of this report completes a mejor subtask of Generic Task K-1.

The remaining suhtasks are expected to be completed in 1980.

a

7 ASYEltlC BLCWCCWN LCACS O's THE aEaCTCR CCCLANT SYSTDi (GBTt!C TASK A-21

!n :ne very unlikely event of a rupture of the pr* mary c:olant sioing in lignt water reactors, large nonuniformly distributed loacs would te incosec acon tce reactor vessel, reactor vessel intamals, and otner components in :ne reactor c clant systas. These newly icentified asycestric leads, unien result from the rapid ceoressurization of the reactor coolant syster, were not consicereo in the original design of some facilities. The forces associated with a postulated breat in the reactor coolant pioing near tne reactor vessel, for example, c:vid affect the integrity of the reactor vessel succorts and reacter pressure vessel internals. A significant cegree of fatture of the reactor vessel swooort sys-tem, along with 1::cacting the internals, has a potential for (1) camaging sys-tems cesigned to c:ol tae core follcwing ne postulated pioing break. (2) af-fecting ne capantlity of the centrol rocs to function procerly (3) camaging Other reactor c:olant system casconents, and (a) causing other ruptures in tne initially uncreaen reactor coolant systern piping locos and attacned syster s.

The NRC staff's review of this safety issue has been incorporated in the NRC Program for Resolution of Generic Issues as Generic Task A-2.

This program in-cluding the NRC staff's Task Action Plan for Task A-2 was described in a report (NURE3-0410) to Congress submitted in January 1978.

This issue wat originally identified in May 1975 by the Virginia Electric and Power C :rcany in relation to its North Anna Units 1 and 2 nuclear ;cwer plants.

A survty of all operating PWR reac.crs was concucted in Oct:bar 1975 which stewec that asyneetric blowdown loads had not been considered in the design of the reac-tar vessel supports for any ocerating PWR facility. In June 1976, the NRC staff requested all operating PWR licensees to assess the acequacy of ne reactor ves.

sel supports at their facilities wi.tn respect to these newly-identified leads.

e mst licensees with Westinghouse plants initially proposed an augmented inservice inscection program (!$!) of the reactor vessel safe-end to end pice welds in lieu of providing tae detailed analysis requested by the MRC staff. Licensees with Ccmoustion Engineering plants subsitted a ;robability study in support of a conclusion that a break at the location in the pioing necessary to produce the postulated load ' tad sucM a low ;rocacility of occurrence that no further anal.

ysis was necessary. Licensees with Saccock and Wilccx plants tcot an ao-preach similar to Comoustion Engineering licensees.

- 3 The NRC staff's revien of these proccsed alternatives to datatied plant t;ecific analyses has been c:rcleted with the esnclusion nat the af ternative precosals shculd not be ac:epted in lieu of the requested analyses.

Accordingly, the NRC staff sent letters on January 25, 1978 to afJ Phit licensees and applicants stating that an analysis must be uneartaken to assess the design adequacy of the reactor vessel supports and other structures to withstand the loads. hen asynnetric loss-of-coolant accident forces are taken ints account. As part of Task A-2, the NRC staff will review and approve analytical meets and ctmsuter c: des fe.

vele:ed by reactar vendors 'a calculate asytwetric bicwdown foadings grior to their use my licensees and appit: ants in plant.s ecific analyses. In accition.

the staff will deveico explicit guicelines' and acceptance criteria for t3e asym-metric 1 cad analyses and will c:nduct a pi;e break pretacility study to c:nfirm the acequacy of staff decisions related to the continued ::eration of plants for the interim period while Task A-2. 31 ant. specific analyses, and necessary plant fodifications are necessary.

Plant modifications to assure that the postulated leads are ac::nnodated Save been iriplemented late tc the construction stage of several plants and have been preocsed and are unjer staff review for some operating plants. For plants still uncer coerating license review, tne NRC staff requires inat plant-specific analyses be c::moleted and any necessary plant eccifications frolemented ;rfor to issuance of an operating license. The generic efforts for ;ressurf:ed ater reactors uncer Task A-2 are currently schecuted for comoletion in early 1979.

The NRC staff has been investigating this ;nenemena as it a:011es to boiling water reactors and has determined that asynnetric icads are also significant and therefer need to be evaluated for these icwer pressure systems. The staff is currently seveloping plans for excanding Task A-Z to resolve this issue for boiling water reacters.

.9 FRE!5L1t!!E3 WATER 4EaC*02 STEAM GENERATOR TUSE IMTE 2ITY f GE':ERIC ta5KS a-3, a a. a.51 ne hest amouced in tne react:r at a nuclear power plant is used to convert water into steam =nten will drive the turnine. generators. In plants eectoying pressurized water reactors, the primary c:olant water whien extracts heat by circulating through the reactor core is kept under pressure sufficient' to pre-vent deiling. *his hign. pressure =ater passes through tubes areund.hien a secondary c clant (also =ater) is circulating, under scesewhat icwer pressure.

Be water in the sec:ndary system is a11 cued to boil and produce steam to drive the turnine-generators. De assemely in unicM the transfer takes place is the steam generator. De tutes within it are an integral ; art of the ;rfrary

solant Soundary, keeping the racicactive primary coolant in a closed system and isolated frem the envircranent. The primary concern is the capability of steam generat:r tuces 1.' aintain their integrity during nomal o;eraticn and postu-lated accident c:nditions. Ir. addition, the recuirements for increasec steam ge:erator tune inspections and recates uve resulted in significant increases in occu:4tional ex;osures to worters.

A detailed discussion of the specific grenless associated with steam generat:-

tume integrity that were occurring at Ocerating reactors was provided in the 1977 1RC Annual pecort, page 95. ne infomation tetow is provided to suople-meet and 4;cate that information.

'Corrosien resultingin steam generatar tute -all thinning has teen eserved in several Westingneuse and Cemeustion Engineering (CE) plants for a nuceer of w jor changes in their secondary water treat:ient process essentially a

years.

eliminated this fers of degracation. Another major c:rresien-related pee-no:nenon its also been caserved in a nu::cer of plants in recent years, resulting frem a build-up of su; port plate cornston preducts in the annulus betwen the tuces and the support plates. Bis buildwo eventually causes a diametral re.

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cuctica of tuces, called 'senting,* and defomation of the tuce succort plates.

Bis ;cene:nenon mas resulted in other associatec events including stress c:r.

resion cracking, leans at the tace /su; port plate intersections and 'J-cend sec.

tion cracaing of tuces wnica were signly stressed Decause of succort plate cefor atien.

In %y 1977, tune centing =as af scovered at Millstone '.; nit 2 ano t.w %ine Yanaet At:mic Pcwer plant soth of =nica had operated exclusively with an til

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. 10 volatile treatment (AVT) of the secondary coolant. It had been thougnt that this type of treatment signt preclude the denting phenomenon from causing signifi.

cant degradation. The significant developments in Westinghouse and Combustion Engineering steam generators, since June 1977, were the following:

Continued tube denting at Indian Point Unit 2, San Onofre Unit 1, Surry Units 1 and 2, Turkey Point Units 3 and 4, and less-unounts of denting at a number of other 'destinghouse designed reactors. 4 team generator replacement is planned for early 1979 or 1980 at Surry Units 1 and 2.

Replacement or retubing is also being considered for Turkey Point Units 3 and 4 In the interim, the units are coerating under restricticns imposed by tha NRC.

. Discovery of su: ort plate tracking (related to denting) at Indian Point Unit 2 and San Onofre Unit 1.

Removal of several tu:es and a section of su: port plate at Indian Point Unit 2 to investigate the potential for steam generator cleaning revealed continued active corrosion of the support plate.

Continuation of tube denting at Millstone Unit 2 and Maine Yankee and discovery of danting in St. Lucie 1.

Millstone Unit 2. Maine Yankee, and Arkansas Nuclear One Unit 2 have removed lugs and portions of the solid rim in the uppermost su; port plates to recuce the susce:tibility of the plates to denting-related cracks (CE designs).

Palisades Mur' ear Power Station is sleeving degraced cuces instead of plugging them. This process restores the structural integrity of the tubes while keeping them in service (CE design).

Steam generator tube degradation in Sabcock and W11com (BW) steam generators has been limited to the Oconee Nuclear Plant where the first tube leak occurnd in July 1976. In the last quarter of 1976 and the first quarter in 1977, there was a total of seven plant shutdcwns to plug leaking tubes in the three Oconee anits. To-date,14 tuce leaks, all at the Oconee units, have occurred in SW steam generators. The siajority of these leaking tuozs were located adjacent to the open inspection lane. Laboratory examination of removed defective tunes indicated that the tuce failures were caused by the presagation of :ircumfer.

ential fatique cracks, of antnewn origin, by flow-induced vibration.

e It -

The significant developments in SW steam generators, since."ay 1977, were the following:

Continued tube leaks at the Oconee units.

Initiation of a deconstration tube sleeving program by Duke Power Cor:many at the Oconee units. The tube sleeves will not serve as part of the priory coolant boundary but af t1 he installed to change the vibrational charact -

istics of the tubes and decrease the dynamic stresses and the susceptibility of the tubes to fatigue cracking.

Fo11cwing inspections by licensees of their steam generators and the completion of any necessary repair programs, the.1RC usually must approve or concur in the re-start of each of the severely affected facilities. To-date, the units severely affected by the tube denting have comoteted inspection and repair programs and received NRC approval for operation for ifmited time periods. Safe operation is assured by the incosition of strict condition

• an licensed operation, re-quiring the plugging of affected tubes and restricting allcwable leak rates Qing operetion.

As the 'tRC staff continues to closely venitor, evaluate, and approve the ac.

cestability of continued oceration of plants exceriencing steam generator tube proolems, it nas undertaken a number of generic reviews and studies as part of three generic tasks in the NRC Program for the Resolution of Generic Issues; specifically, Generic Tasks A-3, A 4, and A-5 eaca directed at the particular problems of Westinghouse, Combustion Engineering and Sabcock and Wilcox plants, respectively.

Under these tasks generic studies will be conducted *.o (1) evaluate intervice inspection results from operating reactors (2) evaluate the consequences of tube failures peer postulated accident conditions, (3) evaluate tube structural integrity, (4) establish tube piugging criteria based on new infomation, (5) define the requirements for monitoring secondary coolant chemistry.

(6) evaluate inservice inspection methods, and (7) review design fmorovements proposed for new plants. Dese studies will be used to revise current NRC staff requirements and quicance regarding these suojec*.s. In addition, under Task A-3, the NRC staff wi*i review and evaluate the first proposed steas generator replace-ment operation to establisn acceptance criteria and guidance 01 a generic basis for use in the review of subsequent replacament operations. These generic

• asks are currently scheduled to be completed in early 1980.

12

!WR "A2K ! AnD " ARK !! 22 ESSURE SUPPeESSICM CONTAINMENTS (GENERIC TASKS A-6, A 7, 4-3, A-391 In the course of perfoming large scale testing of an adva iced :esign pressure-suppression containment (Mart !!!), and during in-plar.t testing of. art I con-w tainments, new suppression pool hydrodynamic loads were identifico.nfch trad not explicitly been included in the original wart I or Mars II containment design basis. These additiona! Ioacs result from dynamic effects of drywell af r and steam being rapidly forced into the suppression pool (torus) during a postulated LOCA and from suopression pool response to various modes of safety relief valve (SRV) operation generally associated with plant transient operating conditions.

Since these new hydrodynamic loads had not been exolicitly considered in the original design of the Mart I a.id Mark II containments, the NRC staff detemined that a detailed reevaluation of these contairment system destqns.as reouf red.

As a result of the need for this reevaluation the affected utif f ties formed ad hoc Mart I and Mart !! Cwners' Groups and esca has engaged the General Electric Co=pany as its program manager. Zotn Owners' Groups developed two-onase ;rograms consisting of a short-term program and a long-tem program for resolution of the The Owners' pool dynamic concerns for their respective containment designs.

Groups' programs consist of among other things, a numoer of comprehensive ex-perimental-aad analytical programs to establisa generic pool dynamic Icads. Icad coccinations ~4i.1 design criteria.

The NRC staff has identified and initiated a numoer of generic tasks to review and evaluate the results of the Part I and Mart II Owner's Gecup short-tem and long-tem programs to develop tecanical positions for use in licensing actions on These individual plants utilizing the Mart I and Mart II containrient designs.

generic *. asks are ircluded in the NRC 8cogrts for Resolution of Generic Issues (described in ?!UREG-0410 as noted above). Speciffcally, they are Task A-6, Mart ! Short-Mm Program; Task A-7, Mart I Long-Ters Program; Task A-4, Mark II Contaimeent Program; Tast A-39, stemination of Safety Relief Valve ($2V) Pool Dynamic Loads and Temoerature Limits for 3h2 Containments.

The cafectives of the Mart ! Short-Tem Program were: (1) to examine the contain-rent system of esca BWR facf11ty with a Mart I containment design to verify

• hat

13 it nould.aintain its integrity and functional capant11ty when subjected to the " lost probable hydrodynamic loads induced by a postulated design basis loss-of-coolant accidentt and (2) to verify that Itcensed Mart I EWR facilities may continue to.perate safely, without undue risk to tPe health.nd safety of the public, wn;te : zet*!odical, c =prenensive Long-Tem Program is conducted. The NRC determined that, for the Short-Term Program,

• maintenance of containeant integrity and function" would be adequately assured if a safety factor to failure of at least two were camonstrated to exist for the weakest structural or mecnan-ical component in the Mart I containment system (i.e if the calculated stresses in all c:mcenents of the affected containment structure were snown to be less than one-half the stress which sculd cause the cceponent to icse its structural integri ty). The NRC concluced that the ocjectives of the Short-Tem Program had been satisfied and documented the basis for this c:nclusion in the

• Mart !

Contaiment Short-Tem Program safety Evaluation Report,* 4LRE3-C4C8, dated Oecameer 1977.(1.e., Task A-6 =as completed in Oecemcer 1977).

The objectives of the Mart I Long-Tem Peceram are: (1) to estaclish design casis loads that are appropriate for the anticipated life of esca Mark I !WR facility, and (2) to restore the original intended design safety margins for esca Mart I containment system. The Mart ! Long-tem Program consists of a series of major tasks and sustasks wnten are designed to provide a detaf f ed tasis for hydrocynamic load definition and the metnodology and acceptance criteria for the structural assessments. The generic aspects of the Part I Long-Tem Pr) gram will te described in a Plant Unique Analysis Applications Guide, seneculed to be completed in Oct eer 1978, and in the Load Oefinition Report, senecuted to se c:molated in Oecember of 1978. Sunsequently, eacn utility with a Mart I plant will perfom a plant-vique anaissis using approved Icad deffnitten and struc-tural analysis tecnniques to deceristrate conformance with the Mart I L-ng-Term Program structsral acceptance criteria. These analyses are currently scheduled for c:moletion in October 1979.

The schecuted completion cate for the Mart I Long-Tern Program (Tast A-7) in-cluding the issuance of license amencments and the fealementation of any plant modifications necessary to satisfy the " art ! Long-Ters Pregeam struc*. ural accept-ance criteria, is ecemoer 1980. In recognition of this sesecule, a numcer of facilities are acopting their own senecules to imolement anticteatec plant mdi-fications and sinimize the potential for extended plant outages or anscheduled outages.

1s n, sjective of De MC staff's ef' orts uncer Generic Tast A-8 related to es

" art II Shcrt-Tem Pr: gram (577).as to review and evaluate ce ; col :ynamic Icacs associated.ith a :estulatea large less-of-c:clant ac:icent ;re:csed my the

" art II 3.cer's 3rcus ts setemins their ac::: tan 111ty for use in ;lant uni se analyses. The " art II Short-Tem =cogras.as c:mpleted in Oct:ter 1978 and ccu.

mted in *; REG-0487. ** art II C:stairment Lead Plant Program Lead Evaluation and Ac:e tance Criteria." With regard ta the Mart !! L:ng-Tem Program (LTP), tSe MAC staff will evaluate t3e results of t3e " art !! c nfir at:ry experimental and analytica! ;regra=s to atsess the margin *:r selected !cacs. The " art II L:ng.

Tem Pr: gram is :urrently scaetaled f:r c:;ricletien in Oct::er 1980.

"ncer 3eneric Tast A-39, tse 4RC staff will review anc evaluata t'e results of t*e Nrt I and " art II Owners' Groun's ex:er* mental are aaalytical rograes to estantism ar4 justify Me safety relief valve-related ; col yramic icacs 'er T.R " art ! ar'

• art II c:ntainnent cesigns. Se results f Seneric Tast 2-39 mill :e an integral : art of tre final ac:estan111ty of ee.9rt I and N et II pressure sue:ressica ::ntainment designs. This ;eceric tast is cur-ently sc*ecuted for emletion in Oeceecer 1979. An interia assessment of miti:le-censecutive 52V discaarges is currently being :er# reed *cr t3e c;erating Mart I facilities to su::cr ceferral of tais issue until tse c:.cletion of ne

" art I L:ng *em :mgram. The review of these assessments is screcuted f:<

c:=cletien in icve-ter 1978.

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15 ANTIC PATED ToAMs !NTS WITkOUT SCRAM (GENERIC TASK A.11 Nuclear plants have safety and control syste.rs to limit the consequences of temporary abnormal operating conditions or "anticfpated transients." Some deviations from normal operating conditions may to minor; others, occurring less frequently, may impose signU ncant demands on plant equipment. In some anticipated transients, rapidly shutting down the nuclear reaction (initiating a

• scram *), and thus rapidly reducing the generation of heat in the reactor core, is an important safety measure.

If there were a potentially severe " anticipated transient

• and the reac'Jr shutdown system did not " scram

• as desired then an

• anticipated. transient.without. scram,"

or Aid 5. euld have occurred.

• his issue has been discussed throughout the nuclear industry for a number of years.

Historically. the regulatory staff has excluced very low probacility cents from the design basis. At issue in the ATd5 discussions is unether or not the procability of an Ard$ event is sufficient'y low to warrant the continuance of the current staff practice with regard to AW5. i.e.

continued exclusion from the design casis for nuclear power plants laecause of its low protability.

Because of the perceived potential for serious consequences resulting from AWS events a numcer of studies have been undertaken to assess *he procacilities and consequencas of such events. These studies mave teen performed 3y vendors.

'atility groups, and by the AEC and NRC regulatory staff. The ATW5 issue was in.

corporated in the NRC Program for Resolution of Generic !ssues (described in WRES.

3410 as noted above) as Generic Tast A-9.

In Septancer 1973, the then.AEC staff punitshed WASH.1270. " Technical Recort on Anticipated Transients Without ! cram for ' dater Cooled Power Reactors,'.nich set forth staff

• acceptance criteria

• to protect against Ard$ events. Ouring the Dvo.

year period folTowing pubitcation of the staff report, each of the four reactor

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manufacturers submitted analyses and supporting information on AWS wnich.es re-viewed by the NRC staff and addressed in four status reports publisned in Oecameer 1975. The staff reports evaluated the infomation for conformance to the WASH.1270 critsria and noted unere design changes.and additional analyses ere recuired.

The vencors and owners have questioned nether the NRC staff's requirements are neces.

sary and justified. The induttry contentts that the probaattity of an ATd! event is significantly less t.* tan estimated by the NRC staff and so tow as to maze AWS events sinor safety concerns in If ght water reactor coerations.

- !6 He_ause of the continuing controversy over the NRC staff position since its publication in WASH-1270, a staff review and evaluation of all the infor.ation available on the subject of ATds, and in particular, the material develooed subsequent to the publication of the staff status reports referred to above, was undertaken in the latter part of 1977 and early 1979. A report, i.:$tEG-0460, was published in April 1978 providing ' ' results of this review and evaluation.

It was concluded in WREG-0460 that considering the expected frequency of transients. the reliability of current reactor scram systems necessary to meet the safety cofectives has not been comonstrated and may well nave not been attained. WREG-0460 recomended that means of mitigating the con-sequences of ArdS events te provided in plant cesigns.

The recomendations presented in WREG-0460 have been criticized by industry and some membe9 of the staff as unnecessarily conservative and therefore too costly. The staff is nowevaluating alternative means of reducing the probability or consequences of Aids events, other than that recomended in WREG-C460. The effectiveness, cost and other factors such as the effect on the licensing pro-cess of these alternatives is betn1 evaluated. Based on this evaluatien, the staff will reconsnend to the Cxenission the alternatives unica provide the best balance between safety and cost for new designs, plants unoer construction and operating plants. The staff eigects to provide its recomendations to the Comission in early 1979.

17 SWR N0ZILE Cp>CXING (GENERIC *Att a 10)

Over the last several years, inspections at 20 of the 23 boiling water reactor (SWR) plants licensed for operation in the U.S. have disclosed suit degree of cracking in the feedwater nozzles of the reactor vessel at aII Dut two facil-ities. The exceptions were a plant with less than one year of operation and a plant with welded nozzle thermal sleeves. The three other facilities M - 1ot yet accumulated significant operating time and have not yet been insp-The feeowater nozzles, part of the " pressure vessel," are an integral part of the primary pressure boundary of the reactor coolant system and the second bar-rier (after the fuel cladding) to the release of radioactive fission procucts.

All of the repaired SWR fee <h ater nuzzles met the ASPI pressure vessel code Ifmits, however, and no icinediate ac* ion was necessary. Because relatively small amounts of base metal have been renoved by repair operations, there has been no significant reduction in safety margins. Several plants Mave removed the stainless steel nozzle cladding as a means of eliminating crack initiation since the clad thickness was not necessary to twet code reinforcement require-ments. Nevertheless, the cracking is potentially serious because:

- Excessive crack growth could lead to impairment of pressure vessel safety margins requiring more c: implicated recair were tnan siepte grinding.

- The design safety margin could te reduced by excessive removal of $4se metal.

- The exposure to radiation of the personnel performing inspection and recair tasks can te considerable.

- The recair of tasse kinds of cracks can result in consideraole shutdown time at the plant affected.

The reactor vendor (the General Electric Coacany) and the NRC have concluded from their respective studies that the cracking is caused by fluctuations or

' cycling

• of the temcerature on the inside surface of the nozzles; that the stainless steel cladding exhibited less resistance to crack initiation tPan the underlying low-alloy steel; and that, after initiation in the stainless steel c!accing, cracks can be propagated by operational startup and shutoown cycles or other ocerationally-incuced transients. The vendor has perfor ad extensive analysis and testing to confirm the suspected cause af the cracking and to ancover possible long-tars solutions - a newly designed siaeve, removal of the

18 stainless steel cladding, reduction of the temperature oifferential at the noz.

21e. or some comoination of these. D e licensees involved have increased the number and extant of inspections of feedwater nozzles, with careful repair and reinspection where cracks were found. The vendor advised these licensees to closely monitor startup and shutdown procedures in an effert to substantially reduce the time during which cold feedwater is being injected into the hot pressure vessel.

In a closely related area, the NRC was informed in March 1977 by the General Electric Company that a crack had been found in the nozzle of the " control rod drive (CRD) return line" in a reactor vessel in a foreign country. The CRD return line~ nozzles are the openings in SWR pressure vessels through which the high pressure water in excess of that needed to operate and cool the CRDs is returned to the pressure vessel. Later in March, the Philadelphia Electric Ccmoany reported that similar cracking had been found in the CRD return line nozzle at its Peach Bottom Atomic Power Staticn, Unit 3.

The cracks reses01ed those found i, the feedwater nozzles and seemed to be the result of the same kind of cy..fc therssi stresses that were causing feedwater nozzle cracks. Both the foreign reactor and the Peach Bot *.ca Unit 3 reactor are reoresentative of a small nur.cer of SWRs which do not have a thermal sleeve in the CRD return line nozzle.

The licensee removed the cracks in the Peach 3cttes CRD nozzle by grinding out the cracked area, the maximum crack depth being 7/5-inch, and returned the unit to operation with the CRD return line " valved out" and with the flow and pres-sure in the CRD hydraulic systes modified.

Inspection of*other CRD return line nozzles which incorporated tnermal sleeves indicated that these sleeves may not be effective in preventing this cracking phenomenon. The Georgia Power Company found a crack in the CR0 return Ifne nozzle at i s Hatch Plant, Unit 1 which did have a thermal sleeve. (~he t

e.ack was removed, the nozzle capped, and the return line rerouted to the reactor water cleanuo systan.)

- 19 The NRC staff efforts related to the resolution of these two similar issues regarding nozzle cracking in boiling water reactors were consolidated into a single staff effort, Generic Task A-10, in 1977. Under Generic Task A-10, the staff issued interin guidance to operating plants in a report entitled,

• Interim Technical Report on 8WR Feedwater and Control Rod Drive Return Line Nozzle Cracking" in July 1977. The staff is often requiring inservice inspection using liquid penetrant examinations at operating reactors in accordance with the frecuency, procedures and acceptance criteria described in the above report.

Additional efforts under Generic Tast A-10 include following and reviewing ad-vancements in (1) the development and testing of effective feedwater nozzle thermal sleeves and spargers, (2) life-cycle testing of certain CR0 system valves. (3) the development of various feedwater system and CR0 system *iodi-fications, and (4) the developrant of viable ultrasonic system tec.niques by the nuclear industry to allow reliable and consistent early deternination of cracking from positions exterior to the reactor vessel.

Generic Task A-10 is scheduled for completion in lata Ig77.

+

e

23

E C R YE!!EL " ATE 4fal! *0f M ES$

(GDERIC TASK A-11) tasistance to crittle fracture, a rapidly =rcoogating catastro:nic fatlure moce, for a corconent c:ntaining flaws is :escribed tuantitativeiy my a Mterial Orcoerty generally denoted as fracture toupness. This resistance to fracture, or fracture touganess. Mas cifferent values and characteristics cepending upon the material being censidered. For nuclear reacter pressure vessel steels, three consicerations are important. First, fracture *.ouganess increases.ita in-c'tasing temoerature. Second, fracture touganess decreases mita increasing lead rates. Third, frac *.ure toughness decreases with neutron irractation.

In etecqnition of these consicerations, powee react:rs are ::eented within restrictions ircosed by tse Tecanical 5:ecificaticns en t?e :ressart turieg meatus and coolcewn coerations, nese restrictions assure t3at the reactor vessel will not de si.njected to that coscination of =ressure and tescerature t34t could cause brittle fracture of the vessel if significant flaws in :ne vessel meterial exist. The effect of neutron radiation on t3e *racture tougeness of tRe vessel meterial is accounted for in develocing and revising tSese Tecanical Scacification limitations over the life of t3e :lant.

Fce t~e service times and coe* sting conditions tyli1 cal of current ::erating 21 ants, reactor vessel fracturt *.ougnness procerties =revice acecuate Sargies of safety against vessel failure. Furtter, for mst slants ce sessel.ater*al

rece-ties are saca that acecuate *cacture tougnness :an :e maintatred over t3e life of tse plants. %ver, results from react:r vessel surveillance :r: gram indicate t*at up to 23 oicer operating pressurized.ater react:rs ere *acricated wita saterials that will have marginal ::ughness after comparatively snort

eriods of Operation.

Ne :njective of Tass A-11 is to evaluate saterial :egracation recsanis=s re-sui:1eg fr:m wat-on treaciation and detemine a :recr4 ate licensing criter4a and c:r-ective actten *cr Iow ::ughness react:r vesset 9ater als in rese

h 21 -

currently Itcensed plants. Task A-11 is currently scheculed for comoletion in July 1979. This emeletion date is well in advance of the date needed to

&Ssure that adequate fracture toughness is maintained in these older plants.

2

22 FUC-"eE *0'JGKiEIS 30 SCW!AL FOR LW Ls2 'U2!Z OF WR 5*UM 3DEU*02 20 *EAC"02 *0CLL1T *W SUPDCE*5 (GD DIC TASK L12)

Ouring tne course of liceir ig review for a scecific Pressur* ec.ater Reactar (NR) a nweer of tuestions were raised as ts (1) tre acecuacy of tMe fracture tcugnness prncerties of the : material usea ts facricate :e reactor coolant pumo and steam generator supports and (2) the potential for failure due to la::lellar tearingef these same suoports. The safety concern is tnat, althougn these succorts are cesigned for acrst-case accicent cancittons, poor fracture tougaress or leiellar tearing could cause the succorts to fail if severely loaded during sucM accicents. Saccort fa11ere could c:nceivanly f acair the ef'ectiveness of sysm cesigned to sitigate the c:nsequences of the acci-cent. An esancia of a postulated event sequence of ;:otential c:ncer9 woula :e a large nice treat in t*e reactor c:clart system enica severely leacs the succorts, follewed by a support failu-e of sufficient =agnituce inat a major c cconent such as a steam generater is severely disalaced resulting in failure of tMe emage-y care cooling system pfoing unich is neeced to ;rovice cooling water to the c re.

Two different steel specificattens (A53 A36-704 anc A53 A572 7Ca) covereo rest c" the material used for the succorts of ne M in cuestion. To accress t*e fracture taughness at.astien D.ellar tearing is discussed secarstely te10w) tests not ortgtnally spects**ed and not in !?e relevant A53 s:ecifica-tions were race on those nests of steel for unica excess material =as avail-anle. The toughness of the A36 steel was found ts te adecuate. Out

• e tougn-ness of the A572 steel was relatively poor at an operating terce-sture of 30*F.

In :Pe case of the M in :uestion, the acclicant agreed to a license c:ncition anich stated Mat 9e muld raise t*e temcerature of tse A53 A572 teses in the 4

s steam generator succorts to a siniram terceriture of 225'F erior to reactor coolant system pressurization to levels above 1000 psig, assuring adecuate tougnness in the event of an accident. Auxiliary electrical neat will te ned to sucolarent the heat :erived from tMe react:r c:olant loco to sotain tre re-sutrec sonesting tancersture of tne sa: port saterials.

Because similar ater*als and esigns 9 ave :een usea in other slants and the-efere si: sitar reciers.ay exist, =e save f nc:r:oratec tse eview of tais issue in t*e 1RC *mgram 'er Resolutten of Generic :ssues as Gecerte Tass 4-12.

e

I 23 A consultant was engaged to reassess the fracture toughness of tre steam generator and reactor coolant pump succort materials for all operating M plants and those in the later stages of operating license review.

De staff has ccmoisted a review of the materials utilized in the supports of 34 potentially affected PWRs. Based on the consultant's pre-li31 nary evaluation, we have detereined that there are accroximately 15-20 31 ants unose suCQorts have cuestionabl' Mugnness. We exDeC0 that these plants may be required to utili2e inservfCe inspection or ausiliary heating of adequate tougnness properties cannot be demonstrated.

"pon comaletion of our generic study, we will document *Me generic

nase of the frscture tougnness program and will begin to implement the results on a plant-scecific tasis. De generic solution will result in changes to the Standard Review Plan to incorporate the lessons lear 9ed for use in future license reviews.

De staff has concluded that continued operation (and licensing) of PWRs is justified pending congletion of this task anc implementation of the task Msults. Support failure is not excected to occur except uncer t'a unlikely carcination of:

(', De occurrence of an initiating event (e.g., a large mi:e creat) which has been determined *4 he of low ;ropasility (normal operating stresses on Diping are very low).

(2) *be existence of non-recuncant and critical sucocrt structural mercer (s) with low fracture tougnness (r.any succorts contain re-cuncant meeners).

(3) De existance of support structural mercers at operating teccer-atures low enough that the fracture tougnness of the succort material is recuced to the level that brittle failure could occur if a large

" law existed.

(4) ~Pe existence of a ' law of sucn large size that *ne stresses im

ar*ed :ur*ng the initiating event :oula cause tne flaw to essioly retagate Msulting in 3rittle failure of the iercer(s).

24 The second potential concern (i.e.. Iamellar, tearingb)sayalsosea problem in those succort structures similar in design to the aforementioned WR.

However. continued operation of WRs during our continuing generic review of this concern is acceptacle, based en the fact that a review to date of approx.

imately 400 related technical documents revealed only one instance of known failure from lamellar tearing. This fatture occurred in often-strassed truck brakes. In addition, the fac'ars considered above for the fracture tougn-ness concern, sucn as low stresses during normal operstion and the Icw proca.

3111ty of an initiating event ecually apply to this concern.

The geceric fracture tougnness program is expected to be coc:aleted in August 1979. The lamellar tearing evaluation is a longer tar; effort and is exoected *4 be concleted in 1981.

1'Lamellar tearing is a cracking onencuene unich occurs beneath welds and is princisally found in rolled steel plate faericaticas. The tearing always lies within the sarent plate, oftsn outsice the transforned (visible) Neet-effected zone (HAZ) and.is generally parallel to the weld fusien boundary. Lamellar tearing oc:urs at certain :ritical joints usually within large welced structures involving a %1gn gegree of stif* ness and nstraint. Restraint may te defined as a restriction of the movement of the various joint coreanents that wuld wrmity oc:ur as a result of excansioned contraction of weld metal and adjacent regions during weloing ("Lamellar Tearing irt Welded Steel Faorication*, The Weiding Institute).

. Sys*EwS IM*EuC*! CMS IN W EAR 3CWEM SLANTS (GENERIC *A$K a 17)

In Novemter 1974 the Advisory Co:inittee on Enactor Safeguards requested the 1RC staff to give attention to the evaluatten of safety systems f*om a multi-disciplinary coint of view to icentify potentir.lly undesiracle interactions between plant systems. The concern arises because the casign and analysis of systems is frequently assigned to teasts with functional engineering specialties such as civil, electrical, mechanical or nuclear. The questien is whether the wort of these functional specialists is integrated in the design and analysis so as to identify adverse interactions bet een and among systems.

w e

These adverse events signt occur because designers signt not, for example, as-i sure that redundancy and ineependence of safety systems are provided under all conditions of oceration where reduncancy and incecendence is recuired because the ' unction 41 tavs may not ce aceouately coorcinated. Simoly stated: the left nand may not know or understand what the rignt hand is doing in all cases where it is necessary for the hancs to be coordinated.

The NRC staff telieves thetits current review procedures and safety criteria provide reasonable assurance t!tet an accent 201e level of reduncaNy and inde-

endence is rovided for systems that are required for safety. Nonetheless, in mid-1977 this task (Task A-17) was initiated to investigate systems inter-action frem the point of view of confirming that our cresent ;rocedures acceptably actount for potentially undesiracle interactions netween and among systems.

• he 1RC staff's cur-ent review procedures assign primary responsibility for various technical areas and safety systems to specified tecnnical review organizational units and assign secondary rispensibility to other units enere there is a functional or interdisciplinary relationsnia. Designers follow somewnat similar procedures and pmvide for interdisciplinary reviews and analyses of systems. Task a-17 will provide an inde:endent investigation of safety

• unctions and systeres required to :er#crs these functions in order to assess the acecuacy of curnnt review rocedures. *his investigation sill to conduc*ed oy $4ndia LaCoratories Jncer contrac* assistance to the NRC staff.

25.

The centract effort Phase I of the task, began in May 1973 and is excected to be c moleted in Sectemoer 1979. The Phase ! investigation is structured to icentify where interactions are possible between and among systems enere these interactions have the potential of negating or seriously degrading the performance of safety functions. The investigation will then 3roceed to identify where our review procedures may not Mave properly accounted for these interactions. Finally, Desed on a deter'sination of the overall sig-nificance to safety, in a follow-on Phase !! of the task, scocific cor.

rective measures will te taken in areas wnere the investigatica shows a need.

As acted above, tMe NPC staff telieves that its review arecedures and ac.

cectance criteria currently provide reasonable assurance *?at an accectacle level of system recuncancy and incocendence is proviced in plant Jesigns and this task is excected to confirm this belief. Mcnetseless, because adverss systems interactions art potentially of large significance to plant

'~

safety tnis issue has been includad as an "Unresolvec Safety Issue." If no significant system interactions are identified in the Phase I investigai. ion described above, as is expected, this issue will not ce treated in subse-quent recorts as an "Unresolvec Safety Issue.*

27 ENVIRC'MNTAL O'3LIFICATIC': CF

!AFE*Y aELATE') EtEC*oICst EOU!pw!17 (GENERIC TASK A.24)

Cespite the conservative design, construction and operating practices and quality assurance measures required for nuc! ear ; der ;Iants, safety sys.

tems are installed at nuclear plants to eitigate the consequences of postu.

lated accidents. Some postulated accidents could create severe environmental conditions inside of containment. De most lietting of these accidents are hign energy pipe breaks in the reactor coolant system pipin, or :n a main steam ifne. In either of these cases, the release of hot pressurized.ater and steam to the containment creates a hign tem:erature environment (250 to aCO'F) at hign humidity (including staare) and pressure (as Sign as atout 50 pstg).

For some applications, chemicals are added for fissien product removal to tne containment sprays that are useo to reduce the pressure in :ne containeent.

Additionally, some electrical equipment is predicted to be suceerged following a large pipe break. Bus, the safety equipant is exposed to suca envirovental conditions and needs to remain opersole during this period, as well as for the long-tene post-accident period.

De NRC requires that electrical equioment in safety systems, principally the emergency core cooling syste!e and containment isolation and cleanuo systems.

De environmentally qualified to assure that this equi; ment will ;erferar its re.

5:e-quired function in the envircreent associated with suca severe ac:fdents.

cific electrical equipment of concern during postulated accident ccnditions includes (1) the instrumentation neeced to initiate the safety systems and provide diagnostic information to the plant ocerators (e.g., electMcal ;ene.

trations into containment, any electrical connectors to :aoting.nica trans.

mits signals, and the instruments themselves). (2) control power to -otor operators for certain valves (e.g.. ECCS and containmeret isolation valves located insit'e containment), and (3) f an cooler motors for those plants that utilize fan coolers for containment heat renoval.

• he current NRC safety review p-ocess for nuclear ;cwer ;Iants incluces cri-teria *eleted to the valification of cert.in electMcal equipment. *hese criteria recuire that electrical equipment imoortant to safaty sust :e :ual.

ified to function in the envircnment that signt result from various accicent

23 c:ndi tions. Althcugn such criteria have been acclied to varying degrees since the early days of c:reertial nuclear powr t?r? deta11s of these criteria have teen mere clearly defined over the years.

These clarifications of the criteria have raised some questions es to:

(1) the degree to unich electrical equi;mnt Jsed in alcer plant designs (those new :erating) is capable of withstanding the envircreental conditiens (pressure, tercerature. Mumidity. Steam, c?emicals, vitra-tion and radiation) cf various accident conditions ur. der =nich it 1

rust function (i.e., the *;ualification of equi;eent" in these alcer plants), and (2) tne adecuacy of test or analyses conducted for electrical equierent f a newer plants to " qualify

• such equipment as capable of withstanding the e editions of the envircnment created 3y various accidents during wnics the equipment must function (i.e., the

• adequacy

• of qualification tests).

Witn regarc "J alcer plants, the follewing actions Mave teen place in recent

.Tnths.

As a result of a Sandia testing ;regras seing :encucted for the *ffice of 'twelear 2tquiatory Researca, a generic safety concern with the acecuacy of environ.

antal qualification of certain electrical equi;rnent.as identified. *his issue was hignlignted by a 'lovemcer 4.1977 petition frem the Union of Oncer*

Scientists whip requested truediate action regarding ecerating ;cwr reactors and licensing actions for other ;recosed plants.

m

. Swosecuent NRC staff investigations in resconse to this issue have led.

as of June 30, 1978, to seven plant shutdowns for corrective action and extended outages for tw other plants to sake modifications. These actions were for the most part a result of a lack of conclusive infomation regarding the qualification of cartain safety equipm4nt.

Having identified problems associated with qualification of electrical equip-ment, the NRC conveyed its information to the licensees of all operating reactor facilities through an Inspection and Enforcement Circular =hich was issued on May 31, 1978. The purpose of this Circular was to ensure that the knowledge gained by the NRC staff and th, lessons f earned wovid be accro-priately factored into future actions. The 'tRC staff also ha$ initiated an aus ented inspection effort as part of the nemal BRC activities. This effort =1 1 concentrate on the inspection of installed safety-related elec-trical equipment and on an audit of the recorcs for environmental qualification.

1 Additionally, a review of the envirornental qualification of safety-related electrical equignent has been initiated for 11 operat!ng reactor facilities in the Systematic Evaluation Program (SEP). (See Chapter 7. "Abnorme1 Occur-rences - 1978.*)

With regard to the second question acove, the NRC staff has worked with the industry to develop standards for ecutement qualification and documentation unich would assure tar high level of equipment reliability required for nuclear apolications. This effort has culminated in the development of IEEE Std. 323

• IEEE Standard for Qualifying Class IE Equipment for Nuclear Pcwer Generating S tations." This standard and its ancillary standards have provided the focal point for the developt of environmental qualification requirements in recent years.

IEEE Std. 323 was first issued as a trial use standard (IEEE 5td. 323-1971) in 1971 and f atar, after substantial revision, as a final standard (IEEE Std. 223-1974) in 1974 Both versions of the Standard set forth tasic requirements for environ-mental qualification of electrical ecuipment sut do not provice scecific 1etails for imolementation of these reovirements. Soecific qu?.lification tecnniques } ave been reviewed and approved by tse NRC staff on a case-by-case basis as a part of individcal licensing actions. These licensing actions include initial con-struction pensit and operating license application reviews and requalification actions for operating reactors wnere documentation of the initial qualification was not available.

The evolutionary nature of the process of developing environmental qualification remirements and the case-by-case implementation of these requirements has re-sulted in a diversity of :nethods in use and different levels of documentation of the extent to which equipment is qualified.

Several aspects of eouipment qualification are being pursued at this ti*e by the NRC staff and the nuclear industry on a generic basis to achieve a more uniform implementation of the general qualification requirements establisned in IEEE Std. 323-1974 One such activity is the development of interim NRC

. staff positions regarding how the requirements of IEEE Standard 323-1974 can be met. This activity is a part of Generic Task A-24, "Ernrironmental Qualifi-cation of Safety-Related Electrical Equipment." in the NRC Program for the Resolutich of Generic Issues and is sc'eduled for completion in Further efforts under Generic Task A-24 involve the review of the environmental cualification programs of reactor sendors and architect / engineers as a barts for qualifying safety-related electrical equipant to the requirements of ;EEE-S tanda rd 323-1974 Performing these reviews on a generic basis rather than on case-by-case licensing reviews will provide resource savings for the NRC staff and the industry. This follcw-on portion of the generic task will be seneduled

~

following completion of the development of the interim NRC staff positions re-ferred to above.

AUC*02 VE!!C. 7t!!StaE *U'tSIENT *toi!CTI N f TE1!C *a5K A-25)

Over the past several years incidents identified as ;ressure transients

• ave oc:urred in pressurized water reactors (M). To-date, Dere P. ave teen taf ety.

three suc!t events. Half of these events oc:ur-ed befort ce plant acaf eved in-itial criticality (i.e., before initial speration of ee react:r). -he -aferity occurred during startup or shuta:=n operations. All of t"4 ;ressure transients were sucs that fracture rechanics and fatigue calculations indicata cat the reac*.or vessels were not dameged and c:ntinued operatica af Nevertsafess, t.e staff concluced that these vessels mas ac:ectatie.

a:prepriate regulatcry acticas were aecessary (1) to reface the f-equency of pressure transient evenu and (2) ts provice ecutanent vnica =culd restrict future transients to ac:sstatie pressures. nfs ac* ion was necessary because reactor vessel safety margins =culd be eeduced caring the Itfetime of ee vessel t e to neutr n irndistica causteg reduced aterial tougneess.

The WC staff's review of this safety issue was inc:rporated in tm MC 7-gram for Resolut1=n of Generic :ssues as "eneric Tast A-25. De final report, MURE.-

C224, *Reac:sr Vessel Pressure Transient Protection for Pressurned Water Rese.

tors," =as issued in Septe cer 1978. This ;reviously *"nresolvec Safety Issue

• Pas teen -esolved.

"; graced precadural c:ntrols were festemented at ccerating - t 'acilities N

Se

.nica significantly reduced tM cc:urrence of ressure transient events.

few events nnica c. ave octur-ed were not significant and =ere of t'e ty;e t34t will be precluded by equissent enanges.

The ajor4ty of tne etufprent c3anges interented at :cerating M. facilities invelve t."e addition of a second lower set point en existing ;cwer egerated relief valves, the addition of new spring-loaced relief valves, or modifications to allow ase of existing spring-loaced relief valves. A few newly licensed facilities sust c:rclete sisitar design c.anges by tMir first refueling saut::=n; Se extenced e:uisnent igle-entatien senecule for new facilities was based acon t.'e reduced fretuency of occurrence af ressure transient events tse to fe:reved

rocecursi

:ntrols and tne large safety nargins for new pressure vessels.

_ RESI:X AL MEAT 2DCVA: SHUT *JC' M REOUTRE.wENTS (GENERIC WK A-311 The safe shutdown of a nuclear power plant following sn accident not related to a loss-of-coolant accident (LCCA) has been *ypically interpreted as achieving a hot-stanchy condition (i.e., the reactor is shutdown, but system temperature and pressure are still at or near normal operating values). Consecuently, con-siderable emphasis has been placed on the hot-standby condition of a power plant in case of an accident or abnormal occurrence. A sta11ar degrea of emphasis has been placed on long-term cooling, which is typically act.ieved by the residual heat removal (RHR) system. The RHR system starts to operate wnen the reactor coolant pressure and teraperature are substantially lower than their hot-stanchy condition values.

Even thougn it may generally be considered safe to amintain a reactor in a Nt-standby condition for a long time, experience shows that there have been events that required eventual cocidown and long term cooling until the reactor coolant systen was cold enougn to perform inspection and repairs. It is therefore 30-vious that the ability to transfer heat from the reactor *.o the enviro.vnent after a shutdcwn is an lay.crtant safety function for both PdRs and I'WRs. Con-saquently, it is essential that a power plant have the capacility +4 go from hot-standby to cold-shutdown conditions (when this is determined to be the safest course of action) under any accident condittons.

This issue was adooted as a Category A issue and designated as Tast A-31, *RHR Shutdow Requireents" in 1977. It was described in the NRC Report to Congress, NUREG-3410, "MRC Program for the Resolution of Generic !ssues Related to Muclear Power Plants ' issued on January 1, ' '8.

In accordance with the Test Action Pun for this task, Je staff's views on reouf rements fty residual heat resoval systems were translated into proposed changes to Standard Revit a Plan Section 5.4.7.

These proposals were considered by the Regulatory Requirements Review Coenittee (RARC) during its /1st setting on January 31, 1978.

The RRRC recormended appmval of the proposed changes and fur *>er rec :cences thac (1) the changes x applied on a case-ey-case basis to all operating reac-tors and all other plants (custom or standard) for wnica the issuance of the oper-ating license is expected before January 1. 1977, and (2) the changes be bactfitted

33 to all plants (custom or standard) for nich construction penitt or prelfminary design approval applications were socketed before January 1,1978, and for nafen the operating Itcense issuance is expected after January 1,1979. These recom.

msedations were approved by the Director, NRR and are being implemented.

Subsequently, the staff positions on design requirements for residual heat removal systems were incorporated into Regulatory Guide 1.139,

• kidance for Residual Heat Removal *, which was issued for public careent in May 1978. Coments were received during the latter part of 1978 and it is expected that this Regulatory hide can te issued in its final form in late 1979 or early 1980.

O O

34 CONToCL OF "EAVY L0aCS MEAR ! PENT AJEL (GDERIC ?ASK A.36)

Overneed handling systens (cranes) are used to lift neavy cejects in the vicinity of spent fuel in PWRs and BERs. If a heavy object, e.g., a spent fuel shipping cask or thielding block. were to fall or tip onto spent fuel in the storage pool or the reactor core duMng refueling and damage the fuel, there could be a release of radioactivity to the envirofsient and a potential for radiation over. exposures to innlant personnel. If the dropped ooject is large, and is assumed to drop on fuel containing a large amount of fission products with minimal decay time, calculated offsite cases could ex.

caed the siting guideline values in 10 CFR Part 100.

The NRC staff's review of this safety issue has been incorporated in the MRC Program for Resolution of SeneMc Issues as Generic Task A 35.

The objective of the task is to develop a revision to the Standard Re.

view Plan (SRP) based on a reevaluation of current MRC reavirements and procedures curretnly utilized at operating plants. If found to be necessary, the revision will provide criteria to further reduce the catential for heavy loacs causing pacceptacle damage to s:ent fuel in a storage pool or in the reactor core duMng refueling. The revised SRP will provide the basis for implementing additional requirements and procedures in existing plants unere 1

mareanted and can be used in future reviews of new plants.

i It is the NRC staff's view that continued coeration during our review of this geneMc issue presents no undue Msk to the health and safety of the pun 11c.

Coerating facilities use a variety of design and aministrative measures to einimize the potential for dropping a heavy ocject over tM reactor core or over the spent fuel pool. These design and aministrative seasures have been effective since no neavy load handling accidents resulting in damaged fuel i

have occurred in over 330 reactor years of U.S. operating excertence. Additionally, for facilities that have recuested increases in scent fuel pool storage cacacity, the NRC has prenibited the movement of loads over fuel assemolies in the spent fuel pool that weign more than the ecu1 valent seight of one fuel asserrely.

Also for those plants wnere a review of cask drop or the crane handling system is not conolete. movement of shielded casks over or near spent fuel nas oeen pronibited.

2.

25 Concurrent with our review. Itcensees have reviewed their current procedures for the movesent of heavy loads aver spent fuel to assure that tne potential for a handling accident that could result in canage of spent fuel is minimited untle our generic evaluation proceeds. The majority of the licensees' submittals of their reviews have been received and are under review.

Generic Tast A-36 1s expected to be completed in early 1979. The Tast util result in the develonnent of generic criteria however, fmalerant.

ation of these criteria util be Mgnly decendent on plant design characteristics and the specific crocedures in effect at eaca particular clant.

l I'

M i

t e

25 SEI!w!C OE5:3 9:TTt!A

[1DEt!C *Afr a Jo)

MRC teostations require that nuclear power plant structures, systems and :ce :crents incertant to safety be cesigned ta of tastand t*e effects of natural phemnrena suen Cetailed requirements and ;utcance regarcing the saf smic desip as eartN;uakes.

of nuclear plants is provided in the NRC regulations and in Regulatory Lices ts-Sued by the Cornission. However, there are a nacer of plants with c nstructica persits and operating licenses issued before the NRC's current rg;14tices and regu-latory guidance were in place. For this reason, reresiews of t'.e seismic design of various plants are being undertaken (;rincf; ally as part of the Cx:rissica's Systematic Evaluation Pecgram) to assure tr.at these plants do not present an edue rist to the public.

De NRC staff is c:nducting Generic Tast A 40. as : art of the MRC Program for Resolution of Generic Issues. Tast A 40 fs, f a effec *. a corcendium of sacrt ters efforts to su;;crt t.e reevaluation of tse seismic design of c;erating reactars.

De cbjective of Task A 40 is. in part, to investigate selected areas of t.e sais-sic casign secuexa to determine their censervatism for all types f sites. to investigate alternate approacPes ta ; arts of tse design sequence, and :s quantify t*e overall c:nservatism of the desip sequence. In tr.is manner tais ;regram will aid the NRC staff in ;erforming its reviews of the saf ssic design of ::erating reactors.

Generic Task A 40 is separated itta ten separate sustasas. *he W orsty sf t e su: tasks are schecuted for caroletten in Sectarter 1979. However, taree :f t e suctasks relatec to teveleping state-of-tse-art retNesicqy ts :etter deffre eartacuase ground motien near eartscuake sources are Ionger term ef'erts. 7ese three sustasts are scheouled for csuoletion in 1981.

O

37 Pl#E GActs sf 80fL*NG WATE1t 2EAC*Ct3 (GDERIC TA$r a.421 Pipe cracking has occurred in the heat affected renes of welds in prirary system piping in' boiling water reactors since the mid-1960s. These cracks have occurred mainly in Type 204 stainless steel that is being used fn most operating his. The major probles is recogni:ed to be intergranular stress corm ston cracking (!GSCO) of austenttic stainless steel c moonents that have been made susceptible t2 this failure mode by being " sensitized

• either by post. eld heat treatment or by sensiti:ation of a narrow heat affuted gone near welds.

• !afe ends" (shert transition pieces between vessel nozzles and the piping) that have been highly sensiti:ed by furnace heat treatment natie attached to vessels during fabrication were very early (late 1960s) found to te susceptible to IGSCO. Because they ere susceptible to cracting, the At mic Energy Cos.

sissien took the position in 1969, t.*.at furnac3 sensitized safe ends should not be used on new applications..tst of the furnace sensitized safe ends in older plants have been removed or clad alth a protective material, and there are only a few.hlte that still have furnace sensitized safe encs 11 use. Most of *hese, however, are in smaller diamatar 11aes.

Earlier reported cracks (prior to 1975) occurred primarily in 4* diameter re.

Circ:lation loop-by. pass lines and in 10* diameter core spray if nes. More recently cracks were discovered in recire114 tion riser piping (12".14*}

in foreign plants. Cracting is most often detected during Inservice Inspec.

tion using ultrasonic testing tecaniques. Scme piping crsets haea teen dis.

covered as a result of primary coolant leaks.

Because of theses,ccurrences of ht primary system cracking, there has t:mn a variety of actions uncertaken by the NRC. These actions included:

issuance of Regulatory Guide 1.14 on

• Control of the use of Iensitized Stainless Steel" issuance of aegulatory Liee 1.45 on " Reactor Coolant Souncry Laat i;etection Systesis*

closely foliewing the incidence of cracking in hts, including foreign O

experier.ca encouraging replacement of fur iace sensitized safe ends e

I

33 requiring augmented inservice inspection (additional rore frequent ultrasonic examination) of

• service sensitive" lines, f.e.. those that have experienced cracking e

requiring uporading of leak detectica systems Pipe cracking and furnace sensitized safe and cracting has been recently re.

ported in large* (24" diameter) lines in a GE.destgrad SWt in Gemany with over 10 years of service. Because the safe ends on that facility had been furnace sensiti:ed during fabrication. IG3C was suspectsd. As a result of concerns regarding these furnace sensitized safe ends. a safe end was removed in order to perform destructive examination. Curing i *boratory examination of the re.

4 roved safe end. including a small section of attached phe, cracks were dis.

covered at various locations in the safe end and in the weld neat affected rene of the pipe. The cracks in the pipe weld area sere very shallow witn the a:aximum depth less than about 5 ne (about 1/8*). Cracking in the furnace sensitized I

safe end was screwhat deeper. The Geman experience was the first known or. cur.

rence of IEC in pipes as large as 24* in diameter.

In June 1975, a througn. wall crack was af scsvered in an Inconel recirculation riser safe end (10* diameter) at the Ouane Arnold facility. The crack has been attributed to !G3C0 although the natorial in this instance is different frem the Type 304 stainless steel that has teen historically found to crack. Sun.

secuent ultrasonic examination discovered indications in six of the other saven safe ends. Following their removal, erscAing =es discovered in all eignt safe ends. The cracking appeared to have originated in a tight crevice between the inside wall of the safe end and an internal t*erwl sleeve. 5 ch crevices are known to enhance,1G3CO. Of fferences in materials. geometry, stress tevels and crevices appear to make the problem at0uane Arnold unique to a particular type of recirculation riser safe end (Type !). As a result Jf this event. ultrasonic examination of the other Type ! safe ends in 'J.S. SWts. i.e.. at the 3runswict I and 2 facility, was conductad. No significant indications were found in

. 33 Unit 2 and one indication. was icentified at Unit 1.

Although this indication is relatively minor and is not

• reportable' pursuant to the NRC Regulations.

it is continuing to be evaluated. The ultrasonic indication which was found a

will be reevaluated at another plant shutdown scheduled for later in 1978.

In addition to discussior.s with General Electric (the reactor vendor) regarding recent pipe cracking experience. General Electric was asked to provide an in-depth report on the significance of recent events regarding current inscection, repair, and replacement programs. They were also asked to address any new safety concerns related to the occurrence of cracking in large

' main recirculation piping. Sased on inforretion presented by General Electric to date, and extensive staff evFuation it was concluded that the recent xcurrences did not constitute a basis for innediate conce n about plant safety, nor require any new innediate actions by licensees.

The staff briefed the Comisshn on pipe cracking in GWRs on August 31 1978. and on Septecter la.1978. re-established an MRC Pipe Crack Study Group.

The Study Group will specifically address the following issues:

the significance of the cracks discovered in large diameter pipes relative k

to the conclusions and reconnendations set forth in the referenced report and in its imolementation doc:enent NUREG-0313.

resolution of concer es raised over the ability to use ultrasonic techniques to detect cracks in austenitic stainless steel.

the significance of the cracks found in largs diameter sensitizeo safe ends.and any recomendations rtsarding the current MI program for dealing with this matter.

the potential for stress corrosion cracking in PWRs. and the significance of the safe end cracking at Quane Arnold relative to similar mata dal and design aspects at other facilities.

~

The study Grovo is scheduled to courolete its evaluation and report in January 1979. In addition to the Stady 3roup effort. the NRC has underway several generic technical review efforts regarding flaw detection unich are aimed at improving piping inspection techniques and requirements. *hese generic efforts and any folicw-on efforts resulting from the Study Gr vo's evaluation will be incorporated into a new Category A generic task. Tast A 42

" Pipe Cracks at Boiling water Reactors.*

mm

=

ao.

CCNTaINMENT EPE9GENCY $UtdP QElfABILITY (GEN GIC Ta$K a.4 )

Following a postulated loss.of-coolant accident, i.e.. a break in the reactor coolant system piping, water flowing fran the break would be collected in the emergency sumo at the low point in the containment. This water would be recirculated through the reactor system by the emergency core cooling peps to maintain core cooling. This water would also be circulated through the containment spray system to remove heat and fission products from the con.

tainment. Loss of the ability to draw water frae the emergency sump could The otsante the emergency core cooling and containment spray systems.

consequences of the resulting inability to cool the reactor core or the con.

tainment atmosphere could be melting of the core and/or breaking of the con.

tainment.

One postulated means of losing the ability to draw water from the emergency sumo could be blockage by debris. A principal source of such debris could be the thermal insulation on the reactor coclant system piping. In the event of a ciping break, the subsequent violent release of the nign pressure water in the reactor coolant system could rip off the insulation in the area of the break. This deoris could then be swept into the susp. potentially causing damage.

Cur.ently, regulatory positior.$ regarding sump design are presented in Regulatory Lide 1.52, "Succs for Emergency Core Cooling and Containment Scray Systems.*

wnich av4resses deoris (insulation). The Regulatory kice recommends, in aci dit"on to providing redundant separated sumps, that two protective screens be provided. A low approach velocity in the vicinity of the sump is required to allow insulation to settle out before reaching the stano screening; and it is reouired that the sumo remain functional assuming thac one. half of the screen surface area is blocked. The NRC staff believes that sumo designs in accorcance with this regulatory guide acceptably resolve this issue. Monetheless, the NRC staff is continuing to study the benavior,of insulation uncer pipe break conditions to gain a metter understanding of how it. sight behave.

~

.m

41 -

A second postulated.9.eans of Wsing the ability to draw water from the emergency sumo could be abnormal conditions in the sump or at the puso inlet such as air entrainment, vortices, or excessive pressure drops. These conditions could re.

sult in puro cavitation, reduced flow and possible damage to the purcs.

Currently, regulatory positions regarding sump testing are con ined in Requ-latory Guide 1.79," Pre-Operaticnal Testing of Emergency Core Cooling Systems for Pressursted Water Reactors," which addresses the testing of the recircu.

lation function. Both in-plant and scale mocal tests have been performed to demonstrate that circulation through the supo can be reliably accomplished.

The NRC staff believes that sur.ps tested in accordance with this Regulatory Guide acceptably resolve tnis issue. As supplemental guidance, the staff, througn a contractor, is studying = nether 'urther guidance for the design and review of energency sucps to assure acequate hyoraulic design can be developed.

The NRC staff init! ally planned to study the issue of containment emer;ency sumo blockage from insulation as part of Generic Task C-3, " Insulation Usage Within Containment." In addition, initial plans wre to study the vertex formation issue as part of Generic Task 5-18, " Vertex Suppression Requirements for Con-tainments." However, containment emergency sump operability is fundamental *a the successful operation of both the emergency core cooling system (neeced to cool the core) and the containment spray system (needed *4 assure cantainment integrity) following a loss-of-coolant accidents For this reason, these portions of Tasks C-3 and 8-18 have been concined and elevated to Category A as Generic Task A 43 under the more general title of " Containment Emergency Suro Reliability."

Because tnis action has r.nly recently been taken, a Task Action Plan and schedule for this task have not ht been developed.

O e

42 STAftCN SLACK 0VT (GENERIC TASK A 44)

Electrical power for safety systems at nuclear power plants is supplied :y tw redundant and independent divisions. The systems used to remove decay heat to cool the reactor core following a reactor shutdown are included among the safety systems that must meet these requirements. Each efectrical divison for safety systems includes an offsite af ternating current (a. c.)

power connection, a standby ertrgency diesel generator a. c. power supply, and direct current (d. c.) sources.

The issue of station blackout was originally included as Generic Task S-57 in the NRC program for tesolution of Generic Issues. The task involves a study of whether or not nuclear power plancs should be designed to accomodate a complete loss of all a. c. power, i.e.

a foss of offsite a. c. sources and both onsite emergency diesel generator sources. Loss of a!! a. c. for an ex-tended period of time in pressurized water reactors accccanied by loss of the auxiliary feedwater pumps (usually one of two redundant pumcs is a steem tur-bine driven pump that is not dependent on a. c. power for actuation or opera-tion) could result in an inability to cool the reactor core with potentially serious consequences. If.the auxiliary feedwater pungs are decencent on a. c.

power to func*1on, then & loss of all a. c. power for an extended period could of itself result in an inacility to cool the reactor core. Althougn this is a lcw probability event sequence, it could be a significant contributor to risk.

Current MRC safety requirements require as a minimum that diverse power drives be provided for the redundant auxiliary feedwater pumes. As noted above, this is normally accomplished by utilizing an a. c. powred electric motor driven pung and a recundant steam turbine driven pump. One concern is the design adequacy of plants Itcensed prior to adoption of the current requirements, e

An initial survey of operating plants has been c:ppleted which indicates that all operating pressurized water reactors have either steam turcine detven or diesel driven auxiliary feedwater puncs (neither of wnich are dependent on a. c.

power). This assures at least that some capability exists for accomodating an extended loss of all a..c. power. Further review of older plants in this re-gard will be concucted as part of the NRC's Systematic Evaluation Program (see earlier discussion in this Chapter on page

).

e

43 Further study regarding this issue will include setemining if requirements beyond diverse power drives for the auxiliary feedwater ;umes are needed. Such requirements m[gnt include specific time requirements for which the plant must te capable of accormodating a station Blackout.

This safety issue was previously included in the NRC Program for the Resolution of Generic Issues as Generic Task 8 57, but has recently teen elevated to Cate-gory A as Generic Task A 44 Because this action has only recently been taken, a Task Action Plau and schedule for this task have not yet been developed.

e e

i ENCLOSURE 2 SUf41ARY STATUS INFORMATION ON REMAINING CATEGORY A TASKS

A-13 Snubber Operability Assurance As part of Subtask 1, a survey of the current axperience with snubbers was completed in July 1978. The results of this survey have been published as NUREG-0467. Based on this survey recomendations for changes in Technical Specifications will be developed. These recomendations are scheduled to be completed in July 1979.

As part of subtask 2, a contractor (Liquid Metal Engineering Center) completed a preliminary report in October 1978 of the sensitivity of snubber perfomance to changes in design and operating conditions.

This study is scheduled to be completed in October 1979.

As input to Subtask 3, proposed requirements for the inservice inspection and functional testing of snubbers were drafted in August 1978 for incorporation in the ASME code.

The final result of the task will be a recomendation for a Regulatory Guide on the application of snubbers, which is sc:1eduled for January 1980, and revisions to the Standard Review Plan and Standard Technical Specifications, if necessary, which are scheduled for June and October 1980, respectively.

A-14 Flaw Detection Management of this task was transferred from NRR to the Office of Standards Development in July 1978 in the FY78 reprogramming effort.

At that time the unavailability of personnel in NRR had resulted in a one year slippage of the original schedule.

Although the developnent of licensing requirements is part of this task, its primary purpose is to monitor and assess ine results of programs being conducted by other off.oes (Standards Development r

ard Research) and industry. Those portions of this plan for which delay might affect two other plans, BWR Nozzle Cracking (A-10) i and Pipe Cracks in Boiling Water Reactors (A-42), have been removed from this plan and included in those plans. The progress of the remaining portions has not been affected by the delay and reassignment of this Task Action Plan. The first Regulatory Guide identified in the plan was completed in September 1978. The remaining three regulatory guides are scheduled for comoletion in July 1979, April 1980, and April 1981.

9

A-15 DECDNTAMINATION primary Caslant System Decontamination The initial phase of the Decontamination Task Action Plan requires the completion of our review of strong solution decontamination of the lead plant, Dresden Unit No. I.

Licensee delay in completing support facilities for the Dresden Unit No.1 decontamination has caused the decontamination date to slip to June 1979. To accomodate this delay the staff has rescheduled its workload relating to the Dresden decontamination to concentrate staff effort on other higher priority work. No other requests for NRC approval for primary coolant system decontamination have been received.

A report from Pacific Northwest Laboratories which will provide generic review of the impact of decontamination on LWR Radioactive Waste Treatment Systems is being prepared in draft form and is scheduled for completion in April 1979.

Steam Generator Chemical Cleanino Chemical cleaning of the secondary side of steam generators involves the use of chemical agents which may degrade the integrity of the primary coolant system boundry (i.e., steam generator tubes). The acceptability of various cleaning agents is a major consideration of TAP A-15.

This generic task had been expedited to support the lead case. However, Consolidated Edison has decided, for the present, to abandon the concept of chemical cleaning of the Indian Point 2 steam generators. Chemical cleaning at other facilities will be considered within the scope of A-3, A-4 and A-5, steam generator tube integrity.

e A-16 STEAM EFFECTS ON BWR CORE SPRAY DISTRIBUTION The task is on schedule, S' ingle nozzle tests have been run in steam for BWR/6 spray nozzles. "Similar" nozzles are being developed which will reproduce, in air, the spray patterns produced by the actual nozzles in steam. The simulator nozzles will be installed and tested in the full scale (air) core spray distribution test facility l

at Vallecitos. Calculation superposition techniques, which will be used in conjunction with the experimental data, have been developed.

The next major milestone will be a meeting in February 1979 with GE and BWR licensees to discuss application of the new techniques to older plants. A more definite completion date can be estimated following that meeting, but the task is scheduled to be completed by the end of calendar year 1979.

A-18 Pfoe Ruoture Design Criteria The draft of the revision of the Standard Review Plan to provide consistent pipe rupture criteria for application to piping both inside and outside of the containment (Task 1) was completed in October 1978.

Task 2 consists of the evaluation of the design and inspection requirements of containment piping penetrations. As part of this task, two contractors are analyzing the loads and stress in pene-trations that would result from pipe breaks. These analyses have been slipped approximately six months and are scheduled to be completed i

in March 1979.

4 Task 3, an assessment of the effect of pipe whip restraints and snubbers on the normal operation of piping systems, is being conducted by a contractor (Teledyne Engineering Service). Three quarterly reports documenting progress to date have been completed.

The schedule for completion of this work has slipped six months to October 1979.

The results of these three tasks are to be combined to develop a revision to the Standard Review Plan and recommendations for a Regulatory Guide. This is scheduled for completion in March 1980.

A-19 Digital Comeuter protection System This task, which has the purpose of documenting and standardizing the technical review procedures 'for computer-type reactor protection i

systems, was scheduled to be initiated in October 1978. Hcwever, review of RESAR 414, ANO 2 and CESSAR System 80 applications has pre-empted the time of the task manager and other specialists in the Instrumentation and Control Systems Branch and the task initiation will be delayed further. The bulk of the task is to be done by a con-tractor (CRNL). Discussions are now in progress to define the contract effort for FY79 and beyond. Cmpletion of the task has been slipped at least into FY 81. This slippage is not of great concern to the staff since our review of the first generation of computer based pro-tection systems will not be completed prior to that time and this task is designed to codify that review practice.

A-20 Imoacts of the Coal Fuel Cycle The contractor (contract issued September 15, 1978), Tecnektron, Inc.,

has completed an in-depth review of the literature, and assessment of the literature is now underway. Over 2100 literature sources have been identified for documentation of which 350 have been chosen for further study. Complete analysis of data and information will be completed by April 1979, with completion of a draft report by June 30, 1979. The final report is to be submitted by the contractor on or before November 1979. At that time, the staff will review the report to determine what action, if any, would be required.

A-21 Main Steam Line Break Inside Containment An interim evaluation and licensing position was completed in February 1978. A survey of the current methods of analyzing this event and the environmental qualification of equipment needed to mitigate this accident was completed in September 1978.

f A contractor (LASL) is new evaluating the methods used to snalyze I

this accident and is expected to complete his work in December 1978.

Development of licensing requirements based on this work and completion of the task is scheduled for February 1979.

t 4

A-22 PWR Main Steam Line Break, Co:e, Reactor Vessel, and Containment Buildino Resoonse This plan was originally scheduled to be initiated in November 1978.

I}

, *^.

However initiation is being delayed until August 1978 because the task manager hast been assigned to Systematic Evaluation Program and the review of the Bellefonte, Sequoyah and Fort St. Vrain operating licenses.

W-

A-23 Containment Leak Testing This task was completed in May 1978 with the forwarding to the Office of Standards Development of requested changes to Appendix J to Part 50 " Primary Reactor Containment Leakage Testing for Water-Cooled Power P.aactors."

i

A-25 Non-Safety loads on Class lE Power Sources The bulk of this task is being performed by a contractor (ORNL).

Four tasks,1) a survey of present load shedding design practices,

2) a compilation of failure rate data of circuit breakers and fuses,

3) an investigation of possibly useful relf.1bility methods and

4) an evaluation of possible configurations of isolation devi:es were completed by the contractor (ORNL) in September 1978.

The original intent was to obtain the necessary failure rate data from the reactor vendors and letters requesting the data were sent. However, the contractor was unable to obtain the dat? in this manner, and was forced to obtain the data from other sources.

Th!s. resulted in a six month delay. The schedule for completion of this plan has slipped to October 1979.

A-27 RELOAD APPLICATIONS The RRRC approved the Branch Technical Position BTP-DOR-1, Guidance For Reload Submittals, on January 31, 1978. The Office of Standards Development (SD) drafted a Regulatory Guide based on the BTP and has circulated it for internal coments. SD is now incorporating the staff coments into the Guide and plans to submit it to the ACRS for review in March and issue the Guide for coment in April.

The BTP will be revised and sent out to the licensees in February.

Work on developing an SRP has not begun. Due to its lower priority, this activity is currently not scheduled.

i A-28 INCREASE IN SPENT FUEL STORAGE CAPACITY This task is nearing completion.

The revision to the Standard Review Plan (NUREG-75/087) was finalized

{

in September 1978, and will be issued by OSS in their routine updating of the SRPs.

Recommended revisions to Regualtory Guide 1.13 " Spent Fuel Storage f

Facility Design Basis," were forwarded to the Office of Standards I

i Development in October 1978. SD is developing a schedule for the t

l formal revision of Regulatory Guide 1.13.

i Based on the findings of the Draft GEIS on spent fuel storage develcped i

by NMSS, no changes are necessary to the NRR licensing activities.

The findings of the Final GEIS are notexpected to alter this position.

.e

l A-29 Nuclear Power Plant Design for the Reduction of Vulnerability to Industrial Sabotace This effort involves the evaluation of various design alternatives for their potential for increasing the inherent protection of nuclear power plants against sabotage.

To date, the followit.g steps towards a resolution of this generic concern have been completed:

1.

An inter-office NRC working group has met, and discussed various potential candidate design alternatives. A summary of the candidate alternatives considered by this group has been distributed within NRC for comments.

2.

A research program has been initiated at Sandia Laboratories to evaluate the feasibility, effectiveness, cost, etc., cf the various design alternatives for a typical standard LWR.

To date, the plant characterization serving as the base-line for the evaluations has been completed.

Preliminary results from the research program are scheduled for October 1979. Task A-29 is scheduled to be completed in 1981.

A-30 Adequacy of Safety Related D.C. Power Supplies Management of this task was transferred from NRR to the Office of Nuclear Reactor Researgh, in September 1978, as part of the FY78 reprograming effcet.

The bulk of the task is being perfomed by a contractor (Sandia) who started work in October 1978. The fault tree analysis being performed by the contractor is scheduled to be completed in February 1979 and a completed report is tentatively scheduled for June 1919.

Completion of the task and revision of the Standard Review Plan, if necessary, is scheduled for November 1979.

A-32 Missile Effects Management of this task was transferred from NRR to the Office of Standards Development in July 1978 as part of the FY78 reprogramming effort. The bulk of the task is to be performed by a contractor.

but redefinition of the scope has delayed initiation. A request for proposal is to be developed in January 1979. Completion of the task is tentatively scheduled for November 1980, but the final date will depend upon the course and outcome of the contractor selection process.

A-33 NEPA Reviews of Accident Risks Work on the various phase reports has continued at the participating laboratories. PNL issued its Phase I report, "A Review of Methodology for Accident Con equences Assessment" in December 1978.

It was expected in March 1978. Work on Phase II continues. BNL Phase I (PWR Short Term Program) is now complete. An informal report was drafted on November 9, 5

1978. BNL is now expected to ecmplete Phase II in February, 1979. It was originally scheduled for September 1978.

Completion of Task A-33 is new expected to be ccepleted in July 1981.

~

A-34 Instruments for Monitoring Radiation and Process Variables During Accidents To date there has been major industry opposition to implementation of Regulatory Guide 1.g7, Revision 1 (Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident). This opposition has been based on both technical and philoso-phical concerns about the various aspects of the Regulatory Guide. In response to these concerns the staff has been working with an ad hoc sub-committee of the Atomic Industrial Forum that has been established to I

coordinate industry connents concerning implementation of this Regulatory Guide.

It is currently planned that Position C.1 of the Guide concerning instru-ments for monitoring Design Basis Accidents will be implemented beginning with the New Haven application. Based on the experience gained during the New Haven review, the staff will begin backfitting this Position on a case-by-case basis. In addition, Position C.3.a through C.3.c of the Guide concerning instruments for monitoring accidents that degrade beyond those assessed in Chapter 15 of the SAR, will be implemented on all plants in the near future. Implementation of Position C.3.d concerning monitors of activity release rates from identifiable release points will be delayed for at least one year while a study of release rate monitor capab.ilities is undertaken.

A-35 ADEQUACY OF 0FFSITE POWER The grid frequency decay study being performed by ORNL is nearing completion with a report expected in January 1979. Results of the study may necessitate examination of the capabilities of PWR designs to withstand possible frequency decay events. Review of complete offsite power losses and partial offsite power losses is about 1/2 complete. Completion of the loss-of-offsite power review is expected in February 1979.

Development of criteria and technical specifications for degraded voltage protection is essentially complete and being implemented.

Review of testing requirements for onsite power sources is complete and Re,gulatory Guide 1.108 is being implemented. Methodology and computer program (s) for analysis of offsite power systems are being developed by ORNL and are expected to be completed in late FY 1979.

~

i i

A-37 Turbine Missiles i

A sur: nary of the history of turbine failures (Task A) was ccepleted in March 1977. An evaluation of the means of improving turbine disk integrity (Task B) was completed in September 1977. An evaluation of the means of reducing the probability of turbine overspeed (Task C) was completed in October 1977. The specification of the staff position i

concerning the probability that a turbine missile would occur and the requirement for assuring this probability (Task 0) are scheduled to be l

completed in January 1979. The staff position on an acceptable method of determining strike probabilities (Task E) was completed in March 1978.

The assessment of the probability of failure given a strike by a turbine missile (Task F) has been deferred by the delay in TAP A-32. However, this will not cause any delay in Task A-37 since the final report will be based on the bounding assumption that this probability is one. The evaluation of computer codes used to determine strike probabilities (Task G) was completed in March 1978. The revision of the Standard Review Plan and completion of Task A-37 are scheduled for March 1979.

A-38 Tornado Missiles This task was transferred frcm NRR to RES in Octcber 1978 as part of.

the FY 1978 reprogrrraing effort.

Subtask 1 - Soft Missile t=cact RES contract is in the process of being let. Work will c nnence in January 1979. This task is about six months behind criginal schedule but this delay is not expected to impact the overall completion date.

for Task A-33.

Subtask 2 - Ceta_rministic Assess. nt w

This subtask has slipped due to transfer to RES, but is near ccepletten.

A draft rec:maending a revised missile spectrum will be circulated for staff review in January 1979.

Subtask 3 - Probabilistic Assessment This subtask is proceecing en schedule. Centract assistance effort at NBS is proceeding satisfacterily.

Subtask 4 - Positien Cevelocment Scheduled to begin in April 1979.

Current estimated cocpletion date fcr Task A-38 is the original One -

fjovember1979.

HI:50tlRCE SilHMARY Page 1 of 3 FullDillG (Ill0llSAlllis 0F $)

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0.2 g

A-17*

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430 0.6 0.18 0.76 g

A-18 35! -

35 0.6 0.04 0.64

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Page 2 of 3 FullDillG (Til005AllDS OF $)

IIAllPOWER (ftANYEARS)

TASK 10/08/78-11/04/78 TASK HAllP0HER FY 77 fY, 78 FY 79 TASK $

01/01/78-10/07/78 (0llLY CllE H0itTil 0F TOTALS 11 0.

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I A-19 75l 85l-160 0.4 0

0.4 I

g A-20 l-75l 35 g -

110 0.2 0.03 0.23 A-21 18l 20 l -

38 0.3 0.03 0.33 A-22 I

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A-29

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- l 185 310 0.0 0.02 0'.82 i

l l

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- l-0.2 Conglete 0.2 A-32 I-l 75 l -

75 0.2 0.02 0.22 A-33' l-l 200

- l 180 380 1.3 0.06 1.06 e

I

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Page 3 of 3 I

FullDillG (Til00SAllDS OF $)

ItAllPOWER (HAllYEAR3)

TASK

[10.

FY,77 FY, 78 FY 79 10/08/78-11/04/78 TASK IWiPOWER

. TASK $

01/01/78-10/07/78 (0;1LY OllE 110i4Til 0F TOTALS l

TA 'RES TA ! RES TA l RES TOTALS py79. DATA AVAILABLE) 01/31/78-11/04/78 9

R I

I A-34 0.4 0.02 0.42 I

A-35

- l -

100l l-100 0.1 0.03 0.13 A-36*

- I l-0.1 0.02 0.12 l

38 0.3 0.05 0.35 25

• 13 l A-37 I

I A-38 60l

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l l

A-39*

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240 l 2192/

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629 1.5 0.04 1.54 I

A-40*

3901 -

186!'472 1

281! 60 1389 0.2 0.02 0.22 A-41

---

[----- ----,r------

Tills l TASKllAS BEEll D t0P P ED --------------

g A-42* g 0

0 flew Task i

I A-43*

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53 0.1 0

llew Task g

A-44*

- l-

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TOTALS 10h8 4d33 39!!2 9573 23.5 IlY 1.64 HY 25.14 HY a

l l

l Denotes " Unrest lyed Safety Issue" N'Programalsoalplicable tb Task A-7 U Program also ai.plicable to Tasks A- ', A-8 and B-10 t

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