ML20117N314

Forwards LER 83-74,associated Licensee Corrective Action Documentation & Evaluation of Safety Significance Re Discovery of Numerous Main Steam Safety Valve Setpoints Outside Test Acceptance Criteria for Transmittal

ML20117N314
Person / Time
Site:	05000000, Palisades
Issue date:	07/16/1984
From:	Jorgensen B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Shafer W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML20114G054	List: ML20115A138 ML20115A202 ML20115A205 ML20115A239 ML20115A266 ML20117M511 ML20117M526 ML20117M570 ML20117M585 ML20117M618 ML20117M633 ML20117M663 ML20117M787 ML20117M807 ML20117M825 ML20117M837 ML20117M871 ML20117M930 ML20117M935 ML20117M980 ML20117N004 ML20117N028 ML20117N038 ML20117N065 ML20117N086 ML20117N154 ML20117N314 ML20117N323 ML20117N374
References
FOIA-84-616 NUDOCS 8505170471
Download: ML20117N314 (11)

Text

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3h UNITED STATES e

NUCLEAR REGULATORY COMMISSION REGION lli e'[

799 ROOSEVELT ROAD o,

GLEN ELLYN,ILLINotS 60137 July 16, 1984 MEMORANDUM FOR:

hief, Reactor Projects Branch 2 THROUGH:

G. C. Wright, Chief, Reactor Projects Section 2A FROM:

B. L. Jorgensen, Senior Resident Inspector

SUBJECT:

EVALUATION OF PALISALES LER 83-74 Attached are Palisades LER 83-74, associated licensee, corrective action documentation, and an evaluation of the safety significance of the matter performed by the licensee. This event involved the finding of numerous main steam safety valve setpoints to be outside test acceptance criteria on the "high" side. The information is provided per Regional Review Committee discussions and directions, for transmittal to DRS for a determination on the safety significance of the event; i.e. the licensee identified, reported and corrected a problem - does the problem represent noncompliance at Severity Category 3 or above? The licensee's evaluation concludes the safety significance was low.

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IEENTIFICATION ANTI Of9CrfPTION !

I PART I g rirm Valves Failed to Meet the 'As Found' Acceptence Criteria Four1d in Technical

_ Sperif4catinnn Test RM-99..

a.] eu.t sests (cro 09110 tevie.ot. cuss (cro 13107 taur,.t r i,.u. RV-0701 - -0724 d onscai,ri p or accons.cc os cc oirio :

On 10-3-83. SP&LS_ began __ testing Palisades main steam relief Y_alves at the BC Cobb Plant.

Between Oct. E aAd Oct.14. tTie~'as fosiiU tests were enmp1cted nn RV-0731 thvnngh RV-072.l; _

RV--0715.vn a fonna within tha *an* a"i* aria-The remining 93 relief vnivna unrn fnund uith lift nnint = thnt eyenaded thn 1/e cr4torin_

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All vnline vn1 van knvn henn retestod nna b ot the_acceptnnna critar4n on? lad far in RV 99_

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r in thin instnnea 1983._ hon nn mnnv_vern fnund high) to datermine (and report to P-CARB on nnma) What._patentin1

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fauna tn be lenking_ badly and recitired resetting to meet the acceptance criteri1. Thnt adluats;ent may be the reason for the high lift noints

_ ___fnuna anrina thn 1083 tantn.

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__ _.yantJB.. EVALUATION _AND_ DISP _0SI. TION B00P CA_USE (cnnt.)

After reviewing the test _ data. I disagree with the Crosby representative.

The valve

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T hn14 eve that thn en11nvine nennnvin 1.

The valve discharge heights were adjusted iii~1981.

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2. -The valves were then tested and~ consistently found to lift at a lover 3

Vresaure thnn reituired.

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The valves were t' hen read.1usted to a higher lift oressure.

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They_yere retested with the adjust.ldischarre ~ heights _J.nd_ tested good.

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In 1983 the valves were retested without adjustment using a test rig with -

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The valves consistently lifted at a Iq, pressure than requirnd.

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_The4adiu.T.ent to the valve discharge __h: light c?.usWiiill the valve 33 N f t at __'. -__

loane procenra uhan tantna-Tha vn1_vn lift najuntenntn nnan*in 10R1 nnly ennpnnnnena for_ the discharge he.icht_ad.justr.ents. When tested in 1983 the valves Derformed as they vatild in Mer.Yice.

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• 83 --20hA I'The main steam relief valves _{RV-0101 thru RV-072h) are cacable of nerfo_r_ ming their.

ilqsiE;n function.

"found' set points.The results from 1981 RM-29 testing indicated extremely lov 'as from predescribed factory settings.These values were obtained after the nozzle ring was adjusted

~~

The nozzle ring adjustment was nececsary to de-

_crgase blevdown due to teqt_Sy_st.ca_limitatio_n L_.iL_3he__ test system did cated that set point was not affected by nozzle ring adjustment.

~~

The valves were inspected, cleaned and retested. The spring tension was adjusted (increased) to.-.

achleve proper lift.

Repetitive lifts were performed to demonstrate reliability.

Upon completion the nozzle ring was returned to the factory setting and lock vired. ---

The 1983 FR-29 testing indicated _extremelv hiFh 'as found' set coints.

These values

_ ve re-obt aine.d_vithqut_ prior _adiusicentS. Tht1981 tent system _stngn_cnpacity vas- ~ ~ -

_imprayed_batyeen 198F and 1981c TherefDIcde_noMle_riUg_ adjustment vas not nec- -

__qss_ary. -The valves vere inspected cleaned. renaired,_and/or relanned under the directi superyision of the manufacturer's field service representative.

~ -

The valves were re-tested-vith-the spring tenr. ion adjusted (decreasedi to achieve proper lift.

J lifts vere performed ~ to demonstrate reliability. 71th thd ~ ~ alves still hot, discharge '

Repetitive v

_ pipe va$._r_ecored~ and_at 40fr_of_5et_ point __prissure the._yalyes were visually fpr_1_qaK_,t_ight ne s s.

instected i

In recent telechone conversations with D. A. Desnover (CPCo - Trail Street) and D, Tuttle (Crosby' Company - engineerine) the__ Crosby cosition on_the no gle ring and l

_ Jet point relationshin was clarifiei,_,_Egraally. the nozzle ring _is not adjusted..If l

the nozzle ring tains _that nqr_ mal _m_inpr_ adjustments vill not affect set point.is adAtsted it_la_psu l

{g logerinct.he_n.qule_ ring _of a safete.xalve cwid imraan varn hhE--and mld a

They added that

( _cause a dicht._incr_q uq_ifdeppir.g _;.r.wsure ' _CgvgM -was in 1131fraising the no l

During the 1981 testing the nozzle rings were adjusted as much as 12dscrease in__pgpping pres-sure.

notches.'

In addition, the instructions bulletin for operation and maintenance of Crosby safety valteuVendor. File _MI.-F-L_ statg s :

"If the_.ynlve nous too lov. the seats may have become damaged Lov popping pressure _is seldom caused by spring _ set. although this is considered a common cause.

_The_above information indicates that set points can be influenced by seat condition and _nqsalg_r_ing_a_diustments although__the exac_t relationship can not be determined.

m Suffice to state that the 1981 RM-29 'as found' test results were in error, influ-enced by leaki_ng valves and nozzle ring adjustments.

The corrective action (spring tensioning)_taken during the 1981 RM-29 test caused the deviations observed / recorded during the 1983 testing.

As stated above the main steam _r_elief valves (RV-0701 thru RV~~6724) are capable of -

1 per_ forming _their_ design function _._Each relief._ralye,_as_ installed upon connletion

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o f_t he ;1983_RM-29_t esL_haa_t he _adj ustment _ holt _Le ont r_als__ spring _1_ ens ionLt.etur_n e d tq vithin_1125_ flat _s of the_ 'as_found' position _qr_the__1981-RM-29_ test.

e they_h ave _in diyidually_hecILin.spe c t e d _fo r _ l e ak_t ight n e s s.

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__In_ addition.

Non-cqnclus_ivq_but

.__gupporf.iye evide_nig_e_gan be shown (see_)ttachment 1).

_numer_ical average of the observed 'as found' lift point during both the 1981 andThe bold line highlights

_19833M:29_.lestm.__lll_ hut __one_lalve_ falls _vithin__the_. recommended set point _t_plerance hnnd_

--he_ aggregate of the above informatil_on is t_he basis for the conclusions sef ~f5rth.

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CONSUMERS POWER Date June 8, 1984 COMPANY Subject REACTOR ENGINEERING RESPONSE TO E-PAL-83-204C, Internal STEAM GENERATOR SAFETY VALVES FAIL TO MEET Correspondence ACCEPTANCE CRITERIA MJB 84-001 Ref 1)

XN-NF-77-18, Plant Transient Analysis of the Palisades Reactor for Operation at 2530 Mwt, July 1977, and

2) Phone conversation, Mel Woolpert, Combustion Engineering,

(

5/31/84.

Between October 4 and 14 of 1983, steam generator safety relief valves RV-0701 l

through RV-0724 were tested for p' roper lift pressures.

Only 1 of the 24-valves was found to lift with'in the ASME code acceptance band of 1%. Event Report E-PAL-83-204, part c. raises the question of the impact of the "as found" ' settings on the existing Palisades safety analysis.

The purpose of this analysis is to determine what effect, if any, is created by this discre-

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pancy.

i I

It is the opinion of RED that the Palisades plant was not in an unsafe condition as a result of the "a's found" safety valve settings. However, the plant was in an unanalyzed and potentially unsafe condition.

Four licensing basis transients (Ref.1) ; challenge the secondary side safety relief valves.

The increased lift pressure would alter the results of the above mentioned

i' analysis. However, for this case', it is our belief that the "as found" safety valve setpoints would still yield acceptable results for the transients in question. Regardless, it should be stressed that

.1) our findings are based on engineerings approximations, and 2) increased drift in these setpoints could create an unsafe condition.

We recommend that this problem be corrected to insure safe operation of the plant.

If it remains difficult to hold these values within the 1% acceptance band, a widening of the band is possible.

(Ref 2)

The four transients which challenge the secondary safety valves are as follows:

A.

Rod Withdrawal from 102% Power, B.

Rod Withdrawal from 52% Power, C.

Single Control Rod Withdrawal, and D.

Loss of Load.

Transients A, B and C reach minimum departure from nucleate boiling ratio (MDNBR) at the time of reactor trip. The secondary safety's don't open until, a considerable time after trip.

At that time, DNBR has increased and is no longer an important factor. Primary system pressure remains low during these transients, never being greater than 2210 psia. Therefore, the safety valves have little or no affect on safe operation in these transients.

j However, the loss of load transient challenges peak primary system pressure (52750 psia),

j Primary secondary AP (51530 psid), and MDNBR (21.30).

I In this transient, the secondary safetys are need to relieve pressure in the early stages, and t.__.._-----..

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..a-2 therefore'are an important element in mitigating the consequences of this event.

Three loss of. load cases.were analyzed in Ref I.

is the worst with respect to peak primary pressure.(See attachment 1) Case 1

. effect of delay safety. valve opening on peak primary pressure. Attachment 2 shows the safety valves.were assumed not to open until the highest "as found"The secondary for group I was reached, clearly a conservative assumption.

setting Peak primary pressure reached 2450 psia, much.less than the 2750 psia technical specifica-tion limit., Case 2 is the worst in' terms of primary - secondary AP. shows the estimated AP as a result of a secondary safety valve setpoint for group,1]of:10372 psia, as in Case 1.

The approx'imated AP = 1424 psid, well below the limit of 1530 psid.

technical specification; limit.on DNBR-of 1.3. Case 3 challenges the Palisades In this~ case,~DNBR'= 1739.

-Reactor. Engineering feels that delayed. safety valve openin,g would. havg. littles._ _ _

effect on-actual -fuel performance'. Other 'fofes bf. sic'on'dgry pressure. relief are: conservatively'rass'unied in'op'ellbTE"fo'r' 'this"cYse".~

.In reality,.the er atmospheric dump' Wlves"and' 'c'oide'n'ser bypass' would-bd. available to relieve pressure. ? Another conservatism ' n ' the' Exxon analysis"XN-NF-77-18, ;is the use i

of the W-3 DNB correlation which has a license DNBR limit of 13 accurate XNB DNBR correlat' ion *ch'an'ge"now' allows a'MDNBR'of 1.17', which The more provides additional operational margins.

- sufficient margin of safety, to offset the change _ in safety valve setpoints.

Therefore, Reactor engineering c,oncludes that the Palisades Plant was not r-an. unsafe condition due to the improperly set valves.

in However, the valves should be maintained at their required setpoint to stay within the bounds of

'the existing safety analysis.

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l lVhile shutdown for refueling a routine refueling duration surveillance test I a 2 m l revealed that fourteen of twenty-four main steam relief valves lifted at i

E l pressures in excess of the urrer limit srecified in TS 7.1.7_ r-N va,vae l gypw l are believed to have been able to provide their intended safety function, Ig 7,,

, 3 d 'lilovever evaluation is incomplete at this time.

Notification vill be vro-I g gvided if final analysis is to the contrary. Reportable per TS 6.9 2.b(1).

"N m m q,e g IIThe valves have been disassembled, rebuilt and successfully re-tested.)

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h ICause is believed to have resulted from adjustrents ende to each valve's blowdown ring which affected the lift setting. These adjustments were made l i

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Palisades Plant LER 83-74.

December.9. 1983

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During the current refueling outage, all twenty-four (24) main steam relief valves were removed and sent off-site for surveillance testing per

' Surveillance Procedure RM-29.

The results of the "as found" testing indicate that fourteen (14) valves lifted at test pressures in excess of the upper limit specified in TS 3.1.7.c.

The maximum pressure at which a valve lifted

-was 1147 psig and the minimum pressure in excess of the TS limit was 1040 psig.

-As described in FSAR Chapter 4.3.4, the design at Palisades is such that_the main steam relief valves are divided into three groups, each group having a different lif t setting..The performance of the surveillance test resulted in a total of twenty-three (23) relief valves lifting at pressures in excess of the upper limits of the FSAR and surveillance test acceptance criteria for each group Fourteen (14) of the'se valves were also in excess of the upper limit imposed by TS 3.1.7.c as described above.

The cause is believed to.have resulted from adjustments made during the

-previous refueling outage to the valve's blowdown ring and subsequent to the

~

final lif t setting : adjustment which was made at that time. Adjustments to the valve's blowdown ring were permitted by procedure. _ Adjustments to the blowdown ring were not anticipated to have an effect on the lift pressure value.

The test procedure will be revised to prohibit any adjustments to the relief valve blowdown ring prior to "as found" testing or subsequent to "as *lef t" testing.

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l General Omces: 212 West MacNgen Avenue Jackson. M6cnigen 49201. Area Code St7 788-05S0 December 9, 1983 James G Keppler, Administrator Region III US Nuclear. Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - LICENSEE EVENT REPORT 83-74 (VALVES FAILED TO MEET "AS FOUND" ACCEPTANCE CRITERIA)

Attached please find Licensee Event Report 83-74 (Valves Failed To Meet' "As

- Found" Acceptance Criteria) which is reportable to the NRC per Technical Specification 6.9.2.b(1).

Brian D Johnson (Signed)

Brian D Johnson i

Staff Licensing Engineer CC Director Office of Nuclear Reactor Regulation Director Office of Inspection and Enforcement NRC Resident Inspector - Palisades Attachments

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JUN 2 71984 MEMORANDUM FOR:

Victor Stello, Jr., Chairman Committee to Review Generic Requirements FROM:

Richard C. DeYoung, Director Office of Inspection and Enforcement

SUBJECT:

RE-REVIEW BY CRGR OF PROPOSED IE BULLETIN 84-XX:

UNDERVOLTAGE TRIP ATTACHMENTS OF REACTOR TRIP BREAKERS

REFERENCE:

Your memorandum for William J. Dircks dated April 16, 1984, entitled:

" Minutes of CRGR Meeting Number 61" Please schedule a CRGR re-review of the proposed subject bulletin (Enclosure 1).

We suggest a CRGR meeting during the week of July 1, 1984.

The referenced memorandum included background information and minutes of the April 4, 1984 CRGR meeting related to the proposed bulletin.

The minutes of the meeting summarized the material presented, discussions by CRGR members, and l

agreements reached by the CRGR, including concurrence in issuing the proposed bulletin subject to two exceptions.

One exception would emphasize the need for improving the design of the undervoltage trip attachment.

The other exception would modify the wording of paragraph 3.(b) to assure that the bulletin does not conflict with the plant technical specifications and that action required by the bulletin would not place the reactor in a less safe condition by requir-ing the tripping or removing of an inoperable reactor trip breaker (RTB) within four hours.

The exceptions raised by the CRGR have been accommodated in a revised version of the bulletin.

However, in addition to the minutes of the meeting, the referenced memorandum included a new caveat that was ireposed during a subsequent CRGR meeting held on April 11, 1984.

This caveat was precipitated by two concerns.

The first con-cern questions whether it is safer to intentionally trip a slow PTB, thereby making the plant more susceptible to spurious scrams, rather than accepting a delayed RTB trip.

This concern has been a' meliorated by revising paragraph 3.(b) of the bulletin.

This revision emphasizes the advisability of corrective main-tenance and allows the plant to be operated in a critical mode for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with an INOPERABLE RTB closed.

This 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limit, compared to the previous 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit, should allow sufficient time for repair or replacement of an INOPERABLE RTB such that the need to trip a slow RTB would be virtually elimi-nated.

It should be noted that this provision only applies to those cases where operation with an INOPERABLE RTB is not covered by existing plant techni-cal specifications.

The second concern questions whether the actions specified by the bulletin would degrade safety.

Accordingly, NRR/IE were requested to perform analyses to demonstrate that the actions specified by the bulletin would not degrade safety prior to re-review of the matter by the CRGR.

This concern is addressed in a memorandum for C. E. Rossi from G. M. Holahan dated rhUN

y Victor Stello.

May 4, 1984 containing the requested analysis (Enclosure 2).

In addition, a.

separate analysis of the safety impact of the proposed bulletin was performed by the Reliability and Risk Assessment Branch of NRR (Enclosure 3).

This analysis was limited to Westinghouse designed facilites because the impact of this 2 breaker configuration should be greater than on B&W and C-E designed facilities.

As summarized in the analysis, the net positive or negative impact oS the proposed testing cannot be quantified.

Although the analysis indicates that the proposed testing would increase the RPS unavailability and the likeli-hood of inadvertent reactor trips, it also indicates that this negative impact may well be offset by the improvement in RTB availability due to the proposed testing.

For example, the timely identification of precursors to RTB failures would improve overall RTB availability such that an overall reduction in the likelihood of core melt may result.

In this regard, a sensitivity study indi-cates that the benefits due to increased availability are proportionally much higher than the negative impacts due to increased inadvertent reactor trips and RPS unavailability due to RTB testing.

In any case, the bulletin is intended to improve overall safety by a set of integrated actions (i.e., assuring ade-quate testing, clarifying RTB OPERABILITY requirements and obtaining informa-tion needed for future staff actions) rather than by the testing actions alone.

I am enclosing a memorandum from D. G. Eisenhut to E. L. Jordan, et al, (and its enclosure, a joint NRR/IE Review Team Report on GE Type AK-2-25 RTBs) dated May 18, 1984 (Enclosure 4).

The enclosed memorandum provides information on how the proposed bulletin fits into the overall RTB program.

In brief, the bulletin will be used as an interim measure aimed at assuring the operability of the RTBs and that appropriate corrective actions are taken prior to imple-menting long-term design changes.

Your attention is called to the reliability evaluation of the RTBs contained in the enclosed review team report.

Tnis evaluation is based on two separate calculations using different analytical methods.

The first calculation indicates that the new preventive maintenance program has had no effect on RTB reliability.

In contrast, the second calcula-tion indicates that the new preventive maintenance program may have resulted in an order of magnitude improvement in RTB reliability when compared to NUREG-1000 data.

As you may recall, the second calculation was included in a previ-ous draft report that was prematurely forwarded to your staff; however, as stated in the report, the NUREG-1000 data used for comparison in this evalua-tion is suspect.

Although neither calculation should be used as a basis for NRC actions, the first appears to have more merit.

Nevertheless, because of the uncertainties associated with the data, the quantitative results of both calculations were deemed as being inconclusive.

Also enclosed is a copy of SECY-84-215, dated May 24, 1984 (Enclosure 5), which includes the staff's technical assessment of licensees' responses to Generic Letter 83-28.

As evidenced by the percentage of incomplete responses, espe-cially those related to the RTS reliability (maintenance and surveillance of breakers), additional actions need to be taken, some of which are included in the bulletin.

Victor Stello.

' Finally, a tabulation comparing the current testing of RT8s being' conducted by NSSS classification with.the additional tests required by the proposed bulletin has been prepared (Enclosure 6)..It should be noted that the comparison includes.

two d.iscrete tabulations of the so-called " additional" tests required by the bulletin.. The first such tabulation.is based on the assumption that the tests required by the bulletin cannot be combined with existing tests.

If such were the case, the additional tests would represent about an 11% increase in RTB testing for a period of one year.

The second tabulation, which we believe more but are measurements of-parameters that could be taken concurrent with most of the on going tests.

On such a basis, the additional tests being requested by the bulletin would represent less than a 1% increase in RTB testing for a period of one year.

In summary, we conclude that the burden of additional testing imposed by the bulletin is miniscule, as is the probability of causing a spurious reactor trip by such measurements.

We believe that the performance tests will identify pre-cusors to RTB failures and therefore the actions being requested by the bulletin would improve overall RTB availability such that the likelihood of core melt may well be reduced.

,0riginal signed by c

Richard C. DeYoung, Director Office of Inspection and Enforcement

Enclosures:

1.

Proposed Bulletin

2..

Memorandum for C. E. Rossi from G. M. Holahan 3.

Safety Impact of the Proposed Bulletin, Reliability and Risk Assessment Branch 4.

Memorandum for E. L. Jordan, et al, from O. G. Eisenhut dated May 18, 1984, Joint NRR/IE GE Type AK-2-25 Reactor Trip Breaker Report 5.

SECY-84-215 dated May 24, 1984, Status Report on Generic Letter 83-28 (Salem ATWS) and

-Related Issues.

6.

Tabulation of RTB Tests By NSSS Vendor Distribution:

DCS EA8 R/F s DEPER R/F RCDeYoung JMTaylor SASchwartz CERossi NVillaiva WSchwink 15 AThadani SNewberry

• SEE PREVIOUS CONCURRENCES i

GHolahan

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ENCLOSURE 1 SSIN No.:

6835 005 No.:

3150-00012 Expiration Date:

IE8 84.

UNITED STATES NUCLEAR REGULATORY C0f0f!SSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.

20555 April

, 1984 IE BULLETIN NO. 84- :

UNDERVOLTAGE TRIP ATTACHMENTS OF REACTOR TRIP 8REAKERS Addressees:

All pressurized water reactor (PWR) power facilities holding an operating license (OL) for action except Yankee Rowe, Fort Calhoun and Palisades; all other nuclear power facilities for information.

Purpose:

The purposes of this bulletin are:

adequate testing of the. reactor trip breakers (RT8s), (b) to clarify the "0PERASILITY" requirements related to RTBs, and (c) to collect information i

needed for future staff decisions related to the need for improving the reliability of the undervoltage trip attachments (UVTAs).

Description of Circumstances The NRC staff has become aware of 41 instances at seven nuclear power plants where the UVTAs failed to trip GE Type AK-2-25 RTBs within the acceptance time 1'

specified by the licensees.

These failures occurred subsequent to actions initiated by licensees in response to IE Bulletin 83-04 and Generic Letter 83-28.

In some instances, the UVTAs did not trip their associated RT8s at all while in other instances multiple RTB failures occurred.

On two separate occasions, all eight of the UVTAs used in a Combustion Engineering-designed plant failed to trip their associated RT8s within the acceptance ti.ite specified by the licensee.

Background:

As a result of the February 22 and 25,1983 Salem anticipated transients without scram (ATWS) events, the NRC issued IE Bulletin 83-01 and formed a task force to' assess the generic implications of these events.

On. March 11, _

1983, Southern California Edison reported that three RT8s on San Onofre Unit 2 and one on Unit 3 failed to open during testing of the UVTA.

As a result of

~

.these failures, the NRC issued IE Bulletin 83-04.

The task force's actions resulted in the issuance of NUREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant" and Generic Letter 83-28, delineating the procedural and plant changes required.

Findings in NUREG-1000 were based, in part, on assurances that improved maintenance of a

l IEB 84-April

, 1984 Page 2 of 5 the breakers being recommended by the manufacturers would improve the rel.iabil-ity of the WTAs to an acceptable level for the short term, after which the 4.

longer term corrective actions would be implemented.

Based on the recent number of faj1ures experienced with GE AK-2-25 type breakers, it now appn rs i

that an acceptable reliability is not being attained at all plants.

A partic-ular concern is that many of the failures occurred after the breakers were reportedly maintained per GE's latest recommendations.

The NRC staff, there-fore, is concerned that maintenance alone may not be effective in providing reliable performance of GE AK 2-25 RT8s on a continuing basis.

In view of the existing small design margin of the UVTA regarding forces available versus those required to actuate the breaker, it appears that design improvements are necessary to achieve the long-term high reliability required for reactor trip systems.

Nevertheless, licensees are expected to continue to emphasize preven-tive maintenance to assure high breaker reliability.

Toward the same end, the NRC will continue to scrutinize any ensuing RTB malfunction in order to pre-ciude an ATWS event.

t During one recent instance, all the RTBs failed to open within the acceptance time specified by the licensee; however, since the shunt trip devices in the RTBs were functional, the licensee considered the RT8s to be "0PERA8LE."

The NRC staff does not agree with this approach and accordingly has developed the interim guidance given later in this bulletin.

Said guidance is to be used until such time as the relevant issues of Generic Letter 83-28 are resolved.

In view of recent operating experiences, the NRC established a special review team to determine which (if any) actions in Generic Letter 83-28 involving 1

GE AK-2-25 RTBs need to be accelerated or changed.

Certain information being solicited by this bulletin will facilitate the staff's final decisions on these matters.

In addition, Westinghouse-designed plants are included in this bulletin because of the uncertainties involving the manner by which their RTBs are being tested and whether any such RTBs have failed to open per their i

OPERA 8ILITY criterica in 1983, and especially because most such plants do not presently include an automatic back-up device for the UVTA such as a shunt trip attachment or silicon controlled rectifiers.

During recent meetings with the staff, both the Combustion Engineering and the Babcock & Wilcox owners' Regulatory Response Groups stated that diverse -

reactor trip features (shunt trip and silicon controlled rectifiers) are currently being tested on a periodic basis at'each plant.

Such tests, coupled with the tests requested by this bulletin, will ensure that the capability of reactor trip is closely monitored pending completion of long-term actions to improve reactor trip system reliability.

The' actions.being requested by this bulletin shall remain in force until the i

relevant issues in Generic Letter 83-28 are resolved or for twelve (12) months from the date of this bulletin, which ever occurs first.

i I

a

-IEB 84-April

, 1984 Page 3 of 5 Required Actions for PWR Facilities Holding Operatina Licenses (Except Yankee Rowe, Fort Calhoun and Palisades):

1.

Performance tests of the UVTA function of each in-service RTB shall be initiated within 30 days of receipt of this bulletin.

Response time tests are the preferred performance tests for plants using GE RTBs.

. Licensees using Westinghouse RTBs may find that other parameters such as drop-out voltage are better indicators of an INOPERABLE RTB than response time. -In either case, licensees may propose alternatives to response time as the performance tests.

In such cases, licensees shall provide the bases for selecting the alternative parameters.

To minimize the impact of increased test time and the likelihood of inadvertent trips by the actions requested by this bulletin, the requested performance tests should be conducted to the maximum degree practicable in conjunction with presently required tests.

The tests shall be conducted monthly until each breaker successfully passes two consecutive monthly tests, after which the test frequency may be relaxed to a two month inter-val.

If a breaker fails to pass its performance test during the two month testing interval, maintenance should be performed and the surveillance testing shall be returned to a monthly interval.

Performance tests shall be conducted independent of shunt trip, if applicable, and prior to any maintenance, adjustment or functional tests.

Testing should be conducted to the maximum degree practicable with the RTBs in their cubicles, provided such testing does not jeopardize personnel or plant safety, and prior to any operation that would trip the RTBs.

The use of lifted leads or jumpers is to be minimized.

If lifted leads or jumpers are used during RTB testing, subsequent steps in the procedure must demonstrate that the OPERABILITY of the tested RTB has, in fact,- been restored.

Performance tests may be conducted on a staggered basis provided that each in-service breaker is tested during the test interval described above.

Plants-for which on-line testability is not provided (i.e.,

plants whose RTBs cannot be tested without tripping the reactor) shall perform these tests prior to resuming operation or if currently operating, during the-next plant shutdown.

2.

Provide the following information:

(a) A brief description of the performance tests being used to verify the operability of the UVTA.

The description should include the accuracy of the test equipment and the location of the RTB during the test (e.g...in-situ, bench).

(b) An explicit OPERABILITY criterion for the UVTA performance parameter (e.g., response time, drop-out voltage) of the RTB based upon the current licensing basis for the plant.

This criterion should conservatively account for the instrument accuracies stipulated in i

l 1

C IEB 84-April

, 1984 Page 4 of 5 item 2 (a), above.

(Note:

Based upon information provided to the, NRC staff, we believe the response time criterien should be 100 milliseconds or less for Combustion Engineering and Babcock &

Wilcox designed plants.)

^(c) Plants are requested to report the failure of any RTB to respond within its UVTA OPERABILITY criterion or other failure during the i

calendar years of 1983 and 1984, including the date of the failure.

3.

Provide written instructions to the plant's operating staff requiring that:

(a) A reactor trip breaker shall be considered INOPERABLE if its UVTA performance parameter does not meet the OPERABILITY criterion of 2(b), above.

(b) IN0PERABLE RTBs shall be promptly restored to OPERABLE status by repair or replacement and, in any case, treated in accordance with the plant's technical specifications regarding time limits for operation with INOPERABLE RTBs.

In the absence of plant technical specifications regarding time limits for operation with INOPERABLE RTBs, the licensee shall specify actions to be taken to ensure that the plant is not operated in a critical mode for longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with an INOPERABLE RTB closed.

(c) The NRC shall be notified via the Emergency Notification System (ENS) of any RTB becoming inoperable, regardless of whether the failure occurred during in-situ or bench testing of a functional RTB or the operating mode of the reactor.

Because of the generic implications and the common mode failure potential associated with UVTAs failing to meet their operability criterion, the NRC staff has determined that for the duration of this bulletin such failures shall be reported via the ENS within four hours unless the number and/or l

locations of the failed RTBs warrant reporting under 10 CFR 50.72 on an acce.erated basis.

Further, an inoperable RTB shall also be reported under 10 CFR 50.73 unless it is determined that redundant breakers would not be susceptible to the same fault or condition.

Should a licensee determine that any action requested by this bulletin jeopardizes overall plant safety, the NRC should be notified of that fact and provided with appropriate justification for not implementing the requested action.

Such notification shall be made within 30 days of the date of this bulletin in accordance with the notification instructions of this bulletin.

A written response addressing each of the items including results of the first l

monthly test shall be submitted within 60 days of the receipt of this bulletin i

to the appropriate Regional Administrator under oath or affirmation under provisions of Section 182a, Atomic Energy Act of 1954, as amended.

The original of the letter and a copy of any attachments shall be transmitted to the U. S.

Nuclear Regulatory Commission, Document Control Desk, Washington, D. C.

20555 for reproduction and distribution.

Subsequent reports shall only be submitted l

l

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IEB 84-April

, 1984 Page 5 of 5 l

following a RTB 1 ailing to meet its OPERABILITY criterion.

Such reports shall be made within 30 days of the malfunction in accordance with the requirements of the original letter report unless such malfunctions are reported under 10 CFR 50.73.

These reports shall identify the INOPERABLE breakar, its devia-tion from the OPERA 8ILITY criterion, remedial actions taken, and the results of said actions.

This request for information was approved by the Office of Management and Budget under a blanket clearance number 3150-0011, which expires April 30, 1985.

Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D. C.

20503.

I i

Richard C. DeYoung, Director i

Office of Inspection and Enforcement I

i Technical Contacts:

I. V111alva, IE J. T. Beard, NRR (301) 492-9635 (301) 492-7465

Attachment:

List of Recently Issued IE Bulletins k -.

s

ENCLOSURE 2

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',A UNITED STATES NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D. C. 20006 k,

my.g g MEMORANDUM FOR:

C; tuttessi, Chief Events Analysis Branch Division of Emergency Preparednass and Engineering Response Office of Inspection and Enforcement FROM:

Gary M. Holahan, Chief Operating Reactors Assessment Branch Division of Licensing s

Office of Nuclear Reactor Regulation

SUBJECT:

PROPOSED IE BULLETIN ON UV TRIP ATTACHMENTS ON REACTOR TRIP BREAXERS The following analysis was perfonned in order to be responsive to V. Stello's concern that additional monitoring of parameters (such as trip time) could be detrimental to safety in that it could increase the frequency of reactor trips. The anal RTB testing; 2) ysis indicates that:

1) few reactor trips are caused by the monitoring of parameters is not likely to noticeably increase the frequency of reactor trips; and 3) that the contribution to core melt from these activities is extremely small.

The event sequence of interest is a reactor trip followed by loss of all feedwater (main and emergency).

The estimated probability for this sequence (for a PWR) for all reactor trips is:

P (core melt) = P(trip)

• P(LO M FW)

• P(Low Aux Fw)

• P (inadeo Feed & Bleed) 10~4 P (core melt) = 7/RY 0.. I

~0.5 P (core melt) = 3.!i x 10-5/RY The probability of such a sequence being initiated due to RTB testing is:

P(cm from RTB testing) = Pf trip from RTB testing)

• P(Lo in FW)

• P(Low Aux FW)

• P(inadeq. Feed & B P=.02/RYx9ix10-}eed) x 0.s P = 1.0 x 10~ /RY Where the probability of a reactor trip due to RTB testing is based on a point estimate from In83 PWR operating experience, (i.e. O trips due to RTB testing in50PWRRY's). Based on INPO statistics for the last six years, only 9% of reactor trips are associated with activities on any part of the RPS and only

47. are a:30ciated with errors during testing.

It is therefore not surprising that there have been no recent examples of reactor trips caused by RTB testing.

M (p WG:'t?@~T F -

l C. E. Rossi Under the most' conservative set of assumptions (i.e. assuming that monitoring would be the cause of all cases of RTB testing resulting in reactor trips) and using a conservative point estimate of the frequency of RTS testing resulting in reactor trtps the core melt probabil associated with RTB's would be 1.0 x 10'f ty from the current monitoring activities

/RY. Since the proposed bulletin would increase the amount of monitoring of parameter during RTB testing by a factor of 2 to 4.

The maxim 1.0-7/RY to 4.0 x 10 ym increase in core melt probability would be from 1.0 x

/RY.

lower, perhaps by several orders of magnitude.The actual contribution is likely to In addition to being the direct cause of reactor trips, RTB parameter monitoring could contribute indirectly to the frequency of reactor trips by keeping a given RTB out-of-service longer than it would be othemise.

During such a period, the plant could experience a reactor trip due to a single failure in another part of the RPS. Currently RTB are out-of-service for about 8 days /

servicing twice per year plus 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of testing each month) year (2 days of For most plants the current monitoring of results in an additional time out-of-service of

.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> / year (.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> twice per year) or.3% of unavailability (i.e., plant susceptibility to trips on single failure). Under the proposed Bulletin, this component of unavailability would be increased by approximately a factor of 4 resulting in a 1.2% contribution. Since all RPS testing only contributes 4% to reactor trips, the contribution to reactor trips of parameter monitoring would be less that 1 would be 1.7 x 10-8 2% of 4% or.05%.

The corresponding core melt contribution

/RY.

An analysis of the improvement in RTB reliability due to early identification of degraded performance is not available, however, it would appear to also have only a very small contribution to changing the estimated probability of core melt since RTS reliability is not the controlling factor in the RPS reliability for most systems and since the probability of a core melt from an ATWS event is already relatively small.

In sumary, the proposed Bulletin will clearly not cause a significant reduction in safety; will probably cause a small improvement in estimated core melt probability and will address an identified deficiency in an important safety related component of the RPS. The monitoring parameters during RTB tests will identify, and result in repairs, of UV devices which are undergoing, progressive degradation and which are likely to fail if not given attention. On this basis I continue to support the proposed bulletin.

kf*

.k w Gary M.Holahan, Chief Operating Reactors Assessment Branch Division of t.icensing Office of Nuclear Reactor Regulation cc:

F. Miragita E. Jordan E. Buche J. T. Beard J. Conran i