ML19220D037

Forwards Final Evaluation of Items Discussed in 790228 Memo Re B&W-designed Nuclear Plants.Requests Addl Info Be Forwarded to Licensing Boards

ML19220D037
Person / Time
Site:	Davis Besse, Midland, Crane
Issue date:	03/28/1979
From:	Moseley N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
To:	Thompson D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
Shared Package
ML19220D004	List: ML19220D003 ML19220D017 ML19220D025 ML19220D031 ML19220D037
References
TASK-TF, TASK-TMR NUDOCS 7905160216
Download: ML19220D037 (10)

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l'IMOPMDUM FOR:

Dudley Thompson, Executive Officer for Operations Supcort, IE FRCM:

flor.an C. Moseley, Director, Division of Reactor Operatior.s Inspectica, IE SU3 JECT:

tiOTIFICATICN CF LICENSING ECARDS On February 23, 1979, six items concerning Babcock and '.lilcox designed nuclear plants were cent tc you fcr for.;ardir; to the ascrcpriate licensing boards.

At that time only a preliminary evaluatica had been done. We have completed cur evaluaticn of each. cf tr'.e items and '. hat information is enclosed.

Inis additi;nal 'nfor.ation should ba foraarded to the licensing beards.

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f Harman C. Vaseley Director

? vision of Reactor Operations Inspec: ion, IE

Enclosure:

Evaluations of Concerns cc:

S. E. 3ryan E. L. Jordan R. P. Heishman, RIII J. C. Stcr.e D. Kirkpitrick LGrC' Gewer V. D. Th: mas CONTACT:

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EXCER?T FRCM ME' CRA' DUM ENTITLED " CONVEYING SEW INFOR"ATIO. TO LICENSINO 30AEOS - DAVIS-BESSE [RIITS 2 & 3 AND MIDLAND UNITS 1 5 2",

DATED JANUARY 3, ic79, FROM J.S. CRESWELL TO J.F. STREETER s

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During a recent inspectica at Davis-Besse Unit 1 information has been attained which indicates that at certain conditions of reactor coolant viscosity (as a function of temperature) core lif ting may -

occur.

The licencee informed the inspecter tha :his issue involves other E62 facilities. The Davis-3 esse FSAR sgates in Section 4.4.2.7:

The hydraulic force on the fuel assembly receiving the most flow is shown as'a function of systen flow in Figure 4-39.

Addit:cnal forces ac:in; on the fuel assembly are :he assembly welght and a hold dcwn spring force, ehich resulted in a net d awrward f orce at all times during nce al station operation.

The l'.enset s tates that there is a 500 F interloc.; for the start-ing of tae icurth reactor coolant pump.

However, no Technical

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Specificarica recuires tha t the pump be startef at or above this t enp era t t.r e. A concern regarding this matter would be if assem-blics reved upward into a position such that control rod nove-ment uccid be hindered.

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w The potert.al fo' core lif ting in 35W plants is a concern wnich has been previously reviewed by NRR.

The concern was first raised in conne: tion with the Ocenee 2 and 3 reac crs, where the primary coolant flow rs'.es were found to be in excess of the design flow rates.

For exampi., the Unit 2 ficw rate was found to be 111.5% of the design f1:w rate.

Since this was very near the predicted core lif: ficw ra:e of 111.9", an analysis was done by 350 to determine what effect. core lifting would have en the previous saf ety analysis f or these plants, This analysis (dated May 2, 1975) indicated that the potential for care lif ting did not result in sn unreviewed safety question. A subsequent review of this 352 analysis by NRR also concluded that an unsafe condition did not exis: (letter frc= R. A. Purple :o Duke Fever, dated 9/24/75).

It should be noted that the potential vertical displace ent of the core is limited to a very small distance by the upper core support structure.

Core lifting at power would result in a slight reduction in reactivity since the rising fuel wculd tend to engage the withdrawn con:rol reds to a slightly greater extent than it would bottomed condition.

The arount of :his change in reactivi;y i' mrse, available for reinsertion should the fuel settle be'

's agina' cosition.

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pc:ential reactivity increase caused in; cf the 16 centrally located con:rol rod assembly elemen

" i. ave been subject to lifting in :he Oconce 2 reactor) was

..a;ed :o be 0.17. a K/R.

This value is insufficien: to have much effec: en the acciden: and transien:

?,afety analyses.

119 169

Sectica 1. An additional concern was the potential for d2: age to the fuel assembly s _end fittings which tight be caused by fretting due to repetitive fuel covement. Consequently, Duke Power was requested by NRR to ecke certain exaninations of the Oconee 2 fuel during the first refueling to confirm that fuel ele:ent notica was not occurring.

The results of this exacination (letter fro: W. O. Parker to R. C. Rusche dated 7/21/76) showed that no fuel lif ting or other type of motion had occurred during the first cycle of operation.

After the core lift concern was identified, 3&w developed newer types of fuel holddawn springs which provide core cargin against core lifting than the previous springs did.

It is our understanding that the newer types of springs have been installed in all 3&e' reactors.

For these reasons, we believe tha: there is presently 1i::12 likelihood that core lif ting will occur during normal pcwer opere'icr.

At lawar temperatures, there is an increased flew inducad lifting force on the fuel due to the higher viscosity of the reactor coolan:.

Consequently, o

we view the restrictica aga! ist 4 pump cperation belcw 500 F as a prudent' pre :ution agains

. eel fre: ing.

Mcwever, since the pctential for cora lif ting has 11 :le safety significance and because critical operatica below 300 ? is not permitted, we have no basis to recontend

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30ARDS - DAVIS-EESSE UNITS 2 5 3 A.':D ':DU.ND UNITS 1 5 2",

':ATED JANUARY 8, 1979, FRC". J.S. CRES'.;ILL TO J.F. S3EETER 2.

Inspection Report 50-346/7c-06, paragraph 4, reported reactivity -

power oscillations in the Davis-3 esse core.

These oscillaticas have also occurred at Oconee and are attributed to stean genera-ter level oscillations. 3 5'a' report 3A%-lC027 states in A9.2:

The OTSG laboratory nodel test results indicated that periodic oscillations in steam pressure, s:can flow, and steam generator pri ary outlet te=peratures could occur under certain conditions.

I: "as shown that the oscilla tions were of the tv. pe associa-ted with the relationships be tween f eedwa:er hea:ing chamber pressure drop and tube nes: pressure drop, which are elimi-nated or reduced to levels of no conscquence (no f eedback to r ea c:cr svs ten) by adjus tnent of t,ne tube nes: 4nie:

r es is tance.

A3 2 result of the tests, an adjust:ble orifice has been installed in the dcunccccr sec:ica of the g

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nest inlet resistance and to provide the means f or elimina-tien of escillaticas if they should develcp during the operating lif etime of the generators.

The initial crifice setting is chosen conservati-tely to minimize the need for further adjust =ent during the startup test prcgran.

k'e also no:e th2: the effect on the incore de:ccccr systen for moni:oring core parameters during the oscillaticr.

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..., A.L..~.. L,v. 4 wb Fower Oscilla:icns of the order of 1.5% of full power have been observed at all of the Oconee plants and are ccasidered nornal.

In 19 7'i the power oscilla:icas experienced by the Oconee 3 reac cr increased := a taxinua of 7.5% of full power. At that tire the proble: was reviewed by

';RR with the conclusion that there was no significant safety considera-tion at that value (No te to 3. C. ? ackley f rca S. D. MacKay, dated January 27,19 7S).

It should be no:cd that the 7.5% power oscilla:icns o

cause about a 1 ? oscillatier. in core average te perature due to the short period of the oscillations. The important core safety parane:ers, which are, the departure fre: nucleate boiling ra:io and the average taximun linear hea: 5'eneration rate are affected verv. little bv cscilla-tions of this a:pli:ude. The primary cause of the power oscillatiens

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is believed to be a fluctuation of the secondary water level in the steam genera: ors.

This can be cinimized by increasing the flow resistance M 4

Section 2. in the downconer region of the stean generators. The corrective effort at Cconee 3 was co plicated by the f act that the orifice plate provided ror this purpose could not be fully closed.

However, the oscillations at other BSU plants have been kept to about 1.5*' of full pcuer by appropriate adjustment of the downconer flow

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resistance. For these reasons, the power oscillaticns at 35W plants are not considered to be a significant safety concern.

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Inspection and Enforce cat Report 50-345/78-06 docc=ented that pressurizer level had gone of f scale f or approximately five minutes during the November 29, 1977 less of offsite power event.

There are sc e indications that other 3LW plants say have prcb-Icas naintaining prassuriser level ind;ca tions curing transients.

In addition, under certain conditions such as loss of feedwater at 100T. power with the reacter coolant pumps running the pres-stri:er may veid ccepletely.

A special analysis ha s been per-forced concerning this event This analysis is attached as.

Because of pressuri cr level maintenance prob-

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scale (less than 52C3F).

In additica, it was noted that the cakeup flow =cnitoring is l'-<

ad to less than 160 gp: and tha t takeup ficw may be substantially great?r :han this va lue.

This informa rion should be exacined in light of the require-cents of GDC 13.

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that no unreviewed safe:y question existed.

The pressuri:er, together wi:5 the resctor coolant makeup systen, is designed to raintain the pri:ary systen pressure and water icvel within their operational limits only during normal operating conditions.

Cecidewn transients, such as loss of of f site power and loss of f eed-wa t e r, scuetimes result in prinary pressure and volume changes that are beyond the ability of this syste to control. The analyses of and experience with such transients show, however, that they can be sustained withcut cc=pr: ising the safety of the reactor.

The principal concern caused by such transients is that they night cause voiding in the prinary coolant sys: 20 that vculd lead to loss of ability to ade-quately cool the reac cr core. The safety evaluation of the 1 css of of fsite pcwer transient shows that, though level indica:ica is lost, sone water remains in the pressuri:cr and the pressure dces not decrease below about 1600 psi.

In order for voiding to occur, the pressure must 2_

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in :his case, since pressure does not decrease to satura:1ca.

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indicates that the water volume could decrease to less tha the syste volute exclusive of the pressurizer. During such an event, the emptying of the pressurizer would be followed by a pressure reduction below the saturation point and the for ation of snall voids throughout much of the primary sys:ce.

This could not result in the loss of core cooling beccuse the veids would ce dispersed over a large volute and forced flow would precent che: from coalescing sufficiently to prevent core cooling.

The high pressure coolant injection punps are started autcratically when the pricary pressure decreases belcw 1600 psi.

Therefore, any pressure reduction which is sufficient to allcw voiding till also result in water injection which will rapidly restore the primary water to nor:21 levels.

For.these reasons, we believe that the inability of the pressuriner and narcal coolan: 24Keup syste: to control sc:e transients does not p ro v.' ' c a-b a o '. s-

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,u-the a: cunt of ~ 2akeup ficv in excess of ic0 gp does not appear to be a sign,.:1 cant, actor in :.ne course at-these occurrences.

The loss of pressurizer water level indicatica could be considered to deviate fro GDC 13, because this level indication provides the principal ceans of determining the primary coolan: inventory.

Scwever, provision of a level indicatica that would ccver all anticipated occurrences may not be practical.

As discussed above, the loss of feedvater event can lead to a =ccentary condition wherein no ceaningful level exists, because the entire primary systen contains a steam water mixture.

It should be noted that the introduction to Appendix A (last paragraph) recognizes that fulfill ent of so:e of the criteria cay not always be appropriate.

This introduction also states that departuras frc the Criteria cust be identified and justified.

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.m JA'OARY 8, 1979, FRCM J.S. CRESWELI, TO J.F. STREETER 4.

.A me=o frc: ESU regarding control red drive syste: trip breaker maintenance is attached as Enclosure 2.

This reto should be evaluated in ter:s of shutdown =argin aintenance and ATVS considerations particularly in light of large positive cderator coefficients allowable with 35W facilities.

DISCL'SSION AND EVALi.'ATION Our investigation of the above circuit breaker problems has revealed that -ight f ailures of reactor scra: circui: breakers :o trip during f

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faciliti2s since 19 73.

In each case, the faulty circui: breaker was iden:ified as a GE type AK-2 series (i.e., As-2A-15, 24, or f0).

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2chanis of the undervaltage trip device (UV) and trip shaf: asse:biv cr an out-of-adj ustnen: condition in the same linkage ne hanist.

35W and CE' determined that the binding and the out-of-adj us t:en t conditions

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. 3 30, 19 78 event, both redundant service water pump circui: breakers failed to : rip as required durin; :he loss of of f-site power test.

These f ailures in turn created a potential everload condi: ion en the eter;ency busses during the sequential bus loading by each diesel generator.

However, both diesel generators successfully picked up their required bus loads without experiencing a unit shutdcwn f rom an overlocd conditica.

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.a issue, both B&W and GE are in the process of issuing alert letters to their custeners. These letters are scheduled for issuance by late March and will describe the causes for failura. and provide recer.nendations to resolve the problem.

Eased on our study findings and on information ob:ained in discussions about the breaker proble: with the knowledgeable people frc: 3 5 n', CE and

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3 thrust of the Circular will be directed toward the need for adequate

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Inspection and Enforcement Report 50-346/78-17, paragraph 6 refers to inspection findings regarding the capability of the incore detector syste: to deter =ine worst case thernal conditions.

The reactor can be operated per the Technical Specifica:icas with the center incore string out of service.

If the peak power locations is in the center of the core (tnis has been the case at Davis-3 esse), fac crs are not applied to conset vatively imnitcr values su:y c

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'.'e do no t b elieve tha t there is a valid basis for requiring the center string of incore detectors to be always operable in 3 &.s' reac:crs.

The pcwer cistribution s for various plant conditiens, :hrcughou: the fuel

cycle, re calculated prior to the operation of the reactor.

The power distribution is verified at the beginning of cperation, and periodically

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2 om the pcwer ratics between such an asse bi; and nearby assemblies tha: have de:20: ors.

These ra:ics can then be cultiplied by the pcw2r in the reasured assemblies to derive th2 power icvel in any specific unreasured a s a e -.' 1 e.

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s pcuered csse:bly in the Davis 3 esse reactor at the beginning of the fuel cycle, this is net the case at all reactors. Nor does the central assembly have the highest power, in the Davis 3 esse reactor, a: the end of the first fuel cycle.

Since there is sc=e variation between the calculated pcwer distributions and the actual ones, an appropriate cargin is assured for this variation 4-establishing the allcwable power peaking factors.

Fixed incore detectors cust functica in an extremely harsh enNirennent and are subject to high failure rates.

In order to ensure that an ade-quate nu ber will survive the fuel cycle, many more detectors are installed than are necessary for the power distributions determinations.

To require the central string to be always operable would likely resul a.

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m Specifica:icns (STS) for 3&W plants nor the ST3 for CE plants (which also have fi:<.ed in: ore cetectors) require the central de:ectors to be operable.

i19 176

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E0ARDS - DAVI3-3 ESSE L7ITS 2 & 3 A;;D MIDISO U::ITS 1 &

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6. describes an event that occurred at a 35W facility which resulted in a severe thercal transient and e:<trere dif-ficult'f in centrolling th'e plant.

The af orementioned f acilities snould be reviewed in light of this infor ation for possible safety 4-14-'tions.

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7 FollowinI the cooldcun transient at Rancho Seco, N~L'. evaluated the event and cc.uchded that r.o structural damage had occurred to the primary coo l;nt systen which would preclude f uture operation of Rancho Seco.

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RR perfor= a generic review of the non-nuclear instrumentation power supplies for cther 35*J units, if design changes to the non-

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events c: failurcs which could cause sicilar significant cooldczn t ransien t's.

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