ML20052C552

Responds to Request for Differing Prof Opinion Re Pressurized Thermal Shock

ML20052C552
Person / Time
Issue date:	10/08/1981
From:	Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Palladino NRC COMMISSION (OCM)
Shared Package
ML20051T417	List: ML20051T385 ML20052B089 ML20052C548 ML20052C552 ML20052C560
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8205050143
Download: ML20052C552 (15)

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0:tober 8,1981 MD40RANDUM FOR:

Chaiman Palladino

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FROM:

William J. Dircks Executive Dire: tor for Operations

SUBJECT:

DIFFERING PROFESSIONAL OPINION OF D. BASDETAS ON PRESSURIZED TrEPFAL SHOCK At your meeting on September 17, 1931, with Mr. D. Basdekas, he discussed '

.with, you a series of his concerns related to pr.essurized themal sho:k of PWR pressura vessels.' After that meeting, you asked Dr. Denwood Ross to provide staff views on the points raised by Mr. Basdekas.

In response to'your request, responses as provided by the NRR and RES staffs are enclosed.

EtnedWFilam J.Dir:b William J. Dircks-Executive Director fo'r' Operations

Enclosure:

Answers to Mr. Basdekas' Concerns cci Comissioner Gilinsky Comis'sioner Bradford Comissioner Ahearne

. Comissioner Roberts D. Basdekas, RES R. Minogue, RES H. Denton, NRR SECY Mc::Cu cam

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COWINT 1.

"Some of the steps taken or proposed by the Staff may be necessary, but they are not sufficient to provide an acceptable level of protection to public health and safety.

Contrary to the staff's position, this matter is urgent.

It is also more extensive than the Staff states it.to be."

RESPONSE

The staff regards pressurized thermal shock of pressure vessels as a very important ssfety issue and believes that the actions taken and planned are appropriate on the basis of our current knowledge.

In the judgment of the staff; there is sufficient time to' evaluate the infomation to be submitted in response to the letter to eight licensees and in the PWR Owners Groeps' generic studies before deciding what further 'rs9ulatory action is needed.

The' staff report to the Commission, SECY-81-286A (dated September 8,1981) con-veys a sense of urgency..A complete reading of Enclosure 1 to SECY-81-28SA, which is the minutes of meetings held July 28-30,'1981, indicates that the staff position is for prompt, positive. actions to prevent p6tentially damaging transients or mitigate their effects'.

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COMMENT 2.

The precursory operational experience on pressurized thermal shock has been noted, but not heeded by the Staff. The probability of severe' overcooling accident sequences (particularly outside the design basis envelope) are substantially greater than those given by the Staff. The industry's " bounding" cases are based on design basis accidents.

htESPONSE.

TE~siaff has made.a study of PWR operational experience to look for severe overcooling' transients and their precursors.

In 1980, the RES ~ staff reviewed

-the operat.ing experience of B&W plants and found that there had.been a number of overcooling transients in B&W plants.

The most serious transient was that' at Rancho Seco on. March 20, 1978, in which the coolant temperature dropped from

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550*F to 280*F in about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> while the system pressure first dropped, then returned to near its original value.

Based on this experience, a probability of 3 x 10-2/ reactor-year was estimated for a B&W plant to experience an over-cooling transient as severe or more severe than the' Rancho Seco event was estimated.

In addition, B&W plants 'have. experienced several similar, but less severe transients such as occurred at Oconee-3 in November 1979 and Crystal River-3 in February 1980.

These and other small er transients that occur. red.in B&W plants lead to an estimate of 5 x 10-I sina11 transients per year in B&W plants as described in M.. A. Taylor's memorandum of October 29, 1980.

Since.these occurrences, operators 'have received special training in transient.

response.

Babcock and Wilcox plants have added a back-up power supply to the nonnuclear instrbmentation bus, whose failure initiated the three transie,nts above. The NRR staff examined the impact.of the improved power supply and operator training and suggested that these improvements might have reduced the probability to 10~3/ reactor year for an overcool.ing transient as severe as the Rancho Seco event for B&W plants.

ComENT 2 (continued)

The operating experience of CE and Westinghouse plants has also been examined.

There have been no events like the Rancho Seco transient, but thsre have been scme precursors.

These are events which typicall.y led to secondary steam dump valves or steam bypass' valves sticking open, but which did not result in steam flows iarge enough to produce very severe overcooling transients. The most

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severe of these transients occurred at Arkansas Nuclear One-2 (a CE plant) on December 27, 1978, where a main steam relief valve ' lifted and failed to reset, thereby causing the reactor coolant temperature to drop by 107'F in 52 minutes.

Based on this experience,-the staff estimates the probability of.a severe over-cooling transient in a CE or )' plant due to a large steam line break or its

-4 equivalent to be no greater than about 10 / reactor yea'r..

In sumary, the staff estimates the probability of a severe overcooling transient e

during which the primary pressure remains ~ high is about 10-3/ reactor year for B&W

-4 plants and 10 / reactor year for CE and Westinghouse plants.

There may be a f5.ctor of iO uncertainty associated with these estimates.

Although some of the PWR Owners Groups' calculations have been based on design basis ' ccidents, we have asked for analyses of a wider range of overcooling a

tra'nsients.

Thess analyses should cover transients resulting from multiple failures or operator errors.

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I or CDfEENT 3.

The uncertainties i.n critical parameter values are substantial culminating in RTNDT uncertainties far greater than those given by the staff.

Domestic and foreign experience indicates a trend -

for higher than estimated RTNDT.

s RESPONSE.

4 The sta.ff is aware that there are uncertainties in estimating RTNDT. principally from the following sources:

(1) uncertainties in the initial RT

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,(2) uncertainties in copper content of the weld metal; (3) uncertainties in vessel fluence estimates; (4) uncertainties in irradiation temperature; shift to (5) uncertainties in the Regulatory Guide 1.99 curves relating RTHDT fluence and copper content of welds In calculating RTNDT, the uncertainties have been accounted for by using con-This is servative estimate,s of the factors that are used in calculating RTHDT.

especially true of the initial RTNDT. and~ th~e pr'ediction of RTNDT. shift based on Regulatory Guide 1.99.

Vessel " fluence calculations will $e carried out with a 'well benchmarked and l

- calibrated neutron transport code to yield fluence accuracy not worse than +20%,

and'the remaining uncertainty will pave only a small effect on RTNDT. The copper content of the weld meta 1 is believed to be known to within +0.03% copper (base'd

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'on measured data) and this uncertainty will have only a small effect 'on RTNDT*

i shift from sur-There is a substantial amount of information from measured RTHDT l

veillance tests of, welds which shows that the curves in Regulatory Guide 1.99 for Based on this shifts are conservative for shifts above 150*F.

predicting RTNDT information, the, staff believes that there are substantial conservatisms in the for shifts above 150'F.

. predicted values for RTNDT

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_u C0tNENT 4.

The scenario related thennohydrodynamic assumptions are not

- realistically conservative.

The designation of the Rancho Seco event as the bounding benchmark for I&C systems initiated overcooling transients is not realistic.

RESPONSE.

The staff is evaluating a wide range of possible overcooling transients, as presented hy the licensees as wel1 as those done by us*, and will assess their expected frequency as well as severity.

Until this evaluation is completed,

'the staff has selected the March 20, 1978, Rancho Seco transient as a bench-mark for use in' fracture mechanics calculations of pressure vessels. The Rancho Seco transient is not intended to be a bounding overcooling transient.

The final choice of overcooling transients to use as benchmarks for fracture

' mechanics calculations will be made on the basis of the staff's judgment of t

the probability of the transients occurring and their severity of overcooling with appropriate consideration of 'the uncertainties associated with such estimates.

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• The staff uses several system analysis' nodels to predict the temperature and

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j C0!NENT 5.

"A request for design infomation on, control systems to selected utilities was blocked by NRR management."

RESPONSE.

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Mr. Basdekas recommended that the following information be requested in the August II,1981, letter to eight licensees:

" design descriptions, including system functional block ' diagrams and single-line schematics of the plant's control systems, and 9'

the' associated P&ID's and system flow diagrams."

Mr. Basdekas believed that this infomation was required for the staff to perform a complete an'alysis of the potential for pressurized themal s'h' ock incidents occurring at these plants and also to provide some of the data for a research contract to investigate the safety significance of control system failures for which he is the NRC contract monitor.

During development of the final version of the letters, NRR Managers concluded that this request for control system design details, which are in excess of those nomally reviewed in the licensing process, was not appropriate'at this time..in the context of the letters 'on pressurized themal shock.

(The letters do request the licensees to " provide any failure modes and effects analyses of control systems currently available or reference any such analyses already

.submi tted...'")

After the staff has reviewed the information to be submitted in response to the August 21 letters, and the reports of the Owners Groups due by the end of the year, a decision will be made regarding the depth of staff review of control systems,needed to resolve the issue of pressurized thermal shor.k.

c As to obtaining this type of infomation for the research contract to investigate the safety significance of control system. fai. lures, a different, more appropriate inethod 'to obtain this infomation will,be utilized.

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It should also be noted that a Task Action Plan is being developed for Unresolved Safety Issue A-47, " Safety Implications of Control Systems." Consideration is being given in the development of that plan to the identification'of control system failures that can contribute to reactor vessel overcooling transients, and to the' development 'of criteria for plant-specific reviews of control systems.

Mr. Bas'dekas' coments.have been, and will continue to be, requested and con-sidered in the development of the plan.

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The signif.icance~of the synergistic effect of Nickel on the rate of increase of RT is n t acc unted for in Regulatory Guide 1.99.

NDT RESPONSE.

o The chemical content (including nickel)is known for all specimens from which data were derived and used in generating the curves in Regulatory Guide 1.99.

l Thu:;, tne effect of nickel on the RT shift is accounted for in Regulatory NDT Guide 1.99.

Past and recent information from measpred RT shifts for sur-NDT v'eillance tests of weld material shows that the Regulatory Guide 1.99 curves for predicting RT shifts are conservative for materials with high and low NDT nickel content for shifts above 150*F.

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COMMENT 7.

One assumption on which the staff based the in-vessel materiels surveillance program was that the welds were not going to be the

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critical elements for embrittlement considerations. Hence, the circumferential placement of sample capsules were designed with that in mind.

It turns out that in most instances, welds are the data are critical elements. Furthermore, fluence and RTNDT obtained in cycles of,5-6 years.

RESPONSE.

Surveillance capsules which consist of weld, heat affected zone and base mate-rials, are placed in reactor vessels to provide lead time information on RT NDT shifts and to. provide data' to benchmark heutron fluence calculations. It is' not necessary th'at the surveillance capsu'les be placed at the weld ' locations,

.since the data obtained can be extrapolated to the cor, rect vessel. locations by means of calculations.

In fact, locating the surveillance capsules. at the

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longitudinal welds would. interfere -with inservice inspection of the welds, and would actua11'y increase the rate of ' embrittlement of the weld. because there i.sia. peak in.the fast neutron flux just behind the capsule.on the inside wall

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of the vessel.

With~ respect to timing of. surveillance capsule examination, a capsul e is i

typically pulled at the first or second refueling. Rules for withdrawal schedules are given in 10 CFR 50, Appendix H.

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CO. M NT 8.

There are important aspects of the foreign experience, concerns.

and measuies, taken on this problem that need to be examined carefully.

j RESPONSE.

In the context of th'is answer, foreign. experience refers to overcooling transients at foreign reactors.

The concerns are, of course, the same:

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- stress; undercalculation of fast fluence; loss of f'racture toughness due to irradiation. The countermeasures are the same as being considered in the US:

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reduce fluence, reduce challenge; restore and maintain ductilit.Y.

The RES and NRR staffs have had regular contact with foreign ex;terts on pressure vessel integrity for years.

An extensive set of questions relating to pressure

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vessel themal shock has been sent to foreign regulatory authorities, and the staff has reviewed.the responses received to date. As an example of the type of discussions with foreign experts, the staff. met with representatives of the German Ministry of the Interior (BMI) and the Reactor Safety Comiss' ion (RSK) in Bethesda on September 29, 1981, on this subject.

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CdMMENT 9.

The in situ annealing capability requirement (Appendix G, IV:C).

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is not met by many.PWRs.

J RESPONSE 1-The 10 CFR 50 (Appendix G, IV-C) requires that the reactor pressure vessel be designed to. permit in-place annealing to recover material fracture. toughness properties if calculations predict that neutron irradiation may increase the RT t 200*F or more before the end of plint life. For licenses issued NDT after the effective date of this rule (August 16, 1973), licensees who predict NDT > 200*F have asserted that they have the capability to' anneal.

end-of-life RT Since the requirement was not imposed until August 16, 1973, some older.planti; have' never been asked'to respond to the requirement of Appendix G. 'The.;taff j

judgment however, is that the similarity in PWR designs, is such that there

.should be'no design aspects that would preclude in, situ RPV annealing in any plant.

The eight licensees

• have been specifically asked to provide the basis for dc.nonstrating that their plants meet the requirements in 10 CFR 50, Appendix G, -

IV-C.

When these responses are received in January 1982, the staff will have a better basis for assessing the capabilities of these PWRs for in, situ annealing.

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C0titENT 10.

The PT nr mentioned possible operational limit of 3DD*F is very N

question'ible.

RESPONSE.

The staff has considered the possibility of establishing an upper limit on RTHDT.f r Perating PWR pressure vessels.

Although the value of 300*F has been mentioned as an example, many more analyses of. transient frequency and severity 'are needed before a' limit can be established.

Furthermore, if an

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' upper limit on RT is to be imposed, it. may well turn out that the limit

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COMMENT 11.

Not all reasonable options appear to have been considered by the staff seriously.

The impact of possible shutdowns must be determined if it has not already.

4, RESPONSE.

The staff has concluded on the basis of current information that corrective acti,on' is not necessary at this time.

The actions the staff has initiated and the information we have requested from licensees and owners groups are

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~ ' intended to lead 'to decisions on any required corr'ective actions by the

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summer of 1982, or earl.ier if, indicated.

The option of plant shutdowns is I

always available should th'e staff judge such actions are necessary to protect the public health and safety.

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c0MMENT 12.

For all the above, and in relation to Item No.1 on this list, it is recomended that an ad hoc group, including experts out-side NRC, be charged to study this matter and report to the Comission with short-and long-term recomendations for dealing with it.

In the interim, the Comission may consider all reason-able options available to it to assure an acceptable level of protection of public health and safety, something the staff

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measures do not provide.

RESPONSE.

• Experts outside the NRC have.been assisting the staff on the subje d of pressure

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vessel integrity for a number of years. Major contributions have been made by experts from ORNL, NRL, US Naval Academy, Naval Ship.R&D Center, University of Maryland, BCL, HEDL, and NBS. We have also had critical contributions from researchers in Belgium, Germany, France and England.

Our long-standing practice of involving the most knowledgeable experts continues in this case.

Of course, it is the staff that must be responsible for ultima.te

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decisions.

We intend to discuss the. conclusions with outside experts and with the ACRS before final disposition by.the Comission.

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8 NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 o

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' MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation Edson G. Case, Deputy Director Office of Nuclear Reactor Regulation FROM:

Darrell G. Eisenhut, Director Division of Licensing

SUBJECT:

THERMAL SHOCK TO PWR REACTOR In response to E. Case's note to me dated April 16, 1981, we have coordinatei an interdivisional technical review of the reactor vessel fracture issue to determine if imediate licensing actions should be required.

Our preliminary review has concluded that although no imediate action is requ' ired for operating reactors, the staff should continue to evaluate this issue in the near future with the actions identified herein.

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arr efsn tor Division Licensing cc: NRR DIV DIRs l

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2 PRELIMINARY ASSESSMENT OF THERMAL SHOCK TO PWR REACTOR PRESSURE VESSELS 1

INTRODUCTION During the past few months the subject of reactor pressure vessel thermal shock has received increased attention by the NRC staff. Most recently, on March 31, 1981, NRC representatives met with the Pressurized Water Reactor (PWR) industry Regulatory Response Groups and the PWR reactor manufacturers.

In addition, concerns have been raised regarding the safety of operating reactors.

In order to determine :hether any immediate licensing action is necessary relative to the potential for thermal shocks in pressurized water reactor (PWR) pressure vessels, the staff has evaluated (1) the types of transients or accidents that could lead to overcooling of the reactor system; (2) experience to date with transients that have occurred in U.S. PWRs; (3) the probability that such overcooling events will occur; and (4) the capability of reactor vessels to withstand these transients.

Item 4 focused on the likelihood of a flaw existing in a reactor vessel RV),

the copper content of RV welds, and the extent of RV irradiation fluence).

BACKGROUND Severe reactor-system overcooling events which could be followed by repressurization of the RV can result from a variety of causes. These include instrumentation and control system malfunctions and postulated

acc~idents such as small-break loss-of-coolant accid.ents (LOCAs), main steamline breaks, or feedwater pipe breaks. Rapid cooling of the RV~

internal surface causes a temperature distribution across the RV wall.

.This temperature distribution results in thermal stress, with a maximum tensile stress at the inside surface of the vessel and a compressive stress at the outside surface. These stresses combine with the hoop stress caused by the internal pressure in the vessel. The magnitude of the thermal stress depends en the temperature differences across the RV wall.

As long as the fracture resistance of the RV material remains high, such transients will not cause failure. After,the fracture toughness of the vessel is reduced by neutron irradiation, severe thermal transients could cause fairly small flaws near the inner surface to initiate --

and result in -- significant cracking. The vessels of concern are those with a history of high radiation exposure, which are made of material that has a high sensitivity to radiation damage (such as those made with welds of high copper content).

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4 For failure to occur, a number of contributing factors must be present.

These factors are:

(1) a reactor vessel flaw of sufficient ' size to propagate, (2) high copper content, (3) a relatively high level of irradiation, (4) a severe overcooling transient with repressurization, and (5) a resulting crack of such size and location that the ability of the RV to maintain core cooling is affected.

EVALUATION The staff review of overcooling events anri their probabilities included a review of the Office of Reactor Research (RES) study on overcooling events at Babcock & Wilcox (B&W) plants (Ref.1 attached);

a survey of operating (experience on Westinghouse (W) and Combustion Engineering (CE) plants Ref. 2); a review of availabTe accident analysis in Final Safety Analysis Reports (FSARs) and in vendor topical reports; and a preliminary probabilistic analysis perfonned by the Division of Safety Technology (DST) (Ref. 3 attached). The preliminary results.of these evaluations indicate that there is a probability of about 10-3 per reactor year that a B&W-designed plant will experience a severe overcooling transient timilar or s'reater in magnitude to that experienced at Rancho Seco on March 20, 1978. This transient is the most severe overcooling transient experienced by any PWR in the U.S.

This probability of 10 S per reactor year includes contributions from steam generator control system malfunctions (the dominant contributor); small-break LOCAs; main steamline or feedwater line breaks; and complete loss of feedwater flow.

.The staff estimates that the probability of such an overcooling event in' CE or W-designed reactors is lower, perhaps :by an order magnitude, than for B&T designed reactors. This difference is based on design differences and on operating experience.

fairly high level (1500 psig to 2100 psig) pressure was maintained at a In the 1978 Rancho Seco transient, reactor throughout the cooldown.. The minimum temperature of the reactor coolant (280*F) during the transi' nt e

was high enough to maintain the material toughness of the reactor vessel.

Moreover, this evaluation leads the staff to believe that if this transient were to be repeated at Rancho Seco or any other B&W-designed facility within the next few years, the RV failure would still be unlikely.

Nonetheless, the possibility of vessel failure as a result of an over-i cooling event cannot be completely ruled out.

If an overcooling event such as that at Rancho Seco were to occur, based on the many factors pertinent to an analysis of vessel failure,the staff would expect much less than one failure in the current population of reactor vessels. Even for the vessel with the worst material properties, the staff would not expect a failure.

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The staff conclusion is supported by the analyses of the Rancho Seco event performed by the Oak Ridge National Laboratory (ORNL) (Ref. 4 attached). The ORNL analyses indicate that the threshold irradiation level for crack initiation (that is, small cracks growing to larger ones assuming conservative initial material properties such as RTNDT = 40*F and coppgr content of 0.35%) would be in the range of 0.75 x 1019 to

.l.7 x 1089 neutron /cm2 The highest fluence to date in a B&W-designed facility is less than half the minimum value listed above.

It would, therefore, be several years before any B&W-designed facility reached its threshold irradiation level.

Some reactor vessels in CE and W facilities have somewhat higher fluences; however, other mitigating factors -- such as lower values of initial RTNDT -- provide a significant margin to failure should an overcooling event similar to that at Rancho Seco occur.

CONCLUSIONS AND RECOMMENDATIONS As a result of its evaluations to date, the staff has concluded that the probability of a severe overcooling transient (similar in magnitude to the Rancho Seco event) is relatively low.

For B&W-designed reactors this probability is estimated to be about 10-3 per reactor per year, and for H-and CE-designed reactors, it is lower, perhaps by an order of magnitude.

In addition, the staff has concluded that, based on present irradiation levels at operating reactors, RV failure from such an event is unlikely. Accordingly, the staff believes that no immediate licensing actions are required on operating reactors; however, the staff recommends that the following actions be taken:

1.

Request industry representatives meet with the NRC staff in the near future to discuss:

a.

industry progress since the March 31, 1981 meeting b.

bases for ~ continued' safe operation c.

the letter of April 10, 1981 from D.L. Basdekas to Chairman Udall (Ref. 5 attached)

The Division of Licensing has requested such a meeting with the PWR vendors and Owners Groups. The meeting is scheduled for April 29, 1981.

2.

The staff should continue to refine its understanding of this safety concern. This continuing assessment, taken together with information being provided by industry Owners Groups (including the Owners Group Action Plan due May 15,1981) should permit the staff to define what l

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4-i actions the industry and the NRC must take to resolve this safety concern. The staff's efforts during this short term should' include, but not necessarily be limited to, the following areas:

Development of a better understanding of overcooling transients a.

and accidents.

Factors to be examined or addressed in this continuing evaluation would include:

(1) human factors considerations (2) refinements in the analysis of the probability of such events occurring, including considerations of overcooling events more severe than at Rancho Seco (3) an understanding of improvements in instrumentation and control systems implemented since the event at Rancho Seco and other overcooling events and the effects of these improvements on the probability of overcooling events.

Development of a better understanding of the potential for o.

and effects of RV thermal shock including:

(1) a categorization of the susceptibility of operating RVs to cracking as a result of rapid cooling, considering the combination of irradiation levels, vessel impurity content, and existing flaw sizes.

(2.) a sensitivity study of the effects of fluid mixing and the development of realistic models and assumptions, c.

An assessment of further requirements and of the overall ~ contribution to safety of potential improvenents.

d.

An overall. integrated assessment, and report of conclusions and recceniendations developed in connection with the above items.

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" Insights on Overcooling Transients in Plants with the B&W NSSS "

M. Taylor to S. Fabic, dated October 29, 1980.

2.

Nuclear Power Experience 1980, Bernard J. Verra, Publisher; Nuclear Power Experience, Inc., Encino,'CA.

3.

Frequency of Excessive Cooldown Events Challenging Vessel Integrity.

A. Thadani to G. Lainas, dated April 21, 1981.

Parametric Analysis of Rancho Seco Overcooling) Accidents, ORNL 4.

letter, R.D. Cheverton to M. Baginis (NRC, RFS, 3/3/81.

5.

Letter from D. Basdekas to The Honorable Morris K. Udall, dated April 10,1981.

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