ML20053A649
| ML20053A649 | |
| Person / Time | |
|---|---|
| Issue date: | 10/08/1981 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Palladino N NRC COMMISSION (OCM) |
| Shared Package | |
| ML20053A645 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8205270068 | |
| Download: ML20053A649 (15) | |
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e October 8,1981 i
MEMORANDUM FOR:
Chaiman Palladino
'FROM:
William J. Dircks Executive Director for Operations
SUBJECT:
DIFFERING PROFESSIONAL OPINION OF D. BASDEKAS ON
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PRESSURIZED THERMAL SHOCK At your meeting on September 17.,1981, with Mr. D. Basdekas, he discussed' with you a series of his concerns related to pr.essurized thermal shock of PWR pressure 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.
-p4;nscWrifinal.Dhh William J. Dir'cks-Executive Director fo'r' Operations
Enclosure:
Answers to Mr. Basdekas' Concerns cc:
Comissioner Gilinsky Comis'sioner Bradford Comissioner Ahearne Comissioner Roberts D. Basdekas, RES R. Minogue, RES H. Denton, NRR SECY OGC OPE 6
820527006f'
n-COMMENT 1.
"Some of the steps taken or proposed by the Staff ray be necessary, g
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 safety 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 information to be submitted in response to the letter to eight licensees and in the PWR Owners Groups' generic studies before deciding what further[regul' tory action is needed.
a 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-81-286A, which is the minutes of meetings held July 28-30,'1981, indicates that the staff position is for prompt, positive actions to prevent poten.ti. ally 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
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design basis envelope) are substantially greater than those given by the Staff.
The industry's " bounding" cases are based on design basis accidents.
RESPONSE.
The staff has made a study of PWR operational experience to look for severe overcooling' transients and their precursors.
In 1980, the RES staff reviewed the operating experience of B&W p ants an.d found th&t there had been a number l
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 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
-2 of 3 x 10 / 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 small transients per year in B&W plants as described in M. A. Taylor's memorandum of October 29, 1980.
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Since,these occurrences, operators have received special training in transient 1
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Babcock and Wilcox plants have added a back-up power supply to the nonnuclear instr'umentation bus, whose failure initiated the three transients above.
The NRR staff examined the impact of the improved power supply and l
operator training and suggested that these improvements might have reduced the s l
-3 probability to 10 / reactor year for an overcooling transient as severe as the t
Rancho Seco event for B&W plants.
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COMMENT 2 (continued)
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TheoperatingexperienceofCEandWestinghouseplantshasdisobeenexamined.
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There have been no events like the Rancho Seco transient, but there have been Theseareeventswhichtypicalkyledtosecondarysteamdump some precursors.
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 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, i
thereby causing the reactor co'olant temperature to drop by 107 F in 52 minutes.
Based on this e' perience,-the staff estimates the probability of a severe over-x 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 p>
iity.of a severe overcooling transient
-3 during which the primary pressure remains ~ high is about 10 / reactor year for B&W
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plants and 10 / reactor year for CE and Westinghouse plants.
There,may be a factor of 10 uncertainty associated with these estimates.
1 Although some of the PWR Owners Groups' calculations have been based on design basis accidents, we have asked for analyses of a wider range of overcooling tra'nsients.
These analyses should cover transients resulting from multiple failures or operator errors.
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COMMENT 3.
The uncertainties in 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 RT s
riDT.
RESPONSE.
d The sta.ff is aware that there are uncertainties in estimating P,TNDT. principally from the following sources:
(1) uncertainties in the in.itial RTNDTI
,(2) uncertainties in copper content of the weld metal; (3) uncertainties in' vessel fluence estimates; (4) uncertainties in irradiation temperature; (5) uncertainties in the. Regulatory Guide 1.99 curves relating RT shift to NDT fluence and copper content of welds In calculating RTNDT, the uncertainties have been accounted for by using con-servative estimates of the factors that are used in calculating RT This'is NDT.
especially true of the. initial' RTNDT.and the prediction of RTriDT: shift based on Regulatory Guide 1.99.
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Vessel ' fluence calculations will be carried out with a well benchmarked and calibrated neutron transport code to yield fluence accuracy not worse than +20%,
and'the remaining uncertainty will have only a small effect on RTNDT.
The coppar content of the weld metal is believed to be k'nchi to within +0.03% copper (based
'on measured data) and this uncertainty will have only a small effect on RTNDT' There is a substantial amount of information from measured RT shift from sur-NDT veillance tests of, welds which shows that the curves in Regulatory Guide 1.99 for predicting RT shifts are conservative for shifts above 150'F.
Based on this riDT information, the staff believes that there are substantial conservatisms in the
. predicted values for RT f r shifts above l50*F.
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C0tNENT 4.
The scenario related thermohydrodynamic 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 by 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 the probability of the transients occurring and their severity of overcooling with appropriate consideration of 'the uncertainties associated with such
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estimates.
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- The staff uses several system analysis models to predict the temperature and pressure histories of a primary cooling, system in a transient state.
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C0!HENT 5.
"A request for design information on. control systems to selected utilities was blocked by HRR management."
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RESPONSE.
Mr. Basdekas recomended that the following infomation be requested in the August 21, 1981, letter to eight licensees:
" design descriptions, including system functional block ~ diagrams and single-line schematics of the plant's control systems, and the associated P&ID's and system flow diagrams.'"
Mr. Basdekas believed that this information was required for the staff to perfonn a complete an'alysis of the potential fo'r pressurized themal shock 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.
l' 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 normally reviewed in the licensing process, was not appropriate'at this time..in the context of the letters 'on pressurized thermal shock.
(The letters do request the licensees to " provide any failure modes and effects analyses of control systems currently available or reference any such gnelyses already
. submitted..."")~ 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 shock.
l As to obtaining this type of information for the research contract to investigate the safety significance of control system. fai. lures, a different, more appropriate method 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, 5
I and to the' development of criteria for plant-specific reviews of control systems.
Mr. Bas'dekas' comments have been, and will continue to be, requested and con-sidered in the development of the plan.
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COMMENT 6.
The signif.icance'of the synergisti,e effect of Nickel on the rate of increase of RT is not accounted for in Regulatory Guide 1.99.
NDT RESPONSE.
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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.
Thus, the 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 veillance 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 materials surveillance program was that the welds were not going to be the 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 RTH0T 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 o'n RTNDT shifts and to. provide data' to benchmark heutrcn fluence calculations.
It is not necessary th'at the surveillance capsules be placed at the weld locations,
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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 actua1Ty increase the rate of ' embrittlement of the weld. because there
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is sa peak in the fast neutron flux just behind the capsule.on the inside wall of the vessel.
With respect to timing of surveillance capsule examination, a capsule is l
typically pulled at the first or second refu,eling.
Rules for withdrawal schedules are given in 10 CFR 50, Appendix H.
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COMMENT 8.
There are important aspects of the foreign experience, concerns.
and measures; taken on this problem 'that need to be examined carefully.
RESPONSE.
In the context of th'is answer, foreign. experience refers to overcooling transients at foreign reactors.
The concerns are, of course, the same:
high thermal stress; undercalculation of fast fluence; loss of f'racture toughness due to irradiation.
The countermeasures are the same as being considered in the US:
reduce fluence, reduce challenge; restore and maintain ductility.
The RES and NRR staffs have had regular contact with foreign experts on pressure vessel integrity for years.
An extensive set of questions relating to pressure vessel thermal 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 GermanMinistryoftheInterior(BMI)andtheReactorSafetyCommiss~ ion (RSK) l in Bethesda on Septembe.r 29, 1981, on this subject.
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COMMENT 9.
The in situ annealing capability requirement (Appendix G, IV-C),
is not met by many.PWRs.
RESPONSE.
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 to 200*F or more before the end of plint life.
For licenses issued riDT after the effective date of this rule (August 16, 1973), licensees who predict
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end-of-life RTt1DT > 200*F have asserted that they have the capability to anneal.
Since the requirement was not imposed until August 16, 1973, some older plants have never been asked'to respond to the requirement of Appendix G. 'The staff 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 demonstrating that their plants meet the requirements in 10 CFR 50, Appendix G, -
IWC.
When these responses are received in January 1982, the staff will have a better basis for assessing the capabilities of these PWRs for jn_ situ annealing.
- 1MI-1, Ft. Calhoun, Robinson 2, San Onofre 1, Maine Yankee, Oconee 1, Turkey Point 4. Calvert Cliffs 1
COMt1ENT 10.
The RTNDT mentioned possible operational limit of 300*F is very question &ble.
RESPONSE.
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The staff has considered the possibility of estatriishing an upper limit on RT f r operating PWR pressure vessels.
Although the value of 300*F has NDT been mentioned as an example, many more analyses of. transient frequency and severity are needed before a' limit can be established.
Furthermore, if an
' upper limit on RT is to be imposed it. may well turn out that the limit NDT will have to be a plant-specific limit.
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COMMENT 11.
Not all reasonable options appear to have been considered by j
the staff seriously.
The impact of possible shutdowns must be determined if it has not already.
RESPONSE.
The staff has concluded on the basis of current information that corrective action' is not necessary at this time.
The actions the staff has initiated and the information we have requested from licensees and owners groups are
' intended to lead 'to decisions on any required corr' ctive actions by the e
summer of 1982, or earl.ier if. indicated.
The option of plant shutdowns is always available should th'e staff judge such actions are necessary to protect the public health and safety.
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COMMENT 12.
For all the above, and in relation to Item No.1 on this list, it is recommended that an ad hoc group, including experts out-side NRC, be charged to study this matter and report to the Commission with short-and long-term recommendations for dealing with it.
In the interim, the Commission may consider all reason-able options available to it to assure an acceptable level of protection of public health and safety, something the staff measures do not provide.
RESPONSE.
Experts outside the NRC have.been assisting the staff on the subject of pressure 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 continuet in this case.
Of course, it is the staff that must be responsible for ultimate
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decisions.
We intend to discuss the conclusions with outside experts and with the ACRS before final disposition by.the Commission.
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UNsTED STATES E
NUCLEAR REGULATORY COMMISSION o
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APR 2 81981
- 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 l
coordinated an interdivisional technical review of the reactor vessel fracture issue to determine if immediate licensing actions should be required.
Our preliminary review has concluded that although no immediate action is required for operating reactors, the staff should continue to evaluate this issue in the near future with the actions identified herein.
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tor arr Division Licensing cc: NRR DIV DIRs l
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