ML20052C493
| ML20052C493 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 04/30/1982 |
| From: | Deitrich L, Dickson P, Switick D JOINT APPLICANTS - CLINCH RIVER BREEDER REACTOR |
| To: | National Resources Defense Council, Sierra Club |
| References | |
| NUDOCS 8205050045 | |
| Download: ML20052C493 (55) | |
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4/30/82 N
' t UNITED STATES OF AMERICA d .a 4 wc NUCLEAR REGUIRIORY CMiISSION g; APR 301982 > '-5 Ohc cf the Secretsy DcGcti:rg & 8erscq In the Matter of ) Brad! 6
) 5 % a UNITE STATED IEPARIMENT OF ENERGY ) DOCKER No. 50-437 (D PRCATECT MM&GEMENT CDRPORATION )
TENFmTF VAILEY AlmORITY )
Clinch River Breeder Reactor Plant )
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APPIlCANTS' UPDATED RESPONSE #3 h a /?
'IO NA'IURAL RESOURCES DEFENSE %
COUICIL, INC. AND SIERRA CWB [Ax REQUEST 'ID APPLICANT AND STAET EUR ADiISSICNSA
,y [/j' Qa Pursuant to 10 CER paragraph 2.742 and in accordance with the Board's Prehearing Conference Order of Ebbruary 11, 1982, the United States Department of Enertjy, Project Management Corporaticn, ard the Tennessee Valley Authority (the Applicants) hereby update their responses to the Natural Resources Defense (buncil, Irr. and the Sierra Club Request to Applicants and Staff for Admissicnc, regarding (original) Contention 2, (original) Contention 3, (original) Contention 5 (partial), and (original)
Cbntention 10 dated August 12, 1976, August 13, 1976, Septertber 16, 1976, September 30, 1976 and January 14, 1977.
In these updated responses the following style has been utilized: ;
1 Eac:h statement has been restated and the updated response provided. I Certain of the responses are unchanged frcm those initially fwnished.
Ibwever, for convenien those unchanged responses also have been set forth after the w ur iate statements. 'Ihe responses contained in this Updated Response #3 supercede all prior responses to the statements to Which they are applicable.
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AMISSIWS REIATIVE 'IO (ORIGINAL) CINTENTIN 2 Statenent 1 l
Se following events had significant safety inplications:
(a) Brown's Ferry fire; ]
(b) Absence of an ECCS in the Indian Point 1 reactor; (c) he partial core melt at Enrico Fermi I.
Response 1 (a) The Applicant admit this to the extent that the safety implications were positive. The Brown's Ferry incident deTonstrated the soundness 1
of the defense-in-depth ):hilosophy and related the capability of safety !
systens to resporri to abnor:nal and unanticipated events and ability of those systens to unintain the nargin of safety required to protect the health and safety of the public.
(b) The Applicants cannot truthfully admit or deny this for the reason that the Applicants are not in possessicn of sufficient information cn this subject to draw a conclusion as to safety inplication of absence of m in Indian Point I.
(c) The Applicants admit this to the extent that the safety implications were positive. Se adequacy of the Fermi plant safety systens was I dancostrated. he reactor was shut down in an orderly fashion without any hamti to the health and safety of the public.
Statement 2 If the information now available had been known when each of the above-referenced reactors were constructeS, design changes would have been made to take account of the currently known data.
Response 2
%e Applicants admit this because imnediately following the Brown's Ferry Fire, TVA reevaluated the plant electrical systen design and separation ;
criteria. As a result of this reevaluation, runerous design nodifications '
and improvements to the plant fire protecticn capability have been made to increase the existing unrt31 n of safety.
AC-2 1
i8 . <a , , e Statement 3 Se error (s) which led to the Brown's Ferry fire were (select me or nore):
(a) HLrnan error; (b) Design error; (c) Engineerirg judgment error; (d) Comon node error.
Respcnse 3 he Applicants admit that the factors contributing to the Brown's Ferry fire, as set forth in the NBC Office of Inspecticn and Enforcenent Inves-tigation Report, IE Report Nos. 50-259/75-1 and 50-260/75-1, dated July 20, 1975, can be categorized under the four areas listed in the request for admission. A ccmplete description of the Brown's Ferry incident is given in the N1C Office of Inspecticn and Enforcement investigaticn report, IE Report Nos. 50-259/75-1 and 50-260/75-1, dated July 28, 1975.
Statement 4 he error (s) which led to the absence of an ECX:S in the Indian Point 1 reactor were:
(a) Htanan error; (b) Design error; (c) Engineering judcpent error; (d) Ccmncn mode error.
Response 4 The Applicants cannot truthfully admit or deny this statement. See l Applicants' response to 1(b) abcNe.
Statement 5 he errors Which led to the partial core melt at Enrico Fermi I were:
(a) HL: nan error; i (b) Design error; I
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(c) Engineering ju $ r.d. error; (d) Ccumen rrode error.
Response 5 ;
he Applicants admit that design error and engineering judcynent error may have contributed to the Enrico Fermi I incident. A couplete description of the Fenni incident was provided by the Atanic Power Developnent Associates in APDA-233, " Report cn the Fuel Melt Incident in the Enrico Ebrmi Power Plant on October 5,1966," dated Decertber 15,1%8.
Staternents 6 - 30 Statement 6 he Irobability of a CIR being initiated by sabotage conducted by one or two insiders wto could gain access to vital equipnent, or by a group of outsiders aided by one or two insiders, or by insiders who families were held hostage thereby coerced by outsiders, is greater than the probability of a design basis accident in an IWR.
Statenent 7 he probability of a CDA being initiated by sabotage conducted by one or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by cne or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders, is greater than the probability of a IOCA in an INR.
Statement 8 he probability of a CIR being initiated by sabotage conducted by one or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by cme or two insiders, or by insiders whose families were ,
held hostage thereby coerced by outsiders, is greater than 10 per reactor year of operation.
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Statement 9 '
We probability of a CDA being initiated by sabotage conducted by cme or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by one or two insiders, or by insiders who families were held hostage thereby coerced by outsiders is greater than 10-5 per reactor year of operation.
Statenent 10 he probability of a CDA being initiated by sabotage conducted by cme or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by one or two insiders, or by insiders who families were held hostage thereby coerced by outsiders is greater than 10 per reactor year of operation.
Statement 11 he probability of a CDA being initiated by sabotage conducted by one or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by one or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders is less than the probability of a design basis accident in an IMR.
Statenent 12 he probability of a CDA being initiated by sabotage conducted by one or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by cne or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders, is less than thr probability of a IOCA in an IMR.
Statement 13 l
he probability of a CDA being initiated by sabotage conducted by cme or ;
two insiders who could gain access to vital equipnent, or by a group of outsiders aided by cme or two insiders, or by insiders dxme families sere held hostage thereby coerced by outsiders, is greater than 104 per reactor year of operation.
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Statement 14
'Ihe probability of a QR being initiated by sabotage conducted by cne or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by cme or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders is less than 10-5 grer year of operation.
Statment 15
'Ihe probability of a Cm being initiated by sabotage conducted by cne or two insiders who could gain access to vital equignent, or by a group of outsiders aided by cne or two insiders, or by insiders, or by insiders whose fa'nilies were held hostage thereby coerced by outsiders is less than 10-4 per reactor year of operation.
Statement 16
'Ihe probability of a Cm being initiated by sabotage conducted by cne or two insiders who could gain access to vital equignent, or by a group of outsiders aided by cne or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders, may be greater than the prob-ability of a design basis accident in an IMR.
Statement 17
'Ihe probability of a QR being initiated by sabotage conducted by cne or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by cne or two insiders, or by insiders Whose families were held hostage thereby coerced by' outsiders, may be greater than the prob-ability of a IIXA in an IMR.
l Statement 18
'Ihe probability of a QR being initiated by sabotage conducted by cne or two insiders who could gah access to vital equignent, or by a group of outsiders aided by one or two insiders, or by insiders Whose families were held hostage thereby coerced by outsiders may be greater than 10 per reactor year of operation.
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Statement 19 We probability of a CDA being initiated by sabotage condue:ted by cne or two insiders who could gain access to vital equignent, or by a group of outsiders aided by cne or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders may be greater than 10-5 per reactor year of operation.
Statement 20 he probability of a CDA heing initiated by sabotage conducted by cne or two insiders who could gain access to vital equignent, or by a group of outsiders aided by one or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders may be grea .er than 10 per reactor year of operation.
Statement 21 he probability of a ca being initiated by sabotage conducted by cne or two insiders vho could gain access to vital equipnent, or by a group of outsiders aided by one or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders, may be less than the probability of a design basis accident in an IMR.
Statenent 22 he probability of a CDA being initiated by sabotage conducted by cne or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by cne or two insiders, or by insiders whose families were held hostage thereby coerced by outsiders, may be less than the probability of a IOCA in an IMR.
Statenent 23 he probability of a CDA being initiated by sabotage conducted by cne or two insiders who could gain access to vital equignent, or by a group of outsiders aided by one cr two insiders, or by insiders Whose families were held hostage thereby coerced by outsiders, may be less than 10 per reactor year of operation.
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, Statenent 24 l
he probability of a CDA being initiated by sabotage omducted by one or two insiders who could gain access to vital equipnent, or by a group of outsiders aided by cne or two insiders, or by insiders tJhose families were held hostage thereby coerced by outsiders, may be less than 10-5 per reactor year of operaticn.
Statenent 25 he probability of a CDA being initiated by sabotage conducted by one or two insiders who could gain access to vital equignent, or by a group of outsiders aided by one or two insiders, or by insiders who families were held hostage thereby coerced by outsiders, may be less than 10 per reactor year of operation.
' Statenent 26 .
We Applicant is unable to make a determination that the probability of a '
CDA being instituted by sabotage ccnducted by one or two insiders who could gain access to vital equipnent, or by insiders whose families were held <
hostage thereby coerced by outsiders is less than the probability of a design basis accident in an IRR.
Statenent 27 he Applicant is tnable to make a determination that the probability of a CDA being instituted by sabotage conducted by one or two insiders who could gain access to vital equipnent, or by insiders whose families were held hostage thereby coerced by outsiders is less than the probability of a LOCA in an IWR.
Statenent 28 Se Applicant is tmable to make a determination that the probability of a CDA being instituted by sabotage conducte3 by one or two insiders who could gain access to vital equipnent, or by insiders whose families were held hostage thereby coerced by outsiders is less than 10 per reactor year of operation.
AC-8
Statement 29
'Ihe Applicant is tmable to rriake a determination that the probability of a cn beirg instituted by sabotage conducted by one or two insiders who could gain access to vital equipnent, or by insiders whose families were held hostage thereby coerced by outsiders is less than 10-5 per reactor year of operation.
Statement 30
'Ihe Applicant is unable to make a determination that the probability of a CIA instituted by sabotage conducted by one or two insiders who could gain access to vital equipnent, or by insiders whose families were held hostage thereby coerced by outsiders is less than 10 per reactor year of oper-ation.
Response 6 - 30
'Ihe Applicants can neither admit nor deny the truth of these statements.
The Applicants have not developed ntanerical estimates for the probability of acts of sabotage and it is tmnecessary to quantify the probability of such acts in order to insure public safety. 'Ihe Applicants' position is that acts of sabotage are highly irrprobable and a defense-in-depth approach to effectively preclude such acts has been implemented as described in the Applicants' response to Item IV of the Eighth Set of Interrogatories.
Statements 31 - 71 Statement 31 An accident is considered credible if the accident scenario is plausible.
Statenent 32
" Plausible" is a synonym of " credible".
Statement 33 A reactor accident is considered credible if the accident scenario is considered to be, er believed to be, possible by a consensus of reactor safety experts.
AC-9
Staternent 34 A reactor accident is considered credible if the accident scenario is possible.
Statement 35 A reactor accident is conceivable if it cannot be dertenstrated that the accident is impossible.
Statenent 36 A reactor accident is possible if it cannot be demonstrated that the accident scenario is impossible.
Statement 37 A reactor accident is conceivable if it cannot be demonstrated that the accident scenario is irtpossible.
Statenent 38 A reactor accident 4" considered credible if it cannot be demonstrated that it is impossible.
Statement 39 A reactor accident is considered credible if it cannot be demonstrated that it is incredible.
Staternent 40
'Ihere are plausible accident scenarios that have not been conceived by reactor safety experts.
Statement 41 Ingrabable accidents are credible accidents.
Statement 42 A reactor accident is considered credible if it is not iripossible.
AC-10
i Statement 43 A reactor accident is incredible cnly if it is incapable of occurring.
Statermnt 44 l i
A reactor accident is considered credible if it is not incredible. <
Statement 45 I A reactor accident is possible if it is not inpossible. f t
Statement 46 i A reactor accident is conceivable if it is not inconceivable.
Statement 47 j
A reactor accident is considered credible if the accident scenario is j considered to be possible by a minority of reactor safety experts. (
Statement 48 i i
A reactor accident Whose true probability of occurrence is greater than '
10 per reactor-year of operation is considered a credible accident. >
Statement 49
- A reactor accident Whose true gobability of occurrence is less than 10 '
per reactor-year of operaticm is considered an incredible accident.
j Statement 50 A reactor accident Whose true Irobability of occurrence is greater than :
10 per reactor-year of operaticn is considered a credible accident.
Statement 51 A reactor accident Wxase true gnhability of occurrence is greater than ;
-5 10 per reactor-year of operation is considered a credible accident.
Statement 52 A reactor accident Wxuse true probability of occurrence is greater than 10~7 per reactor-year of operaticn is considered a credible accident. !
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, Statenent 53 1
A reactor accident whose true probability of occurrence is less than 10 per reactor-year of operaticn is considered an incredible accident.
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l Staternent 54 A reactor accident whose true probability of occurrence is less than 10 -$ !
per reactor-year of operaticn is considered an incredible accident.
Statement 55 A reactor accident whose true probability of occurrence is less than 10~
per reactor-year of operaticn is considered an incredible accident.
Statement 56 A reeactor accident whose true probability is unknown but whose probability is plausibly greater than 10 per reactor-year of operation is considered a credible accident.
Statenent 57 A reactor accident whose true probability is unknown but whose probability is plausibly less than 10-6 per reactor-year of operation is an incredible accident.
Statement 58 A reactor accident whose true probability is unknown but whose probability is plausibly greater than 10-4 per reactor-year of operation is considered a credible accident.
Statement 59 A reactor accident Q1ose true Irctability is unknown but whose probability is plausibly greater than 10-5 per reactor-year of operation is considered a credible accident.
Statement 60 A reactor wident diose true Irobability is unknown but whose penhability is plausibly greater than 10 ~7 per reactor-year of operaticn is considered a credible accident.
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. I Statement 61 A reactor accident diese true Irobability is unknown Mt whose probability is plausibly less than 10 per reactor-year of operation is an incredible accident.
Statement 62 A reactor accident dose true Irobability is unknown but whose Irobability
, is plausibly less than 10 -5 per reactor-year of operation is an incredible accident.
Statement 63 A reactor accident dose true Irobability is unknown Mt t*cse Irobability is plausibly less than 10- per reactor-year of operation is an incredible accident.
Statement 64 A reactor accident Wose true Irobability is mknown but whose probability is possibly greater than 10 per reactor-year of operation is considered a credible accident.
Statement 65 A reactor accident whose true Irobability is mknown but whose probability is pzsibly greater than 10-5 per reactor-year of operation is considered a credible accident.
Statement 66 A reactor accident Wiose true Irobability is mknown but W1ose probability is possibly greater than 10 per reactor-year of operation is considered a credible accident.
Statement 67
- 1 A reactor accident dose true Irobability is mknown but W1ose Irobability l is possibly greater than 10-7 per reactor-year of operaticn is ocnsidered a credible accident.
AC-13
, Statment 68 A reactor accident Wlose true probability is unknown but whose probability is possibly less than 10 per reactor-year of operation is an incredible accident.
Statment 69 A reactor accident whose true Ircoability is unknown but whose probability is possibly less than 10-5 per reactor-year of operation is an incredible accident.
Statment 70 A reactor accident whose true Irobability is unknown but whose probability is possibly less than 10 per reactor-year of operation is an incredible accident.
Stat m ent 71 A reactor accident whose true Irebability is unknown but whose probability is possibly less than 10-7 per reactor rear of operation in an incredible accident.
Responses 31 - 71
'Ihe Applicants have inplemented an integrated Reliability Program to further assure that events which could potentially lead to a core disrup-tien have been dealt with adequately. This Pr@mn is in addition to the multi-layered safety approach. 'Ihe CRBRP is designed to conservative standards ard engineering practices and additional features and margins are incorporated in the plant design to protect the public frm certain highly unlikely events.
Within the Reliability Program, runerical pals are set which Ircnide design objectives. Ntznerical evaluations are performed to assure that these design objectives cr aiming points are being adequately apIroached.
'Ihese aiming points are not fixed .1tanbers which must be denonstrated for a particular system cr for the CRIRP as a sole.
AC-14
F l, mny of the statements contained in this admission requrest would require
! adnission or denial based upcn fine shades of meaning of subjective words or phrases. 'Ihe meaning of these words and phrases within the restrictive l context of thew requests for adnissions (credible, plausible, conceivable, probable, etc.) depends upcn an individual's perception of a particular word and its connotations. 'Ihe Applicants have used tenns of this type to ,
provide non-technical descriptions of various accepted qualitative judg-ments. Itwever, the manner in which the subject statements are presented requires the Applicants to prestne an accepted universal definition of these terms and the interrelationship between these various terms ard the fine shades of meaning of these terms. It is not necessary to make these types of judcynents in order to have or evaluate an effective Reliability i
Program and protect the public health and safety. fereover, the qualifi- '
cations that would necessarily have to accmpany the interpretation and !
application of these terms within the context of these admission requests would only serve to cloud these vital issues and bog these proceedings down in exercises in semantics. '
'Ihe above discussion relates to Statements 31-71 inclusively.
More specifically, the statements are addressed as follows:
31 - 34.
The Applicants deny these statenents. See the Applicants' response to r statement Nos. 41-43.
35 - 39. !
These statements are denied, based upon the subjective terms which are used and the recogniticn that an impossibility cannot be demmstrated due to the i infinite time or ntsnber of tests required for such denonstration.
_40.
'Ihe Applicants can neither deny nor admit to this statement.
AC-15
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'Ihese statenents are denied. he terms used in these statements are I r
subjective in nature and presuppose an agreement as to the meaning of each term and a uniform perception of events.
44 - 46.
'lhe Applicants admit these statenents. i i
47 - 71.
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'lhese statements are denied. 'Ihe terms used in these statements are l subjective in nature and presuppose an agreenent as to the exact meaning of i each term and a miform perception of events. !
i Statenent 72 !
'Ihe IMFBR technology has a certain lack of solid in-pile test experience and a lack of maturity of technology which makes preclusion of CIas as !
DBA's, i.e., their prevention, to be next to inpossible. !
Response 72 *
'Ihe Applicants deny this statenent. 'Ihere is an extensive base of IMFBR design, cx:nstructicm, and operating experience which has been developed over the last 28 years, both in the U.S. and in other countries, tich is '
available fcr use in the safety evaluation of CRBRP technology. It may be safely concluded that CIas can be excluded as design basis events consid- '
ering this extensive DFBR technological base in conjunction with the preventive design features included in the CRBRP augmented by reliability activities.
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i Statenant 73 P
CIRs nust be in the design spectrun, they have to be considered as the same t
bases, desicpt bais accidents. '1he CRBR should be designed for these on a reasonable but conservative basis and analyzed specifically, this specific desigri for its specific paraneters.
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i Response 73
'Ihe Applicants deny this statement. See the Applicants' response to Iten III in NRDC's Interrogatory Set 10.
Statenent *Q As the Applicant / Staff (to each as appropriate) understand the technology and sees the safety issues, DCM have to te considered as design basis acridents.
Response 74
'lhe Applicants deny this statenent. See response to No. 73.
Statement 75 It is NRC Staff's objective to ensure that the level of public risk fran CRBR does not exceed the risk fran other nuclear power plants.
Response 75
'Ihe Applicants can neither admit nor deny this statenent since the Appli-cants do not develop NRC Staff policies and objectives. However, it is the Applicants' intent to assure that the level of public risk fran the CRBRP does not exceed the risk fran other nuclear power plants. ;
Statenent 76
'Ihis objective stated in Admission 75 may require adopting criteria nere limiting than canparable IMR criteria to account for the lack of experience and the tacertainties in technology for INFBRs.
Response 76
'1he Applicants deny this statenent. See the Applicants' response to statenent No. 72.
AC-17
Statement 77 Se hypothetical core disruptive accident (HCDA), inder various names, has been postulated in the consideraticn of fast reactor safety by U.S.
regulatory groups for 25 years.
Response 77 he Applicants admit this statenent.
Statenent 78 Sere is a consensus that the codes describing the cnset soditan boiling and subsequent novenent of fuel (durirx3 a 'IOP or IN CDA) give results that are at best seniquantitative once fuel melting begins, and the description of the transiticn frun the onset of fuel and clad melting to the core config-uration described by these codes Wilch calculate energy release, Iroceeds primarily by plausibility argtrnents.
Response 78 he Applicants deny this statenent. In addition to the plausibility anjtrnents, the scditan boiling model, which predicts the onset of soditrn boiling in the SAS3A and SAS3D code analysis of hypothetical core disrup-tive accidents, is shown in Ref. (1) to be in good qualitative agreenent with experimental observations. Also, the analytical models which describe the transiticn frun the onset of fuel and clad melting to the core config-uration described by these codes, which calculate energy release, are based cm first principle fonnulations expressing conservation of mass, mcynenttrn, and energy, as doctanented in Refs. 6, 7, 8, 9 and 10 cn page 11-1 of CRBRP-GEFR-00523.
i Referencer (1) G. Ibeppner, F. E. Dunn, and R. J. Heames, "'Ihe SAS3A Soditan Boiling Model and Its Experimental Basis," Trans. Am. Nucl. Soc.,
Vol. 20, p. 519, April 1975.
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AC-18
Statement 79 Considerably nere operational experience and an extensive search for initiating events associated with actual reactor designs is required to l
demonstrate that the Irobability of a GBR C[A initiating event (such as lose-of-flow acecrnpanied by failure-to-scram) is incredible, l
Response 79 he Applicants deny this statement. See responses to No. 72 and No. 73.
Statenent 80 With the ted.niques and experience yet available, it is inpossible to demonstrate with certainty that a severe (CRBR CER) excursicn could not take place.
_ Response 80 he Applicants deny this statenent. We word " certainty" is asstrned by the Applicants to mean that quality or state of being certain cn the basis of objective evidence and the entire phase " demonstrate with certainty" to mean thoroughly confirmed without reservation or doubt.
W e Project and the NRC agree that HCDAs should not be considered as design basis accidents. he CRBRP Project is cx:rtmitted to provide objective evidence that the plant design will meet or exceed the nulti-layered safety criteria. This objective evidence nust and will " demonstrate with cer-tainty" that CIAs are properly excluded frcm the design basis.
Statement 81 he use of the term " highly inprobable" by the Staff / Applicant (to each as aan,iate) is meant to imply a probability equal to or less than (select one):
a) 10-1 b) 10 -2 c) 10 -3 d) 10 AC-19
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g) sme lower Irobability Response 81
'Ihe Applicants can neither admit nor deny this statenent since the Appli-cants have not precisely quantified an associated :nathematically-rigorous probability for the term "high inprobable" as used by the Applicants in the responses to Interrogatory IV of Set 8, Interrogatories II-3(a), II-3(b),
and III of Set 10, Admissions No. 6-30 on Cbntention 2, and Admissions No.
49 and No. 50 cn Contention 5.
Statenent 82 The use of the term " highly miikely event" by the Staff / Applicant (to each as appropriate) is meant to imply a chance of occurrence equal to or less than (select cne):
a) 10
-1 b) 10-2 c) 10-3 d) 10 e) 10-5 f) 10 g) sone lower Irobability Response 82
'Ihe Applicants can neither admit nor deny this statenent since the Appli-cants have not precisely quantified an associated mathenatically-rigorous probability for the term " highly unlikely event" as used by the Applicants in the response to Adnissions Nos. 31-71 cn Contention 2.
_ Statement 83 With respect to the Um, the determination by the Staff / Applicant (to each as appropriate) that a CDA need not be included in the spectrun of DBAs is a subjective determination.
AC-20
Response 83
'Ihe Applicants deny this statenent. See response to No. 84.
- j Statenent 84 With respect to the GBR, the detennination by th: Staff / Applicant (to each as appropriate) that a CIA need not be included in the spectrun of DBAs is an objective determination.
Response 84
'Ihe Applicants admit this statenent cn the assumption that an objective analysis of the design of systens and ccmpone.cs and application of recognized detenninistic criteria to the design, coupled with reliability and probability analyses and sound engineering jur'm falls within the definition of the tenn " objective detennination" t. lized in the admission request.
Statenent 85 With respect to the CRBR, the determination by the Staff / Applicant (to each as appropriate) that a CDA need not be included in the spectrun of DBAs is based upon qualitative judcynents.
Response 85
'Ihe Applicants admit this statement to the extent that both quantitative and qualitative judcynents fonn the basis for the objective detennination that a CDA need rot be included in the spectrun of DBAs. See response to No. 84.
Statement 86
'Ihe Staff / Applicant (to each as appropriate) is unable to state the probability of a CRBR-CIR in quantitative terms such as, for exmple, less than one chance in 104 reactor years.
AC-21
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Response 86
'1he Applicants deny this statement. 'Ihe Applicants are able to " state this probability in quantitative terms" to the extent shown in the assessments which were Ireviously subnitted providing an estimate of the Irobability of occurrence of certain initiating sequences leading to an ICDPL. See MARD-D-0118, Rev.1, CRBRP Nuclear Island Reliability Assessment of CRBRP Reactor Shutdown Systen (Nov. 1975); NEIM-14082, An Update of the Preliminary Reliability Prediction for CRBRP Shutdown Heat Removal System (Jan.1976).
'Ihese assessments are one of several factors which has led to the Appli-cants' position that HCDAs should be excluded fran the CRBRP design basis.
See response to No. 84.
Statanent 87 It is impossible to design a 350 N(e) INFBR such that, within the time schedule for the CRBRP, the CDPL can be objectively determined to be not credible, where credible is defined in a manner consistent with existing NRC regulations.
Response 87 The Applicants deny this statement. See response to No. 84.
Statenent 88 It is possible to design a 350 N(e) INFBR such that, within the time schedule for the CRBRP, the CDA can be objectively detennined to be not credible, dere credible is defined in a manner consistent with existing NRC regulations.
Response 88
- 'lhe Applicants admit this statement. See response to No. 84.
(
AC-22 l
[ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .
Statenent 89 he Staff / Applicant (to each as appropriate) is unable to quantitatively determine that the probability of a CRBR-CDA is less than one chance in 4
10 reactor-years with a cmfidence level greater than 90 percent.
Response 89 he Applicants can neither admit nor deny this statement. he Applicants have not attempted to place confidence levels cm probability assessments of the initiation of HCIA's.
i Statement 90 he Staff / Applicant (to each as appropriate) is unable to objectively determine that the probability of a CRBR-CDA is less than one change in 10 reactor-years with a confidence level greater than 90 percent.
Response 90 he Applicants can neither admit nor deny this statement. See response to No. 89.
1 Staternnt 91 !
mile sane probabilistic criteria has been arployed, the ultimate decision that a CRBR-CIR does not have to be considered as a DBA is based upon engineering judcynent.
Response 91 he Applicants admit this statenent to the extent that engineering judg-ment, objective evaluations of the design, and deterministic application of design criteria are enployed in concert with p cbabilistic evaluations.
- See response to 84.
Statenent 92 he decisics that GBR-CDA does not have to be ocmsidered as a IEA was made in a deterministic manner as distinguished fran a probabilistic manner. !
AC-23 1
t i
Response 92
'Ihe Applicants deny this statement. Both deterministic and probabilistic '
evaluaticms were used to preclude ICIas fran DBAs.
Statement 93 '
men a decision is made in a deterministic nenner, the decision makire rvcoss involves engineerirg judgment.
Response 93 r
'Ihe Applicants can neither admit nor deny this statment. A decision "made in a deterministic manner" may or may not rely upcn engineerirg judgment depending upon the nature of the decision and upon the information that is available.
9 Statenent 94 Since the decision that the CRBR-CDA does not have to be considered as a CIA was made in a deterministic manner, it is not possible to q.2antify the probability of the occurrence of a CDA. -
Response 94
'Ihe Applicants deny this statenent. See response to Nos. 86 and 92.
Statement 95 Since the decision that the GBR-CDA does not have to be considared as a DBA was made in a deterministic manner, the chance of occurrence of a CR can only be stated in subjective terms such as highly inprobable or extrancly tulikely.
Response 95
'Ihe Applicants deny this statement. See response to Nos. 86 and 92.
AC-24
l
- l. ADMISSIONS RELATIVE 'IO (ORIGINAL) CINTENTION 3 Statenent 1 l
'1he Staff / Applicant (to each as appropriate) uses the SAS3A code in its analyses of I& events.
Response 1
'lhe Applicants admit this statement to the extent that SAS3A is one of the codes used fcr such analyses. Recent analyses had ussi SAS3D code.
Statement 2
'Ihere is no formal or informal docunentation to demonstrate that SAS3A or any of its nodules behave in a stable fashicn or produce accurate solutions to the original equation sets.
Response 2
'Ihe Applicants deny this statement. Refer to Applicants' response to Interrogatory Set 2, Item I(A)-4, (6 and 7), and Set 6 Itan I(A)-2.
l Statement 3 Thc details of the conputational algorithms associated with each of the models in SAS3A an$ the code as a whole can cnly be disecnered by physi- !
cally examining the EORTRAN coding otnprising the cxxle.
Respcnse 3
'1he Applicants deny this statement. A reasonable level of docunentation of the ocmputaticnal algorithms usa 3 in SAS3A and SAS3D are contained in references listed in response No. 2 above.
In response to Interrogam Set 6, Itan I(A), the Applicants state; "A great deal of the algorithmic detail of the nodels is left for the inter-ested reader to discovery by physically examinirg the PORTRAN coding canprising the nodels." (emphasis added) . that was meant by algorithmic detail is a statenent-by-statement flow chart identifying where each and AC-25
. . .. . . i every equaticn is located in the sugc u. This degree of detail can only be discovered by physically examining the FORTRAN coding atmprising SAS3A.
However, the computational algorithms, or mathematical computations
(
(formulas) are derived and doctrnented in the references identified in l response to No. 2.
Statenent 4 [
In the use of SNS3A for the CRBR analyses, 33 subassernblies nust be ltrnped f
into 10 channels ard this is a subjective exercise.
Response 4 2e Applicants deny this statement. he SAS3A nodel uses 10 channels to -
model 198 subasseiblies. Rese 198 subasserrblies can be reduced to a set of 33 groups etnsisting of six subassenblies in each group which are '
essentially identical in their thermal hydraulic and nuclear characteris-tics by virtue of the core hexagonal synmetry. It does require engineering '
judgment to reduce these 33 groups into the 10 channels in the SAS3A code.
However, it is not accurate to characterize this gocess as purely sub-jective since there is a sound technical basis for the process of reducing the 33 groups to 10 groups. he SAS3D code provides capability to model up to 34 channels.
Statenent 5 2e Staff's channel selection for the CRBR in SAS3A analyses differs fran ,
that of the Applicant.
Response 5 h e Applicants admit this statement.
Statement 6 l If the Staff were to use the Applicants' channels together with the other Staff asstunptions ard inputs to SAS3A, the Staff would calculate lower ;
energetics than den using their own channels.
AC-26
Response 6 he Applicants can neither admit nor deny this statement. We Applicants have not perh-3 calculations usirg the Applicants' channel selection together with the Staff's assunptions and input to SAS3A.
Ibwever, as identified in the September 29, 1976 ACRS meeting, two varia-tions of the NBC base case have been performed. We two cases being:
- 1. NRC base case with the asstrnpticn that 50% of the fissicn gas being available instantaneously upon melting in SIJJMPY, and
- 2. NRC base case using the asstanption of SIINFY ccmpressible region coupled to soditsu vapor dynamics as in CIAZAS.
Statement 7 If the Staff were to use their own channels together with the Applicants' inputs and asstanptions to SAS3A, the Staff would calculate higher ener-getics than when using their own inputs and assunptions.
Response 7 he Applicants can neither admit nor deny this statement. %e Applicants have not performed calculations using the Staff's channel selection together with the Applicants' inputs and asstznptions to SAS3A. See response 6.
1 Statenent 8 h e Applicants' channel selection in SAS3A may not be conservative.
Response 8 he Applicants can neither admit nor deny this statement. Given that there are 198 unique subassernblies in CRBRP, each with 217 pins, any reasonable channel selection scheme in Wiich these have to be grouped into 10 SAS3A channels and each channel modeled as a single pin would generally result in a conservative treatment of the incoherency effects in CRBRP during a AC-27
l
.. 1 l .
postulated transient period. Ibwever, the channel selection is generally unimportant conpared to many other SAS3A inputs ard asstmptions in deter-
, mining the level of energetics Iredicted by SAS3A for a particular case. .
hus, it is not meaningful to assign any particular degree of conservatism to a given channel selection. l Statenent 9 he Staff's channel selection in SAS3A may not be conservative.
Response 9 Se Applicants can neither adM t nor deny this statement. Refer to response Ib. 8 above.
Staternent 10 here are a ntrnber of input asstmptions to SA33A analyses that in and of themselves have a minimal effect but which in ocznbination can result in large differences in the calculated energetics of a IM event.
Response 10 Tne Applicants can neither admit nor deny this staternent since the specific input asstrnptions to the SAS3A code to which NRDC refers are not iden-tified.
Statement 11 There is no real evidence that axit.1 fuel expansion will occur en the short time scales required to alter the energetic of a IM event.
Response 11 he Applicants deny this statenent en the basis of the information in ANL-RDP-23, ANL Monthly Progress Report, December 1973, SAND 76-0273, IEJREG-766501, July 1976 and the Applicants' otmputed results. he refer-ences provide direct experimental evidence that fuel axial expansion occurs even daring Ix:wer transients. Fuel expansion effects are inportant during AC-28
l
.= . . .
the relatively slow power changes that occur during the initial period of an IIF event. SAS calculations have shown that fuel expansicn can occur during the slow power changes without generating excessive reaction forces l in the fuel cr clad. Daring an IN event, the early fuel axial expansion is a prompt negative feedback effect that can alter the energetics of the event. This fact is docunented in Section 7.2.3 in CRBRP-GEER-00523 ard in Section 7.2.2.3 in CRBRP-GEFR-00103.
Statement 12 Axial fuel expansion estimates nust be extrapolated frcm non-prototypic ;
experiments.
Response 12 he Applicants deny this statement. We experiments listed in response to No. 11 above are sufficiently prototypic for application to CRBRP.
Statenent 13 It is inpossible to quantify the confidence level associated with the '
calculated energetics of a IM event because considerable engineering judgment is involved in the application of SAS3A.
Response 13 ;
he Applicants deny this statement. It is not inpossible to quantify the confidence level associated with calculated energetics. Refer to CRBRP-1.
P Statement 14 2 e Staff and the Applicant utilize identical SAS3A codes.
Response 14 he Applicants cannot admit nor deny this statement. h e Applicants have perwided a copy of the SAS3A code and subroutines to the Staff, but have not verified that the code used by the Staff is identical to that used by AC-29
l the Applicants. Se SAS3D codes now used by the Applicant and Staff ;
consultants are known to not be identical.
gaternent 15 he Staff / Applicant (to each as appropriate) have checked the EURTRAN coiing and determined that their SAS3A code is identical with that of the Staff / Applicant (to each as w @riate).
Response 15 he Applicants deny this statement. he Applicants have provided the Staff with a copy of the SAS3A code ard its subrustines but have not performed a statement-by-statement conparison of the cade used by the Staff to verify that the Staff's/ Applicants' ocdes are identical.
Statement 16 l
%e Staff / Applicant (to each as appropriate) uses the VENLE-II code in its analyses of IM events.
l Response 16 he Applicants admit this statement.
Statement 17 Were is no formal or informal docunentation to demonstrate that VENUS-II or any of its modules behave in a stable fashion or proSuce accurate solutions to the original equation sets.
Respcmse 17 he Applicants deny this statement. Ibcunentation describing the nunerical stability of the VENUS-II code is provided in Reference 5 in CRBRP-GE:ER-00523. Also see response to Interrogatory Set 2 Item I(B) (6 and 7), (8).
AC-30
1 '
Statsnant 18 he details of the conputational algorithms associated with each of the models in VENUS-II and the code as a whole can only be discovered by l physically examining the FORTRAN coding acrnprising the code.
Response 18 he Applicants deny this statenent. Doctanentation of the acrtputational algorithus used in the VENUS-II code are cxntained in ANE,-7951. ,
Statenent 19 .
he Staff / Applicant (to each as appropriate) uses the PIITIO code in its analyses of IN events.
Response 19 Se Applicants admit this statenent.
i r
Statenent 20 '
mere is no formal or informal doctanentation to dertonstrate that PIITID or ;
any of its modules behave in a stable fashicn or produce accurate solutions to the original equation sets.
Response 20 Se Applicants deny this statenant. Ibetanentation describing the ntrnerical ;
stability of PIIIIO is identified in Applicants' response to Interrogatory Set 2, Item I(C), 4, 6, and 7 and Set 6, Item I(C) 2.
Statement 21 he details of the cartputational algorithms associated with each of the I models in PIITIO and the code as a whole can only be discovered by phys- ,
ically examining the PIITIO coding ocmprising the code, l 9
AC-31
i i
Response 21
'Ihe Applicants deny this statement. Doctrnentation of the emputational algorithns used in PIUID are contained in reference E-2 and E-3 in CRBRP-GEFR-00523.
Statement 22 For each of the following mdes, subroutines, or cmputational packages identified as (A) through (S) below, respond to the admissions identified as (a) through (k):
(A) SAS3A (K) SIEX (B) SAS/ECI (L) HOPE (C) FXVARI (M) Source Subroutine for VENUS (D) SILMPY (N) VENUS-II (E) PIUIO (O) HAAFM-1 (F) CIAZAS (P) HAAIN-2 (G) TS000L (Q) HAA-3 (H) SSFUEL (R) REXCD-1EP (I) DEEDRM-II (S) Decay Heat (J) CDMRAIEX-II
- a. 'Ihe code (subroutine, or cmputational package) has not been inde-pendently validated by the Staff / Applicant (to each as appropriate) to deTonstrate that the code's output is the correct nirnerical calculation that should result frm a given set of input data and model asstruptions.
- b. 'Ihe code (subroutine, or cmputational package) has been independently validated by the Staff / Applicant (to each as appropriate) to demonstrate that the code's cutput is the correct nirnerical calculation that should result frm a given set of input data and model asstsuptions.
c.
'Ihe code (subroutine or cmputational package) has been independently validated by the Staff / Applicant (to each as appropriate) to demonstrate ,
that the code's cutput is the correct nirnerical calculation that should result frm a given set of input data and model asstmptions, but the 7
AC-32
l i
l \
i l Staff /Ag>licants' (to each as appropriate) validation is not sufficiently rh'==nted to enable an outside independent reviewer to confirm that the i validation was adequately performed.
- d. 'Ihe code (subroutine, or emputational package) has not been validated against actual experimental data.
- e. 'Ihe code (subroutine, or computational package) has been validated against experimental data, but the experimental data is insufficient to determine the accuracy of the cade, or nodel.
- f. 'Ihe code (subroutine, or conputational package) has not been validated against experimental data to dancnstrate that the code, or nedel, can be exterded to the CRBR CM's analysis.
9 'Ihe code (subroutine, or canputational package) has been validated against experimental data to denenstrate that the code, or nodel, can be extended to the CRBR, but the experimental data is insufficient to deter-mine the accuracy of the code, or nodel, when applied to the CRBR.
- h. All the input data used in the codes (subroutine, or canputational package) for purpcses of assessing CRBR CDA's, has not been validated to denonstrate that the cade or nodel results will not result in either an underestimate of the CDA work potential, or an overestimate of the CRBR contairmnent capability with respect to a CDA.
- 1. 'Ihere is no doctynentaticn to confirm that the code's (subroutine's or conputational package's) output is a correct runerical calculation that should result in a given set of input data and model asstmptions.
- j. 'Ihere is no doctanentation to verify that the code's (subroutine's or ocmputational package's) ntunerical algorithns have been sugam.ed cor-rectly.
AC-33 l
e I
- k. It formal doctznentation of the code exists other than an input de- '
scription which is an integral part of the code in that it is contained cm PORTRAN-language (or other cxrnputer language) cxmment cards. [
I Response 22 '
(A) SAS3A (a) The Applicants deny this statement.
(b) The Applicants admit this statement. Refer to Applicant's }
response to Interrogatory Set 2, Item I(A)-4 and Applicant's i response to Interrogatory Set 6, Item I(A)-3. i (c) The Applicants admit this statement.
i (d) 'Ihe Applicants deny this statement. Refer to the response to I Interrogatory Set 2, Item I(A)-6 & 7. !
p (e) The A,rplicants deny this statenent. Refer to (d) above. , !
(f) The Applicants deny this statement. Refer to (d) above. ,
(g) 'Ihe Applicants deny this statement. See sutpart f of this respotme.
(h) The Applicants deny this statement. Refer to (b) and (d) above. l (i) 'Ihe Applicants deny this statenent. Refer to (b) and (d) above. i (j) The Applicants admit this statement. i (k) 'Ihe Applicants deny this statement. Refer to Applicants response l to Interrogatory Set 2, iten I(A)(1).
i
\
(B) SAS/ECI (a) The Applicants deny this statement. i (b) 'Ihe Applicants achtit this statement. Refer to response No.
22(A)b.
(c) 'Ihe Applicants admit this statement.
(d) 'Ihe Applicants admit this statement. 'Ihe verification of SAS/FCI, I fuel coolant interaction nodel has been based cm ocmparisons with l the PUTIO 2 code which has been validated with experiments. t I
(e) 'Ihe Applicants deny this statement. Refer to (b) and (d) above. !
l !
l
! sc-34 ;
i i
i
l (f) The Applicants deny this statement. Fefer to (d) above and to response to Interrogatory Set 2, Item II(9-20), (SAS/ECI Stmmary) for additional subroutine doctrnentation and use for CRBRP QA analysis.
(g) The Applicants deny this statement. Refer to (d) and (f) above.
(h) 'Ihe Applicants deny this statement. Refer to (b) ard (g) above. ;
(i) The Applicants deny this statement. Refer to (d) and (f) above.
(j) The Applicants admit this statement.
(k) The Applicants deny this statenent. Refer to (b) and (d) above.
(C) FXVARI (a) The Applicants deny this statement.
(b) 'Ihe Applicants admit this statement. Refer to response to Interrogatory Set 3, Item I(A)-4.
(c) 'Ihe Applicants admit this statement.
(d) The Applicants admit this statement. See Applicants response to Interrogatory Set 3, Item I(A)-6 & 7.
(e) The Applicants deny this statement. Refer to (d) above.
(f) The Applicants a& nit that the code has not been validated against experimental data but deny that it cannot be extended to the CRBRP CIA analysis (refer to response to part (d) above).
(g) The Applicants deny that the code has been validated against experimental data but a& nit that it can be extended to the CRBRP.
(h) The Applicants deny this statement. Refer to (b) and (d) above.
(i) 'Ihe Applicants admit this statement.
(j) 'Ihe Applicants admit this statement.
t (k) 'Ihe Applicants admit this statenent. Refer to Applicants response to Interrogatory Set 3, Item I(A)-1.
l (D) S[1 MPY l
(a) The Applicants deny this statement.
l AC-35 I
i*
(b) he Applicants admit this statement. Refer to response No.
22(B)-b. Further doctanentaticn and analysis is presented in the l Applicants' response to Interrogatory Set 2, Item II(63-69) SIUMPY Saman.
(c) The Applicants admit this statenent.
(d) The Applicants deny this statement. Refer to Applicants' response to Interrogatory Set 2, Item II (63-69), SU.NPY Suntnary.
(e) The Applicants deny this statement. Refer to response (d) above.
(f) The Applicants deny this statement. Refer to response (d) above.
(g) The Applicants deny this statement. Refer to response (d) above.
(h) The Applicants deny this statement. Refer to response (d) above.
(i) The Applicants deny this statement. Refer to response (b) ard (d) above.
(j) The Applicants admit this statement.
(k) The Applicants deny this statement. Refer to response (b) and (d) above.
(E) PIUIO (a) The Applicants deny this statenent.
(b) The Applicants adnit this statement. See Applicants' responses to Interrogatory Set 2, Item I(c)-4 and Set 6, Item I(c)-3.
(c) The Applicants adnit this statement.
(d) %e Applicants deny this statement. Refer to response to Interrog-atog Set 2, Itan I(c)-6 & 7.
(e) The Applicants deny this statenent. Refer to (d) above.
(f) he Applicants deny this statenent. Refer to (d) above.
(g) The Applicants deny this statement. Refer to (d) above.
(h) he Applicants deny this statement. Refer to (b) and (d) above.
(i) The Applicants deny this statement. Refer to (b) and (d) above.
(j) The Applicants adnit this statement.
(k) h e Applicants deny this statement. Refer to (d) above.
AC-36
(F) CIAZAS (a) The Applicants deny this statement.
(b) We Applicants admit this statement. See response No. 22(A) (b).
(c) The Applicants admit this statenent.
(d) he Applicants deny this statenent. We experimental verification of CIAZAS, the clad notion nodel in SAS3A has been established in the Applicants' response to Interrogatory Set 2, Item I(A)-6&7.
See also response to Interrogatory Set 2, Item II (60-62) for further doctznentation.
(e) The Applicants deny this statenent. Refer to (d) above.
(f) We Applicants deny this statenent. Refer to (d) above.
(g) The Applicants deny this sMtenent.
(h) The Applicants deny this statenent. Refer to (b) ard (d) above.
(i) The Applicants deny this statement. Refer to (b) and (d) above.
(j) We Applicants admit this statement.
(k) The Applicants deny this statement. Refer to (b) and (d) above.
(G) TSCOOL We Applicants can neither admit nor deny the truth of admission statenents (a) - (k) since TS000L, a subroutine in SAS1A, is not explicitly incorporated in its entirety into the SAS3A, as stated by the Applicants in their response to Interrogatory Set 6, Item I(A)-1, "Likewise, a part of the codiry fran subroutine TSCDOL in SASIA, went into subroutine TSCl in SAS2A." Subroutine TSCl was incorporated in its entirety into SAS3A.
(H) SSEUEL We Applicants' response for admission statenents (a) - (k) are the same as (a) - (k) in 22 (F). .
(J) CXHRADEX - II (a) The Applicants deny this statement.
AC-37
(b) The Applicants admit this statement. In their response to Interrogatory Set 5, Item I(A)-4, 5, the Applicants stated 4.) Independent hand calculations were made to check interms$iate calculations of the code. Simpson's Rule integration routines were checked by otmparison with m1culations which can be reproduced analytically.
5.) All COfRADEX programning has been checked as indicated in (4) above to verify that the code performs the correct numer-ical calculations." !
(c) he Applicants admit this statement.
(d) The Applicants deny this statenent. Refer to Applicants' response to Interrogatory Set 5, Itans I, II, & III.
(e) - (k) The Applicants deny these statements. Refer to (d) above.
(K) SIEX (a) The Applicants deny this statement.
(b) %e Applicants adnit this statement. Se SIEX code has been validated to dancnstrate that the code output is the correct ntrnerical calculaticn that results fran a given set of input data and nodel asstrnptions (see Reference 1).
(c) %e Applicants deny this statenent. We code is sufficiently docunented (Reference 1) to enable an outside independent reviewer to confirm that the validaticn was adequately performed.
(d) The Applicants deny this statenent. Se SIEX code is based cm experimental data as shown in Reference 1.
(a) The Applicants deny this statement. Refer to (d).
(f) We Applicants deny this statenent. We SIEX code has been validated against experimental data in Reference 1 to demonstrate that the code is applicable to CRBRP CIR analyses.
(g) The Applicants deny this statement. Refer to (f).
(h) h e Applicants deny this statenent. Se SIEX code was not used in the analysis of CIA events W11ch resulted in the most severe CIR work potential.
(i) he Applicants deny this statement. Reference 1 shows that the SIEX code output is the correct numerical calculaticn that should result fran a given set of input data and nodel assunptions.
AC-38
l i
i (j) The Applicants deny this statement. %e SIDC code numerical l algorittrns have been correctly pwanmM as shown in Reference 1.
l (k) %e Applicants deny this statement. Refer to (j).
(Q) HAA-3 (a) The Applicants deny this statenent.
(b) The Applicants admit this statement. Refer to Applicant's response to Interrogatory Set 5, Item I(B)-4, 5.
(c) The Applicants deny this statement. Refer to Applicant's response to Interrogatory Set 5, Item (b)-7.
(d) Se Applicants deny this statenent. Refer to (b) and (c) above.
(e) - (h) The Applicants deny these statenents. Refer to (b) and (c) above.
(i) - (k) The Applicants deny these statements. Refer to (b) and (c) above and Items I, II, and III in Interrogatory Set 5.
(I) DEEDIN - II (L) HOPE (S) Decay Heat *
(O) HAAIN - 1 (P) HAAIN - 2 he Applicants are not oognizant of items (I), (L) and (S) and do not utilize itens (O) arri (P) in CRBRP analysis. h erefore, the Applicants can neither admit nor deny the truth of statements (a) - (k) with respect to these ccmputer codes.
(M) Source Subroutine for VIME (EXPAND)
Se Applicants asstune that the source subroutine for VDUS which is mentioned in this admission refer to the Irogram EXPAND. We DCPAND ocde is used to conpute the area under the Pressure Voltane (P-V) curve that would result fran isentropically expanding the fuel ternperature distribu-tion that results fran a VENUS-II calculation down to 1 atmosphere. Since the EXPAND code is a sinple mathematical function, its verification has not been formally documented. A copy of this code will be made available for inspection and copying.
(a) The Applicants deny this statement.
- It is asstuned that " decay heat" is a ocmputer code, subroutine or ccmputational package.
AC-39
l (b) The Applicants admit this statement. Se EXPAND code has been validated against hand calculations.
(c) Se Applicants admit this staternent.
(d) The Applicants admit this statement; however, given that the code performs a simple well defined mathematical calculation based on first grinciple, it is not necessary to validate the code against experimental data.
(e) The Applicants deny this statement. See response to (d) above.
(f) Se Applicants admit that this code has not been validated against experimental data but deny that the nodel cannot be extended to CRBRP.
(g) The Applicants deny that the code has been validated against experimental data but admit that it can be extended to CRBRP.
(h) The Applicants deny this statement. Refer to (b) and (d) above. ,
(i) he Applicants admit this statenent.
(j) - (k) The Applicants admit these statements.
(N) VENUS-II (a) The Applicants deny this statement.
(b) he Applicants admit this statenent. See Applicants' responses to Interrogatory Set 6, Item I(B)-2, and Set 2, Item I(A)-4.
(c) Se Applicants admit this statement.
(d) The Applicants deny this statement. Refer to Applicant's response to Interrogatory set 2, Item I(B)-6 & 7.
(e) The Applicants deny this statement. Refer to (d) above.
(f) he Applicants deny this statement. Refer to (d) above.
(g) he Applicants deny this statement. See subpart (f) of this response.
(h) he Applicants deny this statement. Refer to (b) and (d) above.
(i) Se AEplicants deny this . statement. Refer to (b) and (d) above.
(j) S e Applicants admit this statement.
(k) Se Applicants deny this statenent. Refer to (d) above.
AC-40
(R) REXO>4EP (a) Applicants deny this statment. !
(b) Agplicants a& nit this statement. Refer to Applicants' response to Interrogatory Set 3, Item I(B), (4).
(c) The Applicants deny this statment.
(d) The Applicants deny this statement. Refer to the Applicants' response to Interrogatory Set 3, Item I(B), (7).
(e) The Applicants deny this statement. Refer to (d) above.
(f) The A;plicants deny this statment. Refer to (b) and (d) above.
(g) The Applicants deny this statenent. Refer to (b) and (d) above.
(h) The Applicants deny this statement. Refer to (b) ard (d) above.
As stated in the Applicants' response to Interrogatory Set 3, Item I(b),(6),
In the few areas Where experimental validation does not exist at this time, issues were resolved by makirg conservative asstmptions. '
(i) The Applicants deny this statement. Refer to (b) and (d) above. *
(j) 'Ihe Applicants deny this statement.
(k) The Applicants deny this statement. Refer to (d) above.
Statenent 23
'Ihe models and input assunptions used in the PSAR (including the latest versions of SAS and VDKE) are not adequately doctanented, verified, or validated by ocmparison with applicable experimental data, to scientif- [
ically determine that the upper limit, cr upper bourd estimate cn the work energy release in a I& event is less than:
(a) 1 Mj '
(b) 10 Mj (c) 100 Mj (d) 500 Mj (e) 660 Mj .
t (f) 1200 Mj AC-41
(g) 2000 Mj (h) 10,000 Mj Response 23
'Ihe Applicants deny this staternent. In so denying, the Applicants assume that "to scientifically determine" means "to determine with reasonable assurance." As stated in the Sections 2. and 4., Interrogatory Sets 2, 3, 4, 5, and 6, and Admission Response No. 22, the Applicants have used models and input assunptions Wiich have been doctanented, verified, and validated to determine that the work energy release accrrnpanyirg the hypothetical postulate of core disruptive accidents is low. However, in keeping with the safety design philosophy adopted for the plant, structural design margins to acocmrodate a 661 Mj energetic event have been included in the design.
Statenent 24 Since the upper internals structure (UIS) would absorb energy in deforming under an ICDA load, the response of the reactor closure head to an ICDA mechanical loading would not be altered if the UIS had been considered in the analysis.
Response 24
'Ihe Applicants deny this statenent. In terms of the pressure pulse applied to the head. the energy absorption in the upper internals structure would attenuate the loads. 'Iherefore, neglecting the head 1<wls provides a conservative representaticm of those loads.
Statenent 25
'Ihe response of the reacttr closure head to an HCDA mechanical loading would be altered if the UIS were considered in the analysis.
, AC-42 I
O Response 25 he Applicants admit this statenent. As discussed in response No. 24, neglectire the effect of the UIS results in a conservative assessment of loads cn the reactor closure head.
Statement 26 Scaling laws are applicable to systems such as the reactor vessel closure hea3 only whm material properties such as the strain rate, and the temperature effects on such properties are well known.
Response 26 h e Applicants admit this statement. However, in situaticms where material properties of the models such as strain rate and tamperature effects tend to be self-conpensating, the absolute values of these quantities need rot t.a as well known. -
Statement 27 Se shear ring is incapable of retaining the reactor vessel head for the 661 MJ IEDA car .
Respcmse 27 he Applicants deny this statement. In Applicants' answer to NIC question 001.540, the statement is made "the reactor vessel head, including the shear rirgs and riser bolts has margins to auu,mudate all of the loads analyzed, with a substantial margin to failure." me loads analyzed included the 661 MJ case.
Statenent 28 he NRC's analytical model used to assess the structural integrity of the reactor closure head is more accurate than the Applicant's analytical model.
AC-43
i 1
Respcnse 28
'Ihe Applicants neither admit nor deny this statenent since the Applicants are not cognizant of NRC's analytical nodel. ,
l 1
Statenent 29
'Ihe NRC's analytical nodel used to assess the structural integrity of the reactor closure head is less accurate than the Applicant's analytical model.
Response 29
'Ihe Applicants neither admit ror deny this statement. Refer to response No. 28.
Statement 30
'Ihe NRC's analytical model used to assess the structural integrity of the reactor closure head gives results that are contrary to the Applicant's results.
Response 30
'Ihe Applicants neither admit nor deny this statement. Refer to response No. 28.
AC-44
AINISSICNS REIATIVE 'ID (ORIGDRL) COtfITNTICN 5 (PARTIAL)
Statenent 20 It is possible that a CIR could be initiated by sabotage conducted by one or two insiders wto could gain access to vital equipnent.
Response 20
'Ihe Applicants admit this statenent to the extent that is is not inpossible that a C3R could be initiated by sabotage conducted by one or two specially trained insiders who could gain access to vital equipnent. Ibwever, as ,
discussed in detail in the Applicants' response to Interrogatory Set VIII (question 10) it is highly inprobable that this could occur.
Statenent 21 It is possible that a CIA could be initiated by sabotage conducted by a :
group of outsiders aided by one or two insiders. ,
Response 21
'Ihis response is the same as the response to No. 20.
1
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AC-45
REQUEST POR ADMISSIONS REIATIVE 'IO CatfrENTION 10 l
Statenent 1
'Ihe enclosure 4 attached to Applicants' Response to NRDC, g a . 'IWelfth Set of Interrogatories to the Applicants accurately describes the criteria Wich ERIA ertployees were directed to utilize in conducting the analysis of the 24 identified ERIA sites.
Response 1 The Applicants deny this statenent. Ihclosure 4 to the Applicants' Response to NRDC, g 'al. 12th Set of Interrogatories to the Applicants provided guidance to the ERIA Field Offices to assist thc= in identifying and supplyire the site specific information an the enumerated 29 sites which the CRBRP Project Office required in connection with the alternative site analysis.
h Ac-46
1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION -
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In the Matter of ) -
DEPARTMENT OF ENERGY )
DOCKET NO. 50-537 '
PROJECT MANAGEMENT CORPORATION -
) .
TENNESSEE VAL' LEY AUTHORITY .
) _
( AFFIDAVIT OF PAUL _W. DICKSON, JR. .-
baing duly sworn,' deposes and saye as follows:
- 1. That he is employed by Westinghouse Electric Corporat' ion _
as 'rechnical Director, Clinch River Rrceder Recctor Project, Westinghouse ldvanced Reactors Division, Post Office Box W, Orek Ridge, Tennessee 37830
- 2. That he i,s duly authorized to execute the responses on behalf of the applicants to the NRDC's August 13, 1976, september 16, 1976, and January 14, 1976 request to Applicants for Admission concerning dontention _
2._ _
- 3. That the above-mentioned and attached answers are'true and correct to the best of his knowledge'and belief. .
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SIGNA'1TRE
, SUBSCRIBED and SWORN to before me this ff d .
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, 1982.
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My commission expires _
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M) Commission Expires April 28.196d 0019-9E9 Sid 3DOIM >100 dM883 IND v1:03 08E/Wr0 l
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PWWECT MANAGOST C0APGaATION )
TBWESSEE . VALLEY AUTHORITY ) -
l NFI0AVIT F 00 MIS M. SWITICK -
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l Bennis M. Shritick, beleg 61y morn, deposes and says as follows:
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- 1. That he is uployed by the General Electric Company as Manager, safety '
Analysis, Advanced Reactor Systems Department, 310 De Guigne Drive, Sungrvale, Calffornia M006.
2.
That he is 61y othertred to execute the responses on behalf,of the Applicants to the MDC's September 30. 1976 request to Applicants for l i
A mtssion (numbers 1. 4-10, 13, 15, 19, 22K, L and 23) concerntng Centention 3.
3.
That theof the best above-mentioned his knowledge andand attached amissions are true and correct to belief. .
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DOCKET NO. 50-537 (
l PROJECT MAllAGDUT CORPORATION ) i TeslESSEE VALLEY Nml0RITY ,)
N F10AVIT F L. NALTER DEITRICH ,
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l L. Unitor Beitrich, being 41y morn, deposes and says as follows:
1s That he is amployed hy the Reactor Analysis and Safety Division of National Laboratory, 9700 So Cass Avenue. Argonne, Illinois 39, as Associate Divisten Director. .
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- 2. That he is ely authorized to emacute the responses on behalf of the Applicants to the NIBC's September 30, 1976 request to Applicants for Aerission Inumbers 2, 3, 11, 12, 14, 15, 17, 18, 20, 21, and 22 (A-1), M.
B) concerning Contention 3. !
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That the above-mentioned and attached a M ssions are true and correct to the best of his Itaculedge and belief, j i
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, AFFIDAVIT OF STANLEY H. FISTEDIS,
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.t Stanley N. Fistdis, being duly sworn, deposes and sars as follows-
- 1. That he is agleyed by the Reactor Analysis and Safety Division of -
Nattomal Laboratory, 3700 So. Cass Avenue, Argonne Illinois j
, as Manager of the Engineering Mechanics Program. !
- 2. That he is duly authorized to execute the responses on behalf of the l App 11 cants to the NROC's Septader 30.1976 request to App 1tcants for l Admission (ander 22 (A)) concerning Contention 3.
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T w m VALLEY AtmlORITY ) .g 0
. A T10AVIT F LEE E. STRAWBRIDGE ,
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- Lee E. Stradridge, being 41y morn, deposes and-stys as fo11ews:
1.-
That he is employed by-the Westingho se Adv.anecd Reactors Olvision as Nanager, thclear safety ad Licensing, P. O. Box 158, Madison, j Pennsylvania 15663. ,
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- 2. That he is 61y authorized to execute the responses en behalf of the
%11 cants to the HRDC's Septamtwr 30,1976 request to Applicants for
- 9 Admission (sunners 21 J, 0, P, Q, 5, and 24-30) concerning Contention 3.
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That the best theofabove-eentioned his knowledge andand attached admissions art true and correct to belief. I
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UNITED STATES OF AMERICA .
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NUCLEAR REGULATORY COMMISSION In the Matter of -) .
DEPARTMENT OF ENERGY ) ,
DOCKET NO. 50-537 ~
PROJECT MANAGEMENT CORPORATION }
~
TENNESSEE VALLEY AUTHORITY ) , ,,
AFFIDAVIT OF _ PAUL W. DICKSON,'JR. _ _ _
b31ng duly sworn., deposes and says as follows:
- 1. That he is employed byWestinghouse I:lectric Corporation _
cs Technical Director, Clinch River Breeder Reactor PYOject, Westin9 house Advanced Reactora Divi si on, Post Of fice Box W, Oak Ridge, Tennessee 37830
- 2. That he is duly authorized to execute the responses on behalf of the applicants to the NRDC's August 26, 1976 request to Applicants for Admission (number 20 and 21) concerning Contention'5.
3'. That the above-mentioned and attached answers are true and correct to'the'best of his knowledge and belief. -
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SIGNATURE , J, Y '
SUBSCRIBED and SWORN to before me this E Fd ay of AN/ , 1982'.* ~ -
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My commission expires _ , 19 ___ .
My Commission Dpires Apm 28, TM 001'9-939,51d 3DOIM >lWO dM8831WD G1:03 083/Wr0
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U. 5. DEPARTMENT OF ENERGY ., )
DOCKET NO. 50-537 PROJECT MANAGEMENT CORPORATION )
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TENNESSEE VALLEY AUTHORITY ,
.A_E,FIDAVIT OF LAWRdNCE J. KRIPPS
/ Lawrence J. Kripps, being duly sworn, deposes and says as follows:
- 1. That he is employed as a safety and Environmental Engineer, of Energy Incorporated, and that he is duly authorized to answer the response
. on behalf of th,e Applicants to the NRDC's Admissions to old Contention 10, Admission nunter 1., dated January 14, 1977.
- 2. That the above-mentioned and attached answers are true and correct to the best of his knowledge and belief. ,
~
. e . pps SUBSCRIBt0 and SWOffi td pefore me .
this /fL~O_. day of //4mZ . 1982. -
, gNDTARY PUBLIC My Commission Exp!tc3 April 28,1934 0019-9E9 Sid 3D0I8 >lWO dM883 IND #E:2T 062/Wr0 r
.gj','19 si i
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l
)
In the Matter of )
)
UNITED STATES DEPARTMENT OF ENERGY )
)
PROJECT MANAGEMENT CORPORATION ) Docket No. 50-537
)
TENNESSEE VALLEY AUTHORITY )
)
(Clinch River Breeder Reactor Plant) )
)
CERTIFICATE OF SERVICE Service has been effected on this date by personal delivery or first-class mail to the following:
- Marshall E. Miller, Esquire Chairman Atomic Safety & Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C. 20525 Dr. Cadet H. Hand, Jr.
Director Bodega Marine Laboratory University of California P. O. Box 247 Bodega Bay, California 94923
- Mr. Gustave A. Linenberger Atomic Safety & Licensing Board l U. S. Nuclear Regulatory Commission l Washington, D. C. 20545 -
- Daniel Swanson, Esquire l *Stuart Treby, Esquire
! Office of Executive Legal Director U. S. Nuclear Regulatory Commission i Washington, D. C. 20545 (2 copies)
I
- u. ., . . ,
-2.
- Atomic Safety & Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C. 20545
- Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C. 20545
- Docketing & Service Section Office of the Secretary U. S. Nuclear Regulatory Commission Washington, D. C. 20545 (3 copies)
William M. Leech, Jr., Attorney General William B. Hubbard, Chief Deputy Attorney General Lee Breckenridge, Assistant Attorney General State of Tennessee Office of the Attorney General 450 James Robertson Parkway Nashville, Tennessee 37219 Oak Ridge Public Library Civic Center Oak Ridge, Tennessee 37820 Herbert S. Sanger, Jr., Esquire Lewis E. Wallace, Esquire W. Walter LaRoche, Esquire James F. Burger, Esquire Edward J. Vigluicci, Esquire Office of the General Counsel Tennessee Valley Authority 400 Commerce Avenue Knoxville, Tennessee 37902 (2 copies)
- Dr. Thomas Cochran Barbara A. Finamore, Esquire Natural Resources Defense Council 1725 Eye Street, N. W., Suite 600 Washington, D. C. 20006 (2 copies)
Mr. Joe H. Walker 401 Roane Street Harriman, Tennessee 37748 Ellyn R. Weiss Harmon & Weiss 1725 Eye Street, N. W., Suite 506 Washington, D. C. 20006
' 4 , c. , . . ,
Lawson McGhee Public Library 500 West Church Street
. Knoxville, Tennessee 37902 William E. Lantrip, Esq.
Attorney for the City of Oak Ridge Municipal Building P. O. Box 1 Oak Ridge, Tennessee 37830 Leon Silverstrom, Esq.
Warren E. Bergholz, Jr., Esq.
U. S. Department of Energy 1000 Independence Ave., S. W.
Room 6-B-256, Forrestal Building Washington, D. C. 20585 (2 copies)
- Eldon V. C. Greenberg Tuttle & Taylor 1901 L Street, N. W., Suite 805 Washington, D. C. 20036 Commissioner James Cotham Tennessee Department of Economic and Community Development Andrew Jackson Building, Suite 1007 Nashville, Tennessee 37219 A
Geor M . Edg F Attorney for Proj ect Management Corporation DATED: April 30, 1982
- / Denotes hand delivery to 1717 "H" Street, N.W., Washington, D. C.
- / Denotes hand delivery to indicated address.