ML20069L037
| ML20069L037 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 11/12/1982 |
| From: | Cochran T National Resources Defense Council |
| To: | |
| Shared Package | |
| ML20069L032 | List: |
| References | |
| NUDOCS 8211160360 | |
| Download: ML20069L037 (61) | |
Text
s to , j 6
p ..
Y l Q? -6' i
g- gg@4 4td ;
BEFORE THE
.S y UNITED STATES NUCLEAR REGULA T'$SION ATOMIC SAFETY AND LICENS RD'~
CD'
)
In the Matter of )
)
UNITED STATES DEPARTMENT OF ENERGY )
PROJECT MANAGEMENT CORPORATION ) Docket No. 50-537 TENNESSEE VALLEY AUTHORITY )
)
(Clinch River Breeder Reactor Plant) )
)
TESTIMONY OF DR. THOMAS B. COCHRAN PART IV AS SUPPLEMENTED BY NEW INFORMATION IN CRBR FINAL ENVIRONMENTAL IMPACT STATEMENT SUPPLEMENT (Intervenors' Contentions 1, 2, and 3)
DATED: November 12, 1982
, 8211160360 821112 l PDR T ADOCK 05000537 l
Q.1: Please identify yourself and state your qualifications to present this testimony.
A.2: My name'is Thomas B. Cochran. I reside at 4836 North 30th Street, Arlington, Virginia 22207. I am a Senior Staff Scientist at Natural Resources Defense Council, Inc. My i background and qualifications to present this testimony are presented in previous testimony in this proceeding.
I (Tr. 2870-71, Cochran.)
I Q.2: What is the subject matter of the present testimony?
A.2: Part IV of my testimony deals with the potential for severe accidents at CRBR and the adequacy of Applicants' l
and Staff's analyses of those accidents. These are matters that are raised in Intervenors' Contentions 1, 2, and 3. For purposes of this phase of the proceeding, i those Contentions read as follows:
a) Neither Applicants nor Staff have demonstrated through reliable data that the probability of anticipated transients without scram or other CDA initiators is sufficiently low to enable CDAs to be excluded from the envelope of DBAs.
b) [ deferred]
- 2. The analyses of CDAs and their consequences by Applicants and Staff are inadequate for purposes of licensing the CRBR, performing the NEPA cost / benefit analysis, or demonstrating that the radiological source term for CRBRP would result in potential hazards not cxceeded by those from any
l accident considered credible, as required by 10 CFR $100.11(a).
a) The radiological source term analysis used in CRBRP site suitability should be derived through a mechanistic analysis.
Neither Applicants nor Staff haye based the radiological source term on such an analysis.
b) The radiological source term analysis should be based on the assumption that CDAs (failure to scram with substantial core disruption) are credible accidents within the DBA envelope, should place an upper bound on the explosive potential of a CDA, and should then derive a conservative estimate of the fission product release from such an accident.
Neither Applicants nor Staff have performed such an analysis.
c) The radiological source term analysis has not adequately considered either the release of fission products and core materials, e.g., halogons, iodine, and plutonium, or the environmental conditions in the reactor containment building created by the release of substantial quantities of sodium. Neither Applicants nor Staff have established the maximum credible sodium release following a CDA or included the environmental conditions caused by such a sodium release as part of l the radiological socrce term pathway analysis, d) Neither Applicants nor Staff have demonstrated that the design of the containment is adequate to reduce calc'ulated offsite doses to an acceptable level, e) As set forth in Contention 8(d), neither Applicants nor Staff have adequately calculated the guideline values for radiation doses from postulated CRBRP releases.
_4_
f) Applicants have not established that the l computer models (including computer codes) referenced in Applicants' CDA safety analysis reports, including the PSAR, and referenced in the Staff CDA safety analyses are valid. The models and computer codes used in the PSAR and the Staff safety analyses of CDAs and their i consequences have not been adequately documented, verified, or validated by comparison with applicable experimental data. Applicants' and Staff's safety analyses do not establish that the models accurately represent the physical phenomena and principles that control the response of CRBR to CDAs.
.g) Neither Applicants nor Staff have ,
established that the input data and assumptions for the computer models and
- codes are adequately documented or i verified.
h) Since neither Applicants nor Staff have established that the models, computer codes, input data, and assumptions are adequately documented, verified, and validated, they have also been unable to establish the energetics of a CDA and thus have also not established the adequacy of the containment of the source term for post accident radiological analysis.
- 3. Neither Applicants nor Staff have given j
suf ficient attention to CRBR accidents other than the DBAs for the following reasons:
a) [ deferred]
b) Neither Applicants' nor Staff's analyses of potential accident initiators, sequences, and events are sufficiently comprehensive to assure that analysis of the DBAs will envelop the entire spectrum of credible accident initiators, '
sequences, and events. ,
c) Accidents associated with core meltthrough following loss of core geometry and sodium-concrete interactions have not been adequately analyzed.
l l
l d) Neither Applicants nor Staff have adequately identified and analyzed the ways in which human error can initiate, exacerbate, or interfere with the mitigation of CRBR accidents.
The accident discussion at this phase focuses on Appendix J of the Final Supplement to the FES, NUREG-0139, Supplement No. 1 (henceforth "FSFES").
Q.3: Dr. Cochran, are you familiar with Staff's NEPA analysis of the risks of potential accidents associated with the CRBR?
A.3: Yes.
Q.4: Where is this analysis set forth?
A.4: Primarily in Chapter 7 and Appendix J of the FSFES, although some paragraphs from Chapter 7 of the 1977 FES have been retained, including the conclusions in $7.1.4.
Q.5: Do you have general criticisms of Appendix J?
A.5: Yes. The methodology in Appendix J is crude by today's standards, and the assumptions behind it (and the input data) are not supported by any substantive analysis.
While it presents estimates of the absolute probability of CRBR accidents, these estimates are backed up by no calculations and no event tree / fault tree analyses as one finds in risk assessment analyses such as the Reactor
. Safety Study (WASH-1400) and CRBRP-1. No operating data are offered in support of its conclusions, and there are no quantified estimates of the uncertainty associated with the probability estimates. It must be remembered that WASH-1400, which contained an incomparably more detailed analysis of accident probabilities for two actual LWRs (and which is, incidentally, the direct progenitor.of virtually all nuclear risk assessment work) was severely criticized for making unsupported assumptions, for failing to properly assess uncertainty and for its factual
- inscrutability. For these reasons, the NRC ultimately repudiated NASH-1400's absolute probability predictions.
Yet, compared to Appendix J, WASH-1400 was a model of scientific analysis. Appendix J.is not even supported by a plant-specific risk assessment. Its assumptions are not
! just unsupported by rigorous analysis; for the most part, they are not even presented for evaluation. If WASH-1400's probability estimates were unreliable, as the Commission correctly concluded, then the probability
- estimates in Appendix J are far more so. There is no l
l reason to accept these on faith, and very little beyond faith is offered.
Moreover, the Staff attempt to quantitatively assess the uncertainty associated with the estimates for various quantitative accident probabilities and consequences l _ _ _ _ , _ _ _ - -
presented in Appendix J is a one-sentence conclusory statement (FSFES, p. J-24) which is unsupported in the document by rigorous analysis. Probably the most serious criticism of WASH-1400 from the scientific community was its failure to assess or properly acknowledge the very large uncertainties attached to absolute probability predictions. Those uncertainties, which have been estimated to be as large as a factor of 100 in some cases, j must be much greater for predicting CRBR accident probabilities, since the body of relevant operating data for LMFBRs is far less than for LWRs and since, for lack
! of a plant-specific assessment, the report is almost i
totally based on conclusory statements that can most charitably be characterized as " engineering judgment."
j Without some reasonable and scrutable assessment of the uncertainties inherent in these predictions, they are simply arbitrary and meaningless.
Q.6: Do you know whether the NRC Staff performed any calculations, reviewed operating data for other facilities, or did any plant-specific assessment of the l
reliability of the CRBR systems to back up the probability estimates presented in Appendix J?
A.6: According to the NRC Staff, with only three exceptions (WASH-1400 for PWR auxiliary feedwater reliability and the
,--- ,,m-,, ,---e -=--7-,y.,3w-~--,- ,.,..m- - . - - - - -- a,-- ,
probability of loss of offsite power, and NUREG-0460 for the frequency of anticipated transients without scram for typical LWRs), they did not. NRDC asked the Staff in discovery to identify the documents relied upon for each of the principal probability assessments in Appendix J.
(See Staff Response to NRDC's 27th Set of Interrogatories, Oct. 1, 1982, pp. 53-70.) In almost every case, the Staff responded under oath that it relied on nct " specific" documents for any of the conclusions presented, instead relying generally on the " cumulative knowledge" of the Staff and its consultants in general, or a similar response. While " engineering judgment" or " cumulative knowledge" is valuable for many purposes, it is not sufficient to support predictions of the probability of serious accidents in a plant as complex and untested as the CRBR.
Q.7: Have you been limited in your ability to independently assess the probability of accidents beyond the design basis for CRBR?
A.7: Yes, independent assessment has been greatly hindered.
The probability of a catastrophic accident in any plant is ,
a function of the plant design, the potential for 1 I
equipment malfunction and human error, and the reliability of its many complex systems and components. The CRBR is
the first plant of its kind. Applicants have done much work in assessing the reliability of the CRBR design, primarily as part of Applicants' Reliability Program (see PSAR, Appendix C). The document known as CRBRP-1 is another prominent example, Applicants have underway a comprehensive probabilistic risk assessment (PRA) of the CRBR and preliminary results have been presented to the ACRS and the Staff (cf., Letter from John R. Longenecker, CRBR Project to Paul S. Check, USNRC, June 21, 1982, subj:
Probabilistic Risk Assessment (PRA) Program Plan).
However, the scope of this LWA-1 proceeding has been limited to exclude inquiry into what are termed the
" details" of the CRBR design. CRBRP-1 has been expressly excluded from consideration. In my judgment, no reliable estimate of CRBR accident probabilities can be made within i
t the present scope of the LWA-1 proceeding and without reviewing the CRBR design in some detail. This has not been possible at this stage.
Q.8: Do you believe that the analysis in Appendix J is realistic and adequate to support Staff's conclusions
! regarding consequences of Class 9 accidents, namely "that CRBR accident risks would not be significantly different
! from those of current LWRs..." and that "the accident risks at CRBR can be made acceptably low." (Appendix J, p.
. _10 J-25)? l A.8: No.
Q.9: Please proceed to discuss some of the specific probability estimates. To begin, what frequency of occurrence did the NRC staff assign to core degradation due to LOHS (loss of heat sink) events for CRBR and what rationale did the staff give for its estimate?
A.9: Staff assigned a frequency of core degradation due to LOHS events of less than 10-4 per reactor year (i.e., one chance in 10,000 per reactor year). Staff cited three principal factors for this result:
- 1. A " general consideration of typical achievable PWR auxiliary feedwater system reliabilities;"
- 2. The " potential for common cause failures;"
- 3. The potential for achieving "high reliability in final design and operation through an effective.
reliability program." (FSFES, pp. J-3, -4.)
While the three factors above are all listed as the bases for the estimated LOHS probability, only the first -- PWR auxiliary feedwater system reliability -- serves as the basis for Staff's quantified estimate. The role the other two factors play in the choice of the 10-4/ year estimate is discussed only in the most general qualitative terms, i e.g., "... unavailability estimates for ... heat removal i
_ _ _ _ - _ . . - _ - _ - _ - _ . . . . . . . . - - - -~ _ - - _ _ _ _ . . -
, systems have been set high enough to include allowance for potential common mode failures" (Appendix J, p. J-22).
- The choice of auxiliary feedwater system failure as the controlling failure mode is not justified. In other I words, there is no reason to believe that failures in l
systems other than auxiliary feedwater may not contribute significantly to the LOHS probability. A fault tree analysis is necessary to justify limiting the discussion
- to auxiliary feedwater reliability.
1 In order to illustrate the complexity of this issue, consider the generalized fault model for the shutdown heat renoval system for CRBR taken from CRBRP-1, Vol. 2, Appendix II, p. 2-14 to 2-22 (attached to my testimony as Exhibit 1). This fault tree, which is developed to the t
system (or subsystem) level rather than the more detailed i
l component level as in the WASH-1400 case, can be considered applicable to a reactor of the general size and type as CRBR. Clearly, it takes a leap of faith to conclude that the failure rate of the auxiliary feedwater system controls the overall frequency of core degradation due to LOHS events.
I Q.10: Setting aside your view that there is no basis for concluding that the failure rate of the auxiliary feedwater system is controlling, do you agree with the i
Staff's estimate of the feedwater system reliability?
Explain your answer.
A.lO: First, I should note that Staff claims that its estimate of the probability of LOHS events was based on independent analyses, primarily by William Morris of the Staff and Staff consultant Edward Rumble of Science Applications Inc., (SAI), each using a different base of information (Deposition of William Morris, Oct. 12, 1982, pp. 24-25).
Dr. Morris claimed his estimate is based on the reliability of auxiliary feedwater systems in PWRs over the years as documented in the Standard Review Plan for LWR feedwater systems (Morris, Deposition of Oct. 12, 1982, pp. 23-24).
Mr. Rumble also claimed his estimate was based on reliability studies of PWR auxiliary heat removal systems, the Accident Delineation Studies (Phases 1 and 2) (NUREG-CR-1407 is Phase 1) prepared by Sandia for NRC-NRR, and the study CRBRP-1 (which is beyond the scope of the LWA-1 proceeding). Mr. Rumble said these estimates were what he believed should be achievable, not necessarily what has been achieved to date (E.R. Rumble, private telephone communication, July 27, 1982, as noted in T.B. Cochran Memo to Files, July 27, 1982).
( I do not agree with Staff's estimate or Staff's l
underlying analysis. First, LOHS fault trees for CRBR I
l developed in CRBRP-1 dif fer f rom those of a PWR as developed in WASH-1400, and consequently there is no obvious correlation between PWR system reliabilities and the core degradation frequency due to LOHS accident scenarios in CRBR. This can be seen by comparing the generalized fault models for CRBR shutdowr heat removal (see CRBRP-1, Vol. 2, Appendix II) with the fault models for a PWR (see WASH-1400, App. II).
Staff claims that its estimate of 10-4/ year is based on " typical achievable PWR auxiliary feedwater system i reliabilities" (Appendix J, p. J-4). If this is so, there must be wide variations in achievable feedwater system reliability. For example, the RSSMAP (Reactor Safety 1
Study Methodolcgy Applications Program) report for Calvert Cliffs (NUREG/CR-1569) concluded that the probability of l
r
! core melt for Calvert Cliffs was 1 chance in 2400 per i
reactor year, largely due to unreliabilities in the i
auxiliary feedwater system and failure of backup heat removal methods. This result is a factor of 4 larger than the Staff's alleged " upper bound" result for CRBR. No justification has been presented for concluding that he CRBR auxiliary feedwater system will be more reliable than Calvert Cliffs by at least a factor of four. Furthertore, there is a serious question about the comparability of PWR l operating data in this area to the CRBR. It should be l
l
. l noted in this connection that the authors of the Applicants' risk assessment work felt that the WASH-1400
. data could not be applied to the question of unavailability of decay heat removal systems for CRBR.
Instead, a fault tree analysis was conducted to determine the system availability.- (CRBRP-1, Vol. 2, at III-3.)
There is no basis for concluding that CRBR's auxiliary feedwater system will be " typical" in its reliability. The conservative assumption to make at this juncture might be to assume that CRBR's auxiliary feedwater system will be no better than Calvert Cliffs' l system. Moreover, since CRBR's Decay Heat Removal System l
(DHRS) is dependent upon AC electrical power, it cannot be assumed to be significantly more reliable than PWR DHRSs; according to Staff (FSFES, pp. J-3,4), a principal unreliability in PWR decay heat removal systems is not in system failures per se but in loss of offsite and onsite AC power. Thus, if Staff is correct, the ability of the CRBR DHRS to operate at " normal" temperature and pressure (whereas PWR DHRSs can operate only at low pressure) should not have a major impact on overall risk.
1 Q.ll: Are there other CRBR heat removal systems that are important in terms of the comparability between the
frequencies of core degradation in CRBR and PWRs due to loss of heat sink (LOHS)?
A.ll: What I noted above was that one cannot tell the degree of contribution that various compor,ent failures have on the overall failure rate without a detailed fault treo analysis. However, it is evident that there are other CRBR heat removal components whose failure rates are not e
necessarily comparable to PWR systems. The steam generators are an example. There is no discussion whatever in Appendix J of the contribution of steam generator failure to the overall risk of LOHS, nor of the possible mechanisms or modes of failure considered.
Unlike an LWR, the steam generators in an LMFBR, such as 4
CRBR, represent a location where significant amounts of sodium and water are in close proximity. CRBR event sequences can be postulated, e.g., propagation of steam generator tube failures, where sufficient water and sodium can be brought together in such a manner as to create a sodium-water reaction coupled with a hydrogen reaction, resulting in loss of the shutdown heat removal function (see generally CRBRP-1, Appendix VIII).
The General Accounting Office in a recent letter to Congress was highly critical of DOE's failure to conduct complete and thorough tests of the steam generators to be used in the CRBR, in spite of the fact that steam
1 I
i
- 1 l
generators for LMFBRs have had a history of serious technical problems and the fact that development and demonstration of reliable steam generators have been and still are one of the most significant technical problems facing the CRBR project. { Letter from Charles A. Bowsher, Comptroller General, to Congressman John D. Dingell, May 25, 1982, GAO/EMD-82-75, attached as Exhibit 2).
In sum, because of the inherent differences in the shutdown heat removal systems, e.g., steam generators, between PWRs and LMFBRs introduced by the use of sodium coolant in an LMFBR, it does not directly follow that the frequency of core degradation due to LOHS events in PWRs is directly transferrable to LMFBRs.
Q.12: How did Staff treat the contribution of pipe rupture failure as a contributor to the core disruptive frequency?
A.12: The frequency of large pipe breaks (loss-of-coolant accidents, or "LOCAs") is pivotal to an assessment of the risk of accidents at CRBR or a reactor of the general size and type. A large pipe break in the cold leg (and perhaps the hot leg, as well) would likely lead to core disruption and serious offsite consequences. It is an important determinant in whether the CRBR site is suitable. Staff states:
Eecause of the high boiling peint of sodium, the CRBRP primary coolant system would i
l , - _ _ - - - - - _
_17_
operate at significantly lower pressures than LWR primary coolant systems. This reduces the frequency of large ruptures in the i primary coolant system. To further ensure that large breaks cannot occur and cause core damage, implementation of preservice and inservice inspection of the primary coolant boundary and a leak detection system will be required. In addition, a guard vessel will be included to prevent unacceptable leakage from large portions of the primary coolant system. For these reasons LOCAsLare not considered credible (i.e., design-basis) events at CRBRP. The frequency assumed for LOHS adequately bounds the LOCA contributions to core disruption frequency.
(FSFES, p. J.4, emphasis supplied.) When asked to identify every document relied upon by Staff for its conclusicn above that "LOCAs are not considered credible
... events at CRBRP," Staff stated:
The cumulative knowledge of the Staff and its consultants rather than a specific document were relied upon by the Staff for its conclusions in Appendix J regarding whether LOCAs are DBAs for CRBR. This issue was also discussed in the SSR and the Staff's prefiled testimony for the site suitability hearings.
(Staff Response to Interrogatory 33, 27th Set, Oct. 1, 1982, p. 58.) I take this answer to mean that Staff has no documentation or written analysis demonstrating that a LOCA is a low probability event for the CRBR.
In the 1982 SSR, Staff stated:
It is the staff's opinion, based on the following considerations, that the heat transport system can be designed for a high level of integrity and for continued assurance of this integrity throughout the operating history of the plant. The specifications include stringent
. nondestructive examination requirements. The material is characterized by high fracture toughness and corresponding large critical flaw size, a negligible growth rate of postulated defects and the probability of throughwall growth rather than elongation of defects. The system has low stored energy and is monitored by sensitive leak detection instruments. The staff preliminary conclusion is that double ended rupture of the CRBRP primary cold leg piping (an event that could potentially lead to a CDA unless otherwise mitigated) need not be considered a design basis event. This conclusion is conditioned on an acceptable preservice and inservice inspection program, a material surveillance program, continued research and development verifying material degradation processes, and verification of leak detection i system performance. Ths staff considers it feasible to implement programs to satisfy these requirements. The staff intends to continue its review of the sodium cold leg piping to insure that the issues are resolved properly.
Because of its higher operating temperature, the same conclusions have not yet been i reached concerning the hot leg piping (995*
- vs 730' F). The staff has studies underway to evaluate the potential for and consequences of hot leg piping ruptures.
Preliminary results obtained so far indicate that this event has more benign consequences with respect to core thermal conditions than the cold leg rupture. For example, a hot leg pipe rupture followed by a scram and a pump trip and normal flow coastdown does not appear to lead to boiling in the core.
Analyses of this event are continuing and the results will be factored into any future requirements to assure that hot leg pipe ruptures, like the cold leg case, need not be considered as events that would lead to a CDA.
(1982 SSR, pp. II-8 to II-9.)
i f
l
\
l_. _ . . _ __ . _ _ _ . . . _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ ,__ _ _ , _ _ _ _ _ _ _ . . _ .
_19_
Q.13: Do you agree with Staff's assessment, as stated above, of the pipe rupture probability, and, if not, what is the basis for your disagreement?
A.13: I disagree with the Staff assessment. In this regard, it is extremely instructive to compare Staff's analysis with the analyses conducted by D. O. Harris of the Palo Alto office of Science Applications, Inc. (SAI), for the CRBR Project office in the 1977-78 period. SAI was a consultant to the CRBR Project in the development and application of the fault tree / event tree methodology for assessing the reliability of CRBR systems as published in CRBRP-1, March 1977, and continued work for DOE on a variety of CRBR risk assessment issues through early 1979 and perhaps beyond. Staff consultant Rumble is a Vice President of SAI at the same Palo Alto office and has stated to me that he relied in part on CRBRP-1 for his assessment of the core degradation frequency which appears in Appendix J of the DSFES (and therefore the FSFES).
I have not been permitted to address that work in I
this hearing because, of course, it involves the " details" of the CRBR design. Only the most general conclusions have been presented in Appendix J.
In what appears to be a final risk assessment task report, obtained by NRDC under the Freedom of Information Act, D.O. Harris of the SAI Palo Alto office summarized
the result of SAI's assessment of the CRBR pipe rupture probability (Harris, D.O., " Relative Pipe Rupture Probability for the Primary Heat Transport System of CRBRP," Nov. 13, 1978, attached as Exhibit 3 to this testimony).
Harris's analysis appears to be based on the assumption that the primary large pipe failure mechanism is fatigue crack growth due to cyclic stress imposed on defects introduced prior to service, hence other potential sources of failure were not considered. In this respect, Harris's analysis appears similar to that conducted in CRBRP-1 (Vol. 2, App. III, p. III-ll2). In the Harris analysis, calculated relative probability of pipe rupture in CRBR compared to that of PWRs was primarily a function of a) probability of having a defect, which in turn was a function of the number and characteristics of the weld joints, Because the appropriate normalization was not known, separate calculations were made using weld volume, weld area, and weld length as the basis of normalization, b) the initial crack size and depth distribution. Because the appropriate crack distribution was not known, separate calculations were made using four crack distribution expressions.
The differences between Staff's assertions and the SAI anlysis are important. Staff's conclusion that the CRBR cold leg pipe break is incredible (i.e., beyond the design basis) is based in part on the fact that there will be
l preservice and inservice inspection programs. Such programs have been in place for light water reactors for
- some time. The SAI analysis assumed equivalent effectiveness for the inspection programs for both CRBR and PWR in each calculation of the relative probability of pipe break failure of the two. This is the approriate way to treat the subject. Staff offers no evidence that any relative difference in the CRBR and PWR surveillance programs would have a significant effect on the crack distributions in CRBR piping relative to that in PWRs.
SAI found that "[w]ith the present state of I
knowledge, it is not possible to ascertain the controlling a parameters" that govern the relative CRBR/PWR pipe break frequency. SAI found a wide range of values varying from 0.0186 to 11.62 (i.e., three orders of magnitude) in the ratio of CRBR pipe failure to PWR pipe failure depending on the assumptions made. In fully 13 out of 36 cases (36%) analyzed, the probability of CRBR pipe failure exceeded the probability of PWR pipe failure.
( Furthermore, the probability of PWR failure was found to be strongly design dependent, varying by as much as a factor of 14 among the three PWRs analyzed.
i.
I . ..
+
n In conclusion, the Staff analysis of the pipe break '
probability is nothing more than a series of unsupported assumptions that appear to be in conflict with a more rigorous CRBR-specific analysis. The SAI analysis does not support the conclusion that a LOCA is " incredible" for the CRBR. Moreover, as evidenced by the SAI analysis, i.e., the lack of understanding of the controlling factors, the fact that the CRBR pipe break frequency may be as much as 12 times higher than that in a PWR, and the fact that the frequency is a strong function of the nurber I and characteristics of the pipe welds, which are desigrd dependent, the Staff' conclusion that a cold (or hot) leg pipe rupture is not credible in a reactor of the general size and type of CR3Rtis not substantiated by rigorous analysis. It should be rejected.
Q.14: Do you agree with Staff's analysis of common mode failures?
A.14: The one sentence' devoted to common cause failure hardly qualifies as "an analysis." LOHS failures due to common '
causes are but one manifestation of a larger class of failures that fall under the general category of. systems interaction (SI). Systems interaction is presently the subject of two unresolved safety issues (USIs) -- namely A-17, " Systems Interaction in Nuclear Power Plants," and s
1
. A-47, " Safety Implications of Control Systems." The NRC has sponsored four separate evaluations of systems interaction in an attempt to develop an acceptable methodology for reviewing final designs for adverse
! systems interactions. These four studies ares l 1. NUREG/CR-1321, " Final Report -- Phase I Systems Interaction Methodology Applications Program,"
G. Boyd, et al., Sandia National Laboratories, April 1980.
- 2. NURECl/CR-1896, " Review of Systems Interaction Methodologies " P. Cybulskis, et al., Battelle Columbus Laboratories, January 1981.
- 3. NUREG/CR-1859, " Systems Interaction: State-of-the-
-l Art Review and Methods Evaluation," J.J. Lim, et al., Lawrence Livermore Laboratory, January 1981.
[' 4. NUREG/CR-1901, " Review and Evaluation of System Interactions Methods," A.J. Buslik, et al.,
Brookhaven National Laboratory, April 1981.
The NRC Staff's evaluation of these four reports is
% 3 summarized in the periodic "TMI Action Plan Tracking System Report" as follows:
l State-of-the-art review concluded that no l single method presently exists in a form that can be used to perform an adequate review for i adverse SI.
Thus, it can be fairly concluded that an adequate systems interaction review of CRBR could not have been
'; conducted. Moreover, such a review requires a final q
, design, which is not yet available for CRBR. It should be noted that three of the SI reviews above attempted
.' unsuccessfully to evaluate SI in actual past events 0
1 0
. involving SI, including the Browns Ferry fire in 1975, the TMI-2 accident in 1979, the Browns Ferry partial scram failure in 1980, the pressurizer relief valve failure at Beznau in 1974, the temporary loss of decay heat removal at Davis-Besse in 1980, the loss of DC control power and diesel generator fire at Zion in 1976, and the Crystal River LOCA in 1980.
In addition, common mode failures and other forms of systems interaction involve more than just hardware failures. Also involved are external events (such as seismic events and hurricanes), human error (including errors of omission and commission, and including not only operations but design, fabrication, installation, maintenance, and testing), and design flaws. The design of the control room and any auxiliary control panels or remote shutdown locations, and actual operating, emergency, maintenance, and test procedures can also
! impact on systems interactions.
In sum, the effect of potential common mode failures on CRBR accident probabilities involves complex issues that the technical community has been wrestling with for l
years, thus far without notable success. There is no substantive basis for Staff's broad-brush assertion that
"[t]he foregoing estimates of frequencies and risk l
associated with CRBR have included allowances for
,=_ _
1 i
uncertainties. For example, unavailability estimates for '
l shutdown and heat removal systems have been set high enough to include allowances for potential common cause failures." (Appendix J, p. J-22.)
0.15: In estimating the quantitative probability of CRBR accidents, can credit be assigned for an " effective reliability program"?
A.15: In my opinion, it is not possible to assign any particular value to the level of " reliability" to be achieved. No CRBR-specific program has been presented by Staff; no i
precedent is cited for an " effective reliability program" for any other plant and no criteria are presented.
I such assertions about the achievability of Finally, high reliability must be taken in the context of the most j
recent construction and design experience. This body of experience includes widespread problems at Diablo Canyon, Zimmer, and Midland. This experience is scarcely cause for confidence.
For all the reasons given above, I conclude that the l NRC Staff's estimate of the frequency of core degradation i
t due to LOHS events is optimistic, unsupported by rigorous i analysis, and fails to properly account for uncertainties.
i
. Q.16: Turning now to other contributors to the probability of core disruption, what assumption did the Staff make with regard to the probability of simultaneous failure of both reactor shutdown systems?
A.16: The Staff assured that "there are sufficient inherent l redundancy, diversity, and independence in the overall shutdown system designs to expect an unavailability of less than 10-5 per demand," and concluded that "the combined frequency of degraded core accidents initiated by ULOF and UTOP events is less than 10-4 per reactor" (FSFES, p. J-4,5).
Q.17: What is the basis for the Staff estimate?
A.17: Beyond the explanation on pages J-4,5 of the FSFES, Staff claimed the value of 10-4 per year was a bounding value based primarily on LWR experience as published in NUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors." In Vol. 1, Section 4.3 of NUREG-0460, an estimate of 2x10-4 per year for the frequency of ATWS for typical LWRs was given. Staff also stated, "Because the
[CRBR shutdown systems] design and the reliability program are not final they have not been definitive in making the reliability estimate." (Staff Response to Interrogatories 36, 37, 38, 27th Set, Oct. 1, 1982, p. 60.)
Staff Witness Morris claimed that Mr. Rumble of SAI
. may have had a different basis for arriving at the value of 10-4 per year (Deposition of Staff Witness Morris, Oct.
12, 1982, p. 43).
Staff Witness Rumble said the basis for his estimate of the scram reliability of 10-5/ demand at DSFES, p. J-4, was based primarily on NUREG-0460; however, several other studies were mentioned as well. Mr. Rumble stated he was not familiar with the Commission's ATWS Policy Statement. (Edward Rumble, private communication, July 27, 1982, as recorded in Memo to files of T.B. Cochran, July 27, 1982.)
0.18: Do you agree with the Staff conclusion that 10-4 per year is a conservative " upper bound" frequency of degraded core accidents initiated by ULOF and UTOP events in CRBR and, if not, what is the basis for your disagreement 7 l
A.18: I do not agree. I believe 10-3 per year would be a conservative upper bound based on the Commission's LWR analysis in the Commission's Proposed ATWS rule for LWRs (46 Fed. Reg. 57521, Nov. 24, 1981)(see Tr. 2845, t
Cochran). While lO-4/ year might ultimately be shown to be appropriate.- in light of the current absence of the
! detailed CRBR failure mode and effects analysis for the shutdown systems and consideration of effects of common mode failure, including, for example, seismic induced l
T scram failures, there is at this time no basis for selecting a value larger than 10-3 por year.
Q.19: What assumptions did Staff make with regard to the probability of core degradation as a consequence of fuel failure propagation?
A.19: ,
Staff assumed that "the CRBR fuel design will be required to have an inherent capability to prevent rapid propagation of fuel failure from local faults" (FSFES,
- p. J-4) and that the frequencies attributed to LOHS, UTOP, and ULOF events adequately bound the contribution to core i
disruption frequency from fuel failure propagation (FSFES,
- p. J-5).
1 Q.20: Has Staff provided adequate justification for this assertion, and what is the basis for your conclusion.
A.20: I do not believe there is an adequate basis for this conclusion. Staff has not developed the specific 1
requirements or any associated criteria or confirmatory programs to prevent rapid propagation (details of the systems to prevent propagation of fuel failure are not 1
final at this time), and Staff could cite no documentation for the conclusion that the core disruption frequency due to fuel failure propagation is bounded by 10-4 per year (Response to Interrogatory 39, 27th Set, Oct. 1, 1982, pp.
l l
l l
62-63).
Q.21: What assumption did Staff make with regard to the conditional frequency that a CDA once initiated would be energetic?
A.21: Staff developed four categories of primary system failure as a function of the energy associated with disruption (FSFES, p. J-5,6) and assigned a probability of primary system failure by excessive mechanical and/or thermal loads resulting in continuous open venting into the upper containment through failed seals (Category IV) of approximately 0.1 per CDA (FSFES, p. J-6).
Q.22: What basis did Staff give for this assumption?
A.22: In response to interrogatories asking for all documents relied on to support this conclusion, Staff claimed that this estimate was based on "the Staff's general knowledge of and experience with the extensive research on the phenomena that may occur in a core disruptive accident
... , but refused to cite any documents. (Staff Response l to Interrogatory 43, 27th Set, Oct. 1, 1982, pp. 66-67.)
Q.23: Do you have any basis for disagreeing with Staff estimate?-
A.23: There is inadequate documentation to support the Staff's estimate, which may be correct, incorrect, conservative, I
1
. or nonconservative.
Q.24: What assumptions did the NRC Staff make regarding containment integrity in its analysis of CDAs?
A.24: Staff assumes that mitigating systems, principally the containment annulus cooling and vent / purge systems, will have an unavailability of less than or equal to 1 in 100 per demand. Staff also assumes that the unavailability of
. containment isolation will be equal to or less than 1 in 100 per demand. (FSFES, pp. J-6, -7.)
0.25: Do you agree with these estimates and, if not, why not?
A.25: If Staff is correct that loss of offsite and onsite AC power dominates the failure probability for LOHS events, such a failure could also cause the failure of the mitigating systems. Staff has not accounted for this common failure mode.
Staff Witness Rumble stated that the basis for the 10-2 per demand for containment failure was based on estimates of LWR containment failure of 3x10-3 (Edward l
Rumble, private telephone communication, July 27, 1982, as
- summarized in Memo to Files of T.B. Cochran, July 27, 1982). As noted in the Union of Concerned Scientists' comments on the DSFES (letter from Steven C. Sholly to
] Paul Check, 13 Sept. 1982; FSFES, p. N-50), the operating
. - - - - - - . , , - , - . , , - , . , .~._7
,- ,_m. _ , . , , _ - - . . . _ - . _ - . . ,y, , _ . . . , - . ,
. history of PWRs and BWRs in the United States does not support the assumed unavailability result of 10-2 per demand. A review of actual experience through 1980 was reported in Nuclear Safety (Michael B. Weinstein, " Primary Containment Leakage Integrity: Availability and Review of Failure Experience," Nuclear Safety, Vol. 21, No. 5, September-October 1980) and concluded that the overall availability of containment integrity was about 0.65 (i.e., an unavailability of 15 in 100 per demand). This experience base would dramatically affect the Staff's risk analysis of CRBR. Using LWR experience would appear to l increase tha estimate for contalment failure by a factor of 15. Even if the value for PWRs alone is used, the result is only 0.96 (i.e., 4 in 100 per demand unavailability factor). Obviously, if a Category IV CDA (as discussed by' Staff) occurs with a breach in l
containment integrity, a very large release to the environment will occur. Use of actual experience is l certainly to be preferred as contrasted with the very soft l
results obtained from the Staff's " analysis." It has not been shown that there are substantial differences between CRBR and the LWRs that form the present experience base.
l
. In addition, it should be noted that the assumption of the failure of the mitigating systems discussed above (the containment annulus cooling and vent / purge systems) will also dramatically affect source term assumptions for the CRBR plant. Such failures will also increase the failure probability of the primary containment since lack of annulus cooling will cause a more rapid pressure rise and an earlier failure of the primary containment. This allows less time for natural processes to operate to reduce the airborne source term in the containment, and the postulated failure of the vent / purge system will also increase the source term for containment release i
substantially, especially for particulates and aerosols.
Staff's analysis is inadequate in its failure-to ,
address the points noted'above and the concomitant large uncertainties inherent in the Staff's assumptions.
Q.26: Turning now to the estimates of the consequences in death and injury of CRBR accidents greater than the design basis, are the Staff's estimates presented in Appendix J likely to be accurate? Explain your answer.
! A.26: No, and there are several reasons. First, Staff's assumed radioactivity source terms are not supported by analysis or documentation. When asked the basis for Staff's estimate of the head release fractions selected in Table
. J.3 at p. J-10, including all analytical calculations and documentation, Staff stated:
The head release fractions (Table J.3) were selected on the basis of judgement from consideration of general LMFBR research of energetic CDAs involving a bubble of vaporized fuel material rising against the reactor vessel head, giving consideration also to the relative volatilities of different types of fission products and other materials. The selections were therefore not based on a set of analytical calculations or on any specific documents.
(Staff Response to Interrogatory 53, 27th Set, Oct. 1, 1982, p. 77.)
The release fractions associated with CDAs are highly design dependent. The Staff "judgements," based on no analysis or documentation, represent speculations, and the uncertainties in some of the estimates, e.g., Pu release under Category IV, could be at least a factor of 3.
Second, the CRAC model utilized by Staff assumes the LD 50/60 (lethal dose to 50% of the exposed population within 60 days) is 510 rads. In my opinion, this assumption is unrealistic. This dose-response level is associated with a dose-response curve depicted graphically at page 9-4 of Appendix VI of WASH-1400. This dose-response curve, however, assumes that the victims receive
" supportive treatment," which includes barrier nursing, copious use of antibiotics, massive transfusions, reverse isolation, and other special sterile procedures. WASH-
l 1400 estimated that the entire medical capability of the United States could provide such treatment to no more than 2,500-5,000 persons. WASH-1400 failed to address, however, how the victims of the highest exposures would be identified when there will be many others who will be suffering symptoms of radiation sickness (such as prodromal vomiting) from lesser exposures.
There is considerable controversy over the use of the 510 rads LD 50/60 The Risk Assessment Review Group l
, (NUREG/CR-0040, " Risk Assessment Review group Report to the U.S. Nuclear Regulatory Commission," Harold W. Lewis, Chairman, September 1978) concluded that scientific opinion supports a range from 400-600 rads. This range could cause a factor of two change either way in the j number of early fatalities. Moreover, the Risk Assessment i
Review Group concluded with regard to supportive treatment that "the ability to carry out such intervention has not only not been demonstrated, but isn't even well planned at this time" (NUREG/CR-0040, p. 19). Changing the LD50/60 from 510 rads for " supportive treatment" to the level of i " minimal treatment," i.e., 340 rada, could increase the number of fatalities by a factor of two to four (WASH-1400, Appendix VI, p. 13-50; NUREG-0340, pp. 26-28).
Other groups have used more realistic dose-response relationships which are closer to the " minimal treatment" f
1
- - , ,---- --v-, ,---- --
curve used in WASH-1400. The California underground siting study used an LD50/60 f r minimal treatment of 286 rads and for supportive treatment of 429 rads (Subcommittee on Energy and the Environment, House Committee on Interior and Insular Affairs, " Reactor Safety l Study Review," Serial No. 96-3, 1979, p. 366, attachment l
to letter dated 21 February 1979, from Bryce W. Johnson, Peter R. Davis, and Long Lee to Hon. Morris Udall, p. D-7). In addition, the " Accident Evaluation Code" (AEC) used to calculate health ef fects in CRBRP-1 utilizes an LD50/60 of 350 rems (SAI-078-78-PA, Z.T. Mendoza and R.L.
Ritzman, " Final Report on Comparative Calculations for the AEC and CRAC Risk Assessment Codes," Science Applications, Inc., December 1978, p. 3-6 and 3-8).
Third, the CRAC code contains several " hidden" assumptions regarding the cancer risk estimator for latent cancers, including an assumption thah the cancer risk at l low dose is a function of dose rate. The net effect of these assumptions appears to be to reduce the estimate of latent cancer fatalities (exclusive of thyroid cancers) by a factor of 2 to 2.5 compared to the estimate one would obtain using 135 x 10-6 potential cancer deaths per person-rem, which Staff claims to use for estimating offsite health effects (FSFES, p. 5-13). Furthermore, a number of experts, including Radford, Morgan, Gofman,
O Stewart, Mancuso, Kneale, and Tamplin, believe the Staff cancer risk estimator, 135/10 6 person-rem, is low, or probably low. Their own estimates of the cancer risk vary, but range from a factor of 3 (Radford, Edward, Science 213, 602 (7 August 1981), to a factor of 7 l (Morgan) to a factor of 28 (Gofman, John W., Radiation and Human Health (Sierra Club Books, San Francisco, 1981), p.
305) times greater than the Staff's estimate of 135/106 I
person-rem for fatal cancers due to whold body low-LET exposure.
Fourth, the source terms used by the NRC Staff in the CRBR accident consequence calculations appear to ignore any possible common cause failure of the containment annulus cooling and/or filtered venting systems.
Certainly both of these systems are dependent upon offsite and onsite power supplies, and both will fail if all power l
is lost. On this basis, as noted previously, it makes little sense to largely ignore common cause failures involving these systems, as Staff has done. If the containment annulus cooling system fails, this will l
I shorten the time between initiation of a CDA and failure of the primary containment. This affects decay of radionuclides that make up the source term and reduces the time available for natural processes such as gravitational settling and aerosol agglomeration to reduce the source
term. Failure of the filtered venting system shortens the time between primary containment failure and secondary containment failure and also increases the source term when the containment fails. In particular, the source term for particulates and radioiodines will be greater if these systems fail. This scenario will result in a larger source term for release to the environment and will result in more serious consequences than predicted by the NRC Staff analysis.
Another consequence of assumption of the containment annulus cooling and filtered venting systems is a greater release of Lanthanide group radionuclides, including Pu-239. These long-lived radionuclides will certainly have an impact on cancer fatalities and on land contamination (and related interdiction criteria).
l Q.27: What is Staff's position regarding the potential for a nuclear explosion in the CRBR?
A.27: In comments on the DSFES, Ohio Citizens for Responsible Energy (OCRE) asserted that "LMFBRs can suffer criticality accidents that can cause nuclear explosions as shown by The Accident Hazards of Nuclear Power Plants by Dr.
Richard E. Webb" (FSFES, p. N-10).
Q.28: Do you agree with Staff's position? Explain your answer.
A.28: No. Staff is incorrect in this regard as evidence by Staff's and Applicants' own characterizations of CDAs as i
explosions. In testimony before the Senate Subcommittee on Nuclear Regulation of the Committee on Environment and Public Works, (attactsd as Exhibit 3), DOE and NRC Staff witnesses discussed environmental and safety matters related to the CRBR, including " hypothetical core disruptive accidents (HCDAs)," " core meltdowns and energetic disassembly," and design basis accidents.
During the course of this testimony the following exchange took place between Senator Bumpers and Edson G. Case, then Acting Director, Office of Nuclear Regulation at the NRC:
Senator Bumpers: May I ask one question? What is an energetic disassembly? Is that an explosion?
Mr. Case In layman's terms, it would be called an explosion. Yes sir. (Exhibit 1, p. 19)
Later in the same hearings the following exchange took place between Senator Bumpers and Eric S. Beckjord, Director of the Division of Reactor Development and Demonstration at ERDA.
Senator Bumpers: Mr. Beckjord, what are the probabilities by ERDA's estimates of an explosion occurring in a breeder reactor plant?
l l Mt Beckjord: That would be the same order, 10 " per reactor year. I might add that one of the margins that is to be included in this plant design is the capability to withstand a very sharp explosion. The words " energetic disassembly" came up earlier. Maybe that is overly technical, but we hva been in discussions with the Nuclear Regulatory Commission on the amount of energy, the amount of explosive force that must be
accomodated within the structure. That matter is not settled yet. (Exhibit 3,
- p. 29).
These are not isolated references. The energetic disassembly of a fast breeder reactor is commonly referred to as an " explosive disassembly "[see, e.g., Lee J.C. and Pigford, Thomas," Explosive Disassembly of Fast Reactors,
" Nuclear Science and Engineering jj[, 28-44 (1972)] or "a small nuclear explosion " Hicks, E.P. and Menzies, D.C.,
Proceedings of the Conference on Safety, Fuels, and Core Design in Large Fast Power Reactors," Oct. 11-14, 1965, ANL-7120, pp. 654-670], a " low-efficiency nuclear explosion" [Stratton, W.R., and Engle, L.B., " Reactor Power Excursion Studies," " Engineering of Fast Reactors for Safe and Reliable Operation" (1973 Karlsruhe Conference), pp. 1331-1551].
There is no universally accepted definition of the word " explosion." The Webster's Seventh New Collegiate Dictionary defines " explosion" as "a large-scale, rapid and spectacular expansion, outbreak, or other upheaval."
l Cook defines an " explosive" as "any substance or device which will produce, upon release of its potential energy, a sudden outburst of gas, thereby exerting high pressures on its surrounding" [Melvin A. Cook, The Science of High l Explosives (Robert E. Krieger Publ. Co., Huntington, N.Y.)
1971, p.1] Cook groups explosives under three fundamental 1
40 4
types, mechanical, chemical and atoric (or nuclear).
Johansson C.H. and P.A. Persson in Detonics of High Explosives (Academic Press, London, 1970) state (at p.6):
Explosion is basically a rapid expansion of matter into a volume much greater than its original one. The word explosion thus includes the effects following or including rapid combustion or detonation,
, as well as purely physical processes as to bursting of a cylinder of compressed gas. We have chosen not to limit this rather useful wide definition of the word.
By these definitions an energetic disassembly of an LMFBR Core would constitute an explosion. It would not constitute a detonation which is a specific type of exothermic reaction that is always associated with a shock wave. If, as some authors prefer, an explosion is given a more limited definition such as to require the production of a shock wave, then most energetic disassemblies of LMFBR cores would not fit that definition.
i A nuclear explosion is an explosion in which most or all of the explosive energy is derived from nuclear processes, either fission or fusion, or a combination of
- both.* [See generally, Samuel Glasstone, The Effects of l
Nuclear Weapons, 1962 Ed. 1 1.10]. Thus, an explosion in an LMFBR, that is an energetic disassembly following a prompt critical excursion, would constitute a nuclear
- Fusion does not apply to the LMFBR for reasons that are obvious.
f 1
explosion as opposed to a chemical or mechanical explosion.
In response to a series of questions by Judge Linenberger in earlier testimony, I characterized a nuclear explosion as requiring a sufficient rate of energy deposition to result in the generation of a shock wave.
Upon reflection, I do not believe this is the preferred 4
definition. In any case, my previous testimony at Tr.
l 2777, 2779, 2785 and 2789 contains an error in inferring i that the energetic disassembly of a fast reactor would result in the production of shock waves.
For the disassembly to be sufficiently energetic for the mechanical loading to challenge the containment, the nuclear excursion in a large Fast Reactor such as CRBR would have to be characterized by a rapid reactivity insertion and the reactivity exceed prompt critical. This will result in a rapid introduction of energy from the l nuclear process, a rapid increase in rector power, l
elevated fuel temperature and vapor pressure formation.
I l In such an event the core will begin to expand.*
- Core expansion and fuel motion which reduces the material
! density will produce a negative reactivity feedback. Only a l
small expansion of the core is required to produce a larage disassembly reactivity. The reactor rapidly becomes sufficiently suberitical that any continued external reactivity insertion mechanism has no appreciable bering on the ultimate consequences. This marks the conclusion of the neutronic excursion and the disassembly of the accident [Waltar, Alan E.
(cont. next page)
I
An energetic disassembly, or nuclear explosion, in an LMFBR differs from a chemical explosion following detonation of a high-explosive in terms of the pressure-time characteristics of the two. Generally mechanical damage from an explosion or pressure transient can be caused by either a shock wave, which is transmitted rapidly to a structure, or the more slowly expanding bubble of reaction products or vaporized material or both. Pressures in a chemical high explosive detonation build up on a microsecond time scale. As a consequence, much of the damage potential of a chemical high explosive to immediate surrounding structures is likely to come from blast or shock wave effects. In an explosion in an LMFBR the build up is over a millisecond time scale and shock waves are generally not produced. Long-term bubble expansion (at least in the absence of a vapor explosion driven by a molten fuel-coolant interaction) would be the predominant damage mode for the slower time scale pressure build up associated with an LMFBR nuclear excursion.
(See, generally, Walters and Reynolds, ibid., p. 664.)
Q.29: What is your overall conclusion regarding the Staff analysis in Appendix J7 and Albert B. Reynolds, Fast Breeder Reactors (Pergamon Press, N.Y.) 1981, p. 619].
. . . _ _ _ _ _ _ _ _ _ _ _ _ _ ~__ _ _ ._.
i A.29: According to Staff Witness Rumble, Appendix J was done hurriedly because of the severe time constraints (Edward Rumble, private telephone conversation, July 27, 1982, as summarized in T.B. Cochran Memo to Files dated July 27, L
1982). This is apparent from the depth of the analysis presented.
Staff can correctly point to several conservative assumptions made in Staff's analysis. Nevertheless, Staff's analysis of the CRBR accident probabilities and consequences is inadequate and unreliable. Staff claims 4
"the uncertainty bounds could be well over a factor of 10 l
and may be as large as a factor of 100, but is not likely to exceed a factor of 100" (FSFES, p. J-24) As noted previously, the uncertainties in the probability estimates are larger than those of WASH-1400 and the Commission's previous conclusion -- that the numerical estimates of accident probabilities in WASH-1400 are unreliable --
applies equally to the Staff Appendix J analysis.
Furthermore, the consequences (i.e., health risks) of i
" Class 9" accidents at CRBR as estimated by the Staff are based on a series of assumptions with large associated uncertainties. One can find uncertainties of at least two orders of magnitude and consequences. When these uncertainties are considered together (compounded), I believe they result in an uncertainty of at least two or l
4 more orders of magnitude in Staf f's estimate of the acute and delayed health effects. With these large uncertainties in the probabilities and consequences, Staff's analysis in Appendix J does not support Staff's conclusions in the FSFES, Section J.l.3, at J-25.
l l
l I
1
I BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD l
l
)
In the Matter of )
, )
! UNITED STATES DEPARTMENT OF ENERGY )
l PROJECT MANAGMENT CORPORATION )
l TENNESSEE VALLEY AUTHORITY )
)
(Clinch River Breeder Reactor Plant) )
)
)
AFFIDAVIT OF DR. THOMAS B. COCHRAN
~~
, City of Washington )
l ) as:
District of Columbia )
DR. THOMAS B. COCHRAN hereby deposes and says:
The foregoing testimony prepared by me and dated November 12, 1982, is true and correct to the best of my knowledge and belief.
_ . _ "fh~ -
Dr. Thomas B. Cochran Signed and sworn to before me this 12th day of November 1982.
b@ hi $/dm Notary Public My, Commission Expires July 31.198~
l
CLINCH RIVER BREEDER REACTOR HEARING DEFORE TRIE SUBCOMMITI'EE ON NUCLEAR REGULATION OF Tile
- COMMITTEE ON ENVIll0NMENT AND PUBLIC W0ltKS l
UNITED STATES SENATE NINETY-FIFI'll CONGitESS
' Filt3T SESSION JULY 11,1977 SERIAL NO. 95-H27 Printed for the use of the Committee on Environment and Public Works enE f, BT N
V%0
?8n
- "8 "W?
U.S. GOVERNMENT PRINTING OFFICE 48-S$8 0 WASillNGTON : 1977 E
CONTENTS Taa.
Ilart, lion. Gary, U.S. Senator frona the State of Colorado, opening state-
= " "
sue n t o f- - -- - -- - -- - -- -- - -- - -- - - - --- ------ - - --- -
" ' IIST OF WITNESSES COMMITTEE ON ENVIRONMENT AND PUBLIO WORK 8 JENNINGS RANDOLPH, West Virgiata, Chairuses pir. . tor, pirlsion of Itenctor Develolunent and Deinon.
EDMUND 8. MUSKIE, Malse stration, amansgauled L,y Locialin Caffey ------- --
gg Re BEltT T. STAFFORD, Veranoat Preguared hintenaient -- --------- =- -------
MIKE GRAVEL, Alaska IIOWA RD II. BAKER, Ja., Tennessee R latory Cont-LLOYD M. BENTSEN, Texas , , Gohsick, Lee Executive Director for Olieration Ni JAMES A. McCLURE, Idaho , sninsion, aca onsaluauled by Edma G. Ca .. 1 -to Oake of Nu-QUENTIN N. I!URDICK, North Dakota PETE V. DOMENICI, New Mealco 4 clear Iteactor Hegulatiosa. NHC; and I aro1 . eintos . Director, Ditl-JOHN C. CULVRR, Iowa JOIIN H. CIIAFEE, Rhode Island 3 GARY H ART, Colorado aston of Site Safety and Envirousnental Anal aim, NR ---
MAIf0LM WALLOP, Wyonalag Itoisinau, Ahttsouy Z. Staff utforney, Natura WENDELL R. ANDERSON, Minnesota g acrosuguinled ley Dr. Tlaounas H. Cociaran, staff scidMet- -- g DANIEL PATRICE MOYNIIIAN. New York Prepared ntateinent ----------- -- = - - - - - - - - - - - ---
Jouw W. Yaco, Jr., Asaf Directer Young, Will1aan 11 vice g resident Ilreeder Heactor Dirimlon, Burns {goe, Ba:Lar Guaan, miserity asas Director Inc.,a man:Iuint ey . cyan ur Unrou, mentor ccrierate vice laresident Lson G. BsLLamos and IIaaoLa H. DaarwaN (Mlsortly), ___ _,, , 37 fof FUE50EU UE O ~~"~~~
sender prefseedemal assi Meesra Pu Lae T. Couan.mos, Recusan M. Hamasa, Ascuano E. Hsaon (Minority), and *tDillTIONAL MATERIAL Karusanna Y. CooLare (Minority), Associare Cesseste diz : Innrus nud lloe Im>Millon I* lier ou LM FBR project--- -------- 125 Frrfessions! sad resserch sier: East R. Baanzuwa:TE, PacL Cususa, E. Ksvin CoahtLI, g . ord, Eric S renlu.use to a reiluent froua Senator Huuipers for addt-Gsoncs F. Faaron, Jr., RamooLeu G. FLoop, Karnanass It. R FoncLu, Jeux D. Fassu.
uaN, ANN GanaamaaNT, Racueso T. Gassa. WsaLar F. Harpsu, Vssousca A. HULLaM8a. tional inforunation------ ---- -----***-
~
RONALo L Kats KaTMLssa A. EcaroN, JCor F. Paassts, Jous EL Poststos, Jr., Jauss Youns, W5IIIHlu II , YiCU I*IE*idenr' urn & Hoe,Inc.: p D. Ramos (Asesefeat Conneel, Ninerily), JaccusLN s E. Scuarsa, Caranosus A. Attachincuts to statenient-------- = - - - = = - - - - ~ ~ ~ ~ * - -
ETuan:Tra, E. Ersysma swasw Jr., Romaat Van HsovsLsN, SaLLr W. W4LKra, and Articles:
Havsu Wustsaios Wgg3lgigig ton Star Yetw( Menlo M5 Onch Def Breeder Proj-cc[* - . - - - - . - - - _ - - - - . - . - - - - - - - - - - - - ~ ~ - - - - ~ ~ - ~ ~ ~ - 75 New York Tinws. site of Tennessee Reactor Called 'One og y SpacouusTTrz ox NucLr.As RecutArion Worst' in Engineers' Hegert" - - - - - - - - = =
Iturns & Hoe stateuicut relative to newspalier articlea--------------- 7g GARY HART, Colorado, Chairmnas MIKE GRAVRL, Alaska 1III)
JAMES A. McCLURE, Idaho WENDRLL R. ANDERSON, Minnesota HOWARD H. BAKER, Ja.. Tenneswe DANIEL PATRICK MOYNIHAN, New York PETH V. DOMENICI, New Mestco GI) a p
I
10 11 In conclusion, the NRC stas f I "" Alr. Gossics. I know of no wason why we could not ask, depen gecegthie froin an environmen al ak a[af7t 1I Ri[""#t " *I'* upa the natun of ourmuceni.
, assuming that the ERDA p ramn t '- '"" ""*"*8I Senator McCarac. llave you done sof state, ment was dispositive of the need for a de -
Mr. Gossics. I am not aware of any incident in th.is particular case, its timing and ob ectives. .Natu $1 tl ijy, includmg "" '
l 11r.
i# ~'""* **'* I" *I?"' uIor AlcCixnx. I wonder if the ERDA witness mi ht wou oIfi N in of the Clinch bteyer site bsed on considesn**
iN -
t "I"""" nta um tabj'lity Beckjord, twrhaps you could indicate whether at. appla-
"'I If'at time.
Mr. Decxsoan. Senatw, we see waspondence tW .= &.W Thank you, Mr. Chairman. cants' internal memorandums.
. Senator H.urr. Thank you Mr. Gossick I I
- Nuo between the various participants in the project, as well as the b,enator Humpets, who is a ' member of tle E ""*"*" "*
C""Imndence that we receive directly. We participate in desie h:s mdicated an intenst in this subject and 1 J '"V m the general question of the nun project, f ynks.for Ved th]N" .i T review meetings. We ask questions to which there I"Usk* t*
- receiving it. As mgards internal memorandums, if they sie sent to us, ,
We are pleased to have him with us. ' ""' I 8 we an awam of thein. H they am not sent to us, we m wt awm M 9"C8 would Lo h setslike to go tothat of anelists,if the ERDA testimonfefo eveho is agreeab
"* "" 'l* .Iiator lleurras. 3Ir. Chainnan, I don't want to interrupt the pro-in y ur st mo y yo sa Q t! e hhesti n,staf Mr. Gossick.
obtained from That ERDA is, ceedings, but I wouhl like to ask a quedion. Due to the magmt a copy of the Burns & doc a uh 6, 2. That the quntion rui. sed in the internal memo about alw suitability o '
is almost 4 years after the da f yn u 8n y u uplain to us.
site which states that it is imleed the worst site ever sele cecordisig to your own procedu to [" "W y 8 nwnmnuuluin ,
nuclear powerplant, I am curious as to whether at that time or sub ,
i calling into serious question a uno'us counts as you de-scribed in your statement, was not made available to the NRC in
! iIuent to that time, Iturns & Roe called it to yo 4 yearsf t iought it was the worst site ever selected.Mr. Hrcxa Mr. Gossicx Si "I n a he status of the document, it We certainly were aware that there were matters which had to be was an internal meIio and a par the material filed investigated with rnpect to the technical suitability of the site. I am by ERDAin the proceeding $t tha,ll i - g mgi e verthat m my testimony. d i Senator Hurr.Wasit i" ERDA "' S'"8!or Huurras.
indicatmg Ilut you don'about theirconcern ef tinssd,t have any di l Mr. Gomicx. I don't know, sir . Mr. Hecxaonn. There is emisiderable correspondence regardm, g the Mr. BecxJoan. No sir. We v d it about 2 weeks ago, after the i
technical suitability of the site. Senator, work that was done at that doemnent was release'd to ti' "* 8P81*r8-time, particularly site borings, that type of infonnation we were aware Senator H'urr. So' 8t '88 85L
- your p ssession during that period of. A complete evaluation of all that mformation was done before the of timef N
Gnal placement of the plant was decided upon, and considerable an l Mr.
Senator BecxJonbumsf.
M.c 8MtY ur custanary pr cedure with se, ysis was terforened io support it. Senator Heurum. I won't p spect to th " "" n ractine agenciest Would you normally " un ist is 6e n rmal Senator Hurr. I think it is obvious that the concern of the conuni I is k d of. ""8" and many of us is whether the architect-engmeer was saymg one thm, g flow of that'internalinfonn n ia e e ntractw, with a argula. mternally and another thmg to the appropricte Government agencies.
tory agency or ERDA f -
That is what we want to pursue. We would like to go forward with I wouhl po.int out in this case, of courre, that Mr. Beckjont, accompanied by Mr. Iachlin
%Ir Gossicx.
const tuted t Sir, he apph, cant to NRC. I think the question is the EHDA testimony, Clinch River project.
Catfey, Dimrtor of the d ddirss. We are not involved with o t acto ina segul i ry .
STATEMENT OF ERIC S. BECKJORD, ACCOMPANIED BY jynat r McCr.Unr In the regulatory sense, then you would not I,0 CHI.IN CAPFEY gna see tho mt,ernal doc,ument of the applicant 8 icx. No, Mr. Un xJono. Mr. Chairman and members of the comndttee,1
.unt sir, Exceptunless thoseat wouhl gortions become of t ieirainternal \iart of docu.the me i$ -[ichli y hoose to pawnt to the regulatory agencyi appreciate this opportninity to discuss the environmental and safety iliatterS related to the Clinch River Breeder Reactor Plant. project Mr Goss C x. Y es, sir'
' which wererecently raised ci in,ted in the press.the July 6,1973, Mr Gojf b* ir- s pp rt of theirapplicationf memoramium 3 n . If these are questions in the minds of what is now With your permission, Mr. Chainnan, I will submit my written NINq, " uhli ise customary for you to ask for internalih>cmnent3! te3tinumy fu the wcord aml reduce the part that I give to you.
I
5 12 13 l
. nator IInrr. Without objection, that would be very agreeable based on the infunnation available to me as a result of research done in the interim. Some of the issues raised were speculative and others Senat Do "* *'*rifying question before he were founded I found oneitincombilete er that the or incorrect information. Of th testifiest In il' Praet I field of contractmg, what does an internal ruemorandum y were t iey writmg this tof What was the ing workissues, towar d pmper resolution is un erway in conjunction with licen pu of itf }f this ur in the day-to-day business of ing netivities as reqaired by N,RC. ,
sv ng that kind o eI Comments on the specific issues raised by the Burns & Roe memo-xsono. nator, any understanding from the infonnation randum nie as follows: I refer now to numbers in the original namo-
, Mr.gB and also page 3, item
- g u y inenuninduni and the statement which randum, in the summary section, page 2, item 5,d the associated costs Burns & R hen this was released,is that this was 5. The issue here is the suitability of the site an t eninterna[ moran "" h* E"Y ** f whicli was to advise the of site development.
directors of B, urns & Roe'of the situation of the project with some The plant site was selected following consideration of several d
~
sible alternative sites. In late 1971, the AEC appointed a Sen -
j mco"" .kati na regardmg their subsequent business actions toward 4 thTP3w"* , ity Steering Committee and Senior Utility Technical Advisory Panel to assist them in selecting a utility partner to design, build and oper-i ate the demonstration plant. Proposals were submitted to the Steering l and !Mgrnafhe purpose neemo. Evidently gar Munient available to theproject.
they didof notthe intend memo. AsPar-to make that indicated,
,. it was a private Conunittee and AEC by groups of utilities interested in participating i
genator Douzwicr. It was their own assessment, directed at their in the demonstration plant progism.
, There were in fact three sites that were considered. The Steering pege,fEcxJoRD. At the tune,3 the:s they proceeded were evidently toim.
a nuinber of evaluate their jobf Committee found that the proposal from Commonwealth Edison and flexibilityover ga, is usiness decisions that the company intended to make. I think ered in that clarifymg statement.The purpose of the memo the Tennessee the other proposals.Valley This Authonty otteredthat was the proposal incewased fins sitinky was a ress t ose decisions. by the Steenng Committee, and following that, by the AEC. I wdl 4
- 3,nator Doumwicr. Thank you, Afr. Chairman. not go over the details of the site comparisons that were made.
The soundness of that original decision was supported by the com-j r. Decusona. I reviewed the Burns & Roe memorandum in detail ~ prehensive and detailed site investigation program conducted during 1
3gI ab s on it,ans, based on information available to me.
Ib p Pmject u m,jant government-industry cooperative 1973, subsequent to the Burns & Roe memorandum. In contrast l
mular crrangentent iw emonstrating a hquid metal fast breeder awacto Bunis & Rue to others utilize for nuciear apJ rehension, powerpiants in ihe region andthewassitedem- w Qr p nt, 3 as' authonned by Congress on June 2,1970-Public Law onstrated to be fully acceptable from all standpoints.
E*' "*'* m this pmject are the Ener The Nuclear Regulatory Commission also confirmed the acceptabil-Developjn y t ,Admmistrat y 3 ty, and Proj, ton, Commonwealth ect Afanagement Corp.
ison, Tennessee Eky Research and ity of this site based on their independent review and assessment as Tlne ohjectives of this project are to design, heense, construct test documented in the final environmental statement for the CRBHP
'" *I te an L&fFBR demonstration plant. In Afay 1978 E$DA issued in February 1977, and the site suitability report issued in Afarch a
y8}umeyull
- , ry sup ort and management control of the pmject with\, continue'd 1977. utility tion In the site suitability report, NRC concluded that the founda-conditions were generally good and there were no subsurface rticipation.
sp n y fw this project since March conditions expected which would preclude the suitability of the site 19 6,. ring that time, project accomplishments have been good
' or the construction of the pmposed plant.
d*8 O As the nuclear powerplant siting criteria have undergone very sub-
{ ,, 'Kgr,
,r w )vb40 percent, complete, all of the an longlead equi ,
uneng .
witti stantial evolution over the past several years, the continued accept-l final enviromnental statement and site suitabilit# ability of this site further remforces the soundness of its selection.
j "{0[.t .avissued by the Nuclear Regulatory Commission. With regard to the co,a of preparing this site, any additional costs i
exammed project records reviewed the mimerous re
- and lnearings concernmg the projec,t, and inquired extensively to $ris
- . incurred for preparation of Ihis site compamt to a hypothetical"opti-j inum" site will be small when considemi in the coritext of the many
{rj ect procedures and status, >articularl in environmental, safet other factors influencing site selection. For example, the cost of high-Me enera y, can say that the project his also inad 8 " 8 cens8ng areas during ihe pa3,t ways that are necessary ao t ransport equipment, can be a major vanable year worki t quired undeNhe 1[E. PA fa.c y[i
- Sal r f hunted work authorization as re.
1970, until the recent suspension of the in the cost of site preparation and tins could vary considerably frtnn site to site.
environmental heann m A ril. The environmental hearings sus. I refer now to the issue of compliance with licensing requirements.
l The statement in the llurns & Roe memorandum, page 8, item 5, page
- g(ns' R t sto termmated M y M A, pendisig a final decision on whether proje orcontmued. 9, p ragraph 1 and page 9, paragrap 3, concerning compliance w,th i yg *" " i * *f**"*"" "" " ""**
!, hecent"*av"inie* o'ine 1"ut $ w*ee"ks aTo.Y"s' tit,U,$,$,",,i5 ,
14 15 S.nator lirure.us. You were saying here you have talked to every-gi ents established by the AEC for this project in material sub-to tl,te Congress pnor to nuthorization. body who mi~ht have tohl Burns & Roe that they would not, have In i r ginal program justification data arrangement for this proj- to coniply wit some of these basic safety requirements and that all of
'* 'h** 6"y, you say each of timin amured you, almre never was a hey that "all On August 11,1972 it was clearly stated or a, practice of avoiding compliance with the AEC Divisi 38*8 au ngulations including those to AE. C 1 censmg, and regulations, will b,e complied with'I'*riaining latmn licensing requirements.
infePe$t hirement, uglated to reflect theJustification establishnient h.unIftoyou talked in theto the writer of the memo,did you ask him DataofAthe f
, is in tho 'aed Program put that memol Mr. II :exaoun. I did not ask him that question, Senator.
roject at this time
- nijgement
. s to my comment No.77-106, on it, the unteswho owrs the f>the Project Steerini" Co'"'Senat or lleu re:ns. Thank you, Mr. Chainnan.
- mittee have been reviewed aus - > cord was found to su Senator II.urr. Proceed.
s tement made by Burns &1 wieing compliance witbi rt ttCF$
' M r. IIEcEJoRD. I Will read this in its entirety requirements. ** The minutes of the Project Steerin Committee have been reviewed nator Boweras. Did yu W . to e, man who wrote the memot and no mconi was foumi to support the statement made by Burna and Hm e nurning emnpliance with to CFR 50 requirements. In addition,
- r. BacxJoan. I have not had dctn: led conversations with Mr* I have personally called a number of men who were leaders in the
- . ung concerning the memo. I conduded that that was not Proper in . .
toleheld. days of the some Nemrek, former Ibiroject. These Reactorare Mssrs. Milton Divisioni Sh j
view of Ilars. this hearmput you have had some talks with him or con- rectors of ERDA's Development
["j{umt I ,
Mr. Wagner of TVA, Mr. Wallace Behnke of Commonwenhh Edison, Mr. BECEJokD. Oh, yes, I have had contacts with Mr. Yo""# be'. Messs. .I hn TayMr and George IIsrdigg of Westinghouse.
, c:use ne is respons,ible for the project for Burns & Roe' I me Each of Ihem n>sured me there was never either a pohey or a practim grd to tlas specific memo,I have not had detailed dircussions b'th of avoiding compliame with the AEC Division of Regulation,heen'.ag th requirements. It was, m fact, the pohey to go throg.. the entire afety 3,nator Haar. But you have had some discussions with himf amt licensing process as part of the project objectives.- ,
- r. It was understood by the pi lii neuono. I have had some discussions with him* of the 10 CFR 50 general des,oject leadu.5 that mu f cat g' IIART. About gnator e
,the memoI simply because of the technical diferemas between light water re-ths test scusson concerned whether we . :shed to see actors, for which the general design criteria were originally written, copy of t$e" test'"""Y-'# yhich he planned to give. He indicated he wouhl sendand a the Clinch River breeder reactor, for which general design entena 3,nator IfanT. But you didn't discuss the substance of the memot were not yet written in 1973.
- r. r.cxJoan. Beyond a few comments, there was no detailed d *' These modifications were develo >ed within the licensing process and
' are consistent with t he evolution o ahe licensing pmcess for LMFBR's.
cugion of ahe substance of his testimony or the memo. -
It should be noted that much work and discusskn was required to re-na or Haur. What was the nature of his comments t -
wire the diferences of technical opinion prior to the finalissuance of 10 ChR 50 mluirementa.- -# "**'"ed this passage regarding compliance with CHilHP general design criteria by NRC on January 9,1976.
3,na The fact that there were significant diferences of technical opinion y orHAaT.What wasin.e natute of that discussion f - -
- during-this eKort, however, does ,not lead to the conclusion that the indicInte[1ts[tYecla9 "".would be "" n as I what was intended. Ile .
project was trying to avoid comphance with safety requirements.The m his testimony, Senato 11
- n g int it at thatloint t safew rystuirnnents were properly established whets NHC issued, and hir. Hr nas. o-the ,roject accepted, these entena.
94 IrH * """'.i I <1 n't want to internipt his test . T io olijective of the design criteria and the net cWect of the CRilRP mm fi rti i y CHiti!Si ienni as safe as a iight water I nt but I think this is very cru. licensing process is same to make cial. si the,te. To suggest, as the Burns an y g,, You ,g s[gthat (1us is one of the mos,t en,tical parts of the memo . reactor h>cated at the nunnett. You say you have talked to Messiv. Milton inemorandum does, that there was an intent not to comply with h.
Tl eOfr. Wagner of TVA, Mr. Wallace Hehnke of censing requirementa or that the AEC desired to avoid including Shsw'mn Conm ve$ltli "N'"'""U O'* U"rdig needed sifety featun s because of cost consideratiova is simply not sup.
Westinghouse,and, I$'cl of Item has assured "'
you there w*as nevereb, of er ported by the facts, I can further testify that during my association with the project, a
ogpolieIy;o" p, P;'."CI80.e of avoiling com >liance with the AEC Division the policy has been, is now, and will continue to be, to comply with T" '"I 8- 9uestion is did y u ask him where lie got that in ormation and whether he based that mfonnation the Nuclear Regulatory Commission's licensing requirementa.
,,, g The three level defense-in. depth safety philosophy currently bem, g Mr BEcuJuan. Did I talk to Mr. Youngt used for design of LWR's was also adopted for CRBRP. This re-
17 ,
~
i quires design measures to prevent accidents, to provide protection -
j egamst either anticipated or unlikely faults that might occur, aml What that means III"'g*"""" design pi{ int for exam!>le , , I will attempt g, gg to beyond this, to provide appropriate engmeered safety features in the PI 2 5 '" " """P " y e & @ lis kr a '
r 1 design to safely accommodate extremely unlikely faults,if they some- We "'" '* "C#""j* sif ta a "'"jc"'tn l$1 occurred these would be no defor-should occur, in order to protect the health and safety of the iijejl
, ba ,e . es gn limit requins the design e ssesimt he exceeded. So the bar mig i defor i, t when th e nt y
, Fu 1thermore,,ERDA and NRC have agreed that, for the CRHRP, it is prudent to include additional measures in design to further limit the bar would be. chysic am t wou g gg I can consider accidents u g Potential consequences to i ich an accident occurred that
- cordingly, the project has m,the health and safety of the public. Ac-eluded C"5c' Imargins,"'II "V"I""'" heyond
' !"I the j'pi' s
necessary
, g'r w yMwde- w mW m M it i did -o tryomi the de3 gn, , ,
sign basis, thetical accidents m order to core involving reduce me the !tdown and energetic disassembi>ostulated .
woufd not return to its nutial connequences of hypo- condition. I, hat doesu't nuan that any-I- thing has happened. It does not mean that the a a e At the time of the Burns and Roe memorandum, there were ongoin d,scussions i between RRD and DEL concerning whether hy >othetical '
core disruptwe accidents IICDA should be include t 9"e"C"* It """ ply means that the sy. tem is capaone time in the design.
design basis for LMFBR's. The resolution with DRL was that, to .
So the question hear is whether flu. s llCDA $1iouId I e a,d
- basis avon1 schedule delay, two CRBRP designs would he submitted for .i-accident ami the entur,nystem should acco!mnot ate t wi i 88, concurrent review, one without and one with IICD.Us in the desi n margins or whether gomg beyond design Inmta shout be permi 4 basis, In a Maythe 1976 reference design letter, the NRC and a parallel agreed that II design'CDA's can and shouhl with the accident contained in other wap That is a one t,une event.
be exchideit from the design basis. Subsequently,the project withdrew I believe that convenly summarizes the dittennees between a e gn I saa the parallel design from further consideration by NRC, but it was basis accident and an accident which goes beyond a design
! imitually agreed that margins wouhl be provided in Ihe plant design in accident. . .
Senator llairr. A hypotiiet.ical core d.isruptwe accident, that means order to reduce the postulated consequences of such hypothetical acci.
dents so that the CRBRP would be comparable to cu'rwnt LWR's. a core ineltdown f Senator IIairr. Let me nm through that in English so I understami Mr. litexJonn. It means a core meltdown or .t could mean a co i disassembly Ihrough some means of sudden energy release which wou what, that means. It seems to me what happened here was in ihe dis, ,
t cussions withm ERDA and with the contractors, that it was decided ca""e it to di>perse.
I for purposes of determining the safety of the project.that there wouhi Senator llairr. Is it nafe to say that that phrase m. eludes the worst be two hypotheticals, or there would he two critical paths followed, possible thmgs that couhl happen t one which meluded the so-called hypothetical core disruptive acci, Mr. llocuaoup. I beheve it does. , ,
dents which I assume, are core meltdowns and things of that sort' Senator llan. So in this case, in order to save time, it seem and one which did not.
two decision, wem made, that you wouhl go on twoyour paths m r Decause in an effort to avoid what are called whedule delavs-you planning, one which incimled these mo>t sencus accidents for design
- took the two path method to avoid the delavs. Later on in an aeree, purposes and one which did not. That went on for awhile,it is a lit le ment hy May !!)76 letters, the NHC ngreed.1 don't know with whom unclear for how long.
that the path inchiding the hypothetical core disruptive accidents Mr. liccxaoao. I think it was midsummer of 1974 until the May 6 which is an interestmg phrase in itself, would be excluded from what letter, of 1976 Senator 1 h basis, one is called ihe design basist pimmiably the basis upon which Ihe deci. with the ace,Ian. Sowithout, ahnpst 2 years, you went on a two pat smn to go forwant would be made. - nlents,one i
So the project withdrew the so-called parallel design. including the Mr. BecxJoan. Yes.
it Senator IIairr. Then NRC i hypothetical core disruptive accident presumptions. Then ou pro- is unclear, that the planm, agreed, w,th whom and for ceed, it was mutually agreed that margins wouhl be providfd in the design to reduce the postulated consequence of such hypothetical accidents can and shoul,d be excluded. Then the pro iect withdrew the ccendents. path with the most entical acenlents meluded from f urther considera-What does that m :an f tion by the NRC, but it was mutually agreed that margins would be Mr. Hocxaoim. It means this, Mr. Chairman, and I will try to ex. provided.
plam ,ti m English. A design basi 3 accident is an accident which is as. What does that mean, margins would be prov,ided f ,
numed,to happen and the comse of the accident is evaluated. liv the Mr. liressoun. As when I was trying to come up with a s,unple ex-definition of design basis accident. what is meant i.s the particular part aniple, margins wouhl be provided to accommodate the consequences of the system or the p of a hypothetical core disruptive accident, but not as a matter of a sequence.5,of thatlent accu,lant in its and control thementirety with nohas advera. to accomumdate conse, the con. design basis--that is, not to say with the bar returnmg to its mitial quences withm design hmits. ongmal position.
19 ,
18 Senatm AlcCanr. So that there was no element of saying time,or' Specifically, these margins have to do with the structure of the meeting a selwdule that was motivating in your decision not to require pl:nt and the containment, the heat capacity in the base of the floor the most uvere accident containmentt and in the ability for heat removal from the con- Alr. Cee. That is correct, Senator.
under t:inment.theso reactor,if that this accident still occurred, it would reach an Senator thaien.us. Alay 1 a3k one question t What. .is an energeh.e equilibrium pomt. disassemt lyi Is that an explosion f It is a very low probability accident, but nonetheless, these design 31r. Case. In layman's terms, it woi1d be called an explosion. yes, mxrgins would make it possible to control the accident.
Senator llanr. There are two questions that come to mind. First Senator 31cCwne. It sounded like some OSHA language.
of all, who defines the margins, specifically and, second, why not, for Senat or 11.urr. 31 r. Beckjord, proceed, please.
purlmses of public safety accept the design course that included the -
Mr. Thex4oap. All of the relevant CRBRP safety issues, m. eluding most serious accidents! W,hy go on the two-path method in the first i those rai:ed by the Burns & Roe memorandum are bemg pmper y phce f Why not take the worst case basis for a design study 1 i-am) thor.mghly analyzed during the course of the bcensmg pr Mr. Case. Senator Hait, may I respond to that I a Most of the issues have been resolved in a manner mutually accepthese First, one should understand there are two aspects to this hypo- to ERDA and NRC. Work is continuing on the temamder o thetical core disruptive accident. First, you do everything you reason- issues at this time. No unusually diflicult problems m design have been ably can do to prevent the accident.There has been no change with the
- identified ~
NRC requirements with regard to that. In other words, we still re- To dafe, the project has made design changes est.imated to ulti-quire all the features necessary to prevent the occurrence of such an mately cost $60 million in order to meet additional h,cenam requ cccident. ments which have evolved during the mteractions,with N , an The other side of the coin is to assume, nevertheless, having all the is possil>le that other changes may yet be regmred. You ma a- ,
fe:tures, the accident occurs anyway for some hypothetical reason. On assured, however, that we have always been, and are at pitsent, thtt score, we took a course of action in between the so-called refer- cated to meeting all necessary licensing sequirements.
ance design and parallel design requiring the plant be designed to Referring again to the llurns & Roe memorandum, on page ld, i eccommodate some of the effects of this accident, but not all of them.
In other words, we contmue to require that the 5, and on page 17, item C, there is an issue rai mamtam its mtegrity for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> followm,contanunent requests for slweial licensing variances.The CRBRP project g ihe occurrence system of for no speciallicensing variances.
this same hypothetical core-disruptive necident. Consistent with one of the major CRBitP proj.ect obj.ectives og The reason we don't require all of the other features of plant be- the CRBRPis yond that time is, simply that we don't, ihink it is necessary from a demonstrating the licensability of the LMFBR concept, NRC as would being subjected to the identical licensing process by the asfety standpomt m view of the very low probability of the occurrence any commercial nuclear powerplant.
in the first place due to the features required to prevent the accident At the time of the llurns & Roe memorandum, the proj.ect was ex-occurrence.
Senator HAirr. On the breeder reactor program, are the same stand- lweting to request an exemption to conduct certam ards used for the light water reactors in this regard,in both regardst the AEC regulations under 10 CFR 50.12(b)., ,,
Mr. Case. The standards here are more severe than those for the light t. ion ag
]Iowerer. that procedure was changed, I beheve, m, antic water reactors. For the light water reactors, we require all the features the establishment of NRC. That attempt was dropped an, mstem to prevent such accidents. We do not require it to accommodate the we began to pursue the limited work author,ization, which is tlye end accident,in the event it should occur. In this plant, we do- of the lowess arquired under the NEPA Act of 1970. The hmited Senator McCwar. Might I just ask this questioni Some reference ~
work authorization wouhl permit us to begm site preparahan ac-has been made to, m order to save time or to avoid schedule delay, that g;y;,;e3' refers, if I understand it correctly that iefers only to the parallel When the CRllRP envimnmental hearing activity was suspended design feature for a period of time,and not to the ultimate decision, in April of this year, we were in the process of pursu,mg the itquest Am I correct t -
for a lim *ted work authorization. My point is that, with t,he excepuon Mr. Case. That ,s i not correct. That factor did I have just expremed, t here have been no requests fori,pecial variances.
d,ecision. Our dec,sion was strictly based onudgment i our j,notthat enter theinto our Reganting other NRC requirements, the pmject will meet all of the risks of this reactor should be comparable to light water reactors. of the Senator McCwne. As a matter of fact, you required a design be- nonlicable nnpiiivments. Ilowever, as aheady stated, some,ther no yond that required of thelight water reactorf NRC nmuirements were formulated for LWR's and have ei applicability or only partial applicability to the CR,HRP. In these Mr. Case. Yes, but I want to make the record clear that light water cases, the project will imet the intent of the LWR requirements hv de-reactors have some mh,erent features iljat tlus plant does not. In re- veloping modified or new mluirements in emIwration with NRC; qmrmg features on this plant, our objective was to make the risks that is, o7 of the 56 general design criteria were modified, plutonium comparable.
20 21 ". .
dose guidelines were developed, and new containment criteria were That concludes my statement, I.fr. Chainnan. I would be glad to l
developei,l. answer any onher questions that you have.
Referrmg aga.m to the Burns & Roe memorandum, on page 17, item Senator llurr.'l hank you, Mr. lleck[ord.
D, on page 17, item 7, on I think wo will direct questions to soth Air. Goss.ick an \ Afr. Case the first full paragraph,page 18 items A through D,and on page 18,the technical as well assuitability the earlier of the plantsite witnesses. id that concems me, hir. Gos-One thing discussed. I won't read through all of those points. sick about your statement is the tone and passive character of some These apprehensions of Burns & Roe about the site were based on 24 of the sentences where you talk about the site and the project as being core borings at the proposed site, of which only 4 were in the inune. not inconsistent wit h NhC objectives and standa Tls.
diat,e vicimty of the plant location. After a comprehensive and detailed At very few points in your statement do you go out of your way to o,te mvestigation program,d.the final plant location at the Clinch River give extraordinary assurances to us for the A,merican pubhc about i site was proven to be soun this project. For example, you say the staf review has been anned at This site investigation program included over 100 additional core i* assuring that these concerns were resolved in a manner consistent with borings, a test grouting program to confirm the homogeneity of the
, a safe facility design and onwration.
I foundation stratmu, detailed geophysical studies. and other extensive That is a very canfully worded statement. You use words like
- enalyses and tests. All the points raised by the Burns & Hoe memo- " aimed at" and " matters consistent with." In matten of this sort, j rendmn were fully and thoroughly reviewed with NRC prior to their ~
what the American people want, at least what I want,is something a
, issuance of the final environmental statement and the site suitability little more than that, how safe these facilities are and that the way report that, the Clinch River project has been going is not mconsistent with condit,for mns the are CRBHP.The good and that NRC the site staf is concluded that the foundation suitable for construction of the other projects and things of that sort. ,
plant. There is lacking, I think, a kind of itive note m, your testimony
, Referring to the Burns & Roe memorandum on page 22, item F, the that I think we would like. Is that a pr lem foryot t issues presented are safety approaches and plant beensability. This Afr. Gossicx. Sir, I think I must addnss the pomt w,th i d to comment,on the licensmg process was made at an early point in the the site as we have concluded in my statement. We arethat.
mced cony,n plant design. the site is a satisfactory site. We have not finished the safety review, As has already been explained,one of the key objectives of this proj- the other part of the review of the CRBR, hir. Chairman. It na still ect has been to b,eense this plant m the same manner as a conunercial undergoing stati review. lt is a pmcess not yet complete.
IMR plant. Many of the specific altproaches aml featums which were Themfore, we must not speculate about. the outcone of that unt,il 1
ultimately mcorporated into the design required extensive study, anal- the hearing is finished on the safety aspects and the staf action is ysis, and develop, ment. completed. All I can say is it is going along as any other application, The problems identified m. the Burns & Roe memorandum have cach recognizing it is a first of its kind.
Leen addressed in the beensing process as the design has evolved. So there is no intent to indicate either jessimism or, for that matter, Either they have been resolved or appmpriate work is underway to no particular grounds that I can cite at this point. for saying that, we resolve them. are convinced that it will be a safe design. We have just not. completed In conclusion, I w.ish to emphas.ize the following points: the goal of that process.
the CRBHP design has been to provide a plant which is at least as Senator ILurr. The tentative nature c f your statement is attributable safe as an LWR located at the same site. Since the commencement of more to the fact that you are still in the pmcess and not that you have the project, it has been the pohey to go through the entire licensing process and,to comply willi heens,ng i requirements established by the linkrering ir. Gossics.That hesitancyis itself.correct. . .
1 AEC Ihvision of Regulation and its heir, the Nuclear Regulatory . Senator II.urr. S tweilically, in your testimony in connection with Conumssion. All NRC licensing requirements are being fulfilled in the Burns & Hoe statement about it bein one of the worst sites ever the, project im lementation. you have the following language, Reduction in accident risks
'I he interna Burns & Roe memorandum is over 4 years ohl. Some selected,le achievab with remote location"-talkmg about the staf balance-of the issues raised m it u,cre speculative, and we have not found a . "against the resulting costs and inability of the demonstration plant, biais for them. The reman to accomplish its goals on a tune frame compatible with the pasent dressed m our detailed des,ung i issues have cach been properly ad-NRC m the beensmg process.gn aml site investigations amt with the timing guilsof the LMFHR program."
Each issue has been fully and completely What that says to me is there is a balancing of risk against cost and resolved or appropriate work toward resolution is currently time. You nsolve it slightly, at least. in the direction of cost and time.
- )toceedmg. correct me. I want to quote in that connection the con-NRC has agreed that ti,ie comprehens,ve i s,te i investigation progrant jg texty ,,from ,,,,nh,c,h wh that assurance came or that statement came, the Anal has established that the site meets NRC requirements. Good progress environmental statement dated February 1977 has the following sen-was inade by the project in,the h,ee,nsing area durmg the past year tence in it, or pa ragraph, that I will extract:
isntil the siispension of licensivig hea rings nii April. Anouser measure of alie relative alli ferences among the sites was obtained by estiina:Ing ilie relative twnsequences in tenus of overall population exposure out
I
- 23 .
22 -
Afr. Geien. Yes, sir. Certainly, that will le under continuing re
- to 50 milles. The radiolosical dosage at the alternative sites would tse roughly a view and scrutiny by the NHC. It would continue if the project con-fictor of 30 leu then the Cunch Blver site by shis ineasure.
tinues and, certa' inly, I assure the committee that will be looked at I think the question is in this balancing:IIow much does the risk go very carefully.
up in order to keep cost an 1 time down f benator lluurnus. Would the chairman yield at this pointI afr. Gossics. Sir, I would like to ask Afr. Case to address the details What is pressure groutingt l
of this. I would say at the outset, however, as I have already imlicated, Air. Gossics. Sir, it is an injection, as I undentand it, of cement
} '
thtt the objectives of the CRHR program within the context of the or concrete into the subsurface. into the areas where it is suspected or overall LMFBR program and the ability to meet, those objectives known that there are cavities that have been formed by erosion. ,
i on a certain time scale have tren stated by ENDA as required, and Senator licuerns. I believe it was in your testimony that you said hzyn been discussed with the Congress and the administration, and that would ho a possible suitable solution to solving the cavities wtro taken into account in that balancing pmcess. problemst Specifically how that was treated, I would like to ask 31r. Case to . air. Gossics. Yes, sir. It is a common technique. As I understan j rddress. many of t he dams in a he Tennessee area, one in paiticular ihar I i Alr. CAss. First risk is a product of probability times consequences, with, have used that technique.
! Your question rea,lly was what i Senator Iluurrus. The one I am familiar with is the Teton Dam.
i Senator HAirr.Say that again. . They used that technique there. ,
Mr. CASE. Misk is a pmduct of probability times consequences-the 3Ir. GossicK. I aln not falniliar with tilat,Init Ihat is Putt,ng ,t into i i j probability of an accident times consequences of an accident. The fac. Ihe rocks. I think that was dealing with an earth dam. We are talkmg i tor of 10 which you mentioned, which comes from our envimnmental here about rock.
striement, deals only with the consequence side. Senator liruerns. Are you not aware of the fact that that is precisely It is indeed true, taking into account the population distrilmtion at what canned the Teton Dam tc the alternative sites considered, that the consequences, shouhl a serious Senator 31cCwne. I would say to the Senator that is not what caused eccident occur, would be 10 times higher at the Clinch River site as the Teton Dam failing. The pirssure grouting worked. They didn't do compared to these alternative sites. some ot her thinge that shouht have been done.
i SenatorIIART. Because of popidation densityI Senator llatr. Mather than debate the Teton Dam, Mr. Case,I think hir. Case. Primarily because of population density. you nferred to the atmosphere in connection with consequences. Do I Senator HAirr. I think there is a quarter of a mdlion people living understand that among the alternative sites that the atmospheric con-
- within 50 miles of the Clinch River site. ditions at Clinch River are such that any escap,mg radioactivity would Afr. Case. Yes; the element we must also consider is the probability irmain in the area longer than the alternative sites t ,
of the accident. In both cases, due to the design irquirements, the Alr. Case. The ditrusion conditions am worse at the Clinch River probability will be very low, site as compared to the alternative sites, so, the answer is yes.
However Senator llatr. On the i is a 10 times, the consequences wouhl be 10 times higher although this of core meltdown, since abnestion ofwe contamment change in a small risk. That is the point which had to be at has come u e, will c,uantify and that,ifthe we cons balanced against in accordance with the Conunission's decision, the may. If you conhl, describe very briell how such an ine. dent would
, efect. of moving,from the Clinch River site to these alternative sites occur, or accident. It is my understan ing what happens is the core l
the efect.on the timing goals set forth by the EHDA Administrator iin' cats its way down through the containment, possibly, and would po-his programmatic statement, since there would be some delay involved tentially nicase large amounts of radioactive materials.
going from one site to the other. Second,in view of the seriousness of those consequences, what is the Taking that timing,into account,it was our view that you couhl not - justification for exchuling the so-called CDA from the requital design meet, the programmatic goals as set forth by the ERDA Administrator. criteria f Senator HART. What, about the degrees of probability among the 31r. CASE. Yes, to your first question, a possible way of violating con-various sitest eaimnent integrity following an extensive core meltdown would be for Alr. CAso. Essentially no diference at all. . the core to melt down through the concrete and then violate integrity l Senator HART. Probahility remains constant, consequences increase by moving into the ground.
I by virtue of staying at Clinch Riveri An important consideration before that sequence of events is another air. CAsn. Yes. possible method of losing containment integrity. That would be to Senator HART. You mentioned, Afr. OosSick the need for finding jiterally blow the contaimnent up due to overPiessurization during a where " solution,i cavities exist" at the site. Can y,ou assure the commit- much shorter period of time.That is our principal concern with regard tee that this will take place or has taken place, talking about the site to the Clinch River tractor.
questionst
' Our requirements are to avoid loss of containment integrity during Afr. Gossics. With regard to the possible cavitiesI the first 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> due to overpressurization, admitting the possibility, Senator HAar. Yes.
l 4
j 24 ~
25
- as is true in light water reactors, that you might lose containment in- information missing or information that needs to be clari6ed for the tegrity after that time through this meltdown process which you have o desenhed. The advantage of maintaining the integrity through thc 21 pM>r.sesof ourstutY review. Case. This is the usual case for us t Lour period is to reduce the potential consequences of accidents due to r:dioactive decay during the 24. hour period. ciencies in a tendered application and to require that the deficiencies The basis for accepting the small risk of the loss of containment in- Le remedied in the uppheation to be docketed for review. There in nothing unusual in this case.
te-rity due to the meltdown phenomenon is the low probability that we .
Senator nicCs.une. What I am interested m. is whether or not the beTieve of such an accident due to other design provisions.
Senator Haar. Does the Clinch River design include a so-called core information which was in the Burns & Roe memoranduin was by one catcherf means or another made with known to or made a concern of the NRC., t Mr. Case. The concerns regard to grouting, solution cavi Mr. Case. The specific method by which they would assure this re-were made known to the NRC, and ;were followed, up m our review.
quirement of 24. hour containment integrity, I don't helieve the pmj- The concerns relating to the physical characteristics of the sde were i ect has figured it out yet,'nor submitted it for our review. ~
made known to us, yes.
l Senator Haar. It hasn t included or excluded it f Senator Mces.une. Even though the memorandum was not furnu. hed M r. Case. Hight.
' to you and you didn't know of it until 2 weeks agoI Senator Haar. The French and British do inchide that featuref Mr. Case. That is correct.
Mr. Case. Yes. * .
4 Senator Doutuici. Did you say it had to be a core catcheri Senator McCs.une. Nevertheless, Ihe design criteria or the s.te i selee-Mr. Case. The method used to satisfy this requirement has not been ion problems that were outlined in that memorandum were either proposed by the apphcant. known to you or discussed by you over the period of the last 4 yearst Senator bovenici. Thank you. Mr. Case. Yes, sir. .
Senator H4=r. Senator McClure t Senator McCouse. I guess the bottom line would be,is there any-Senator McCt7me. Thank you, Mr. Chairman. thi in the Horns & Moe memorandum which would change the NR3 position on the sitet Can you gentlemen tell us how long we have had liquid metal fast Mr. Case. No, sir.
breeder reactors in, operation f .
Mr. HecxJoap. Smce 1951. Senalor Mcci.une. Mr. Beckjord,you we 3 asked the quest. ion if you ;
Senator McCtene. EBR-1 went operational in 1951 and EBR-2 had discussed in 1963. There are othe 3 in the worhl besides those two experimental You said youwith hadMr.
notYoung the background discussed that with Mr.of Young.
his assertions.,te In of breeder reactors in ihe United States,is that correct f tho fact that you have not discussed it with him disectlyince s, spi the Mr. Becusono. There are,I behere, eight that have been placed into memoramhun was called to your attention, do you have any knowledge l operation, Senator,in the world. of the background for his assertionf Senator McCs.une. Some of the design criteria in Clinch River are Mr. Hecuaoim. No, sir,1 do not. ~
not necessarily just dreamed i,ip out of engineers' dreamsf They are Senator McCoone. I suppose one thing that would concern me is based upn some experance with a breeder reactor of this kind f the complexity of management of a plant of this kind, particularly Mr. 04se. Yes, sir. with the way in which it was originally conceived.
Senator McCouse. The diference between th.is and those experi- As I understamt it, and correct me if I am wrong, Consolidated mental breeder reactors is that of scale and the problems on scahng Edison and TVA were copartners with ERDA in the development of up to a demonstration plant und applymg new techniques learned this durmg the experimental breeder reactor operation. Is that correctf Aff.kut originaHy.
Decxaonn. Conunonwealth Edison.
Mr. Case. Yes, sin. . . Senator Mcci.une. Excuse me, Commonwealth Edison and TVA.
Senator McCot ic They were the essential prime participants in the Project Manage-formational defic!ne.,Mr.
iencies Goss,.
were identifiedk, m. yourbystatement, youa say, the staK in letter"In-of that as the cost overruns began to mount and the inent g og ,3, Corp.' rojec,t and the delay of the pmject increased,in Ma fueyou'e n ared those informational deficiencies with the al- -
. 1976, ERD took over the management of the project,is that correct f legations of the Burns & Roe memorandum f Mr. Gossicx. Let me check with the staK. If[y, [',[, "
g7917,*,','
responsibility now, although Senator McCLUFe. I see a n0imber of heads shaki ng behind you. Commo.nweahh Edison and TV1 a;nre still involved in supe .
t Mr. Gossicx. I am advised that some of the questions were involved the rojectt ,
and relate to the matte 3 we have discussed here this morning that
,lfy,'8{#y"$y '"8[you j3 see any diference in the disiculty of i ab[r$a o elaborate. but this is a normal part of our overseeing t,he project froni liRDA's stamlpoint f Was t se g ter licensing process where the application is received and there is needed di cult pnor to ay of 1 11 an tl e at ep i gi
27 .
26 "
Mr. Case. That is correct. The risk is acceptable in either location.*
Senator McCa.ums. Simply because them were more cooks stirring There is less risk at these alternative sites. But takin6 that smaller ths brothf risk must be balanced against. meeting the program objectives.
Mr. EscxJonD. Yes, sir, at this point ERDA is solely responsible Senator Doutsici. .Iust one last summary question for myself. I 2 for the project. and ERDA can act. There wem Imssiile situations have been through the Clinch River project m the Energy Comnuttee before the eknge in May of last year where the activity could have as a new meml= r for a couple of months. In the piecess, I find we become deadlocked because of disagreement. have been on this project for years with all kinds of difering scien-If a disagreement had occurred among the principals, activity could tific positions.
ht to a stop. But that can't happen now. There have been scientists on both sides of this issue from its have Senatorbeen Mcbroudi.caz. Mr. Gossick, could you comment on the same inception. There have been energy people on each side of this issue.
question, from the NRCatandpointf '
Is there anything about the internal memorandum which you now Mr. Gossicx. Senator McClure, from - have in your possession which in any way changes your decisions to our staff does not consider that the ect Proj,the Management information constitutes that a we have, this point in time about its valueI safety issue as far as the di8iculty in managmg the program is con- ! Mr. Gossica.There is not, sir.
cerned. We consider that, purely EHDA's concern. Senator DoneNici. Ilow about ERDA f Senator McCi.uam. Agam, the bottom Ime, I assume, from the stand- Mr. BEcuaono. None, sir.
j pomt. of the hearing today is that there is nothmg m the Ibirns & Roe .
Senator DoMENiC . If you had known about the memonndum memorandum of 1973 which you have not dealt. with or are not deal- 6 months after it was wntien, can you tell us that nothing would Yr. c at is correct, sir.
have changed with reference to the way you have proceeded with f
Mr. C4as. Restricting it to those thin that afect safety. There are a numberof a thg. xaomo.There might have been a lot of activity when we dis-is that don't afect sa ety that we didift even follow. covered it as there has been over the past 2 weeks, Senator. I think Senator M uns. They would not be your responsibilityf Mr. 04am. That is correct. th(nator Douewici. Would we be when we am today with this Senator McCs.umm. Mi .
There are some aspects tgat are not sim lt I address,the same quest. ion toP,jec.t, ERDA. wi.th the same revIuirements imposed at this pomt and the safety, that NRC would not, be involveifwith, y from theERDA that standpomt mightof sannelicensingpswll' edure,f hat 'is ai fa,r statesnent, Senator. I believe so.
- be concerned with. Mr. Becusono.
ERDA has dealt with or is dealing with all of the items that are Senat r DoHENicr. IIow aboulthe NRCf listed in the Burns & Roe memorandum of 1973 f Mr. Gossics. I would concur in that. The mattem in the memo-Mr. Bacusona. I would ask Mr. Cafey to comment on that. randun3 that deal with the site have been brought out. So, them Mr. C4rrer. I would say, Senator McClure, that all aspects and is ist!nng that wouhl chan e matters as far as I can, see.
apprehensions and concerns listed in the Burns & Roe memorandum . Senator Dommu i. Has t u re been a recent companaon of the three, this is aside from business matters of Ibirns & sites inun the point of view of the allegations m the mternal memo-which Roe, have affect the project,ly dealt with except for those individualitems been adeguate rundum f Do we have that kind of evaluation somewhen m the recont of safety issues which we are still interfacing with NRC about. of the Federal (iovernment f All of the management aspects have been adequately dealt with.
~
mtert, the final environ-Mr. litexani:n.1 mental 8tatementaness I would he vhich refer to the,tes the Shernate si were evaluated. The Senator HART. Senator Domenici)
Senator DourNici. Thank you, Mr. Chairman Just a few questions. general cuidernise were looked at at the alternative sites,es well Mr. Case with reference to your statement defining risk as proba. . as Clinch River, the ditterence is that,I don't beheve extensive new bility tiraes, consequences. Could you enlighten me with some specifies f borings were taken at ahose uhernative sites. ,
What kind of probability are you talking about in the two areas If serious consideration were to he given at a future time to a differ-thit have been discussed here todayI ent site, then ihat is the kind of work that you would do to establish Mr. Case. The probability thai we are talking about, in our judg- - that it is in fact. suitable. , ,
ment, for a core disruptive accident is about 1 in 1 million or less per General considerations were looked at, at alternative antes, but not reactor the specifie st nictural mechanics of the sites.
believe year. In other words, the probability of such an accident, we Senator Donexici. Is it true that when you did do the spec,fics is less than one in a milhon per reartor per year. i Senator Don on this site, it proved out satisfactory with reference to meetmg the quences of van,arNicr. ous alternative You,would sites to be multiplying arrive that times conse-at your riski necessary safety requirements?
M r. CAar. Yes, sir. Mr. Hecusono. To the Int of my knowledge, that site is wholly Senator DouzNicr. You made the conclusion then that because the acceptable.
probability is so small, when it is multiplied times a higher conse. Senator DoMENaCs. Thank you, Mr. Chairman.
quence, the nsk is not meressed that much in terms of other considera-tions. Is that correctI
28 29 -
Senator HArr. Senator Bumperst As an example, the core of the Clinch River reactor consists of fuel Senator Buurmas. Mr. Chairman, I just have one item I want to material and it is encased in cladding material and structural material.
, pursue at the expense of going over terntory we have already covered. In order for the worst IICDA to occur, they would have to develop I would like to ask Mr. Beckjord this: the thing that has caused me some way in which the cladding material and the structural matenal more concern, I think, over the Burns & Roe memo than anything eh,e would fall away. It might melt, but the fuel,would stay in, place. I is the statement here, for example, where Mr. Young says: do,n't know of a way that this can happen, so it is considerations like The overall approach to reactor asfety matters has to date been based upon the dus. , ,
Fest Flux Test FaciHty approaches, the poHelen established by Str. Shaw in RitD, I am trying to desen,be it m a very simple fash,on i which has been l which are in many ways contrary to those of the AEC Nuclear Commission. studied extensively. It is by reasonmg such as this that the probability j For example, Westinghouse and Burns & Roe have been told orally that it could happen is nduced; and 10-* is a very small number. ,ll by RRD and PMC that we should not comply with the requirements of What do you do about iti Do you conceiv I
' accommodate this very unlikely event with,e m design of aordesign hmits van- w l 10 CFR 50. They cite the DRL safety considerations and would not '
necessarily provide a simple reliable plan. Then he goes ahead to say ables or do you find some other way, to handle itt The path that has gle between the AEC and so on. been chosen is to build of her margms into the plant. ,
thisis In yourpartof the powerstrufeckjord, testimony, Mr. I you say you started developing Senator BuMrERs. Mr. Beckjord, what ,am the probabilities by parallel systems; then you say, to cover hypothetical core disruptions, ,
ERDA's estimates of an explosion occurrmg in a bmeder reactor and then you drop that, planti Mr. IlecxJoan. That would be the same order,10-8 per reactor year.
e a andIhou e x$iSde I frN 'heNn't aN.*IiuNNOfy'.'the proNtTi,7 I might add t,hmt one of the margins that is to be included in this drew the parallel dealga from further consideration by NRC. but it was mutually Pl ant design is the capability to withstand a very sharp explosun,i.
egreed that naargins would be provided in the plant in order to reduce the postu. The words " energetic disassembly" came up earIIer. Maybe that is j lated consequencem of ouch bypotheticat accidents. Overly technical, but we have been in discussions with the Nuclear It really seems to me, and I admit that I may be in error and I may Regu'latory Commission on the amount of energy, the amount of ex-be inferrmg something here that is in error, but. it occurs to me that plosive force that must be accommodated within the structure. That what Mr. Young has been told orally is precisel matteris not settled yet.
we have cut corners on the safety specifications. y what happened, that Senator Buurens. Incidentally, the one that Senator McClure l Mr. Bacusoas. Senator, I don't believe that is the case. Let me take referred to that was put in operation in 1951 did explode, didn't itt
. your second question first, relating to the HCDA. The question that Mr. Becxaona. No ; it did not.That was a meltdown.
relates to the HCDA is whether the HCDA is to be acconunodated Senator Hairr. I think that was the original question. You say you
- within design basis. are using tigures of I out of 10-8, when in fact six breedem have been That comes back to a discussion which I was trying to clarify earlier developed when two of them have had meltdowns which I understand this mornmg, how a design is accomplished; as to whether the accident to be contained in the definition of a core disruptive accident.
l is fully contained and controlled within the design limits. When you use he term hypothetical because you can't conceive of
, In the case of the HCDA what has been decided is that the HCDA it ever hag 3pening. it has happened twice, at the Idaho Falls plant and is not accommodated with,m design limits; it is acconunodated in the phmt in Detioit. Am I missing something heref
, another way with margir.s built into the plant design so as to mitigate Mr. Becxaonn. Yea. The hypothetical accident we am talking about the consequences of that accident. hem is a lot more seveie.
i Mr. Case was explaining what the rationale for this is, namel Senator Hairr. Let's talk about one that is not so severe because I probability of an HCDA is very low. My figures are somewhat lower 'y, the ,
understand the definition of hypothetical care disruptive accidents
, than his. I would say that, the probability of an HCDA is reckoned to includes core meltdown and it has happened two times out of six.
be of the order of 10 to the minus 8 per y, ear or less. So it has a very low Mr. Itecusoan. We are talking about a total core meltdown.
probability of occurrence. The question is, what do you do about it. Senator Hairr. Well,let's talk about a little core meltdown.
Senator Bourzas. You are not suggesting that you are entirely - . Mr. lirexaonn. Osie occiirred at the plant in Detroit. Past of the eccurate on the probability, are you f subassembly did melt.
I Mr.Becxsona. No, sir,10-sorless. Senator Hairr. Does HCDA include a little ccre meltdownf
! Senator Buurcas. OK. Mr. Brcx.mno. No; that is a big one.
l Mr. BEcxJOaD. This accident has been studied extensively. For it to Senator Hairr. What do you call a little onef occur-let me just say a little more about it. I know of no mechanistic Mr. Ih:cusoun. A lit tie one is a core snelt, way that it can happen. It is called hypothetical because for the pur- Senator Hairr. A hypothetical core disruptive accident poses of analysis and discussion, we assume it can happen, but nobody Mr. Docx.ioun.That is t he hig accident.
nis come up with a mechanism by which it could logically occur. Senator HAltr. What, is the dividing line between big and littlet
!. M **
31 *
,' Mr. BrcxJoan. A little one,I wouhl define that as the accident that
- occurred at the Fermi plant. Part of the assembly melted. The reac. Senator llaar. I apologize for interrupting, Senator Bumpers.
, tor was shut down. It was safely shut, down without activity released Senator Hunerus. I am about finished anyway.The term meltdown
+ to the environment or m, junes to the public. could not have occurred if we had used the so-called core catcher j benator Buu rens. It is still shut down, isn't it, technology-I am sorry, the pool technology which the French and 1
Afr. Becuaoau. After that British are usingt of the accident was determ, med,accident, the vessel the deficiencies werewas opened, corrected, andthe cause Mr. Hecxaoun. Yes, sir. Could you repeat thall
, th:t plant was placed back into operation. It operated, I don't know' Senator Iluurtas. Could the Fermi meltdown have occurred if we for 2 or 3 years. It was finally shut down based on economic consid. were using the so-called liquid sodium pool technologyI Mr. Ih:cuauna. The 1mol or the loop would make no diffeience. That l
' crztions; but. the Senator II4ar. planttodid It seems meoperate again there is little after ciwulai that accident' here reasoning it would not have an eft'ect on meltdown-it could happen. If there
.t,is little if nothmg bad happens. If something bad happens, it is i i was the same design defect in the pool system,it could have happened big: but a big one can't happen. there.
Afr. litex.loun. That is certainly not the impression I am trying Senator BUnrEks. Do you penonally feel as far as you know, any-to convey,1fr. Chah;.um. body in the agency feel that the loop method which we are gomg to Senator II4ar. The Fermi meltdown, little because nothing got use is preferable to the pool techniques t away from it, the operatort Afr. Hocxaoau. Let me give you a short answer on that, Senator.
Afr. BecxJoan. Yes. I believe that a safe ,,ystein can be built using either approach. Each Senator AfcCi.cas. I thought he said the Fermi could be charac. one has advantages and disadvantages. I thmk that imm a safety tenzed aslug. point of view, they can and will be equivalent.
Af r. Becxionn. No. What we don't really know, what nobody knows is which one . is Senator ifAar. It can't be big because a big one can't happen. going to be more economical in the end. The French cite important ad-h vantages for their system. There are important advantages for ours.
des,fr.
ign.HEcxionu.
It happened Theone Fermi dayaccident that theoccurred.
coolant flowTherechannel was a flaw wasin One which we think is imtertant and wriich the Germans also think blocked. That is what happened at, the Fenni reactor. With no flow is important is the ability to inspect the entire system durmg periods pennissible in that channel, there was melting. When the assembly of shutdown. That is not totally possible with the pool system. That was meltmg, the plant was shut down right away. It, was detected. is an advantage for the loop type of system.
Senator HourEns. Have you seen this memo dated June 20,1976, Senator Ifaar. What we are trying to get at is what a hypothetical core disruptive accident is. submitted to ERI)A und the Electric Power Research Institute 1 It has Hurns & Roe and Hockwell International at the bot, tom of that.
Afr. HEcusoun. A hypothetient core disruptive accident is the wont recident that can be conceived of for this reactor. Ifave you seen thati lt is NRB 76-1. I assume that this is somethmg Senator HAirr. But it, can't hapgen, but j( can be concejved o(( that came to EHDA from RockweH and Burns & Roe. Their conclu-sion is that the pool concept is favond over both the hybrid and the Mr. Hocxaonn. No; it can be conceived of; but what. I am saying is that I can't give you a mechanism by which it could happen. In other loop Mr.designs Becnaonn.and they Yes; setaware.
I am out numerous reasons I recall now t k why' hat repost wonts, we assume that something like that could haptwn and we look that I will stand on my statement. I think that most of the people m ct the consequences; but I am telling you I don't know how it, could the business in this country will agree that either system can be made, hrppen. I can't come up and give you a sequence of events that, will that, the two systems can be made equally safe, Senator; but as I say, lesd to that accident. there is this controversy over which one will ultimately be more It is typical in the accident analysis of nuclear reactors that we ,
ible eceomical.
do,n't thmg can always happen.goWe mto try the mechanism.
to figure out a way inWe assume which it mig that the worst 'lbap. Senator Boursus. They,go ahead to say that the total probabilit pen. If we can figure out a way, then we do something about it of the core disruptive accident occurring by the pool cecept is ca -
. Senator HAaT. The key point here is you structure your design stud. . .
culated to be approximately one.fifth and two-fifths that of the loop ies and analyses by a standard called a hypothetical core disruptive and hybrid concepts, nmpectivel . That, is contrary to what, you said accident, but by your own definition, that is a set of circumstances :
a minute ago.These are the peo e that are buildmg it.
which cannot occur or which you cannot conceive of occurringg Mr. HrcxJoan. Can I provi e an answer for the record on that Mr. Hocsaonn. No. sir. I don't know of a way it conhl happen. The point, Senatorf I will stand on my statement. ,
Senator Buurens. Yes. Of coutw, we aw gomg to debate this th,ing studies have, shown that the probabilities of it happening are very this afternoon. If you don't have it to me before 2 o' clock, I will ,
. msll. That is what we am saymg. However, nonetheless, even though tako dramatic liberties with this memo and debate on the floor, they are very small, them are margins in the design to acconunodate Mr. Hrcs.mun. All right. sir,2 0' clock.
such an event and to mitigate its consequences. Those have been re. Senator McCs.uns. Mr. Chairman,I think it might,be helpful if we quired by the Nuclear Regulatory Commission.
would put, in the record a listing of the liquid metal fast breeder
~
1 M '
39~
smser w Emic 3. liax saono, inaccios. Devissos or Exactom Dsvatoeusp reactor plants that have been - '"" ' . " and in- m Nuamma i clude a prototypo whicli has Iblln taill$1 a$'"t"" -
n nt I lY I'"I""" '" N- air. Chairman and asciuhera e f the cominittee. I appreciate this optartunity to It. starts with Clernentine an r 1st 1011; which dincuss the entirunmentat and mafety suutterm related to the Clinch River Breeder Isas becat deconsnii.w.i.onedI E'II,II-I uInicle was espernisota:tl in !!)51 neul Itcactor 1*lant it'ltilH1's l'ruject which were rained les time July 6,1973,laternal gis gseen <yecoinnilssioned. Incidentally, EllR-1 is flie ,lai,e .:l, g- liura.und Itue meuu>rundma recently cited tu the pre ==.
nae enHHi* 1*rojwt le n jutut gusernment4ndumuy cuotenatim arungent produced contenercial electricity. It ligliged a sinall 'c'IY '" I'!"I80 for dranonstrating a I.bi uid Aletal Fast Breeder Heactor power plant sa Ilent gl8e te.St htation. authorized by congress on June 2.1970 tl'un,iic I.aw 91-27ss.The lisetaers la thia l- - -
projert are lhe Energy hemearch and Development Admisduiration (ERDA),
l g The Ferini.ptant Wats t1econtinissioned, but bee' ann opa-Mjml b Conimonneahh Edison (CHI Teane ee valley Authority (TVA) and Project
. - In ID63 SEFOIt which was located in .hkai '" 28"""x""#": n,rpwauon 4l' alm. The ,4,juures w this 1wject am to denigu.
in 1969, has been decon;intissione,d : and ilie FFTF' Cl*'""h '*I" R hanse. ",nsirnet. test and opaste an IJ1FHM deuunntnHaat idenh In May IM reactor;If that list miglit be niade . t " " " * " " * " * " ' ' " " ' " ' " " " " " " ' " " * " ' " ' ' " ""3*** """ **""""* "##"##
Senafor H.urr. Without ob*ection" 3 - lad"*' 'I ""DD"" ""d 8 ""'3 8"'"*"*
I liare had the ERDA res.pussaibilit) for this Projet aluce liarch 1978. Durlag
[The list fo11ows:] that time, l'ruject accoinplinhanents have beest goud, with design now over 40 i ti.s. Lastaa rtants percent cumplete. all of the long lead equlgunent ou order, and the Final Environ-i tuental Statement and Site Sultability Heport lemued by the Nuclear Regialetory seis.es , Cota?ula61ou (NHC). I have examined Project records, reviewed the nusmerous no,, reports and Iscarings concerning the I roject, asad insluired exteam'vely lato Ty88 tecehoe re.w seves M$ e g"""'"*" Project prowdures and statuni, particularly laa enCronsuental, mately and related
=
tapeessiemed..... La asesia abnalng matters. Generally, I can any that the Project his also saade good
- "%.............. en i g"[r$
a sI*" 8 I'e,",u,m,e,n,t,s
'""* and Inesiones ade'eT t.2tsN.".Z ". $ Y**"*d-progreau in theme licenalog areas during the past y ear. norking toward its goal of a I.linited Work Authorization (LWA) as regrared under the NRPA Act of j fa % 8*** 1970. until the recent huspension of time esaviromuental hearings la April. The I ["e,".D ***---
Deed'k 'inha's .----~ ammises...'.maiss I. scene.. ,a,u g ... $ u emetMaies'.
opAes-environinental hearnings suspension was requested by EEDA pending a Baal 1 weer sa (ten u . decintos uns whether the Project is to be tersulanted or contlaued.
Sifo8-.. ~.. ...t M",e %fenegg.
..,..se
- e. ;;;; ,,,y,,,, ,7.. gum y .. .. . no6s Decomasueeed. I have reviewed the Hurus and Hoe memorandona in detail alace It becaem savailable to me aluent two witLs ago. Sly statenients ou it are based on the f afor-
$ "j (csaat" "8"
j e k g'se", h**-
nuallon avaitalile to nie as a result of rehearch done in the laterian. Scene of the preen,pe ones triesk twen Sreeder g g,,,, meed me. ... lessredM, see,ssee.
age....... . iges under ceasepenasTennessee. 22 mem,ae:Z.*[* : inuijes rained were s.l eculative and others were founded on laconiplete or lacor-
- "8" rect luforamtion. Of the reanaladng lusues, I fotend either that they have already
) been resolved or that uork t toward proper resolution is underway la conjunction j wHh Heensing activilles na ruiuired by NRC.
h338 tot
- UAltr. (,sentle3nell, thank you, very lunch' mannents on Hie siawine im6ues mised by the Burns and Roe memoranduas am benator Dostmarr. If you will su ily that. unmer diat you um as fuHou :
In the " Summary" section, pages 2 and 3, Burus and Roe stated :
going to I rovide fo,r Senat 11 ~rhe nHe wiected is Hi.eir to be vny costly to prepare and could seen be Alr.11 Sciuosun. ) es, sir, before 2 0' chick.
""*"l'""***~
, . [The infonnation requested b Senator BunWers and Alr. licek- The cost of preimring the Ciinch River site will have been proven to be sub-Jord's prepared staternent follow: stantially more llania estimated. The site conta and problesas could be auch as to indicate a change of hite."
i ilvrorHETscAL Coms DsaapPrivE AcespEara roa LilFHR's Foot. VEasus laaor _ The plant hite uan wlected following consider:ation of several posalble alter-nadve kurs. In late Mt. the AEC appointed a Senior Utility Steerlag Commit-The riak aamoclated wills a pontulated IICDA is th u " H8e Ct8Hdequelaces tee used Henior Utility Tes huical Advisory Panel to amnist thean la molectlag a (magnitmie) and probability of owurrence Tl uuHty lertner to design, build and operate the demonstration plant. Proposala j event is dependent on the wre composithu$ aut! "" 8 8 la'*8ulaW were nubmitted to tlie Steering Couamittee aind AEC by groups of utilities later-quences are not affected by whether a pool I "" Nfwe cosame-b En'is amanmed. A reinart ented in fuerticlinating in the desconstration plant prograan. Each of the principal dated June:5,1076 FHR-Tebl* by a singt '# W ( - miles advanced tu tlie progan.nla received appeared to sneet the general requiressent that the probability of occurresace of an IICy H3) concludes that the proposed adle whouhl require no tuaumual denigu features or special con-puoi type L3IFHR these for a manlatrable h [" **
- I"Clur f "i of 88te less HHitfor a midanHon in Hwnsing. The Stenlag Caumanittee found, hownn, that the papanel Iloth the loop and lusol concepts are safe a elti appropdate kmn WW utued luncanal blHug Sesibility over the other proposals. This safety requirements.The compeerimons made by AlfHity h Wnau . (MVA protumal for bnHaung mind operating the IJtFHR demonstration plant design advantages and not to compasw absolute safet was ulHinately accepted by the Steering Comhalttee amt the AEC.
, The AI/HHI conclusions that the pool demi I d e lou'er probal.il- Three candidate alten uHhin the TVA area were considered: Widow's Creek, ity for occurrenace of aas IICDA than a lou I*i u is hamed sailely osa the larger John Sevier and Clluch Hiver. Analymiu of the relevant alting, enviromanental and sudiuun Inventory lausuediately surroundin I pents such direct umt factors for the three miten discloued no clearcut or overriding advan-ca loss of ohile immer or large earthipuskes [I "F I"8aIHla I 8 "U dheaus loss d Wes for anny mingle >lte. Much diferences an existed wete considered namensW core coollug, they compute that about 14 hour1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />
- C e acHots to treatmeist in the design within the linalts of enfating technology. An a pract: il for the pool planat an opposed to four hours for the I "" naatter, the three camildate siten were found to be equivalent froan site che
- 8" Hme is muill-cient in take mrrecthe actions but this dHfe"" " "" ect Hae Isrotnabilities teristic uiul environnaental standpulats.
unisiewhat. Ilowever, bince they are com inrin Althougle cosuluariawms of direct alle cost blightly favored the Wessow'm Creek in a mlHlon to I < hance la a billiosi, uncertalu I i H M e Ichanace I I I" # #tou great and John Hevier sites because of the avullability of mosne alte arrvices, the dif-l ta clalus factor of live diference between pool an I I IEH8- ferences were within the range of uncerta,nty inherent in auch cost cottsmates.
A _ . _ _ _ . _ _ _ _
34 '.
35 .
Aa overall analysis of the three altes,lacluding considerations of meetlug project .
end program objectives, showed Clinch Elver to have a decisive advanta;:e, be- 32essrs. 3litton tihaw and Thomas Nemsek, formier directors of RRbA's reacto" i cause the new site eersicos to he provided at Clinch River would be tuore cum- develusauent division, hir. Waguer of TVA, Mr. Wallace Behake of Pa==aa-patible with the nuclear steaum supply erstem- nealth Ediscu,31e==rm. John Taylor aind George flardigg of Westinghoumet Each The soundness of that ortglaat decisloan was supported by the comprehenstre of thesu has ammured sue iliere umm sieser either a policy or a practice of avehung >
and detailed site lavestigatlee prograni conducted during 1913. subsequent to the coaut diance with the AEU Division of Hegulation Ikeasing requir====*= h was Rurns and Roe seemorandusa. In contrast to the Burns and Roe apprehension. in fact itse policy to go through the endre befety and liceamlag preeems as part the alte was actually found to be slailler to others utilised for mainclear power of the project objectives. It was understoud by the project leaders that anodiace-phats la the region and was desmonstrated to be fully acceptable froni au stanad- tions to no.ne of the loCFit50 Gesacrat Demiga Criteria would need to be developed, polits. The Nuclear Regulatory Commisalon also conermed the acceptability of minuply because of the technical differences between IJght Water Reactora this site based on their ladependent review and namensnaent au documented in the g I, Wits), for w hkh the Ue aeral Design Criteria were originally writtes, and the Fi::al Environmental 8tatesment for the CRBHP issued in February 1977 and the Cituch Ittver Isreeder Reactor, for which general design criteria were met yet Sits Sultability Report lasued la liarch 1977. In the Elle Sultability Report, ARC , u ritten is 1973. Tinene moditications were developed within the licensing process concluded tiaat the foundatles conditions were generally gonal and there was ino and are conalatent with the evolution of the licenotag process for LMFBus, it tubeurface conditione expected which would preclude the auttability of the site or the construction of the proposed Ident. As the nuclear power plant ulting ', should be noted that suuch work and discusmien use reituired to resolve the dif-feretaccu of teclasdcal oposalott prior to the Anal 1smuance of UltisitP genesal design criteria have undergone very sehetantial evolution over the past several years. criteria by hMC on Janm.ry 9,197tl. The fact that there were algalacant diSer-the contissued acceptability of this alte further reinforces the moundness of its ences of technical opinion during tblu esfort, however, does not lead to the een-telection. clusion that the 1*roject was tr}ing to avoid compliance with safety requiressents.
With regard to the cost of preparing this alte, any additional costs lucurred for The matety rmulrements were properly estamaad when NRC lemoed, and the preparation of this site compared to a hypothetical" optimum" mile will be small . Project acivpted, these criteria.
when conaldered in the content of the many other factors laduencing site selection. The objective of the design criteria and the met efect of the CRBRP licensing In the " Background" section, poses 8 and 9, the Burns and Roe memorandum process is to inske the CHilHP at lemht as safe se a light water reactor located stttes: at the same tite. To suggest, as the Burna and Roe memnorandum does, that there
- The overall approach to Lif FBR reactor aarety matters has to date i een based was an lutent not to wmply with licenal ig requirements or that the AEC deelred an FFTF [ Fast Flus Test Facility) approaches and policies establisheit by Air. to avoid including needed safety features because of cost comalderatione, la '
Shzw and RRD [Diviolon of Reactor Reneerch and Development) which are in utmply not supported by the facts.
m:ny ways contrary to those of the AEC Dirlalon of Regulation (Ditt.l. For I c.us further testify that during my amenciation with the Project, the poHey example Westinghouse and Burma and Roe have been told orally by RHD and has been, la now, and will continue to be, to comply with the Nuclear Regulatory I' llc that we abould not counply with the requirements of 100FR50 Appendix A Commisalon's licensing requirements.
(General Deelsa Requiremental for EntFBR where suela requirements arine from Tlae three level defenne.in-depth hafety philosophy currently helag used for theoretical DRL anfety considerations and would not necessarily provide a almple, designs of LWHs was also ad6pted for CHBRP, This requires deutga anessures reliable plant. * * * *This approacts is being fontered in full knowledge that it to prevent meildents, to provide protection against either anticipated or unlikely assy not result la meeting DRL's licenalog requirements and that many isnues faults that ninght occur, and beyond this to provide appropriate engineered safety would have to be taken to the AEC Comminoloners for resolution. It is part of features in the design to safely accommodate extresuely unlikely faults,if they a rower struggle between parts of the AEC. The LMFHH Demonstration Plant ,,,yg. nimouhl mur, in order to protect the heekh and safety of the public, is viewed as a test case in which RRD and PnIC can knock out many theoretical Furthermore. ENDA sud NHC have agreed that, for the CRBRP. It la prudent safety-oriented design features which complicate commercial plants and make to incluele additional measures 1:4 design to further limit potential consequences them anore expenalve, and in which a new approach to safety and llevuxing can to the healtle and safety of the public. Accordingly, the Project has included be estabilshed. In addition, the Demonstration Plant is vieweil as having to he margins beyond the necessary design bauls in orJer to reduce the peatulated com-conhintenst with FPTF la order to justify the approaches on that project. I:nfor- acquentes of hypothetical accidentM Involving core meltdown and energellC dI8*
e assmembly. At the time of the Burns and Hoe memorandum, there were on-gelag ere b th t project in. . discussions hetween HHD and DHL concerning whether hypothetical core disrup-
.g number of existing approaches based on FFTF practices are already known five acrldents (llCDAs) should be included in the design basis (level threel u luntentist problem areas. These include the lack of speelAc safety criteria for fur LilFISHa. The rewtution with DHL was that, to atoid schedule delay, two e project; present emergency core cooling provisions and natural circulation CittlHP designs would be mutamitted for concurrent review, one without and assumptions: the current samumpHon that a double-en led pipe break is not a one with IRCDAs In the elemign hamis Ithe reference denigu and a parallel design).
credible accident; the assumptions as to the extent of the Hypothetical Core Dis- In a May 1976 letter the NHC agreed that HCDAs ran and abould be excluded ruptive Accident (HCDAl and features needed to contain it : the effects of undluna frous the design baste HuhuesIuently, the Project withdrew the parallel design rpills and dres; radioactivity release above the operating door; plutonium leak. I'""#"'"'"'"""*'""""I " ' """ "" ""la#
- d il t I" tEs and levels at the alte bounderles; and the ability to design an effective system '
woulel be provided in the plant design in order to reduce the postulated comme-toThiscontain a corecom,erning statement and reactor veswl meltdown'100FH50 compliance with requirements appenrn to queniva e,f much hylm>H:etical acciskats no Hiat die NW wd b CW@
lee tu direct condlet with the requirements estatellahed by the AEC for this Prnject I" *"*"I I'U"*
In material submitted to the Congress prior to authorisation. In the origiust
- AII'# N"" *""I "" *""I#'"*"*"* # ""
Program JuntlAcation Data Arrangement for this Project submitted to the JCAE and Itin suemorandtuu. are being properly and thoroughly analysed during 4he on August II,1972. It was clearly stated that. course of the licenslug pramens. Ilost of the lusuen have been resolved la a ananmer "All appIlcable laws and regulations. Including those pertaining to AEC 11 mutuaHy actWalde to NA and NM Wawk is mnHnuing mi the MW of tiaese lainues at this tinae. No u=usually dilAcult problems la design have been censing and regulations, will be compiled with " .
identhhd To date, Hae 14isject has piade alemign clianges haw to uMW -
This same requirement. updated to redect the establishment of the independent cet p mHHon in order to meet additional licenulag requirennents which have NHC, is in the Revised Program Justlacation Data Arrangement No.77-10d which evolved during the interactions with NHC, and it is pommilde that other changes covers the Project at this time * *
- The tulautes of the Project Steering Committee have been reviewed and no ""Y I'I '"* '"4" I"" "'" Y I'" ""*" '"****' "
- record was found to support the statement made by Burns and Roe concerning and are at present, dtWested to nneHug'aH necchm'ary HeaM Wrm4 compilance with 10CFR50 requirements. In addition. I have personally called a in the "Blackground** section, pages 14 and 17, the Burns and Roe memorandum gumber of men who were leaders in the early days of the Project. These are sta
' I e licensing approach involves numerous variance requests and subsalttale not originally included. * * *
^
- 37 '.,
M te=ts. All these points ralaed by ths Biarna sad Roe saemorand*m wire tztly rM "It appears likel.* that the Regulatory group of the AEC will be snude inde- thoroughly reviewe.1 with NRC larlor to theIr issuance of the Final Environm,en-g , ,g pendent of the development part of AEC moon. This would mean far tema c ance tal Statement and the site liuttatellity epo and that the alte was auttable of early and unique licensing approvals. * * *" concluded that the foundation condittuns are soo The CRBHP Project has asked for ano special licensing variances. Conslutesat for construction of the plant.
with one of the major CRBHP Project objectives of dennoinstrating the liceuma- In the "St.orkground" section, page the urns and Roe memorandum states' bility of the LMFBit concept, the CH15pH la being subjected to the identia al Slauy safety approaches laturporated in FPFF and planned for the LMFBR licensing process by the NRC as would any cosumercial nuclear poner plant. Demonstration Plant may not be commere el y I censable. Thesep plantg,,,g features At the time of the Burn and Roe memorandum, the Project was expecting to could be addreamed and resolved during requent an exemidion to cosiduct certalu site preparaGoa activities istlor to process."
receipt of a Construction Permit, as was perniitted by the AEC Regulations This conament was made at an early poiat in the plant design. As has already under 10 CFR 50.12(b). At that time, the AEO was granting esemptions for been explained, one of the key objectives of th a Project has been to license g, this cousmercist nuclear power plants under tble regulattosa sluce this was prior to plant in the same nianner as a coanmercial ted into the dealga inttitullon of the use of LWAs. When the regulations were cham:ed to incor- proaches and featuren whicts were ultlaustely Ineo I g,
por:te the LWA procedure, the Project abandoned consideration of an exemp- required extensive study, analysis and deve pmen og proe-tion request and oriented licenslag activities toward obtaining an LWA. a the Durus und Rue snenwerusadum hc e each Regarding other NRC requirements, the Project will meet all of the applicable ess as the design has evolved. Elther they have been resolved or appropriate requirements. However, as already stated, sonne of the NRC requisements were wort,1. underway to resolve them.
formulated for LWRs and have either no applicability or only partial epidic* .
In conclublon, I wish to emphamlze the following poln s.a I,lant which la at least chility to the CRBHP. In these cases, the Project will ineet the latent of the The goal of the CRilRP design has been to prov LWR requirescents by developing nuall8ed or new requirements ist couperation
- as unfe as an 1.WH located at the name site.
with NRO (e.g.,27 of the 56 Generr.1 Design Crlieria were modilled, plutonium gguce the commencesuent of the project, it has beea the policy to go through done guidelines were developed, and new containtuent criteria were dercluled). the entire Ilcensing process and to comply with I ce , the puclear Regulatory
' In the " Background
- mection, pages 17 and 1H, of the Burns and Itoe suenao- Ilshed by the AEC Divimlon of Regulation and its ,, g rzndum, additional statementa concerning the site appear :
Comueunion. All NRC licensing requircinents are ng 1
"The alte conditions described below may delay estabilahment of the suitabt!!tF cf the mite.* *
- Impleasentation.The internal Burns and Roe memorandum is over four years old.
"The Clinch River site selected for the LhlFBR Demonstration Plant is one imunes rul=ed in it were nieculative and we Lave not fou of the worst sites ever selected for a nuclear power plant based on its topography reinalning lximes have each been prolerly addressed n ou and rock conditions. The suitability of the alte will not be coudrmed until after ures.
investiga lo.as and with the NRC la the licenslag pr town en extensive soll boring program. There le a possibility that the site may not fully and completely resolved or appropriate wer be eweptable. As a minimum, site developonent costs will be high. The reasons rently prde* JADE-f:r the above conclualone are as follows : NRC has agreed that the comprehensive alte lavestigation program has estab-
"te) The alte has varying rock conditions. The rock on uhtch we are atten pt* li.hed that the alte ineets NRC requirements.
lag to place the plant is known to be somewhat nonhomogeneous and to be Good progre.s was made by the Project la the licensIog area durlag the past rubject to posalble solution activity probleans and perhaps volds and entitles. year until the muspension nf licensing hearings in Apr til be glad to answer any ThIme conditions may require some rock treatment such as grouting, and verld- That concludes my statement, lit. Chairman.
attion of the resuRs by an added soll boring program. Previous alten with additional questions the Committee may have on this subject. ,
sinulla r less have been dlGlcult to license and have been disticult and costly Senator IIAlrr. The next witness wil,1 be Mr. Wi is n hung, me
"(61 The areas surrounding the present estimated plant location are known oe.
to have an as yet undetermined degree of volds and cavities. Heraume of tble Eresident,of the Breeder Reacter D,tvision of Burns .
condition and the large amount of excavation required by the dealgn depth of TVould Iou identify for the record, those who are accompanymg containment at the present time, an estensive rock treatment (grouttug) effort youI aplears to be required, followed by a detailed soll boring program to verl(y thrt the results are satisfactory. This effort is anticipated to be required to STATEMENT OF WILI.IAM H. YOUNG, VICE PRESIDENT, BREEDER avoid posalble severe subuldence roblems, which coukt be the equivalent of REACTOR DIVISION, BURNS & ROE, INC., ACCOMPANIED BY DR.
a seismic event. The AEC has ins sted on such actions for previous sites with less extents of volds and cavities; conalderable costs and delays have been SEYMOUR BARON, SENIOR CORPORATE VICE PRESIDENT POR lavalred- *
"(c) Slope stability will be a. problem during construction due to the nature of ENGINEERING AND TECHN0IAGY thulte material. tr. "I am William H. Young. This is Dr. Seymour
"(d) Extenalve excavation, including much into bedrock, and backilli is pres- Mr.T.0100 cally estimated to Ise required because of the hilly terrain and subsurface condi.
- Unron, who ,. Yes, s.IS t'entor corporute vice pre 8 dent fo tionn at the site. '
"The resuus of the above could mean a minimum of more than six inonths'de- teclinogb to re.ad through my pttpated statement which has b 12y and sulllions of dollars la cost increases. In addition, final location and I WOH orientation of the plant will be delayed pending results of the soil burlug nubuntled along with a nutnbeF oi detailed attaclunents Which I W3II he ap ensions of Burns and Roe about the site were based on twenty-four [e Ijoint that Senator Hum just f core borings at the proposed site, of which only four were in the immediate pared hintenu nt, I certainly wouh ng to answer bequeSitonS wilh, brou On vicinity of the plant location. After a comprehensive and detailed site lurentiga, tion program, the Saal plant locatlou at the Clinch River site was proven to be that document that he he ul). I t d t nu ht be quite important, sound. This site investigation program included over one hundred additional Senator ]I.urr. At un appropriate t*une, wouId encourage you ti core borings, a test grouttug program to confirm the homogeneity of the founda- gggggg3o Acre possible your prepared Statement.
tlou stratum, detailed geophysical studies, and other extensive analyzes and
/~N