ML20052B611
| ML20052B611 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 04/28/1982 |
| From: | Deitrich L, Fistedis S, Switick D JOINT APPLICANTS - CLINCH RIVER BREEDER REACTOR |
| To: | National Resources Defense Council, Sierra Club |
| References | |
| NUDOCS 8205030371 | |
| Download: ML20052B611 (163) | |
Text
4/28/82 c
82 IF 28 N ~ "
UNITED STATES OF AMERICA NUCLEAR REXIIIA'IORY CCPHISSION In the Matter of
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UNITED STATES IEPAR'IMENT CF ENERGY
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Docket No. 50-537 PROJECT MANAGEMENT CORPORATION
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TDNESSEE VA11H AUDERITY
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D Clinch River Breeder Reactor Plant
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iwce/g 3-J 1932w Te APPLICANPS' UPDATED RESPONSE #1 'IO Nmm mmms DEFmSE cem, mC.
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AND SIERRA CWB INTERROGA'IDRIES (SIDCND, dN
'DIIRD, EDURIE, FIFIH AND SIX'IH SETS) d3 Pursuant to 10 CER paragraph 2.740b, and in accordance with the Board's Prehearing (bnference Order of February 11, 1982, the United States Department of Energy, Project Managenent Corporaticn, and the Tennessee Valley Authority (the Applicants) hereby update their responses to the Natural Resources Defense Council, Inc. and the Sierra Club Second, 'Ihird, Fburth, Fifth and Sixth Sets of Interrogatories to the Applicants, dated Decenber 23, 1975, Deceit >er 31, 1975, January 14, 1976, February 12, 1976 and April 7, 1976, respectively.
In these updated responses the following style has been utilized:
Pbt each set of interrogatories the Preamble to Questions has been set forth. 'Ihereafter, each interrogatory within the set has been restated and the updated answer provided. Certain of the answers are unchanged fran the responses initially furnished. Itwever, for convenience those tnchanged responses also have been set forth after the appropriate interrogatories.
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SETF II AA-1 hD\\
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'1he answers contained in this Updated Response #1 supercede all prior answers to the interrogatories as to which they are applicable.
l In many instances, interrogatories specifically related to the previous parallel design covered in Appervlix F to the PSAR. Appendix F was withdrawn from the applicaticn in 1976. Applicants have attenpted in these updated answers to provide tpdated responses to those questions relating to Appendix F where such questions appear to Applicants to be potentially applicable to the current design. 'Ihis has :neant a substantial anant of additional effort by Applicants since the parallel design has not been the subject of attention by Applicants during the past five years and since the interrogatories needed to be interpreted in light of the current design.
Where Applicant believes the interrogatories are related to Appendi:: F and the previous parallel deslign and are not appropriately applicable to the current design Applicant has so noted.
e SET II AA-2
SECIND INTERROGA'1 DRY SET PREAPELE 'IO QUESTICNS With respect to the following requests for information we are concerned with four distinct validations relative to the nodels and cczn-puter codes:
1)
Validation that the code's output is the correct ntrnerical calculaticn that should result fran a given set of input data and the model assumptions; ii)
Validaticn of the nodels against actual experimental data; iii) Validation that the models can be extended to the CRBR; and iv)
Validaticn that the input asstrnptions for the CRBR case are adequate with respect to the CDA analysis, i.e.,
are support-ed by experimental evidence.
By " adequate", here arri below, we mean that the calculations will not tavlerestimate the CDA work potential (i.e.,
forces and resulting energetics of a CDA) or overestimate the containment capability of the reactor with respect to CDA.
QUESTICN I With respect to each of the following codes and each sub-routine of each of the followirx3 codes:
(A) SAS3A (including SASBIDK),
(B) VENUS, (C) PIDIO, please provide the following information [Where appropriate, the parts of the questicn have been restated to reflect the protocol for discovery agreed to by Applicants, Staff, and Intervenors NRDC et al.]:
1)
Ctznplete, current doctanentation (i.e., a writeup) of the codes and the subroutines:
SET II AA-3
2)
Identify, by name and affiliaticn, the author, or authors of each model, subroutine, or portion of each subroutine, which each contributed or worked on;
- 3) Identify by name affiliation (including organization, division, branch, title, etc.) each applicant employee, or consultant, that has intimate working knowledge of the code and each subroutine, or parts thereof, including its validity. Where more than cne person is involved, delineate which portion of the code or subroutine with which each has an intimate working knowledge;
- 4) Describe fully the procedures by which Applicant has assured itself and continues to assure itself, that the various computer programs (codes) accurately reprcx3uces the nodels as described in the PSAR and its refer-ences (see Validaticn (1) above);
5)
Indicate which models (including subroutines, or portions of subrou-tines) have not bem validated as described in Validaticn (i);
- 6) Indicate the models (incitxling subroutines, or portions of subroutines) or asstanptions that have not been validated as described in Validation (ii);
7)
For each nodel, portion of the nodel, or asstanption that has been validated (against experimental (or other) data, see Validation (ii) above) describe fully the procedure by which it was validated, and the results, including all uncertainties and limitation of the validaticn. Indicate the source of the experimental, or other data, that was used in the validation 8)
Explain fully all instabilities in the nirnerical performance in the models, d at causes them, and how they are avoided, and the extent to which this introduces uncertainties in the calculations and limits the validity of the nodel (cf., p.F6.2-10, para. 2).
9)
'Ib the extent that any answers to the above questions are based on referenced material, please supply the references; SET II AA-4
- 10) Explain whether Applicants are presently engaged in or intend to engage in any further research or work which may affect Applicants' answer. This answer need be provided cnly in cases where Applicants intend to rely upon cn goirg research not included in Secticn 1.5 of the PSAR at the DR or ocnstruction permit hearing cn the CRBR. Failure to provide such an answer means that Applicants do not intend to rely upcn the existence of any such research at the DR or construction pennit hearing cn the CRBR.
- 11) Identify the expert (s), if any, whcm Applicants intend to have testify on the subject matter questioned.
State the qualifications of each such expert. This answer need not be provided until Applicants have identified the expert (s) in question or determined that no expert (s) will testify, as long as such answer provides reasonable notice to Intervenors.
ANSWER I(A)
'Ihese answers provide the information requested relative to the SAS3A (includirg SASBIIE) ccmputer code ard have been revised to address the current application of SAS3D.
(1)
References 6, 7, 8, 9, and 10 on page 11-1 of CRBRP-GEFR-00523, "An Assessment of HCIA Energetics in the CRBRP Heterogeneous Reactor Cbre",
S.K. Rhow, et al., describe the SAS3A code, the fuel-coolant interaction nodel, the clad moticn model and the fuel moticn nodel in the SAS3A code.
'Ihe SAS3A sodium film notion model is doctznented in: G. Ibppner, "Soditra Flcw Moticn Ndel of SAS3A," ANL/ RAS 74-22, 1974. 'Ihe SAS3A primary loop nodel is doctrnented in:
Ref. 30 in CRBRPm3EFR-00103. '1he SAS3D code, now beirg used, evolved frcm SAS3A whidi evolved frcm the SAS2A code which evolved frcm the SASlA code. Doctrnentaticn for the SAS3A code is applicable to the SAS3D code. 'Ihe SAS2A code is doctznented in Reference 7 in CRBRP-GEFR-00523, and SAS1A is doctanented in ANL-7607, "SASlA, A Chnputer Cbde for the Analysis of Fast Reactor Power and Flow Transients," by D.R.
MacFarlane et al. The SASBIIE algorithms used in the SAS3A and SAS3D codes are doctrnented in CRBRP-GEER-00103, "An Analysis of Hypothetical Core i
i SET II AA-5
Disruptive Events in the Clinch River Breeder Reactor Plant", J.L. McElroy, et al.
(2) 'Ihe SAS3A and SAS3D codes are otrnplex code systems which have been developed over a period of years by the Reactor Analysis and Safety Div-ision of Argonne National laboratory. 'Ihe SASBIOK algorithm was developed by the General Electric Cbnpany. The principal contributors to the SAS3A and SAS3D code development are identified as authors of the references in Response 1.
(3) The following staff members of Argonne National Iaboratory and General Electric have a working knowledge of the codes, includirx3 their range of applicability and the efforts that have been made to validate then:
L.
Walter Deitrich, Associate Director, Reactor Analysis and Safety Division, Argonne National Iaboratory; David P. Weber, Manager, Accident Analysis Secticn, Reactor Analysis and Safety Division, Argonne National Iaboratory; Dennis M. 9,ritick, Manager, Safety Analysis, General Electric Advanced Reactors Systems Department.
(4) The entire SAS3A and SAS3D codes, including all subroutines, have been checked ard rechecked to assure that the ntznerical algorithns which are implemented in them to solve the equation sets which constitute these codes, behave in a stable fashicn (both individually ard collectively) ar.d produce accurate solutions to the original equation sets. 'Ihis was carried out by conparing SAS3A and SAS3D results with the output fran other codes, with the results of hand calculations, and with what sound engineering judgenent desned to be physically reasonable.
(5) All models have been validated as discussed in (4) above.
(6) - (7) The experimental basis for the SAS3A cxade as of April 1974 has been docunented in the paper, " Current Status and Experimental Basis of the SAS IJf1R Accident Analysis (bde Systen," Proc. Am. Nucl. Soc. Fast Reactor Safety Conf., Beverly Hills, California, CONF-740401, pp. 1303-1318.
SET II AA-6
Additional canparisons of the SAS3A code with experiments have been made since that time and are doctanented in the following references:
(1) Ref. 32 in CRBRP-GEFR-00103.
(2) Ref. 59 in CRBRP-GEFR-00103.
(3) Ref. 8 in CRBRP-GEFR-00523, pp. 54-62.
(4) Ref. 28 in CRBRP-GEFR-00103, pp.64-100.
(5) L. W. Deitrich, " Analysis of Transient Ebel Failure Machanisms, Selected ANL Programs, " Presented at the International Working Group on Fast Reactors Specialists' Meeting cn Fuel Failure Mechanisms, Seattle, Washington, May 11-16, 1975.
(6) E. Barts, et al., "Str mary and Evaluation, Fuel Dynamic Ioss-of-Flow Experiments (Tests L2, L3, and IA)," ANL 75-57, Septanber 1975.
The experimental basis for the SAS3A is applicable to SAS3D and additional experimental basis is doctraented in the following:
(1) Ref. 35 in CRBRP-GEFR-00523.
(2)
" Final Report on the SUSF In-pile Experiment P3A," T.E. Kraft and L.R. Kelman, ANL/ RAS 81-20, June, 1981.
(3)
W.A.
Ragland, "INFBR Ioss-of-Flow Simulations in the Soditrn Inop Safety Facility," ASME Paper 80-C2/NE-22, presented at the Century 2 Nuclear Engineering (bnference, San Francisco, Aug. 19-21, 1980.
It should be noted that many of the models used in SAS3A and SAS3D are par-ametric in nature and justificaticn for the particular parameters used in the analysis is given in CRBRP-GEFR-00103 and CRBRP-GEER-00523. Because of this parametric nature of the SAS3A and SAS3D codes, they can be used to draw valid conclusions relative to the course of hypothetical accidents in an INFBR eve though each subroutine may not have been ccznpletely validated by experiments, since parameters can be varied to determine the sensitivity of the results to variations in parameters.
(8) Mathernatically, practically all of the nodels in SAS3t' and SAS3D consist of sets of coupled ordinary differential or integrc>-differential SEP II AA-7
equations in time or of coupled partial differential equations in space and time.
Ntrnerically, these equaticn sets are salved by applying appropriate linearizaticn arx3 finite-differencing techniques.
Scme of these temporal finite-differencing techniques are fully inplicit arx1 are uncanditionally stable.
Other rmdels, such as that which treats the time-dependent radial heat transport fran the fuel pin into the coolant, have their equation sets solved by seni-implicit tanporal finite-differencing techniques.
It is well known that solutions cbtained by semi-inplicit differencing can exhibit bounded oscillations if time steps which are too large are taken.
'Ihirdly, scme equation sets, such as the SLUMPY a:rnpressible hydrodynamics equations, are solva3 with fully explicit methods. Here, taking time steps that are too large can Iroduce solutions which becane unstable.
'Ihroughout SAS3A and SAS3D provisions have been made to insure that the time step sizes being used for advancing the various solutions in time are kept sufficiently small so that the solutions behave stably and are ac-curate. %ese time step sizes are chosen by nonitoring both the solutions and their t
- w of change and applying step size selection criteria based on both known analytical constraints, Where they are available, and cn experience gained in aIplying the code to a variety of situations.
Wese step size selection criteria are explained in detail in the references provided in part 1 above.
It is still possible, however, to occasionally force a model in the SAS3A or SAS3D code to utilize a time step size which is so large that stability problens result.
It is also possible for the user to try to utilize SAS3A or SAS3D to analyze cases which are not intended to be modeled by SAS3A or SAS3D.
In these cases, the results predicted by SAS3A or SAS3D may tend to beccrne tnrealistic and physically meaningless.
Both of these problerns can and are generally dealt with by carefully scrutinizing the computer output and acrnparing it against engi-neering judgnent.
(9) Se reference doctanents have been or will be made available for inspec-tion and copying.
SET II AA-8
(10) 'Ihe Applicants are currently analyzirg this area ard have doeurnented the planned Irogram of research in Appendix A to CRBRP-3, Vol.1.
Appli-cants have not yet determined whether they will rely cn the results of future analysis.
(11) At the present time, the Applicants have not determined the experts, if any, whon they intend to have testify cn the subject matter questioned.
ANSWER I(B)
These answers Irovide the information requested relative to the VDRE conputer code.
(1) The VENUS-II code is doctznented in Ref. 5 in CRBRP-GEFR-00523.
(2) 'Ihe principal contributors to the VENUS-II code are identified as the authors of the reference in response 1.
(3) 'Ihe followirg staff matbers of Argonne National Laboratory have a working knowledge of the code, including its rarge of applicability and the efforts that have been made to validate it:
L. Walter Deitrich, Associate Director, Reactor Analysis and Safety Division, Argonne National Labora-tory, and David P.
Weber, Manager, Accident Analysis Secticn, Reactor Analysis and Safety Division, Argonne National Laboratory.
(4) 'Ihe entire VENUS-II code has bem thoroughly checked to assure that the equation sets and alcprithms given in Ref. 5 in CRBRP-GEFR-00523 are accurately gup.
mai into VENUS-II.
Because these equation sets are relatively simple, this was done by omnparing output frora the various subroutines against hand calculations.
(5) All models have been validated as described in (4) above.
1 (6) - (7) The VENUS-II code has been validated against the KIWI ' INT experi-ment and SNAPPRAN-2 AND !NAPTRAN-3 experiments. See: "Imgrdvanent and Verificaticn of Fast Reactor Safety Analysis Techniques," Progress Report, SEP II AA-9
Jan. 1, 1977 to Mar. 31, 1977, C00-2571-2, by Dee H. Barker, Terry F. Bott, Paul A.
Weeler, Iarry Larronica, and James F.
Jackson, Department of Chenical Engineerirg, Brigham Young University, Provo, Utah; also T.F. Bott ard J.F.
Jackson, " Experimental Q2nparison Studies with the VENLE-II Disassenbly Code," Proc. Intl. Mtg. cn Fast Reactor Safety and Related Physics, Chicago, October 1976, pg.1139.
(8) The numerical algorithn utilized in VENUS-II to solve the aanpressible hydrodynamics equations in twtxiimensional cylindrical gecmetry involves an explicit finite differencirg of the tenporal derivatives.
A satisfactory method has been inplemented in VENUS-II to control time step size so that its calculations renain stable and accurate.
This metha3 is sumnarized in Ref. 5 in CRBRP-GEER-00523 and is described in detail in the paper, J. F.
Jackscn, R. B. Nicholscn, and W. T. Sha, " Numerical Stability Problens in the VENUS Disassembly Cbde," Proc. of Cbnf. cn New Developnents in Reactor Mathematics and Applications, C NF-710302, Vol. 1, pp. 152-165, 1971. This reference will be made available for inspection and copying.
(9) The reference documents have been or will be made available for inspec-tion and copying.
(10) The Applicants are rot doing develognent work cn VENUS-II.
No such develognent work is currently planned.
(11) At the present time, the Applicants have not determined the experts, if any, what they intend to have testify cn the subject matter questioned.
ANSWER I(C)
'Ihese answers Irovide the information requested relative to tne PLUID 1 and PIITIO 2 Ccmputer Codes.
(1) References 21 in CRBRP-GEFR-00103 and 25 in CRBRPMEFR-00523 describe the PIITIO 1 ard PIDIO 2 Codes respectively.
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SEP II AA-10 l
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(2) he author of the PIUIO 1 and PIITIO 2 codes is Hartmut U. Wider, Asso-ciate Section Minager, Accident Analysis Section of the Reactor Analysis and Safety Division, Argonne National laboratory.
(3) The following staff mernbers of the Argonne National laboratory have a working knowledge of the code, including its range of applicability ard the efforts that have been made to validate it:
L. Walter Deitrich, Associate Director, Reactor Analysis and Safety Division, Argonne National laboratory and David P. Weber, Manager, Accident Analysis Section, Reactor Analysis ard Safety Division, Argonne National Laboratory.
(4) The PIRIO 1 and PIITIO 2 codes have been checked and rechecked to assure that the numerical algorithns which are implenented in then to solve the equation sets have been programned correctly.
FurtNnnore, test calcula-tions were perfonned to assure that these ntrnerical algorithns behave in a stable fashion and Iroduce accurate solutions to the original equation sets.
his was carried out by acrnparing PIUID 1 results with the output frcun another ocde (see Ref. 21 in GBRP-GE:FR-00523) with the results of hard calculations, ard with what sourd ergineering judgement deemed to be physically reasonable.
PIUIO 2 results have been ocznpared with PIDIO 1 results (see H. U. Wider, PIRIO 2:
A Ocznputer Cbde for the Analysis of Overpower Accidents in IMFBRs, TANSAO 27, p.
533, 1977) and with EPIC results (see H. U. Wider, et al., We PIUIO 2 Overpower Excursion Code and Ocznparison with IPIC, Proc. of the International Meeting cn Past Reaactor Safety Technologies, Seattle, 1979, p. 120).
(5) All models have been validated as described in (4) above.
(6) ard (7) %e soditan voiding rates calculated by PIUIO 1 ard PIUIO 2 strongly depend cn a few input parameters cxxcerning the fuel-ccolant interacticn ard the fuel pin pressures at the pin failure time.
Wese input parameters can be chosen such that voiding rates similar to those in
'11 TEAT in-pile experiments are calculated.
I 2e rapid sodium voiding Qtich occurred in the H4 'IREAT test was analyzed with PIITIO 1.
Wis led to important infonnaticn concerning fuel-coolant SEP II AA-ll
interacticn parameters.
(See, H. U. Wider and A. E. Wright, " Analysis of a Soditan Reentry Event in the IM TREAT Test," Trans. Am. Nucl. Soc., TANSAO 22, p. 428,1975.) PLUIO 1 has also been used to analyze the soditm voiding in the E8 and H6 tests.
PUJID 2 has also been used for analyzing part of the H6 TREAT test (see Ref. E-3 in CRBRP-GEFR-00523).
PUTIO 2 has also been used for analyzing the voiding in the L8 TREAT test (see Ref. E-4 in OtBRP-GEFR-00523 ).
(8) Mathematically, PUJIO 1 and PUTIO 2 consist of sets of coupled hyper-bolic partial differential equations which are of first order in tinn and of first and second order in space. All the equations but one are solved with fully explicit methods. Stability of the solution is ensured by using a time step Which satisfies the (burant criterion in all nodes.
'Ihe empressible calculation in the purely liquid scxlium slugs in PIUID 1 always requires the smallest time step and it is also very much the same for all times.
Therefore, no autcmatic time step control is necessary in PIUIO 1 and a constant time step which initially satisfies the Cburant criterion in the liquid soditan sltgs is beirg used in PUTIO 1.
In PIRIO 2, however, with its inompressible liquid sodium slugs, an autanatic time step control is enployed.
Durirg the developnent of the PUTIO 1 and PHTIO 2 codes it has been recog-nized that the explicit solution of the mcmenttan equaticn for the light soditrn/ fission-gas mixture can lead to instabilities of the mixture veloc-ity. This is causal by the action of two large but opposite forces (pres-sure gradient and drag) cn the light mixture. A semi-implicit solution of the soditzn/ fission-gas mcmenttu equaticn resolved the above-menticned, stability problen (see, Ref 9, CRBRP-GEFR-00523).
(9) 'Ihe reference doctrnents have been or will be made available for inspec-tion and copying.
(10) The Applicants develcynent work has been identified in Appendix A to OtNtP-3, Vol. 1.
SETF II AA-12
(11) At the present time, the Applicants have not determined the experts, if any, whcm they intend to have testify cn the subject matter questioned.
QUESTIONS II (General)
Request for the following information is based cn our concerns with respect to Validations (iii) and (iv) above.
In the Applicant's answers to the generic questions (b) and (c) below, the Applicant is requested to be responsive to these concerns.
With respect to each statement, assertion or assumption (fron Section EE.2 of the PSAR) identified belcw, please provide the following information (unless noted otherwise).
(NCTTE:
the follow: Lng nirnbered Interrogatories are identified by the page and/cr paragraph number frcm the PSAR in paren-theses). [Where appropriate, the parts of the question have been restated to reflect the protocol for discovery agreed to by Applicants, Staff, and Intervenors NRDC et al.]
a) Identify, by name, title and affiliation the primary Applicant en-playee(s) or consultant (s) that has the expen krruledge required to support the statement, assertion, or assumption.
b) Describe in detail the supporting evidence for the statement, asserticn, or asstnption and where appropriate the rationale for the approach taken.
c) Provide any additional informaticn requested following each statement, assertion, or assumption.
d) 'Ib the extent that any answers to the above questions are based cn referenced material, please supply the references.
e) Explain whether Applicants are Iresently engaged in or intend to engage in any further research or work whi& may affect Applicants' answer.
Identify such research cr work. 'Ihis answer need be provided only in cases j
where Applicants intend to rely upon cn going research not included in l
l Ser II AA-13
Section 1.5 of the PSAR at the IRA or construction permit hearing on the CRBR. Failure to provide such an answer means that Applicants do not intend to rely uptn the existence of any such research at the IRA or construction permit hearing an the CRBR.
f) Identify the expert (s), if any, whcm Applicants intend to have testify cm the subject matter questicned.
State the qualifications of each such expert. 'Ihis answer need not be provided until Applicants have identified the expert (s) in questicn or determined that no expert (s) will testify, as lcng as such answer provides reasonable notice to Intervenors.
ANSWERS II (General)
'Ibe following responses are identical for all interrogatories except where supplcnentary information is prtwided in responses II-l through II-69 below.
(a) See the attached affidavits.
(b) and (c) See responses ntubered 1-69 below.
(d) The reference doctments, except as otherwise noted hereinabove, have been or will be made available for inspecticn and copying.
(e) The Applicants' program of IED is identified in Section 1.5 of the PSAR.
Additional R&D work has been identified in CRBRP-3, Voltne 1, Appendix A for the SMBDB area and in CRBRP-3, Voltzne 2, Appendix A for the
'IMEE area.
l (f) At the present time the Applicants have not determined the experts, if any, whcm they intend to have testify on the subject matter questioned.
l IG'E: Questions II-l through II-3 pertain to SAS3A Introduction.
SET II AA-14 I
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l QUESTION II-l l
(EE.2-6, pr. 2)
'Ihe assunption that the core can be adequately repre-sented by te channels each containing one pin.
ANSWER II-l (b) and (c)
'Ihe SAS3D code acu mudates up to 34 channels while SAS3A is limited to ten channels. The core subassenblies are assigned to a group of subassemblies with similar neutronic, thermal, and hydraulic characteristics. Technical judgment is necessary in making the selections such that each member of a group may be expected to respond to a transient event in a similar way.
Each such group of subassemblies is assigned a " channel number" and given properties representative of all subassemblies in the group. 'Ihis procedure is standard engineering practice and is comonly used in analysis of nultiple parallel channel flow systes.
Ebr the effect of varying the nunber of channels on a similar system see:
Ref. 4 of CRBRP-GEFR-00523, and " Multi-channel Grouping Techniques for Conducting Reactor Safety Studies," A. E. Waltar, N. P. Wilburn; ANS Transactions, Vol. 22, November 1975, page 375.
QUESTION II-2 (EE.2-6, pr. 2)
'Ihe assurption that the solution can be adequately represented by point kinetics equations with reactivity feedback obtained by sunming over stationary fuel worth curves.
ANSWER II-2 (b) and (c)
'Ihe use of a point kinetics model with fuel displacement feedback obtained by stamirg over fuel worth tables is judged adequate, so long as small, local displacements are considered.
Gross relocation of fuel in large segments of the core can be addressed by recmputation of the fuel worth tables with a nulti-grotp diffusicri code when such reemputation is jtx3ged SElP II AA-15
necessary. See answer to question II(l)(a)-(c) in the Sixth Set of Inter-rogatories to the A plicant (p. AA-138).
GEErrIN II-3 (P6.2-6, par. 3)
The assumption that the iteration algorithn linking the I
SAS3A subroutines adequately represent the time sequence of events during a CIA. In this regard it is noted in para. 5 that, "At the present time, the claddiry and fuel moticn nodules cannot operate simultaneously in the same channel with the ECI nodel, etc.
ANSWER II-3 (b) and (c)
The two distinct nodes of failure identified by the terms "Sitraping" and
" Fuel-Coolant Interaction" are the mutually exclusive extremes of a mn-tinuous spectrtan of failure nodes. There is an area between these extrees where it is recognized that neither of these nodels precisely pra31 cts the physical h
I encmena.
'Ib handle this area with Iresent analysis tools requires technical judgment to ensure that limiting conditions for the available subroutines are applied to bound the resulting mergy release.
SAS3A and SAS3D are sufficiently flexible to control selection of the appropriate nodel cr introduction of results frcm external parallel calcu-lations, as necessary to assure that the resulting energy release is bounded.
NOTE: Questions II-4 through II-8 pertain to the Ebel Pin TOP Pailure Model.
0 2STIN II-4 (P6.2-7, par. 2) 1he asstunption that the deterministic " burst" nodel suggested by Stuart and fornulated for SAS3A by Snith in an' adequate representation of fuel pin failure phencznena.
l SET II AA-16 l
ANSWER II-4 (b) and (c)
'Ihe deterministic burst nodel suggested by Stuart and formulated for SAS3A and SAS3D by Stnith is a mathenatical model of fuel pin failure developed frun analysis of 'IREAT transient overpcwer experiments.
'1he behavior of many of the CRBR mixed oxide fuel pins durity the hypothetical transient overpower accidents is generically similar to fuel pin behavior during the transient overpower experiments fran which the mechanistic burst nodel was developed. Uncertainties which may exist in the fuel pin failure nodel are accounted for by performing parametric calculations in which pin failures are conservatively forced at the core midplane.
'1he applicants are cur-rently evaluating preliminary results fran the Scrlium Icop Safety Facility W-2 test as they may be applicable to the hyInthetical 'IOP accident analysis.
QUESTPICN II-5 (EE.2-7, Par. 2)
The Statenent "'Ihe slope of the cladding strength as a function of tenperature significantly influences the degree of bias in pin failure toward the upper part of the pin."
(c)
What is meant by "significantly"?
How is pin failure influenced by inhanagenieties in the fuel and cladding, e.g., pin hole leaks, fabrication errors, corrosicn, swelling, fuel-cladding gap ccnductance, migration, cracking, and similar phenonena?
ANSTER II-5 (b) and (c)
(b)
'Ihe clad streryth vs clad temperature curves shown in the figure on page 3-43 of CRBRP-GEER-00103 exhibit a change in slope at about 600 C.
Beyond 600 C the steeper slope of the curve indicates that clad failure is more pensitive to increases in tenperature than to increases in stress.
'Iwnperatures exceeding 600 C are attained in the typer part 'of the pin during transient overpower accidents as analyza3 in CRBRP-GEFR-00523.
s SElr II AA-17
?
l
Therefore, failure is expected to occur in the upper part of the pin due to the clad axial tmperature distribution during the transient, the clad loading mechanism, and the nature of the clad failure property.
(c) The adjective "significantly" in this context means that beycnd 600 C, clad failure is much more sensitive to increases in ternperature than to increases in stress as quantified by the clad failure curves on page 3-43 of CRBRP-GEFR-00103.
Inhanagenieties in the fuel and cladding are accounted for by perfonning parametric calculations in which pin failures are conservatively forced to occur at the core axial midplane.
QUESTICN II-6 (F6.2-7, par. 3)
For cladding strength the use measured data fran 20% CW 316SS specimens irradiated in EBR II.
(c)
Wat is the basis for extrapolating the EBR II data to the CRBR environment?
_ ANSWER II-6 (b) and (c)
'Ihe data fran Reic +4%g.., % CRBRP-GEFR-00103 has been used as the basis of data for fast flux irradiated cladding prototypic of that to be used in CRBR. CRBR fuel pin cladding operating parameters outside the range of the EBR-II data are accounted for by perfonning parametric calculations in which pins are conservatively forced to fail at the core axial midplane.
l' acurrIm II-7 l
(F6.2-8, par. 1) The use of the Gruber's simplified correlation to FRAS.
i l
t
{
SEfr II AA-18 1
4 ANSWER II-7 (b) and (c)
It is shown in Reference 13 in CRBRP-GFR-00523 that Gruber's simplified correlaticn is an adequate representaticn of FRAS results.
QUEErrICN II-8 (M.2-8, par. 3) The fuel-cladding relative heating rates are ckninated by the fuel-clad gap corductance, fuel thermal ccnductivity, and reactor power.
ANSWER II-8 (b) and (c)
'Ihe fuel-cladding relative heatire rates during reactor transients are calculated in the SAS3A and SAS3D code by usirg coupled heat transfer models of the fuel, gap, clad, coolant and reactor structure and values of physical properties and heat transfer coefficients based on experiments using prototypic CRBR materials.
The statement that the fuel-cladding relative heating rates are (knineted by the fuel-clad gap conductance, fuel thermal conductivity, ard reactor power is a qualitative statement which means that heat transfer calcula-tiens show greater sensitivity to these parameters than to the other heat transfer gameters in the nodels.
NorE: Questions II-9 through II-20 pertain to the SAS/EC Stmmary.
QUEErrICE II-9
( M.2-8, par. 3) The use of Ibss-Stoute heat transport model.
Ser II AA-19
t ANSWER II-9 (b) and (c)
'Ihe Ibss-Stoute gap conductance type nodel for application to fast reactor fuel rods is given by Dutt (D.
S. Dutt, R. B. Baker, and R. J. Jackson,
" Interim Fuel 'Ihermal Performance Mxlels for LIFE-2," W/ETTF 73518, January
[
15, 1973).
'Ihe contact conductance algorithn of this correlaticn was adopted for use in the CRBR IETA analysis as the best available formulation of this couponent of fuel-cladding heat transfer.
j QUESTION II-10 (EE.2-8, par, 3)
'Ihe assumption that cladding hardness is inversely suvun.ional to the cladding yield strength..
P ANSWER II-10 (b) and (c) l
\\
The word " inversely" was a misprint cn page EE.2-8.
Appendix F has been deleted fran the applicaticn.
It is correctly stated in CRBRP-GEFR-00103 that the cladding hardness is assumed proportional to cladding strength.
QUESTION II-ll (EE.2-8, par. 4)
'Ihe asstanption that a cne-dimensional nodel using Lagrangian cells can adequate Ly represent interaction of fuel ard coolant.
ANSWER II-ll (b) and (c) i In SAS/EUI, it is asstuned that the interaction between fuel and coolant is occurring unifonnly throughout the interacticn zone.
This is an obvious simplifying asstription Wtich makes it possible to treat the interaction zone without resorting to one-dimensional ccrnpressible hydrodynamics treatment.
It is because of this sinplifying asstanption and others made within SAS/ECI that auxiliary calculations were made with the PWIO 1 and PWIO 2 (Ref. 25 in CRBRP-GEFR-00523) codes in order to more accurately SET II AA-20
determine the voiding rates and fuel relocation dynamics associated with these pin failures into non-voided subassenblies.
In PIlTIO 1 it is indeed asstmed that a one-dimensional nodel using Lagrangian cells can adequately represent interacticn of fuel ard coolant.
PIlJIO 2 also uses a one-dimensional treatment but it utilizes the Eulerian approach.
As pointed out in the above reference, the one-dimensional treatment is adequate, based cn the long length in the axial direction of the coolant channel in conpariscn to the relatively small distances between adjacent pins.
QUESirI N II-12 (EE.2-8, par. 4) (Generic answers (a) and (b) are not required.)
(c) Define more fully what is meant by rip lergth ard hov it is determined.
ANSWER II-12 (c)
Within the context of the SAS/ECi model, the rip length is that length of the pin over which the cladding is initially assumed to fail.
The rip length determines the length of the initial interaction zone and defines the length over which the fuel-fissicn gas mixture continues to be ejected fran tN failed pin into the coolant channel.
The location of the rip is determined by centerirg the rip length cn the center of the SAS3A or SAS3D axial node at which the failure criterion is exceeded by the greatest amount at the time step when the failure criterion is first exceeded at one or nore nodes in the channel.
A discussion of the rip length used in the studies is doctrnented cn pages 4-16, 7-27 and 7-28 of CRBRP-GEFR-00103.
SET II AA-21 l
QUESTICN II-13 i
I l
(EE.2-8, par. 5)
(Generic answers (a) and (b) are rot required.)
(c)
Please clarify this statment:
"Three failure groups based upcn time fr m initial failure reflects for the incoherence of fuel-pin failure within fuel subassenblies."
ANSWER II-13 (c)
There was a typographical error in the last sentence of par. 5 on p.
F6.2-9.
Appendix F has been deleted frcm the application. CRBRP-GEFF-00103 correctly states that "Moreover, three failure groups based upon time fran initial failure reflects the incoherence of fuel pin failure within fuel assenblies."
On p. 7 of Ref. 9 in CRBRP-GEFR-00523, the manner in which these failure groups are treeted is discussed in nere detail.
QUESTICE II-14 (EE.2-9, par. 1) The assumption that the Cho-Wright nodel is an adequate representaticn of fuel to sodium heat transfers.
(c)
Identify all alternative nodels (including MECI models) that have been considered ard rejected.
What is the basis for rejectirg these other models? (Fbr each nodel rejected explain in detail.)
ANSWER II-14 (b) and (c)
A stmmary of models to calculate fuel to soditra heat transfer is given in the followiry reference:
Hans K. Fauske, "CSNI Meetirg cm Fuel-Coolant Interactions," Nuclear Safety,16, Ip. 436-442,1975. 'Ibe Cho-Wright nodel is representative of the state-of-the-art; and since it is a parametric model, a wide variety of situations can be simulated by variation of the particle diameter, the mixirg time constant, ard the other ' parametric additions used in SMi/ECI. 'Ihis nodel is adequate for simulating the mild SEP II AA-22
interactions diich have been observed in experiments, as well as nere hypothetical situations.
A further discussion of the canparison between analytical nodels and the experimental data base cri fuel pin postfailure transient behavice is given in the followirg references:
(1)
H.U. Wider and A.E. Wright, " Analysis of a Sodium Reentry Event in the H4 TREAT Test," TANSAO 22, p. 428,1975.
(2) Ref. E-3 in CRBRP-GEFR-00523.
(3) Ref. E-4 in CRBRP-GEER-00523.
b QUESTION II-15 (EE.2-9, pr. 1)
The assumed values for the thernodynamic properties of fuel, cladding and sodium.
ANSWER II-15 (b) and (c) he sodium thennodynamic properties for both single-phase and two-phase soditrn used in SAS/ECI are given in Chapter 6 of Ref. 9 in CRBRP-GEER-00523. %e thernodynamic Ireperties of cladding and fuel used in SAS/FCI are identical to those in the renainder of the SAS3A and SAS3D code. %ese are sunmarized in Section 4.0 of CRBRP-GEFR-00103.
QUESTICE II-16 (EE.2-9, Mr. 2)
We assumption that the sodium void reactivity can be determined adequately fran the average, smeared scdiun density of the interaction zone.
ANSWER II-16 (b) and (c)
One purpose of doing auxiliary PIDIO 1 and PIITIO 2 calculations was to check the SAS/PCI canputed material relocaticri reactivities with a more detailed nodel.
%e PIDIO 1 and PIITIO 2 calculations indicated that the SEP II AA-23
magnitude of the sodium void reactivity following pin failure in the hypothetical accidents considered is not a daninating factor in determining the hypothetical accident progression in CRBRP-GEFR-00103 and CRBRP-GEFR-00523 since the fuel motion reactivity quickly daninates the progression of the hypothetical accident after initiation of a fuel-coolant interaction.
QUESTICN II-17 (FE 2-10, par. 1)
(Generic answers (a) and (b) are not required.)
(c)
Please explain what is meant by " Reactivity feedbacks are determined fran the projection of the cell lengths and masses onto the normal SAS3A axial node lengths."
ANSWER II-17 (c)
In Section 7.2.2 of Ref. 9 in CRBRP-GEFR-00523 the omnplete fuel notion reactivity feedback model, including the statement to be clarified by this response, is described in detail.
QUESTICH II-18 (FE 2-10, par. 2)
(Generic answers (a) and (b) are not required.)
(c) What is meant by " input default opticris"?
ANSWER II-18 (c)
'Ihe original SAS/ECI model, as cbetrnented in Ref. 9 of CRBRP-GEFR-OOS23, required that a ntsnber of parameters it used be supplied by the user as input to the SAS3A and SAS3D code.
After gaining experience with the nodel, it was determined that it would be more appropriate to actually calculate the values assigned to sane of these parameters based on condi-tions within the SAS3A or SAS3D channel at the time that each parameter was SET II AA-24
actually required.
Se parameters discussed in par. 4 on p. 3-5 of CRBRP-GEFR-00103 are those pirameters for which provision was nude within the SAS/ECI subroutines of SAS3A and SAS3D to calculate them for purposes of the cases discussed in CRBRP-GEFR-00103 and CRBRP-GEER-00523.
QUESTICH II-19
( M.2-10, par. 2) The assumption that the rip area is the cross-sectional area of the internal pin cavity.
ANSWER II-19 (b) and (c)
Once the fuel and fission gas which are located innediately behtnd the rip in the claddiry (the SAS/ECI reservoir) are deposited into the interaction zone at the time of failure, any additional fuel and fission gas which are ejected through the rip must cane fran the central cavity in the pin. This cavity is quite long ccrnpared to its cross-sectional area.
Bus the rate at which the fuel-fissicn gas mixture can be transported to the rip and into the interaction zone is controlled by the cross-sectional area of the cavity.
Thus, it is appropriate to utilize this area as the rip area in the Bernoulli equation which is used to ccrnpute the ejection rate.
QUESTION II-20 (EE.2-10, par. 2)
The asstunption that the flow is adequately represented by a one-dimensional Bernoulli equaticn ftr unsteady flow.
ANSWER II-20 (b) and (c) he one-dimensional Bernoulli equation as used in SAS/ECI is a parametric algorithn to calculate fuel ejecticn fran the central pin cavity into the coolant channel.
Se nodel is described in Ref. 9 in CRBRP-GEFR-00523.
Een the fuel pin cavity can greatly change in size followinig fuel pin
- failure, e.g., for a core midplane failure in a pronpt-critical situation, SEP II AA-25
the nodel can significantly cuerpredict the mass-flow-rate of fuel within the pin to the failure locaticn.
'Ibe nodel utilized is judged to enhance the conservative nature of the analysis.
A less conservative treatment would result in a decrease in the calculated accident-energetics in many of these situations.
NCTTE: Questions II-21 through II-27 pertain to the PILTIO Strmary.
QUESTICN II-21_
(M.2-ll, par. 3)
The fuel-fission gas flow in the nelten pin region is treated as hcrogeneous, conpressible, cne-dimensicnal flow with a non-uniform flow cross section.
ANSWER II-21 (b) and (c)
'Ihe assumption of a cne-dimensional flow is adequate because the nelten pin cavities are several tens of cm's long and cnly a few tenths of a cm in diameter. 'Ihe assumption of hcrrogeneous crmpressible tw>-phase flow (i.e.,
no slip betwem fuel and gas) is supported by H.
J. Willenberg and A.
Padilla, Jr., " Analysis of Transient Cbnpressible 'IWo-Phase Flow with Heat ard Mass Sources Using the Method of Characteristics," 02nputational Methods in Nuclear Ergineering, 00NF-750413, Vol.1, p. II-107,1975.
QUE57ICN II-22 (M.2-11, par. 3)
The representation of all failed pins in a given sub-assembly by a single pin nodel.
ANSWER II-22 (b) and (c)
'Ihe assunption of treating all fuel pins in a given subasse5>ly with a single pin nodel is made necessary by ocmputational limitations. This i
l l
l l
SET II AA-26 i
approach is consistant with the one dimensional treatments of fuel and coolant dynamics used in SAS3A, SAS3D, PIDID 1, AND PIITID 2. Use of one pin per channel "which may represent several subassenblies" results in a coherent treatment of fuel coolant interactions, sodium voiding, intra-pin fuel moticn, and fuel sweepout. For short times after failure, the im-portant effects cpverning reactivity are intra-pin notion and sodium voiding. 'Ihe single pin treatment will tend to accentuate these effects and, if positive feedback is predicted such as would be the case for mid-plane failure, lead to a conservatively high positive feedback. It is the short-time positive feedback which would terd to produce an energetic transient. In a longer time, fuel sweeput due to hydraulic forces beccres important. 'Ihe coherent single pin treatment may over-estimate the 1.y-draulic forces available for sweepout. Overall, the single pin treatment will enphasize positive reactivity feedback effects in the short-time after failure, and yield a conservative result.
QUESTICN II-23 (EE.2-11, pr. 4)
The axial notion of material in the ocolant channel is treated as two-conponent slip flow.
ANSWER II-23 (b) and (c)
'Ihe density, velocity, and internal energy changes of both conponents are calculated by solvirg a set of two mass, two manenttan, two energy equations and an equation of state.
'Ihe two nonentum equations contain the inter-active (or drag) forces which couple the flows of the two ccmponents (see Ref.13 in CRBRP-GEFR-OO103).
l i
QUESTION II-24 l
(P6.2-ll, par. 4)
'Ihe liquid soditrn, cr the mixture of liquid coolant, vaporized coolant, ard fissicn gas (Na/EU) is regarded as one ccinponent and its flow is modeled with canpressible lagrangian hydrodynamics.
SETr II AA-27
i ANSWER II-24 (b) and fc)
Slip between liquid sodium and sodium vapor or fission gas can be dis-regarded because the flow under consideration is nestly in a sP.g flow regime. 'Ihe flow is nodeled using Lagrangian hydrodynamics in PlI7IO 1 and Eulerian hydrodynamics in PIITID 2.
QUESTION II-25 (EE.2-ll, par. 4) The other acrnponent is the fuel, which is assumed to be in the form of particles.
'Ihe motion of the fuel particles is calculated by solving the momentum equation for representative (or " master")
particles.
ANSWER II-25 (b) and (c)
'Ihe assumption that the fuel is in the form of particles can be justified, as larg as liquid soditan is close to the fuel.
If the coolant channel is voided, the assumption of annular or bubbly fuel flow, which is made in PIIIID 2, is nore appropriate.
QUESTION II _2,66 (1 i.2-ll, par. 5)
Although soditra vapor condensation on cold cladding is accounted for in PIITIO, the mndensate is currently asstaned not to adhere to the cold wall but rather to be torn off instantaneously and mixed with the scditan in the coolant channel at the same axial location.
ANSWER II-26 (b) and (c)
Hot fuel particles noving through narrow coolant channels should quickly vaporize a liquid soditan film.
Wreover, the gas ard vapor streamirg in the coolant channels will lead to flooding of the liquid soditan film since SET II AA-28
the shysical vapor and gas velocities calculated by PIITIO 1 and PIIIIO 2 far exceed the necessary flooding velocity of about 15 ft/sec. Flooding of the liquid film leads to an mstable edium film interface and a significant increase in film to vapor frictional couplirg. Hence, any sodita film will travel in the direction of the noving vapor with cpod interfacial friction-al coupling, ard the PIITIO 1 treatment is reasonable. PIITIO 2 incorporates the treatment of a liquid edium film which can be evaporated or entrained by high gas velocities or it can be torn off by fuel flows.
QUESTION II-27 (EE.2-ll, par. 6)
Ebr the fission-gas temperatures, mass-weight averages betwem liquid sodits and fuel are used.
ANSWER II-27 (b) and (c)
If there is nuch sodium at a certain location the fission gas ternperature will be close to the so31tn temperature and if there is much fuel, the fission-gas temperature will be close to the fuel temperature according to PIITIO 1 ard PIlTIO 2.
IUTE: Questions II-28 through II-47 pertain to the SASBIDK Stmnary.
t QUESTICN II-28 (E6.2-12 par. 1)
(Generic answers (a) and (b) are rnt required.)
(c)
Explain fully why neither SAS/ECI nor PIITIO is capable of treating fuel blockages, er of continuing the calculation beyond FCI initiation for more than a few hundred milliseconds.
SErr II AA-29
ANSWER II-28 (c)
Neither SAS/ECI nor PIRIO 1 consider " freezing" of Itolten fuel to cold surfaces.
Both rtedels asstane a cx>nstant, average particle size which may be a liquid drop or a solid particle.
PIUIO 2 incorporates a freezing model (Ref. 25, CRDRP-GEFR-00523).
SAS/ECI will not continue calculations beycxid a few hurdred milliseconds because of codirg limitations within the model.
As the transient pro-r gresses, vapor bubbles may form in the channel below the FCI zone and may i
attenpt to merge with the zone. Se present SAS/FCI does not have the coded logic and ntanerical Itodels to treat this occurrence. We calculations are continued with the SASBIIE option in the SAS3A and SAS3D cele. PIRIO 1 and PIDIO 2 are not limited to running time of a few hundred milliseconds.
he PIUID 1 and PIUID 2 codes can only be rtn separately frcm SAS3A and SAS3D using cne of then to generate input for PIUID 1 and FIFIO 2.
8 QUESTION II-29 (EE.2-12, par. 1)
The asstrnption that the hydraulic effect on assembly floLi can be adequately represented frcm packed particle bed correlations and represented as a local hydraulic loss.
t ANSWER II-29 (b) and (c)
I he flow resistance characteristics of the packed particle bed used in the blockage model were taken fran experimental data (Ref. A-1 in CRBRP-GEFR-i 00103) in Which the staterial used in the test was a packed bed of U02 i
having a particle-size distributicri similar to those found to result from molten-fuel-sodium contact.
Results of these tests showed good agreenent with the correlaticn used in the SASBIIE analysis. Representing the packed l
l bed as a local hydraulic loss is a standard technique used in flow testing where conplicated pressure losses are represented by an equivalent local
[
SEP II AA-30
hydraulic loss coefficient.
'Ihe use of these coefficients in the SASBIOK analysis produced an equivalent effect upon the flow within the core.
QUESTION II-30 (EE.2-12, par. 1)
Power generation in the damaged region is proportion-ately reduced by fuel loss.
ANSWER II-30 (b) and (c)
Pin failure results in fuel loss fran the pin.
'Ihe pwer produced in the damaged channel must be reduced in proportion to the amount of fuel ejected frun the pin.
Ebr ntanerical simplicity, the power level is reduced uniformly in the core region of the modeled channel although fuel would actually be retoved preferentially fran the higher worth core midplane region. 'Ihis simplified model underestimates the power reducticn and is therefore adequate for the purpose of analysis.
QUESTIN II-31 (FE.2-12, par. 1)
(Generic answers (a) and (b) are not required.)
(c) Explain fully what is meant by quasi-steady formulation.
ANSWER II-31 (c)
'Ihe term " quasi-steady" was used in reference to the temperature distribu-ticm within the blockage arri the coolant temperature increase due to the heat generation within the blockage.
'Ib ocqx2te these values, a constant heat generation rate, taken at the time of initial formulation of the blockage, and steady state formulas for the tanperature distribution within the blockage were used.
'Ihe short duration of the ficw transient in the 1
f SEP II AA-31 l
1
channel relative to the change in the heat generation rate justified the use of the quasi-steady formulation.
QUESTION II-32 (EE.2-12, par. 1) (Generic answers (a) and (b) are not required.)
(c) Is it currect to assune that SASBIDK is not utilized in analysis of IDF CIA events?
ANSWER II-32 (c)
Yes, SANIX was not used for analysis of any LOF-HCDA event reported in CRBRP-GEFR-00103 or in CRBRP-GEFR-00523.
QUESTICN II-33 (EE.2-12, par. 2) (Generic answers (a) and (b) are not required.)
(c) Explain fully the purpose and operation of the " tabular fission gas release nodule".
ANSWER II-33 (c)
The tabular fission gas release nodel is an option built into the SAS codes Whidt permits construction of user specified bubble growth arx3 collapse within a channel.
Time dependent bubble interface locations nust be specified.
Its purpose is to allow an explicit investigation of the effects of voiding at specified locations and times during the course of an accident. 'Ihe input bubble interface locations and pressures are supplied by auxiliary calculations or estimates.
A conplete description of the opticm is givai beginning cm page 79 of Ref. 7 in CRBRP-GEFR-00523.
SEP II AA-32
QUEErrION II-34 (PE.2-12, par. 2) (Generic answers (a) and (b) are not required.)
(c) Explain fully the "progranmable reactivity opticn."
ANSWER II-34 (c)
The gwimsuable reactivity option in SAS is a table look-up operation to provide user specified reactivity vs time.
Input tables of reactivity and time are provided by the cxxle user.
'Ibe reactor net reactivity is the swimmieci or driving reactivity plus the reactivity ccmponents calculated by the cxxle. Interpolation is performed in the code to find the progranned reactivity at any particular time.
QUESTTION II-35 (EE.2-12, Par. 2) The exit loss coefficient is increased to the blocked
- value, t
(c) What does this mean?
i ANSWER II-35 (b) and (c)
During a SAmfE calculation, the value of the channel exit loss coeffi-cient is increased linearly frcm the initial input value to the calculated blocked value over a period of time equal to twice the fuel ejection time, beginning at the time of initial fuel ejecticn.
'Ihe purpose of the ramp change in exit loss coefficient is to represent the buildup of the blockage over a pericx1 of time.
'IVice the ejection time is judged to be a reason-able estimate of the build-up time.
SET II AA-33
QUESTION II-36 (PE.2-12, par. 2) Explain fully how the locus of coolable blockage config-urations for each degree of hydraulic disturbance is generated and justify each assumption.
ANSWER II-36 (b) and (c)
The approach and calculations for determining the locus of coolable blockage configurations were presented in detail for the ED T Reactivity Insertion Base Chse (Section 6.1.1.2 of CRBRP-GEFR-00103). The assumptions have been justified in Appendix A and Section 6.1.1.2 of CRBRP-GEFR-00103.
QUESTTICN II-37 (EE.2-12, par. 2) Tha assumption of cx2nplete core outlet blockage for an unstable two-phase soluticn.
(c) What is meant by unstable two phase solution?
ANSWER II-37 (b) and (c)
An assumption of atmplete outlet blockage was made in order to result in an analysis which is (a) conservative and (b) ntraerically stable.
This asstanpticn implies a ocmplete subassait>1y meltdown and reactivity estimates of fuel-sitznping are perfonned as described in the response to Question 38.
GJESTICN II-38 (P6.2-12, par. 2) (Generic answers (a) and (b) are not required.)
(
(c) Itw are reactivity estimates cn the effects of fuel sitznping perfonned?
~
l I
SET II AA-34
ANSWER II-38 (c) h reactivity estimates cn the effects of fuel sitznping were performed using diffusion theory models.
Values of K,ff were couputed for various positions of the fuel in the coolant channels, starting fran the point at which it formed the blockage, to the point of maximtzn reactivity change after it moved downward. 'Ihe core nodel and cross sections used were the same as for the other neutronic calculations that were done for CRBRP-GEFR-00103. 'Ibese are described in Section 5 of CRBRP-GEFR-00103.
QUESTPICN II-39 (M.2-12, par. 2) (Generic answers (a) and (b) are not required.)
(c) It:w are the "various analysis path" selected?
ANSWER II-39 (b) and (c)
'Ihe analysis path determination is based on whether reactivity estimates of relocating fuel in the uncoolable assemblies would lead to a critical or supercritical reactor state. 'Ihese two analysis paths are indicated in the lower left hand portion of the SASBIIE flow chart in Figure 3-7 in CRBRP-GEFR-00103.
QUESTICN II-40 (M.2-13, par. 2) The asstrnption that porous blockage is adequately rep-resented by a mean particle size of 4209 ANSM!:R II-40 (b) and (c)
'Ihe mean particle size is a parameter that characterizes the particle distributicn used in the packed bed analysis.
'Ihe 420 value used in the SASBILK analysis was taken frcm Ref. A-1 in Appendix A of CRBRP-GEFR-00103.
SErr II AA-35
i
'Ihis value characterized the particle distribution used in flow tests of packed beds.
'Ihe 420p value was calculated by methods recamended by the correlation which showed good agreement with the test data.
'Ihe particle distributicn used in the tests was similar to those fourd to result frce molten-fuel-sodium contact.
'Iherefore, the porous blockage is adequately represented by the mean particle size of 420p.
QUESTION II-41 (M.2-13, par. 2) The assumption of a friction factor that correlates with tests involving water through uranita dioxide.
ANSWER II-41 (b) and (c)
'Ihe friction factor is a dimensicnless parameter which depends upon rela-tive roughness and the Reynolds nunber. 'Ihe correlations of test data with friction factor can be seen in Appendix C of Ref. A-1 in Appendix A of CRBRP-GEFR-00103 where pressure drop versus velocity has been plotted for both the test data and the Ieva correlation using a mean particle diameter of 420u.
QUESTICN II-42 (M.2-13, par. 2) (Generic answers (a) and (b) are not required.)
(c) Explain fully what is meant by the second sentence beginning with
.since the particle bed correlation..."
ANSWER II-42 (b) and (c)
'Ihe particle distributicn used in the packed bed test (Ref. A-1 in Appendix A of CRBRP-GE2R-00103) was similar to those fotrd to result frcm molten-fuel-sodium contact.
'Ihis similarity forms the basis for using' the corre-lation in the SASBILK analysis.
j SET II AA-36
QUESTION II-43
( M.2-13, par. 2) The use of the Leva correlation.
ANSWER II-43 (b) ard (c)
The leva mrrelation was found to be in good agreenent with test data (Ref.
A-1 in Appendix of CRBRP-GEFR-00103) and this agreenent provides a basis for its application in the SASBIOK analysis.
QUESTION II-44 (EE.2-13, pir. 3) The asstrnption that flow blockage can be adequately modeled by the average loss coefficient representation.
ANSWER II-44 (b) and (c)
We use of an average loss mefficient to simulate a flow blockage is a standard engineering practice in flow nodeling where overall ficw resis-tance nodeling is the objective. Chnplicated blockages can be nodeled with si:nple loss coefficients which produce equivalent flow in the systen. We SASBICK calculations incitxled a parametric variation of the loss meffi-cient so that the range of expected porous blockages is adequately covered.
QUESTION II-45 (M.2-14, pir. 4) The asstanption that the blockages form as a contiguous mass of material locata3 in the fission gas plentrn regicn where either the I
original gecmetry has been destroyed or in the molant channel dere the 1
geonetry has been maintained.
SEP II AA-37
ANSWER II-45 (b) and (c)
'Ibe blockage was assumed to fonn a contigious mass because this would represent the worst case analysis, i.e.,
it would be the most difficult ccmdition to maintain in a coolable and stable condition.
On the other hand, distributed blockages could be nore easily cooled due to the presence of greater coolant access to blockages.
The location of the blockage was selected as one of several locations where agg1cmeration of the material could occur, and the cooler regions of the core provide such a location.
QUESTICN II-46 (EE.2-16, par. 3) (Generic answers (a) and (b) are not required).
(c)
Explain fully the elementary methods that were applied to the gravity effects and the generalized discussion of the coolant effects; with respect to the elenentary methods, is this simply the " falling film analysis" and
" falling of liquid drops" analysis described in the following paragraphs?
ANSWER II-46 (c)
'Ihe elementary methods refer to analyses of "the falling film" and " falling of liquid drops" which are described in Appendix A of CRBRP-GEFR-00103.
Both are described in scme detail in pages A-8 and A-9 of Appendix A and further details can be fourd in Reference A-6, page A-15 of CRBRP-GEFR-00103.
QUESTICH II-47 (F6.2-18, par. 1) (Generic answers (a) and (b) are not required),
i (c)
Identify the " considerable uncertainties" in molten fuel penetration l
methods.
l I
SEP II AA-38
l ANSWER II-47 (c)
'Ihe response to this interrogatory is identical to the response ntriber II-27 of the Third Set of Interrogatories (p. AA-75).
IUTE: Questions II-48 through II-59 pertain to the Sodium Voiding Model Sumary.
QUESTIOT II-48 (F6.2-18, par. 3) (Generic answers (a) and (b) are not required.)
(c) What is meant by " axial interface areas"?
i ANSWER II-48 (c)
'Ihe axial int erface area is the cross-sectional area of the coolant chan-nel. Secticn IIB of Ref. 7 in CRBRP-GEFR-00523 provides a detailed descrip-tion of most of the sodium boiling nodel in SAS3A, and Ref. 6 in CRBRP-GEFR-00523 lists the SAS3A additions to the SAS2A boiling rrodel. The physics represented in the SAS3D boiling nodel is identical to that in SAS3A.
QUESTION II-49 (EE.2-18, Inr. 3) The soditzn vapor and liquid film are assumed to be at naturaticn conditions determined by channel pressure as opposed to non-equilibrium super heat conditions.
ANSWER II-49 (b) and (c)
Ref. 71 of CRBRP-GEER-00523, Ip. 20-22, describes the supporting evidence fcr believing the liquid superheat in a reactor will be anali, and have little effect cm the voiding Irocess.
SEP II AA-39
The thin film of soditzn left cn the clad in a voided region is in contact with the vapcr bubble and the film will vaporize before it superheats.
'1he vapor will cnly superheat if it passes over hot, dried-out clad.
Before the liquid film cn the clad dries out, heat remwal due to vapor-izing the film will prevent the clad surface temperature fran rising nuch above the soditzn saturaticn tanperature.
After extensive voiding, exten-sive film dry-out, and significant heating of the dried-out clad past the sodiun saturaticn tanperature, scme superheating will be limited by the relatively poor heat tran.:*er coefficient between hot, dried-out clad and cooler sodiun vapor.
By the time that any appreciable superheating of the vapor might occur in the IDF cases discussed in CRBRP-GEFR-00103 and CRBRP-GEFR-00523, the active core regicn of the subassembly is voided and dried out; and it remains voided, with little or no heat renoval fran the fuel pins to the voided coolant channel.
At this point, many of the details of the voiding in the subassently, including any superheating of the vapor, will be largely unimportant, sin any coolant voiding reac-tivity insertion has already cxxurred, and the sodiun will no longer renove much heat frcm the voided, dried-out core.
QUESTION II-50 (EE. 2-18, Inr. 4) All the assumptions ocncerning reentry of lower liquid slug discussed in the paragraph.
(c) Describe dat happens *en the channel becanes blocked.
ANSWER II-50 (b) and (c)
Reentry of the lower liquid slug is described in Section IIB of Ref 7 in CRMP-GEFR-00523. When a partial blockage is formed in the coolant channel, the hydraulic diameter and coolant flow area are reduced. 'Ihis leads to an increase in the fricticn pressure drcp due to any vapor streaming past the blockage, and causes the vapor pressure to buildup below the blockage, if SEP II AA-40
the vapor is ocming fran below, or above the blockage, if the vapor is streaming fran above the blockage.
'Ibe increase in vapor pressure usually stops liquid slugs before they get to the blockage.
If a liquid slug does pass a partial blockage, then the fricticn pressure drop in the liquid slug is increased due to the reduction in hydraulic diameter.
QUEErfICN II-51
( M.2-18, par. 5) (Generic answers (a) and (b) are not required.)
(c)
Explain fully what is meant by " reasonably gocx3 qualitative agree-ment."
How did the results ocmpare quantitatively?
ANSWERII-51.(cl For a quantitative ocmparision of the results, see Reference 26 in CRBRP-GEER-00103.
QUESTICN II-52
( M.1-18, par. 5) (Generic answers (a) and (b) are not required.)
(c)
Explain fully how ".
.the experimentally observed dryout mechanism is considered."
ANSWER II-52 (c)
A static film dryout nodel was used for the calculations reported in Ref.
26 in CRBRP-GEER-00103 fcr the Karlsruhe experiments of Peppler.
When the calculations were repeated using the film motion nodel described in Ref. 71 of CR!RP-GEER-00523, good quantitative agreement was achieved between the calculated and experimentally observed dryout times (see G. Hoeppner, F. E.
Dunn, and T. J. Heames, "'Ihe SAS3A Sodiun Boiling Mxlel and Its Experi-mental Insis," Trans. Am. Racl. Soc., y, 519, April 1975).
The film SEP II AA-41
motion nedel was used for the CRBR calculations described in CRBRP-GEFR-00103, CRBRP-GE2R-00523 and Ref. 4 in CRBRP-GEFR-00523.
QUESTION II-53 (M.2-19, par.1) (Generic answers (a) and (b) are not required.)
(c)
Explain fully the ccrnparison of the film dryout models results with experimental cbserved dryout time.
ANSWER II-53 (c)
See Answer II-52(c).
QUE3rION II-54 (M.2-19, Inr. 2) ".
. full assenbly voiding would be expected smewhat earlier than the average pin model predicts..."
ANSWER II-54 (b) and (c)
'Ihe radial growth of voiding within a subassembly, and the applicability of the average pin model are discussed in detail in Section 4.2 of Ref. 71 in CRBRP-GEFR-00523.
QUESTICN II-55 (M.2-19, par. 2) "However, the difference would not be expected to be great."
ANS6ER II-55 (b) and (c)
See Answer II-54(b) and (c).
SETF II AA-42
i QUESTICN II-56 (N.2-19, Inr. 3)
"'Ihe present SAS3A voiding tredel cbes not include a meanirgful consistant treatment of gas release into a voided channel."
ANSWER II-56 (b) and (c) he predictions of the boiling and plenun gas release models in SAS3A are used only for three main purposes in the CRBR analysis.
'Ihese purposes are:
(1) the calculation of the rate at which voiding reactivity is inserted, (2) the calculaticn of heat ramval fran pins in voided regions, determining the time of melting and rate of melting of clad and fuel, and (3) determiniry the impact that the presence of sodiun vapor or scxiiun liquid could have cm the relocation of molten clad or fuel in voided regions.
Because of the strergth vs. temperature characteristics of the clad, the cladding will not fail and release plenan gas during a hyp>-
thetical ILF accident until the sodiun has boiled extensively, the film on the clad has dried out, and the clad tanperature has risen considerably above the sodiun saturaticn temperature.
At this point, failure of the pins and release of plenan gas would have scrne influence on the voiding Irofile; and it could affect Ir-series test voiding measurenents; but it would have very little irrpact cn the voiding and voiding reactivity in-sertion because:
(1) the core region of the subassembly would already be voided before gas release, so the coolant voiding reactivity insertion would have already occurred, (2) the liquid film would already have dried out from the clad in rrost of the hotter parts of the pins, and no heat ra m val fran these areas would be predicted, whether the plenun gas is released or not.
l l
I r
i l
SET II AA-43 i
QUESTICN II-57 (P6.2-19, par. 3) However, at the present time, release of plentrn gas into a voided channel is expected to have only a slight effect cn the overall voiding behavior.
ANSWER II-57 (b) and (c)
See Answer II-56(b) and (c).
QUESTICN II-58 (F6.2-19, par. 3) This should not greatly affect the voiding process cnce full-assernbly voidirg has occurred and cladding film has dried out or has been stripped off.
ANSWER II-58 (b) and (c)
See Answer II-56(b) and (c).
QUESTION II-59 (F6.2-19, par. 4)...this has been suggested to be due to the presence of noncondensible fissicn gases.
l ANSWER II-59 (b) and (c) l i
See Answer II-56(b) ard (c).
tere: Qaestions II-60 through II-62 pertain to the fh $ ding Relocation l
Model (CIAZAS) Stamary.
SET II AA-44 i
/
\\
QUESTION II-60 (EE.2-19, par. 5) How are rnolten clailing-coolant interactions similar to troltm fuel coolant ~ interactions?
ANShER~II40 (b) and (c)
~
CIAZAS was written to analyze the reactivity and thermal effects of clad-dirn relocation before oxide fuel pin disrupticn.
In this phase of a CRBR hypothetical IN accident, the nelten cladding-coolant interaction question is not pertinent, since a long length of hot urrnelted claddirg and a zone producirg large quantities of sodiun vapor separate the nolten cladding 7
from the nearest liquid sodium.
( N ION II-61 i
(F6.2-20, par.1) The assumption that effects of clad melting and reloca-ticn can be adequately rredeled with three radial nodes conprising 25%, 70%
and 5% of the clad respectively.
ANSWER II-61 (b) and (c)
'Ibe clad radial mesh was chosen due to ntrnerical considerations involved in cladding-liquid sodiun heat transfer.
During cluViing meltirg and re-location, the cladding thermal ocnductivity and thickness suggest that thrce nodes are more than adequate, arti perhaps evm one node would be sufficient.
Ebr example,' the thernal diffusivity of nelten cladding is 0.05 on2,,c-1,
. Ibis leads to a thermal response time for 0.038 on thick clackling of 'O.03 sec.
'Ibe time for absorbing the cl.widing heat of fusion is usually a few tenths of a second, or approximately an order of magnitu3e grunter.
Hence, the cladding will melt essentially as a mit and the details of the radial meltiry profile are unimportant te the analysis.
1 SET II AA-45 i
QUESTIm II-62 (M.2-20, ptr. 3) 'Ihus, it would appear that CUCAS overestimates the degree of uped claddirx3 relocation, and that, if the time interval between clad nelting and fuel melting is as short as expected for GBR, very little, if any, net clad notion would result.
ANSWER II-62 (b) arx! (c)
'Ihe basis for this jtr3gment is the R-5 experiment discussed on page 3-28 of GBRP-GEIV-00103 and Ref. 32 in CRBRP-GEFR-00103.
NCfrE: Questions II-63 through II-69 pertain to the SUNPY Stamary.
QUESTICN II-63 (N.2-21, pur. 4) (Generic answers (a) and (b) are not required.)
(c)
Explain fully what is meant by " detailed initial conditions" supplied to VENUS-II. Describe the limitations of one-dimensional notion.
ANSWER II-63 (c)
If the otznputer ;.edels suggest that a hydrulynamic disassetly calculatim should be performed, the SAS3A and SAS3D codes with SUNPY contain all the informaticn, with respect to material location and material internal energy, that is required to set-tp the VDRE-II gecmetric core represen-tation.
One-dimensional notion refers to the restriction of cnly allowing axial fuel relocaticn within the confines of any SAS3A or SAS3D channel.
%o different hypothetical transient initiating accident situations aust be examined, that of Icw power and of high gwer.
In the context of this question, low power is defined as a few times nminal reactor operating ser II AA-46
leuls cr lower.
Here, one-dimensional notion cbes not allow for intra-subassmbly incoherence effects ard hence tends to yield an exaggerated description of material relocation, e.g., any fuel sitznpire is predicted to be too ooherent, and calculations cannot be done cn material noving upward at one radial location and dowrward at another. As the hypothetical accident power level increases, the radial and axial power profiles guar-antee that fuel will nove coherently away frcn the locations of peak puer.
At high power (defined as disassmbly power levels or approximately several hundred to nore than a thousand times ncminal power), the main limitation of axial cne-dimensional moticn concerns the lack of a description on how the hypothetical accident energetics will be mitigated by fuel expansion in more than one directicn.
Here axial one-dimensioral motion is too conser-vative and a twcx11mensional capability such as VENUS-II is desirable.
QUESTICH II-64 (E2-21, par. 4) In most SWMPY calculations, fuel notion is asstrned to begin when meltirg begins in unrestructured fuel.
(c) W at is meant by "equiaxed region"?
ANSWER II-64 (b) and (c)
(b)
Ref. 8 in CRBRP-GEER-00103 describes the experimental basis for assumiry fuel moticn because of fissicn gas release when melting begins in unrestructured fuel.
(c) Fast reactor fuels operate with close to radially flat pcwer profiles and at a high linear pcuer ratirg.
Hence, a large radial tenperature gradient exists inside of the oxide fuel pin.
As the fuel temperature rises above approximately 1350 C, the grains of the fuel pellet exhibit grain growth, without preferred orientation. 'Ibe fuel takes cm an equiaxed grain structure.
At still higher tarperatures, as icw as 1550 C at high burnup, migration of the fabricated porosity to the center of the fuel pin occurs resulting in a "coltarnar grain" regicn.
An examination of scme ser II AA-47
microstructural effects in fast reactor fuels is given in Ref. 53 in CRBRP-GEFR-00103.
QUESTION II-65 (76.2-21, par. 4) ".
.and the nore recent R-and Ir series expriments also seern to verify this latter asstrption".
ANSWER II-65 (b) and (c)
%e assumption mder discussion is nodeling of fuel as an intact coltran durin3 the time following claddirg meltirg but preceding fuel melting.
Restructuring during irradiation is known to cause sintering of fuel pellets.
11ence, this asstanpticn appears to be reasonable for irradiated fuel. We Ie-series 1 css-of-flow tests run with irradiated fuel have been consistent with this asstanpticn.
Figure 11 and Fig. 12 of the following reference:
E. W.
Barts et al.,
"Sturmary and Evaluaticn Fuel Dynanics loss--of-Flow Experiments 'Ibsts L2, L3, and IA," ANL 75-57, 1975, show several seconds between the time of stainless steel meltin3 indications and the tine of axial fuel notion.
Later Ir-series tests L5,14, L7 also shcw results consistent with nodeling the fuel coltam as intact prior to failure associated with fuel melting.
See Ref. 31, 32, 33, pg.11-3, CRBRP-GEFR-00523.
Fresh fuel tests such as L2, and the R-series tests (Ref. 23, pg. 11-2, CRBRP-GEFR-00523) cb not indicate fuel noticn upon cladding melting, but do indicate initiation of fuel collapse once meltirg occurs in the flow tube which provides the radial restraint for the test section.
his loss of radial restraint will not occur in the reactor situation.
QUESTICN II-66 (76.2-21, par. 6) We asstrna3 equation of state cunposed of six p' arts.
SEP II AA-48
ANSWER II-66 (b) and (c)
Details of the equation-of-state used in the SwMPY ccznpressible region are described cn pp.15-19 of Ref. 8 of CRBRP-GEFR-00523.
QUESTICN II-67 (F6.2-22, par. 2) (Generic answers (a) and (b) are not regaired.)
(c) Explain more fully the basis for the conclusions, "A relatively simple single-pin nodel like SWMPY cannot adequately analyze this situation.
Although the model can be forced to fit the experiment, the mechanisms are not presently identified clearly enough to allow extrapolation to another systern. "
Indicate in detail the uncertainties that this introduces into the subsequent calculations nodeling the transition phase and the dis-assernbly (VENW) phase.
ANSWER II-67 (c)
'Ihe SUNPY nodel assumes that fuel moticri occurs cnly in the axial di-recticn.
In additicn, it is assumed in SAS3A arrl SAS3D that the power density in each fuel pin is azinuthally symnetric.
Hence, it is not capable of modeling the pellet stack crtrnbling process, which would be three-dimensional in nature, nor can it analyze the effect of the pcuer density gradient across the pin, as was present in the L2 TREAT test.
Because several of the physical processes nodeled in SwMPY are treated parametrically, it is possible to sinulate observed experimental results by a suitable choice of parameters, so long as the cbninant node of fuel motion is axial motion, theertainties in SwMPY calculated fuel notion have been addressed by parametrically varying an equivalent gravity force and by matching these effects to w vp late test data (see CRBRP-GEER-00523 Sections 7.1.1, SFT II AA-49
7.2.1, and Appendix C). Cbre conditions Wttich result frcm the above varia-tions are carried directly into either the meltout or disassenbly phase i
evaluations.
1 QUESTICN II-68 (EE 2-23, par. 1 (Generic answers (a) and (b) are not required.)
(c)
Explain nore fully (qualitatively and quantitatively) the statement that "this mechanism is somewhat sensitive, and slight variations in the asstunptions can cause the fuel to either rise, fall, or nove in both directions.
ANSWER II-68 (c)
'1he statement should be interpreted in the context of the L3 and IA loss-of-floar TREAT tests. At the steady-state power levels under which these tests were corducted there are several ocznpeting effects tending to cause fuel motion in various directions once fuel nelting starts, e.g., gravity, production of steel vapor Iressure, and any remaining fission gas.
- Hence, calculatal results do beccme sensitive to the rate of heat transfer frcrn fuel to stainless steel, the rate of release of fission gas frcm fuel near its melting point, ard the asstuned radial heat losses.
The quantitative SIUMPY analysis of these tests is presented cn pp. 54-60 of Ref. 8 in CRBRP-GEFR-00523.
It is felt that uncertainties in SIIMPY analysis are covered by the range of parametric cases Which are presented in CRBRP-GEFR-00103.
One initial purpose of the F-series tests (Question II-69) was to provide an increased experimental data base to aid in the nodeling of fuel moticn in a reactor HCIA analysis.
Srr II AA-50
l QUESTICN II49 (P6.2-23, par. 3) (Generic answers (a) and (b) are not required).
(c)
What uncertainties are likely to be resolved by the F-series experi-ment and eat mcertainties are likely to reain?
ANSWER II49 (c)
Experiment F-1 provides insight into the behavior of highly irradiated fuel whm heated to melting at a near naninal pwer level with small radial taperature gradients.
Analysis of this test shows that the observed fuel motion would produce little or ro reactivity increase although the motion was not strt:ngly dispersive (see Ref. 34, pg. 11-3, CRBRP-GEFR-00523).
Experiment F-2 shod that fuel having very small irradiation exposure is dispersive When exposed to a power burst leading to dispersal at elevated
( N6 times noninal) pcwer.
Uncertainties resin in effects of details of fuel characterization (burnup, fission product content) and thermal his-tory. Ibwever, the body of experimental data presents a consistent picture.
See:
L. W. Deitrich, "An Assessment of Early Fuel Dispersal in the Hypo-thetical Loss-of-Flow Accident," Proc. Fast Reactor Safety Meeting, Seattle, Aug. 19-23, 1979, p3 615.
'Ihe applicant is aware that experi-ments F-3 and F-4 have bem conpleted, but it hss not been determina3 to what extent these will be relied on.
SEP II AA-51
'nIIRD INTERBOGA'IORY SET PREADELE 'ID QUESTIONS With respect to the following requests for information we are concerned with four distinct validations relative to the nodels and conputer codes:
- 1) Validation that the code's output is the correct ntznerical calculation that should result fran a given set of input data ard the model asstrup-tions;
- 11) Valdiation of the models against actual experimental data; iii) Validation that the utdels can be extended to the CRBR; and iv) Validaticm that the input asstanptions for the CRBR case are adequate with respect to the CDA analysis, i.e.,
are suprted by experimental evidence.
By " adequate," here and below, we mean that the calculations will not tmderestimate the CDA work potential (i.e.,
forces and resulting energetics of a CDA) or overestinute the contairrnent capability of the reactor with respect to a CDA.
QUESTION I With respect to each of the following codes and each subroutine of each of the followiry codes:
(A) FXVARI (B) REXCD-lEP please povide the following information [ mere appropriate, the parts of the questicm have been restated to reflect the protocol for discovery agreed to by Applicants, Staff, and Intervenors NRDC et al.]:
1)
Conplete, current doctanentaticm (i.e., a writeup) of the co' des and the subroutines; SEP III AA-52
2)
Identify, by name and affiliation, the author, or authors, of each model, subroutine, or portion of each subroutine, Wich each contributed or worked on;
- 3) Identify by name affiliation (including organization, division, branch, title, etc.) each applicant employee, or consultant, that has intimate working knowledge of the code and ech subroutine, or parts thereof, including its validity.
Where more than one person is involved, delineate which portion of the code or subroutine with Wich each has an intimate workin3 krowledge;
- 4) Describe fully the procedures by which Applicant has assured itself and continues to assure itself, that the various emputer programs (codes) accurately reproduce the nodels as described in the PSAR and its references (see Validaticri (i) above);
5)
Indicate Wich models (including subroutines, or portions of subrou-tines) have not been validated as described in Validation (i);
- 6) Irrlicate the nodels (including subroutines, or portions of subroutines) or asstanptions that have not been validated as described in Validation (ii);
7)
Ebr each model, portion of the model, cr asstunption that has been validate $ (against experimental (or other) data, see Validation (ii) above) describe fully the gocedure by which it wtss validated, and the results, includinj all uncertainties ard limitation of the validaticn.
Indicate the source of the experimental, or other data, that was used in the validation.
8)
Explain fully all instabilities in the numerical perfomance in the models, tat osuses then, and how they are avoided, and the extent to which this intro $uces uncertainties in the calculations and limits the validity of the model (cf., p.F6.2-10, par. 2).
SEP III AA-53
1 t
9)
To the extent that any answers to the above questions are based on referenced material, please supply the references.
10)
Explain Wiether Applicants are presently engaged in or intend to engage in any further research or work whid may affect Applicants' answer.
'Ihis answer need be provided only in cases where Applicants intend to rely upon cri going research not included in Section 1.5 of the PSAR at the Im i
or construction permit hearing cri the GBR. Failure to provide such an answer means that Applicants do not interx1 to rely upon the existence of any such research at the IE or ccristruction permit hearing cr1 the CRBR.
- 11) Identify the expert (s), if any, whcm Applicants intend to have testify on the subject etter questioned.
State the qualifications of each such expert. 'Ihis answer need not be provided until Applicants have identified the expert (s) in question or determined that no expert (s) will testify, as long as such answer provides reasonable notice to Intervenors.
ANSWER I(A)
'Ihe processing of neutron cross-section data for the heterogeneous core analyzed in CRBRP-GEFR-00523 did not use the FXVARI or its subroutines.
Accordingly, responses to I(A) I through I(A) 11 are not required. (The CRBRP-GEFR-00103 homogenous core analysis used FXVARI).
'Ihe methods used for pucessing neutronic input data for SAS3D analysis in CRBRP-GEFR-00523 are the same as those used in Section 4.3 of the PSAR.
ANSWER I(B)
'Ihese answers povide the information requested for the REXOO-HEP Q2nputer Code.
(1) Reference 1 on p. 5-45 of CRBRP-3, Vol.1, " Energetics and Structural Margine Beyorx1 the Design Base" is the latest. an3 most ccmplete doctanent describing the REXCNEP code.
SET III AA-54
Doctanentation of the subroutines which constitute the REXCO-HEP otxle are contained in this reference. Other improvements are described in J. Gvildys ard Y.W.
Chang, "REXCO-HEP (Release 4) Users Nnual," ANL/ RAS 78-42 (Septenber, 1978).
(2)
'Ihe REXCD-HEP code is a cortplex system developed over a span of several years. Its developnent started in the Reactor Engineering Division and continued in the Reactor Analysis and Safety Division of Argonne National Laboratory. 'Ihe major contributors are authors of Reference 1 on
- p. 5-45 of CRBRP-3 Vol. 1.
(3)
'Ihe following staff mmbers of Argonne Naticnal Laboratory have a working knowledge of the code, includirg its range of applicability and the extent of its validation:
Stanley H.
Fistedis, ibnager, Ehgineering Mechanics Program, and Yao W. Chang, Manager, Structural Mechanics Section, I:eactor Analysis and Safety Division, Argonne National Laboratory.
(4)
'Ihe hydrodynamic and solid mechanic principles on which the code is based are established scientific facts. 'Ihe individual subroutines and the entire code were checked and rechecked both irxlir Mually ard in its entirety for correctress of results. Extensive checking of the code against other establishal analytical solutions was performed and the ecmparisons were cited in Ref.1 on p. 5-45 of CRBRP-3, Vol.1 and also in J. Gvildys and Y.W.
- Chang, "REXCI)-HEP (Release 4) Users Manual," ANL/ RAS 78-42 (Septaber, 1978).
(5) All of the nodels have been validated as in (4) above.
(6)
'Ihere has been extensive experimental validation of the code as desc.:ribed in Itern (7) below. In the few areas were experimental validation does not exist at this time, issues were resolved by making conservative assunptions.
SET III AA-55
(7)
Substantial validaticn of the RDCCD code systen was performed.
Predictions of the REXOO code system were acrnpared against a variety of experiments. 'Ihe comparisons are doctraented in the following publications:
(1)
"Chnparison of a 'Iko--Dimensicnal lif rodynamics Cbde (RD00) to d
Excursicn Experiments for Fast Reactor Cont.airrnent, " ANL-7911, January 1972.
(2)
"Cbnparison of a 2-D Itydrodynamics Cbde (RD00) to Excursion Experiments for Fast Reactor Contairunent," AMc7911 Supplement 1,
July 1972.
(3)
"Ocrnparison of FFIT Simple-Model Tests with RDOO Predictions,"
ANIe8071, January 1974.
(4)
"REXCO Predictions of Elastic and Elastoplastic Defonnation of Fluid Filled Pipes and Chnparisons with Experiments of 1/10 Scale FFIT Pipe Ndels," ANL 75-61, Septerter 1975.
(5)
Y.W. Chang and J. Ovildys, "Chaparison of RDCO Ovie Predictions with Flexible Vessel Experiments," ANL/ RAS 78-9 (February,1978).
(6)
Y.W. Chang and J. Ovildys, "Chnparison of RDCO Cbde Predictions with Rigid Vessel Experiments," ANL/ RAS 78-30 (June,1978).
(7)
Y.W. Chang and J. Ovildys, "Chnparison of RD00 Cbde Prediction with SRI 91-2 Experimental Results," ANL-78-18 ( August,1978).
(8)
In the explicit integration of the equatients of notion, if the time step is too large, the conputed response may result in numerical instabil-ity. In RDCOO-HEP, the time step used is based upon the M1ite stability criterion. This criterion is explained in detail ard referenced in Ref.1 on p. 5-45 of CRBRP-3, Vol.1.
(9)
'Ihe referenced cbetrnents have been or will be made available for inspecticn and copying.
(10) The Applicants are not (bing developnent work cn REXONIEP. It such developnent is currently planned.
SET III AA-56
QUESITICH II (GDERAL)
Request for the following information is based cn our cxxrerns with respect to Validaticn (iii) and (iv) alx:rve.
In the Applicant's answers to the generic questions (b) and (c) below, the Applicant is requested to be responsive to these concerns.
With respect to each statment, assertion or assumption fran Section M.2 of the PSAR) identified below, please provide the follcwing information (tmless noted otherwise). (NorE: the following numbered Interrogatories are identifie3 by the page and/or paragraph number fran the PSAR in paren-thesis.) [Where appropriate, the parts of the question have been restated to reflect the protocol for discovery agreed to by Applicants, Staff, and Intervenors NRDC et al.]
a)
Identify by name, title and affiliation the priznary Applicant en-ployee(s) cr consultant (s) that has the expert knowledge required to support the statenent, asserticn, or asstrnption; b)
Describe in detail the supportirg evidence for the statenent, assert-ion, or assumption and where wuriate the raticnale for the approach taken.
c)
Provide any additional informaticn requested following each statenent, asserticn, or asstrapticn.
d)
'Ib the extent that any answers to the above questions are based cn referenced material, please supply the references.
e) Explain Wether Applicants are Iresently engaged in or intend to engage in any further research or work which may affect Applicants' answer. 'Ihis answer need be Irovided cnly in cases Where Applicants intend to rely upon cn going research not included in Section 1.5 of the PSAR at the I)& or construction permit hearing cn the CRBR. Failure to provide such an answer means that Applicants (b not intend to rely ugn the existence of any such research at the DR or construction permit hearing cm the CRBR.
SEP III AA-57
f) Identify the expert (s), if any, whon Applicants intend to have testify cm the subject matter questioned.
State the qualifications of each such expert. 'Ihis answer need not be provided until Applicants have identified the expert (s) in question or determined that no expert (s) will testify, as lcmg as such answe_ Irovides reasonable notice to Intervenors.
ANSWER II (GENERAL)
'Ihe following responses are identical for all interrogatories except where supplanentary information is provided in the response.
(a) Dennis M. Saitick, 2 nager, Safety Analysis, General Electric Canpany, Fast Breeder Reactor Department, 310 De Guigne Dr., Sunnyvale, California 94806, has the expert knowledge required to support the responses iden-tified in the attached affidavit.
(b) and (c) See responses 1-29 below.
(d) 'Ihe referenced doctanents have been or will be made available for inspection and copying.
(e) The Applicants are currently analyzing this area and will provide pertinent information as it beccries available.
(f) At the present time, the Applicants have not determined the experts, if any, whcm they interrl to have testify cn the subject matter questioned.
l l
M7FE: Question II-1 pertains to the potential for transition phase occuring from 'IOP events in the DJEC configuration.
l l
l SET III AA-58
QUESTION II-l (P6.2-92, par. 1-4) The last two sentences in the first paragraph beginning with, "If the blockage.
. the first and last sentence in the secmd paragraph; and all of paragraphs three and four.
(c) With respect to the statments in these first four paragraphs concern-ing blockage, relocation and reactivity effects and the conclusion that
" potential for a transiticn phase occurrirg in the IIhr 'IOP event is negligible," depend cn assumptions made in the SAS3A (including SASBIDK) evaluations, to what extent are each of the statements and the conclusion sensitive to asstrnptions concerning each of the following:
(i) the mechanical effects involved in fuel lodging; (ii) the location of fuel blockages above the core:
(iii) the locaticn and degree of fuel failure; (iv) the position of the bulk soditn level above the ejected fuel slug; (v) the dynamic pressure across the slug; (vi) the rate at which soditrn vapor is produced; (vii) the variations in driving pressure:
(viii) the rate of vapor production frcxn various heat transfer processes in conjuncticn with the rate of condensatien of these vapors which play a role in determining the dynamic pressures acting cn the ejected material cn the above core structure:
(ix) the use of equilibritrn thermodynamics; (x) kinematic processes included; (xi) the nelten fuel coolant interacticn model asstmed; (xii) the choice of tncertain reactor parameters incitxling fuel, steel and soditrn reactivity worths arri reactor loading patterns as a function of burnup.
Please provide detailal responses for each case (i) through (xii).
ANSWER II-l (b) and (c)
'Ittis question appears specific to the hcrnogeneous core which is not the current design. Accordingly, it requires no answer.
Section 8 of CRBRP-GEFR-00523 describes the analysis for the current core design.
i l
SET III AA-59 l
NorE: Questicn II.2 pertains to BOEC configuration.
QUESTINS II-2 (N.2-92, par. 5) Generic answers (a) and (b) are not required.
(c)
With respect to the statments in these two paragraphs and the ccnclusion".
it is believed that energy released frcm a partial core meltdown would be bounded by those events which could follow termination of the initiating phase of the IDEX: IDF accidents diich are considered next,"
to what extent are each of the statemer.ts aM the conclusion sensitive to asstznptions concerning each of the following:
(i) the mechanical effects involved in fuel lodging; (ii) the location of fuel blockages above the core:
(iii) the locaticn aM degree of fuel failure; (iv) the position of the bulk soditm level above the ejected fuel slug; (v) the dynamic pressure across the slug; (vi) the rate at which soditm vapor is produced; (vii) the variations in driving Iressure:
(viii) the rate of vapor production frcm various heat transfer grocesses in conjunction with the rate of condensation of these vapors which play a role in determinirg the dynamic pressures acting cn the ejected material on the above core structure:
(ix) the use of equilibritm thermodynamica; (x) kin eatic processes included; (xi) the nelten fuel coolant interacticn nodel asstmed; (xii) the choice of tncertain reactor parameters including fuel, steel ard r.aditm reactivity worths and reactor loading patterns as a function of burnup.
Please provide detailed answers for each case (i) through (xii).
ANSER II-2 (c)
'Ihis question appears specific to the luivge.neous core Wuch is not the current desicp. Accordirgly, it requires no answer.
SL'T III AA-60
i Section 8 of CRBRP-GEFR-OOS23 describes the analysis for the current core design.
NCTTE:
Questions II-3 and II-4 pertain to the ptential for transition phase fran IM events.
QUESTION II-3 (EE.2-93, M r. 2) Generic answers (a) and (b) are not reqaired.
(c) What criteria are used to define tennination of the initiating phases of IM accidents? Of 'IOP accidents?
What criteria are used to define initiation of the transiticn phase of LOF accidents?
Of 'IOP accidents?
Describe in detail each case considered where conditions in the core are such that a true hydrodynamic disassenbly calculaticn is justiCM as a means of continuirg the analysis to permanent shutdown.
ANSWER II-3 (c) mis question appears specific to the hmogeneous core which is not the current design. Accordingly, it requires no answer.
Section 8 of CRBRP4EFR-00523 describes the analysis for the current oore design.
QUESTICN II-4 (P6.2-93, par. 2 and 3) Generic answers (a) and (b) are not required.
(c)
We statenents and conclusions in paragraph 2 beginning at "In these less energetic.
." and the two statenents in paragraph 3, "Wus, the SEr III AA-61
is proceeding gradually into a ocmpletely rrolten state," and core.
"This is judged to be one of early removal of large amounts of fuel.
ard below the core," depend cn asseptions made in the SAS3A IM CR evaluations.
'Ib what extent are each of the statements ard the conclusions sensitive to assumptions concerning each of the following:
(i) the rate at which seditrn vapor is produced; (ii) the use of equilibritu thermodynamics; (iii) the equations of state assumed; (iv) the rnolten fuel-coolant interaction model assumed; (v) the choice of uncertain reactor parameters including fuel, steel, and soditu reactivity worths and reactor loading patterns as a function of burnup.
Please provide detailed answers for each case (i) through (v).
ANDER II-4 (c)
This question appears specific to the htmogeneous core which is not the current design. Accordirgly, it requires no answer.
Section 8 of CRBRP-GEFR-00523 describes the analysis for the current oore design.
tere: Questions II-5 through II-8 pertain to tM potential for the existence of steel blockages.
QUESTIONS II-5 (PE.2-93, par. 4) Generic answers (a) and (b) are not required.
(c)(1) Identify all TREAT tests that were designed to simulate unprotected 12 QR Ihencama; (ii) with respect to these, identify all 'IREAT tests that focused cn coolant and claddity behavior; (iii) identify and supply all ANL doctmentation of results of those 'IREAT tests identified in (ii)
SEP III AA-62 I
above; (iv) identify all TRFAT tests identified in (ii) above where upper cladding blockage was observed; where lower claddin3 blockage was observed.
ANSWER II-5 (c)
Section 10.2 4n CRBRP-GEFR-00103 and Section 8.2 in CRBRP-GEFR-00523 discuss the same material that this Interrogatory makes reference to.
(c)(1) The TRFAT tests that were designed to simulate tmprotected IDF CIA pherci.erman were the L, R, and F series experiments.
(c)(ii) The test series that focused on coolant and cladding behavior were the L and R series experiments.
(c)(iii)
Documentation of the results of the TRFAT tests identified in (ii) above can be fourd in the following doctanents:
(1) Reference 59 in CRBRP-GEFR-00103.
(2)
E. Barts, et al., "Sts: mary and Evaluaticn, Fuel Dynamics Ioss-of-Flow Experiments (Tests L2, L3, L4)," ANL 75-57, Sept. 1975.
(3) Reference 36 in CRBRP-GEFR-00103 (4) Reference 55 in CRBRP-GEFR-00103 (5) Reference 21 in CRBRP-GEFR-00523 (6) Reference 23 in CRBRP-GEFR-00523 (7) Reference 28 in CRBRP-GEFR-00523 (8) Reference 29 in CRBRP-GEFR-00523 (9) Reference 31 in CRBRP-GEFR-00523 (10) Reference 32 in CRBRP-GEFR-00523 (11) Reference 33 in CRBRP-GEFR-00523 (12) Reference 34 in CRBRP-GEFR-00523 (13) Reference 46 in CRBRP-GEFR-00523 (14) R. Sinms, et al., "Fbel Mx.lon in Experiments Simulating INFBR Ioss-of-Ficw Accidents," ANL/ RAS 80-25, Nov.,1980.
(15) R. Sinms, et al., " TREAT 'Ibst L7 Sinulating an INFIR Ioss-of-Flow with FIR-Type Fbel," ANL-80-ll2, Nov., 1980.
SET III AA-63
(c)(iv)
'ntis informaticn is found in the doctments listed in (iii) above.
QUESTION II-6
( N.2-93, par. 4) Generic answers (a) and (b) are not required.
(c)
Clarify how Reference 12 is supportive of the conclusicn, "It is presently believed that such blockages do not form or are ino:nplete in alnest all subassertblies."
ANSWER II-6 (c)
'Ihe reference referred to in the question is Ref. 57 in CRBRP-GEFR-00103 ard the discussicn of this subject has been revised by more recent data from Ref. 49 in CRBRP-GEFR-00523. See Section 8.2.1 of CRBRN-00523.
QUESTICN II-7 (M.2-95, par. 1)
Since incomplete blockages will be innediately melted out whai noiten fuel passes through them, these blockages can be neglected in the analysis.
ANSWER II-7 (b) and (c)
'Ihe discussion of this subject has been revised by nore recent data frcm Ref. 49 in CRBRP-GEFR-00523. See Section 8.2.1 of CRBRP-GEFR-00523.
QUESPICN II-8 (P6.2-93), par. 4) Wat is the basis for asstning a molten steel-coolant interacticn would not occur as the molten claddin3 contacts ligald soditm at the top and botton of the core? If it cbes occur how is this ' interaction nodeled?
SETF III AA-64
ANSWER II-8 (b) and (c)
Movanent of rrolten cladding prior to fuel rrevement would occur into voided regions of tha fuel asserbly where the steel would freeze.
m evidence of steel-sodium interactions were observed in the TREAT R-series experiments in which early cladding relocations occurred (Ref. 23 in CRBRP-GEFR-OO523).
The potential for a dispersed flow regime ejection of noiten fuel-steel-gas mixtures was discussed cn page 10-7 of CRBRP-GEER-00103 ard nere recent evaluations are Irovided in Section 8.2 of CRBRP-GFR-00523.
Neglecting the nolten steel-sodiun interactions is primarily fourded on the dispersed flow regime negating the ptential for large scale, intimate liquid-liquid contact occurring.
Out-of-pile experiments with both simulant materials and reactor naterials support this conclusion (Ref. 63 in CRBRP-GEFR-00103).
NrfrE: Westicos II-9 through 11-16 pertain to extended fuel notion.
QUESTION II-9 (E6.2-95, par. 2)
'Ibe development of fuel and steel vapor pressures strongly suggest that the fuel moticn will be monotonically dispersive.
ANSWER II-9 (b) and (c)
References 57 and 60 in CRBRP-GFR-00103 and nere recent analysis in Secticn 8.3 of CRBRP-GEER-00523 supported by Reference 76 in CRBRP-GEFR-00523 provide the basis for the statement set forth in the interrogatory.
j QUESTICE II-10 1
(FE.2-95, par. 2) Entrainnent of clad could arise.
criteria provided in Ref. 69.
SEP III AA-65
(c)
Under What corditions muld entrairrnent of clad be expected not to occur?
If cladding steel slushing could occur, could fuel sloshing also ocrur?
'Ib what extent, if any, did entrairment take place in the in-pile Im meltdom experiments?
Identify and discuss a'l uncertainties in these in-pile IN meltdown experiments with respect to the applicability of these results to CRBR IN CIR corditions.
In the post-test analysis of the IM meltdown experiments to what extent was the relocated fuel and steel heterogeneous? Identify and discuss all uncertainties in the applicability of the stability criteria provided in Reference 69 to CRBR IN CDA cordi-tions. Discuss.
ANSWER II-lO (b) and (c)
Reference 57 in CRBRP-GEFR-00103 provides the bases for this statenent.
Clad-fuel entrairment is likely to occur for all postulated core-disruptive hypothetical accidents.
Ebel sloshing in the same sense as clad sloshi:v3 is considered very unlikely since follcuirg fuel meltirx3 a large fraction of the neutron heating will appear as latent heat of vaporization.
'Ihe resultant drivirg force is much larger than that represented by the solitrn vapor streaming (which leads to clad sloshing) and therefore results in essentially monotonic dispersal of the fuel.
Even with no entrairment, ccnsidering the fact that the steel-fuel system is largely gedispersed by design, no significant change in the accident segaence as depicted in the CRBRP-GEFR-00103 or CRBRP-GEFR-00523 is anticipated.
QUEETTICH II-11 (FE.2-95, par. 2)
Heat Transfer frcm the fuel to the clad will result in rapid els3 vaporizaticn ard dispersal of fuel.
AN9hER II-ll (b) and (c)
Deference 57 in OtBRP-GEFR-00103 (same as Ref. 50 in CRBRP4EFR-00523) and tpdated analysis in Secticn 8.2 of CRERP-GEFR-00523 provide the' basis for the statenents set forth in the interrogatory.
SEP III AA-66
GESTICN II-12
( M.2-95, p r. 2) only a small fraction of the available clad material is 4
necessary, since the liquid-to-vapor density ratio is in the order of 10.
ANSWER II-12 (b) and (c)
Relatively snall vapor velocities (rv2m/sec) are required to fluidize the fuel for void fractions of interest (40 to 50% ).
Hence, only a small voltrne of liquid steel is necessary to produce these velocities.
Even small puddles of steel left behind cm the fuel pins wuld be sufficient.
(See Ref. 57 in CRBRP-GEFR-00103)
OESTICN II-13 (M.2-95, par. 2)
Ebrthennore, because of the above entrainment processes and since nciten steel is known to wet oxide fuel, the local heat transfer between fuel and clad can be approximated by equilibritrn corditions.
ANSWER II-13 (b) and (c)
'1he final outcone is not sensitive to the assanptions of equilibrium corditions.
Irx al nonequilibritan corditions between fuel ard steel will lead to similar conclusions regarding fuel dispersal and boil-up.
QUESTICN II-14 (M.2-95, pr. 2)
'Ihe vaporizaticn rates are therefore nore than suffi-cient to fluidize ard to maintain a dispersed fuel-steel systen.
SEr III AA-67
l ANSWER II-14 (b) and (c)
)
See Section 8.2 of CRBN-00523.
QUESTICN II-15 (P6.2-95, par. 2) Generic answers (a) and (b) are not required.
(c)
Provide more detail models of possible phencraena and events taking place between clad nelting and fuel dispersal above the gas plenum region, givirg estimates of the time sequence of events, material description, novaments and relocations.
ANSWER II-15 (c)
Presently available nodels of the therrmena and events taking place between clal meltirg ard feel dispersal durirg the initial stages of core disrup-tion are those used by the Applicant in CRBRP-GEFR-00103 and CRBRP-GEFR-
- 00523, i.e., SAS3D, PIlf!O 1, ard PLUID 2. 'Ihe Applicant's updated analysis of fuel penetration into assetbly rod structure is given in Section 8.2.2 of CRBRP-GEER-00523.
QJESTICE II-16 (FE.2-97, Inr. 4)
'Ihis Irocess can continue since experiments (Reference
- 6) (with both simulant materials and reactor systen and a cold liqaid is unlikely to result in sustained interaction Iressures larger than the vapor pressure or systen pressure of the hot fluid.
l (c) mat is the basis for rejecting the analysis and conclusions presented in "'Ihm Role of Spontaneous Nucleation in 'Ihermal Explosions, Frecm/ Water Experiments," S. J.
- Ibard, R. W.
- Hall, G.
E.
Brown, Proc. Fast Reactor l
Safety Meeting, April 24, 1974, Beverly Hills, Calif., USAEC Report CCNF-740401-P2 (1974), Ip. 933-936.
SEP III AA-68
ANSWER II-16 (b) and (c)
(b)
Reference 63 in CRBRP-GEFR-00lO3 supports the conclusion that contact tanperature larger than the spontaneous nucleation temperature is required for sustained Iressure generation. 'Ihis cxandition is not satisfied for the dispersed fuel-steel-soditrn systern.
(c)
The experiments and analysis reported in "We 1ble of Spontaneous Nucleaticri in hermal Explosions, Frecrt/ Water Experiments," by S. J. Board et el., su; port the criteria stated in 16(b) above.
Ebr further classifi-caticri, see the following reference:
R. E. Henry, H. K. Fauske, and L. M.
Methber, " Vapor Explosions with Subcooled Freon," Trans. Am. Nucl. Soc.,
Vol. 22, 1975.
See Secticri 8.0 of CRBRP-GEFR-00523 for recent Applicant evaluations.
!UTE:
Questions II-17 through II-23 pertain to fuel behavior following postulated core plugging.
QUESTICN II-17 (F6-2-98, per. 4) We disruption of the fuel in different subassernblies is relatively coherent across the core due to the high power levels in the initiating phase of the accident. All of the subassertlies experience fuel disrupticx1 within a few seconds of each other.
(c)
Define in nore detail the extent of incoherence that might be in-volved, including the extent of the incoherence in the position of fuel temperatures, reactivity, insertion mechanisms and their spatial listribu-tion, and time sequence of events.
l SET III AA-69
ANSWER II-J7 (b) and (c)
'Ihis question appears specific to the Iw+wous core duch is not the current design. Accordingly, it requires no answer.
Section 8 of CRBRP-GEFR-00523 describes the analysis for the current core design.
GJESTION II-18
( M.2-98, par. 4) Generic answers (a) and (b) are not regaired.
(c)
What happens to the control rods durirry the transition phases of the I& and 'IDP CIas? How are these effects rrodeled?
ANSWER II-18 (c)
'Ihe ocntrol rods are neglected in the transition I ase analyses.
Cbntrol h
rcri material is assumed to be absent frcm the hcrogenized pools that eventually form.
QUESTIN II-19
( M.2-98, par. 4) Generic answers (a) and (b) are not required.
(c) Explain in detail all alternative rrodes of accident progression, if it is asstzned that the melting of wrapper can steel is not faster than the rencval of the postulata$ frozm steel-fuel blockage above the core.
ANSte:R II-19 (c)
'!he identified rrodes of hypothetical accident progression for the requested asstzrpticn that the melting of wrapper can steel is not faster than the removal of the postulated frozen steel-fuel blockage above the core were SET III AA-70
described in detail in Section 10.1 of CRBRP-GEFR-00103 and are uplated for more recent data in Section 8.2 of CRBRP-GEER-00523.
GESTIm II-20 (F6.2-98, par. 4) Generic answers (a) and (b) are not required.
(c) Explain in detail all alternative nodes of accident progression if the core cannot be adequately represented by a single coherent boiling region.
ANSWER II-20 (c)
This question appears specific to the herogeneous core Which is not the current design. Accordirgly, it requires no answer.
Section 8 of CRBRP-GEER-00523 describes the analysis for the current core design.
QUESTION II-21 (EE.2-98, Enr. 5) Generic answers (a) and (b) are not regaired.
(c)
'Ihe basis fcr asstrairg a vanishirg small viscosity, thereby elimi-nating laminar and turbulent flow regimes.
ANSWER II-21 (c) 1he basis for this statement is found in Reference 58 of CRBRP-GEFR-00103.
QJEEFTION II-22 (F6.2-101, par. 2) Generic answers (a) and (b) are not required.
SET III AA-71
(c)
Provide detailed description of the series of experiments being performed to investigate fuel-coolant and steel-coolant interaction postulated if hot fuel and steel are driven out the upper and lower core structure encounter soditrn, ard all writeups (includirg intemal mer:oranda) of results (final and geliminary) of these experiments.
ANSWER II-22 (c)
Section 8.2.6 of GBRP-GEFR-00523 provides an updated sumury of the informaticn requested.
QUESTION II-23 (EE.2-101, p r. 3) Since melt-through of the structure is anticipated well before the power level has dropped to 1%, a significant fraction of the fuel-steel mixture is likely to be rapidly ejected.
ANSWER II-23 (b) and (c)
Evaluations cn the current design in Section 8.3.5 of CRN-00523 indicate that small fractions of fuel-steel mixture are ricre likely to be ejected.
IUTE: Questions 11-24 through II-29 pertain to reactivity effects in a disrupted core.
QUESTICE II-24 (F6.2-101, pr. 5)
'Ihe asstmption that reactivity calculations cn a disrupted core can be adequately modeled usin3 two-d2.mensicn r-z gecnetry?
SET III AA-72
ANSER II-24 (b) and (c)
'ihe types of reactivity miculations mde in CRBRP-GEFR-00103 and CRBRP-GEER-00523 are adequate with twcMilmensional r-z gecnetry because the reactivity changes considered are relatively large, and Irimarily due to axial fuel motions and soditan voiding.
Mare sophisticated axlels (e.g.,
three-dimensional hex-z geometry) are useful primarily in obtaining accurate estimates of physics parameters related to normal operation.
Using such nodels to estimate reactivity changes due to soditan voiding and fuel moticn would not yield significantly different results than would be obtained frcn an r-z model.
'Ihis is a standard engineering approach, the adequacy of which is borne out frcm physics calculations performed for conditions of normal operation.
In such calculations it is standard practice to use simplified nodels to ecmpute reactivity changes, after having verified the approach by periodic checking with three-dimensional calculations.
QUESTION II-25 (FE.2-101, ptr. 5)
'Ihe asstrrption that reactivity calculations cn a disrupted oore can be adequately modeled using the nine group cross-section set.
ANSWER II-25 (b) and (c)
'Ihe types of reactivity mlculations mde in CRBRP-GEFR-00103 and CRBRP-GEER-00523 are adequately modeled with a nine-grotp cross-section set j
because botniaries were chosen to ensure that the inportant physical phencmena were treated properly.
Takirx3 such care results in very good agreenent with calculations made with finer-grote cross-section sets.
l l
e SEP III AA-73
QUESTICH II-26 (P6.2-102, par. 1) Generic answers (a) and (b) are not required.
(c)(1) What is meant by "significant" in".
results has a significant inplication for the transition @ase
."?
(ii)
What are the uncer-tainties involved?
(iii)
Identify the higher pr subassenblies where steel in the core is nearing the Iniling point.
(iv)
What are the uncertainties involved in this estimate?
(v)
Quantify " nearing the boiling paint."
(vi) What are the tncertainties?
(vii)
Quantify what is meant by "imninent" in" rapid production of steel vapor is imninent.
" (viii) What are the tncertainties involved? (ix) What are the uncertainties that lead to the choice of the word "should" in ".
should tncertainties relative to the time of cnset of steel vapor produc-ticn.
(xi)
Quantify what is meant by ".
. delayed for any significant period of time.
" (xii) What are the incertainties involved?
(xiii)
Quantify What is meant by " mild" in "... another mild burst would follow.
.?"
(xiv)
What are the uncertainties?
(xv)-(xxviii) With respect to each of the above questions, (i) through (xiv), relative to (F6.2-102, par. 1), discuss in detail the inplicaticns Lf incoherences in the phe-nonena involved ard consequent lack of synmetry in time sequences and gecmetry.
ANSWER II-26 (c)
(c) (i) through (xxviii).
Sections 8. and 9.4 of CRN-00523 provide the current assessment of the moltout and large-scale Irol phase of the HCDA.
QUESTICN II-27 (P6.2-102, par. 2) Generic answers (a) and (b) are not requi-M.
(c)
Whose " current best estimate" is being referred to in 'the seccnd sentence?
Are there differing views known to the applicant as to what SET III AA-74
T constitutes the "best estimate" as to whether fuel ejected into the blanket will travel through then without plugging?
If so, present in detail the basis for these alternative views.
ANSW:R II-27 (c)
'Ihe extent of penetration of flowing nolten materials through various reactor structures ard its basis are discussed in Section 8.2 of CRBRP-GEFR-00523.
QUEE7 PION II-28 (M.2-102, par. 3) Generic answers (a) and (b) are rot required.
(c) What are the implicaticns of nonuniform renoval of fuel fran the core?
l ANSWER II-28 (c)
In terms of calculation of the reactivity due to fuel relocaticn, the node of fuel removal is insignificant.
'1he actual state of the reactor for which the reactivity is being acunputed has far greater effects upon the resultirg energetics.
QUESTICH II-29 (M.2-lO3, par. 2) Generic answers (a) and (b) are not required.
l (c) Whose "best-estimate path" is beirg referred to here?
Are there differing views known to the applicant as to what constitutes the "best-estinete path"?
If so, present in detail the basis for these alternative views.
i SEP III AA-75
ANSWER II-29 (c)
See Section 8.2.2 of CRBRP-GEER-00523.
SEP III AA-76
FOURn1 INTERROGATORY SET PREA> ELE 10 QUESTIONS In Irevious interrogatories information was requested concerning four distinct validations relative to the nodels ard ccriputer codes used in the analysis of CRBR CRs, namely, 1)
Validation that the code's output is the correct runerical calculaticn that should result fran a given set of input data and the nodel assumptions; 11)
Validation of the rxdels against actual experimental data; iii) Validaticn that the models can be extended to the CRBR; and iv)
Validation that the input assumptions for the CRBR case are adequate with respect to the CDA analysis, i.e.,
are support-ed by experimental evidence.
By " adequate", here and below, we mean that the calculaticns will not underestimate the CDA work Intential (i.e.,
forces and resulting energetics of a CDA) or over estimate the contalment capability of the reactor with respect to a CR.
With respect to the following requests for information we are concerned Irimarily with the fourth validation-validation that the input assumptions for the CRBR case are adequate with respect to the CDA anal-ysis.
Here we are not so nuch concerned with the validity of the nodel expressions as with the uncertainties in the VDRE work energy calculations due to propagation of uncertainties in a) the paraneters used and b) the mcdel input data and due to any synergians among these uncertainties and the nodel asstunptions.
QUESPION I With respect to the calculations identified below, under (A) through (D),
please provide the following informaticn [Where awwplate, the parts of SEP IV AA-77
l l
l l
the questicn have bem restated to reflect the protocol for discovery l
agreed to Iry Applicants, Staff, and Intervenors NRIC et al.]:
i l
l l
1)
List armi identify all model input data (exclusive of coiing
(
flags arx1 inputs that specify codirx3 opticns, criteria, printout formats, etc.) and all model parameters that cane into play in each of the nodels utilized in the cxxpled-code accident analysis calculations, including but not limited to input data arti parameters in SAS3A and VDRJS-II.
Exclu3e parameters ret called into use because a subroutine, or part thereof, was not utilized.
- 2) Describe in detail the basis for the choice of each input datum and model parameter listed abcne.
1)
In each case quantify the tncertainty in the value selected; 11)
In each case indicate whether the value is based on first principles, experimental measurements, tnvalidated by;cthesis, output of a coupled model (eg, VDRE-II input obtained fran SAS3A output), etc.,
iii) In each case indicate whether the etcice of the input datum cr model parameter was selected to represent the "best estimate", or a bounding or " conservative value", W1ere " conservative value" here means a value chosen so as not to underestimate the accident consequences, e.g.,
work potential.
3)
For each input dattra and nodel Imrameter with uncertainty listed in 1) above, indicate in quantitative terms the magnitt:3e of the uncertainty introduced into the final calculation of the work energy by the uncertainty in the input datun or model parameter.
In additicn, discuss in detail any synergistic effects resulting frcm carbinaticms of tncertainties in the input values, model parameters, and model assunptions.
In each case discuss the basis for the estimate of how the moertainties pxpagate, eg, include and discuss all parameters analysed used to test the effect of uncertainties.
SET IV AA-78
'l l
l
4)
Identify by name, title and affiliation the primary Applicant anployee(s) or consultant (s) that has intimate workirg knowledge of the basis for the selection of the parameter or input datum.
5)
'Ib the extent that arry answers to the above questions are based on referenced material, please supply the references.
6)
Explain whether Applicants are presently ergaged in or intend to engage in any further research or work which nay affect Applicant's answer. This answer need be provided only in cases where Applicants intend to rely upon cn going research not incitried in Section 1.5 of the PSAR at the IMA or constructicn permit hearing cn the CRBR. Failure to provide such an answer means that Applicants do not intend to rely ipon the existence of any such research at the IMA or constructicn permit hearing cn the CRBR.
7)
Identify the expert (s), if any, whon Applicants intend to have testify cn the subject matter questioned. State the qualifications of each such expert. 'Ihis answer need rot be provided tntil Applicants have iden-tifis3 the expert (s) in questicn or deternined that no expert (s) will testify, as long as such answer provides reasonable rotice to Intervenors.
(A)
'Ihe three SAS3A/ VENUS-II calculations for the IDP-BOEC configuration considered in Section EE.2.6.2.1 (i.e., 40$/sec, 50$/sec and 100$/sec ramp rates) an1 stmnarized in Table F6.2-2d.
In the latter two cases (50$/sec and 100S/sec) it is not necessary to duplicate information previously provided with respect to the 40S/sec case, e.g., much of the SAS3A input data. Here, indicate cnly those parameters and input data that differ frczn those previously listed.
(B)
The one SAS3A/ VENUS-II calculation for the IDF-E0EC configuration considered in F6.26.2.2 is stmnarized in Table F6.2-22.
l (C)
'Ihe four SAS3A/VEi E -II calculations for 'IOP-EDEC configurations consideral arri stmnarizai in Table EE.2-23.
In the last three cases, (i.e., the namnd SOS /sec, 75S/sec, and LOOS /sec) it is rot necessary to l
i SET IV AA-79 I
duplicate information Ireviously prcnided with respect to the first SOS /sec case.
(D) The four VDRE-II disassably calculations for the IDEC IN innediate re-entry case and three VDRE-II disasser:bly calculations for hcmogenized core re-entry considered in EE.2.6.4 and surmarized in Table EE.2-24.
It is not necessary to duplicate information in each case; indicate only those parameters and input data that differ frcm those previously listed.
ANSWER I
- 1) This question appears specific to the hcnogeneous core which is not the current design.
Accordirgly, it requires rn answer.
CRBRP-GEFR-00523 describes the analyses for the current core design. We SAS3D and VDTJS-II input data used in the calculations have been set forth for selected cases in CRBRP4FR-00523 on page 4-19 and in Appendices D and G.
Riysical descriptions of the ntnerical input data listings are found in References 5 and 6 of CRBRP-GEFR-00523.
- 2) This question appears specific to the hcrogeneous core which is rot the current design.
Accordingly, it requires no anwer.
CRBRP-GEER-00523 describes the analyses for the current oore design.
We basis for the choice of the SAS3D base case input parameter values is presented in Section 4.3 of CRBRP-GFR-00523.
Referring to the VDRE input description in ANL-7951, the following data arises frcm the gecnetric model chosen, and frcm options chosen which were deened q;rvslate for the analyses made:
(a) All data cn cards 2-10, 14, 15, 21, 22.
(b) Data for all regions cn cards 23-27, md 33.
he power densities on cards 11, the material worths cn cards 28, and the Doppler troadening feedback parameters cm card 31 were cbtained frcm calculations similar to those described in Section 4.3 of the PSAR.
We I
l ser IV AA-so l
data cn cards 16 and 17 were obtaina3 fran SAS3D output, as was PZEBO on card 20, and data cn cards 32. 'Ibe rest of the data cn card 20, are desned appropriate for the analyses model.
The data cn card 30 are tased cn experinontal measurements.
'Ihe data cn card 50 are based cn SAS3D output, which was usal to obtain an average core tanperature, Which ws in turn used by VDRJS to obtain an appropriate r-z tanperature distributicn.
Input parameter charges in the base case input decks are described in Secticn 9 of CN-00523.
3)
'Ihe impact of uncertainties with respect to input data is discussed in CRBRP-GEFR-00523. Various parameters were varied to determine sensitivities to data ard nodelirg uncertainties.
4)
Dennis M. Seitick, mnager, Safety Analysis, General Electric 02npany, Past Breeder Reactor Department, 310 DeGuigne Drive, Sunnyvale, California 94006.
5)
'Ihe referenced docunents have been cr will be made available for inspecticn and mpying.
6)
'Ihe Applicants are currently analyzirg this area and will provide pertinent informaticn as it becones available.
- 7) At the present time, the Applicants have not determined the experts, if any, Whan they intend to have testify cn the subject matter questioned.
QUESTION II (General)
Itaquest for the following informaticn is based cn our concerns with respect to validaticn (iii) and (iv) noted greviously.
In the Applicant's answers to the generic questions (b) and (c) below, the Applicant is requested to be responsive to these concerns.
With respect to each statenent, assertion or asstrnpticn (based on Section F$.2 of the PSAR) identified below, please provide the follcwing infonna-tion (unless noted otherwise).
IUTE:
'Ibe following ntrnbered interroga-tories are identified in parentheses by the page and/or paragraph nmber fran the PSAR or by a code ntrrber identifyirg an NIC gaestion addressed to the Applicant. [Where appropriate, the parts of the question have been restats3 to reflect the protocol for discovery agreed to by Applicants, Staff, and Intervenors imDC et al.]
a)
Identify by name, title and affiliation the primary Applicant em-playee(s) cr consultant (s) that has the expert knowledge required to support the statenent, assertion, or asstmpticn.
b)
Describe in detail the supporting evidence for the statement, asser-tion, or asstrnpticn and where appropriate the rationale for the approach taken.
c)
Provide any additional information requested following each statenent, assertion or asstunpticn.
d)
'Ib the extent that any answers to the ab:ve questions are based cn referenced material, please supply the reference, e) Explain Whether Applicants are presently engaged in or intend to engage in any further research or work which may affect Applicant's answer. This answer need be provided only in casec Where Applicants intend to rely upcn cn goirg research not included in Section 1.5 of the PSAR at the IMA or construction permit hearing cn the CRBR. Pailure to provide such an answer means that Applicants do not intend to rely upcn the existence of any such research at the IMA or construction permit hearing on the CRBR.
f)
Identify the expert (s), if any, whcm Applicants intend to have testify on the subject matter questioned.
State the qualifications of each such expert. This answer need not be provide 3 until Applicants have identified the expert (s) in questicn or determined that no expert (s) will testify, as long as such answer provides reasonable notice to Intervenors.
SEP IV AA-82
ANSWER II (General)
'Ihe answers to questions II(a), (d), (e) and (f) are the same as those for questions I(4), (5), (6) arrl (7) respectively.
NCTTE:
Questions 11-1 through II-4 pertain to analysis of hydrodynamic disassemblies.
QUESTION II-l NRC Question 001.497.
In this question to the Applicant the Staff states
"... disassembly calculations results will depend on the results of...the equation of state for disasseubly phase For the disassembly I ase h
perhaps the single nost irmortant uncertainty is the equation of state especially for tarperatures and pressures close to the critical point of fuel vapor.
Generic answers (a) and (b) are not required.
(c)
Daes the Applicant agree with this conclusicn? If not, why not? In either case (if so or if not), discuss in r: ore detail the basis for this conclusicn.
Doctment the uncertainties in the equation of state and indicate the effect these incertainties could have cn the results of the hydrodynamic disassently analysis.
ANSER II-l (c)
Jackson et al., has studied the effects of using various formulations of the fuel vapor pressure behavior proposed for use in disassernbly calcula-tions, and showed that cnly minor differences energed fran using the different models. (J.
F.
- Jackscn, A.
M.
- Eaton, R.
M.
- Hall, T. F.
Bott SErr IV AA-83
(Brigham Yourg University), "The Influence of Equation-of-State Uncertain-ties cm Fast hiactor Disasserbly calculations, Trans. An. !bcl. Soc. 22, p.
368 (1975). They further showM that the Menzies fonnulaticn currently in VENUS-II predicted the greatest energy release.
Uncertainties that might exist close to the critical point are not of great concern, because none of the disasserbly cases presented in GBRP-GEFR-00103 or CRBRP-GEFR-00523 lead to tenperatures anyWhere near the critical temperature.
QUESTION II-2 (FE.2-105, par. 2)
Each of these assumptions contradicts present under-standing of the phencmem and the ccrnbinaticn of all three is highly inprobable.
ANSWER II-2 (b) and (c) 1 This cpestion appears specific to the Inogeneous core Which is not the current design.
Accordingly, it requires no answer.
Section 8 of CRBRP-GEFR-00523 describes the analyses for the current core design.
QUESTICE II-3 (EE.2-106, par. 2)
In cannents on the proposed final environmental state-ment - liquid metal fast breeder reactor (with reference to p. 4.2-148 of the PEES-INFBR), the NRC Staff stated:
"...the bases for concluding that the total energy generated in a series of anall Iower bursts will be no greater than that generated in a single, large, pennanently-dispersive burst requires further study.
Further, the safety significance of such a conclusion, even if justified, is not clear at this tbne.
For exanple, further work is required to evaluate the effective mechanical danage fran repeated pulses if they do occur."
Generic answers (a) and (b) are not required.
SEP IV AA-64
(c)
Does the Applicant agree with this conclusion? If not, why not?
In either case (if so or if not), how are these considerations treated in the CRBR QA disasserbly analyses?
Is it rot pssible that a series of small power bursts could occur in such a fashicn that they would lead to a large reactivity inserticn at or near prompt critical and lead to a sustained supesguyi. critical burst? Is it not pssible that the small pcuer bursts could be due to i encmena having space as well as time, assymetries.
Ebr t
example, in such a fashicn that axial synmetry, as asstrned in VENUS, would be inappropriate for modeling the phencmena?
Discuss in detail the basis for the answers to the above.
ANSWER II-3 (c)
CRBR HCIA disassertbly analyses have not resulted in predictions of se-quences of small pwer bursts.
A series of anall power bursts occurring in such a fashion that they would lead to a large reactivity inserticn at or near pranpt critical arri cause a sustained supeig u yi. critical burst is judged to be highly unlikely.
'Ib obtain such a situaticn requires makiry a continuing series of arbitrary asstmptions regarding coherency of fuel notion.
It is true that the VENUS-II model does not allow for external treatment of fuel motion.
Ibwever, the reactivity effects of fuel notion can be es-timatal and included in the reactivity inserticn rate input to VENUS. 'Ihe resultant energetics are jtriged to be reasonable estimates, provided the void voltane is handled in a conservative manner.
(See R. B. Nicholscn and J. F. Jackson, "A Sensitivity Study for Fast hetor Disassembly Calcula-tions," ANL-7952 (1974).)
te SEP IV AA-85
OJESTICH II-4 (P6.2-106, Mr. 2) In all cases, care was taken to begin the disassably calculaticn early enough to ensure that conservative estimates of the energy generated were made in VI20S-II.
Generic answers (a) ard (b) are not required.
(c) Is it rot true that SAS3A may predict a disasserably ramp rate which is either unreasonable in sign or in magnitude at a specified core fuel tmperature (See Blewbis et. al., Proc. of the Fast Beactor Safety Meeting, Beverly fillls, California CONF 740401, p. 1324).
Itw is this considera-tion taken into account in modeling the transition frcm termination of SAS3A and/or the "transiticn phase to VINUS-II?
Describe hcw the VE20S ramp rate is formulated in light of the above.
ANSWER II-4 (c)
SAS3A or SAS3D, in and of thanselves, do not predict a "disassmbly ramp rate."
It is up to the user to decide when the disassenbly calculation should begin.
With respect to how this is done, a clear description appears cn p.
II-3 and II-4 of CRBRP-GEFR-00103.
The Applicant cannot ascertain how the authors of the referred to paper cbtained the driving reactivities; however, the results seen to be at variance with our experi-ence in CRBR calculations.
NETTE:
Qaesticns II-5 through II-9 pertain to transition recriticality considerations.
GJESTION II-5 (76.2-109, pr. 3)
First, the Iressu e frcan vapor generation in the boiling material in the core could terd to levitate the blockage.
SET IV AA-06
(c)
Provide a time dependent profile of the pressures abcne arx3 beloa the asstaned blockage.
ANSER II-5 (b) and (c)
Detailed calculations of the time dependent pressure above and below the blockage were not performed since a detailed calculational model is not currently available.
Qualitatively, the pressures below the blockage will be higher than those above the blockage.
'Iherefore, the blockage will be levitated.
QUESTICN II-6 (EE.2-109, pir. 3) Any material re-entering the core would most probably not ecme in as coherent slugs but rather ccme in gradually as the upper blockages are melted and " washed" out by the boiling turbulence below.
(c)
Quantitatively ocupare the probability of reentry due to gradual melting and washout with the probability of re-entry due to pressure relief downward.
ANSWER II-6 (b) and (c)
Calculations to quantitatively ampare the probability of re-entry due to gradual melting ard washout with the probability of re-entry due to pres-sure relief downward have not been performed since an appropriate calcula-ticral model is not currently available.
QUESTIN II-7 (P6.2-109, par.
Generic answers (a) and (b) are not required. Msterial injected into the blankets would terd to have a temperature profile that is steadily decreasing away frce the core...
l I
j ser IV AA-87 l
(c)
Provide a gaantitative essessment of the time dependent ternperature profile across the blockage.
ANSWER II-7 (c)
Calculations of the time dependent tenperature Irofil.e across the blockage were not performed since a detailed calculational model is not available.
Qualitatively, the sentence set forth in the interrogatory is supported by the statements which innediately folloa it on page 11-11 of CRBRP-GEFR-00103.
QUESTICN II-8 (FE.2-110, pr. 1)
A second reason that a recriticality is unlikely is that the reactivity of the systern, after the early part of the transition phase, would very Irobably be too low for fuel re-entry to return the systen to critical.
ANSWER II-8 (b) and (c)
Evaluation of the current design in Sections 8.2 and 8.3 of CRBRP-GEFR-00523 indicate that recriticality may occur,, but large coherent reactivity events are not foreseen.
QUESTIN II-9 (FE.2-110, p r. 1) The asstanption that the renainmg inner and outer core fuel is hcnogeniced.
(c) mat are the inplications in terms of possible ranp rates if it is asstuned that the fuel is not hcmogeniza3?
SET IV AA-88 I
ANSWER II-9 (b) and (c)
'this question appars specific to the hmogeneous core dtich is not the current design.
Accordingly, it requires to answer. Section 8 of CRBRP-CEPR-00523 describes the analyses for the current core design.
NOTE:
Questions 11-10 through II-16 pertain to the EDEI' I& inmediate re-entry case.
QUESTICE II-10 (F%.2-110, par. 4)
'Ihe asstrnption that the re-entry can be adequately modeled by limitirr3 the coherent re-entry consideration to the 36 sub-assenblies in the innermost ring of the outer enridrnent zone.
(c)
Itw are the control reds mMeled in the EDED Im Imediate Re-Entry Case?
ANSWER II-10 (b) and (c)
'Ihis question appears specific to the hernogeous core which is not the current design.
Accordingly, it requires to answer.
Section 8 of CRBRP-GETR-00523 describes the analyses for the current core design.
QUESTICN II-ll l
(P6.2-110, rar. 5) Asstrne that, in 75 percent of the subasserrblies in the irrs== t ring of the outer core zone, blockages fonn in the icwer portion of the upper blanket.
(c)
Miat is the basis for the dolce of the location of the blockage in t
the upper blanket?
SET IV AA-89
ANSWER II-ll (b) ard (c)
'Ihis cpestion appears specific to the hcr:ogeous core which is not the current design.
Accordirgly, it requires no answer.
Section 8 of CRBRP-GEFR-00523 describes the analyses for the current core design.
QJESTICE II-12 (M.2-328, Pigure M.2-157) Generic ans.sers (a) and (b) are not required.
(c) Why is the slope of the curve discontinuous?
ANSWER II-12 (c)
'Ihe curve referred to in this interrogatory is now Figure 11-2 in CRBRP-GEER-00103 ard should in fact be drawn as continuous and smooth.
QUESTTICN II-13 (M.2-111, pr. 3)...the inner and cuter core fuels are hcrogenized.
ANSWER II-13 (b) and (c)
'Ihis question appears specific to the hcrogeneous core which is not the current design.
Accordingly, it requires no answer.
Section 8 of CRBRP-GEFR-00523 describes the analyses for the current core design.
GMrTICN II-14 (M.2-111, pr. 3)
Scme 17 percent of the core fuel is located in the blanket.
SEr IV AA-90
ANSWEP II-14 (b) and (c)
'!his question appears specific to the Irrogeneous core sich is not the current design.
Accordingly, it requires no answer.
Section 8 of CRBRP-GEFR-00523 describes the analyses for the current core design.
QUESTICH II-15 (I%.2-111, par. 4).
'Ihe fuel and steel which has been ejected is then postulated to fall out of the blanket in rings two ard three.
(c)
Wat is the basis for selecting the sub-assernblies in rings two and three?
ANSWER II-15 (b) and (c)
'1his question appears specific to tle hmogeous core which is not the current design.
Axordingly, it requires no aruwer.
CRBRP-GEFR-00523 describes the analyses for the current core design.
QUESTICH II-16 (EE.2-111, par. 4) Generic answers (a) and (b) not regaired.
(c) What is meant by "interpenetrate" in the phrase "The material is allowed to interpenetrate the e terial already in the core"?
Describe in detail the gecrnetry of the materials involved and the a nner of penetration.
ANSWER II-16 (c)
"Interpenetrate" means that, een the neutronic calculations were per-formed, the fuel was asstned to be filling space that would otherwise be filled by a liquid fuel-steel mixture. 'Ihat is, in the regions were the 1
SET IV AA-91
fuel as postulated to the re-entering, the fuel density as increased; thus, the neutronia calculations reflect the change in flux shape due to
~
this effect.
1 l
i EEP IV M-92
FIFIH INIEPJ0GA'IORY SLT PREAMEE 'IO QUESTICNS In previous interrogatories infornation es regaested concerning four distinct validations relative to the models ard canputer codes used in the analysis of CRBR CDAs, namely,.
i)
Validation that the cn3e's output is the correct ntznerical calculaticn that should resalt fran a given set of input data and the nodel asstrnptions; 11)
Validaticr1 of the models against acttal experimental data; iii) Validation that the nalels can be extended to the CRBR; and iv)
Validaticn that the input asstrnptions for the CRBR case are adequate with respect to the CDA analysis, i.e., are support-ed by experimental evidence.
By " adequate", here and beloa, we mean that the calculations will not tnderestimate the CR work potential (i.e.,
forces ard resultirg energetics of a CDA) or overestimate the containment capability of the reactor with respect to a CDA.
With respect to the following requests for information we are concerned with these same four validations relative to the radiological source term and site suitability analysis.
QUESTICNS I With respect to each of the following codes and each subroutine of each of the followirg codes:
i (A) CXMRADEX - II (B) HAA - 3 please Ircwide the following information [Where appropriate, the parts of i
I i
the questicn have been restated to reflect the protocol for discovery l
l agreed to by Applicants, Staff, and Intervenors NRDC et al.]:
SEP V AA-93
1)
Chuplete, current cbementation (i.e., a writeup) of the codes and the subroutines;
- 2) Identify, by name and affiliation, the author, or authors, of each model, subroutine, or portion of each subroutine, which each contributed or worked cm;
- 3) Identify by name affiliaticn (including organization, division, branch, title, etc.) each applicant ernployee, or consultant, that has intimate working knowledge of the code and each subroutine, or parts thereof, including its validity. Where nore than one person is involved, delineate which portion of the code or subroutine with which each has an intimate working knaaledge;
- 4) Describe fully the procedures by which Applicant has assured itself and continues to assure itself, that the various ocmputer programs (codes) accurately reproduce the nodels as described in the PSAR and its references (see Validaticn (i) above);
- 5) Indicate which nodels (including subroutines, or Iortions of subrou-tines) have not bem validated as described in Validation (i);
- 6) Indicate the nodels (including subroutines, or portions of subroutines) or asstrnptions that have not been validated as described in Validation (ii);
- 7) Ebr each nodel, portion of the nodel, or assumption that has been validata3 (against experimental) or other) data, see Validaticn (ii) above) describe fully the Irocedure by which it was validated, and the results, including all uncertainties ard limitation of the validaticn. Indicate the source of the experimental, or other data, that was used in the validation.
- 8) Explain fully all instabilities in the ntanerical performance in the models, What causes then, and 104 they are avoided, and the extent to which this introduces uncertainties in the calculations ard limits the validity of the nodel (cf., p. FE. 2-10 par. 2).
- 9) 'Ib the extent that any answers to the above questions are based on referenced material not Ireviously Irovided, please supply the references.
- 10) Explain whether Applicants are presently engaged in or interd to engage in any further research cr work dtich may affect Applicants' answer. 'Ihis answer need be providal only in cases where Applicants intend to rely upon on going research not included in Section 1.5 of the PSAR at the IMA or I
t l
l ser V AA-94 1
l l
construction permit hearing cn the CRBR. Failure to provide such an answer means that Applicants cb not intend to rely upon the existence of arry such
~
research at the IMA or construction permit hearing cn :.he CRBR.
- 11) Identify the expert (s), if any, Whm Applicants intend to have testify on the subject notter questioned.
State the qualifications of each such expert. 'Ihis answer need not be provida3 until Applicants have identified the expert (s) in question or determined that no exp(s) will testify, as Icrg as such answer provides reasonable notice to Intervenors.
ANSWER I(A)
'!he information Irovided in these answers pertains to the COMRADCX-III Cbnputer Code, Which is used for site suitability source tenn analysis.
(1) Reference 1 on page A-19 of the PSAR is the current CINRADEX-III docmentaticn.
(2) The information is available in the docunentaticn listed in Response 1 above.
(3) 'Ihe reference in Response 1 also identifies the contributors to, and the supervisors responsible for, code and subroutine developv_nt.
(4) Independent hand calculations were made to check intermediate cal-culations of the Cbde. Simpson's Rule integraticn routines were checked by conparison with calculations Which can be reproduced analytically.
(5) All COMRADEX-III gwamming has been checked as indicated in (4) above to verify that the code performs the correct nunerical calculations.
(6) None.
(7) All models used in CCHRADEX-III are based cn first Irinciples.
Reference (2), page A-19 of the PSAR provides additicnal discussicn cn this matter.
l SET V AA-95 1
)
i
(8) Program instabilities could result if the tire-step size were too large.
'Ihe code selects the time-step according to predetermined criteria Which assures stability.
(9) References mentioned in iterns (1) and (7) have been or will be made available for inspection ard mpying.
(10) No further research work of Applicant in this area has been iden-tified.
(11) At the ; resent time, the applicants have not determined the experts, if any, Whan they intend to have testify c:n the subject matter questioned.
ANSWER I (B)
'Ihe information Irovided in these answers pertains to the IRA-3B ocrnputer code, which is used for site-suitability source term analysis.
(1) Reference 1 on page A-140 of the PSAR is the current IRA-3B doctrenta-tion.
(2) The information is available in the doctanentation listed in Response 1.
(3) 'Ihe reference in Response 1 identifies the crntributors to, and the supervisors responsible for, ocde and subroutine develognent.
l (4) 'Ite IRA-3B code, including its subroutines, has been thoroughly I
checked to assure that the ntanerical algorithms in the HAA-3B code have l
been Irogramned mrrectly.
In addition, test cases were performed to assure that the code could reproduce previously calculatal results.
(5) All IRA-3B Irogramning has been validated as in (4) above.
SEP V AA-96
(6) None.
~
(7) References 1 through 6 on page A-140 of the PSAR contain detailed descriptions of the validaticn of the HAA-3B code.
(8) 'Ihe integro-differential equaticn of the nodel in HAA-3B is solved by a glying the rxment methd aM using a log-normd particle distribution which results in three, sinultaneous, first order differential equations.
In scme analyses, nore time steps are required to cover the time interval desired than are allowed by array dimensions.
In these cases it is neces-sary to restart the ocde to continue the calculaticn.
Under certain ecnditions, following a restart, a slight input parameter manipulation is required to achieve continuity.
This does not impact the accuracy of the results.
(9) References mentioned in Itens (1) and (7) have been or will be made available for inspection aM mpying.
(10) No further research work of Applicant in this area has been identi-fled.
(11) At the Iresent time, the applicants have not determined the experts, if any, whan they inteM to have testify cn the subject matter questioned.
QUESTICN II With res,pect to the following request for information we are concerned primarily with the fourth validaticn - validaticn that the inpe assmp-tions for the CRBR case are adequate with respect to the source term and site suitability analysis.
Here we are not so much concerned with the validity of the nodel expressions as with the uncertainties in the site boundary doses due to propagation of uncertainties in a) the parameters used and b) the nodel input data and due to any synergisms antog these uncertainties and the model asstanptions.
SEr V AA-97
With respect to the calculations identified belcw, under (A) through (C),
please provide the following information [Where awwiate, the parts of the questicn have been restated to reflect the protocol for discovery agreed to by Applicants, Staff, and Intervenors h%T et al.]:
1)
List and identify all nodel input data (exclusive of coding flags and inputs that specify coding options, criteria, printout formats, etc.) and all nodel parameters that ccme into play in each of the nodels utilized in the site suitability radiological analysis calculaticns, includi:x3 but rot limited to input data and parameters in COMRADEX-II and HAA-3.
Excitrie parameters rot called into tse because a subroutine, or part thereof, was not utilized.
2)
Describe in detail the basis for the choice of each input datum and nodel parameter listed above.
1)
In each case quantify the mcertainty in the value selected; 11)
In each case indicate whether the value is based cn first prin-ciples, experimental measurements, tnvalidated hypothesis, output of other models, arbitrary asstrnptions, etc.;
iii) In each case indicate whether the choice of the input datum or model parameter was selected to represent the "best estimate", or a bounding or " conservative value", where " conservative value" here means a value chosen so as not to underestimate the accident ecnsequences, e.g., site boundary and low population zone radic>-
logical doses.
- 3) For each input datum and nodel parameter with tncertainty listed in 1) above, indicate in quantitative terms the magnittx3e of the uncertainty introduced into the final calculations of the site boundary 2-hour and the icw populaticn zone accident duration doses, respectively, due to the uncertainty in the input datum or nodel parameter.
In addition, discuss in detail any synergistic effects resulting frcm ccrnbinations of urx:ertainties in the input values, model psrameters, and model asstmptions. In each case discuss the basis fcr the estimate of how the uncertainties propagate, eg, include and discuss all parametric analyses used to test the effect of uncertainties.
SET V AA-98
4)
Identify by rme affiliaticn (including organizaticn, division, branch, title, etc.) each applicant er:ployee cr consultant that has intimate working kncwledge of the basis fcr the selecticn of the parameter or input datum.
5)
To the extent that any answers to the above questions are based cn referenced material not previously provided, please supply the references.
- 6) Explain 4 ether Applicants are presently engaged in or intend to engage in any further research or work which may affect Applicant's answer. 'ntis answer need be grovided cnly in cases where Applicants intend to rely upon cn going research not included in Section 1.5 of the PSAR at the Im or ccnstruction permit hearing cn the CRBR. Pailure to provide such an answer means that Applicants do not intend to rely upon the existence of any such research at the IE or constructicn permit hearing cn the CRBR.
7)
Identify the expert (s), if any, whczn Applicants intend to have testify on the subject matter questioned.
State the qualifications of each such expert. This answer need not be provided until Applicants have identified the expert (s) in question ce detennined that no expert (s) will testify, as long as such answer provides reasonable notice to Intervenors.
(A)
The Reference Design site suitability saurce term dose analysis stmnarized in Table 15.A.3-5 (white pages)
(B)
The Parallel Design site suitability source term dose analysis smrarized in Table 15.A.3-4 of Appendix F, Part II (yellcw pages)
(C) Any subsequent site suitability source term cbse analyses based cn source terms (ard other parameters) reccmnended by the NIC Staff.
ANSIER II (GENERAL)
'Ihe Applicants have a single design as described in the PSAR.
'Ihe site suitability source term used by the Applicants in Secticn 15.A of the PSAR is consistent with that rm.um.id-ded by the NBC Staff.
Ctnsequently, it is not necessary to pet:rvide separate responses to Parts (A), (B) and (C).
SET V AA-99
Responses to the list of questions are provided first for COMPADEX-III and then for 194-3B in Answers II(A) and II(B), below.
ANSWER II(A)
'Ihe information govided in these answers pertains to the CDMRADEX-III Otznputer Code.
- 1) 'Ihe following input data were used in CMPADEX III calculations for the Site Suitability Source Tem dose analysis:
Input release fractions of the core inventory of radioactive isotopes released to the RCB as discussed in Section 15.A.1 of the PSAR are:
Ebel Msterial 1.0%
Solid Fission Products 1.0%
Halogens 50%
Noble Gases 100%
'Ihe source tenn is hypothesized as a botnding core related release.
Time dependent clean-up factors used (calculated by the HAA-3B otraputer code) are shown in Table 15.A-6 of the PSAR.
'Ihe leak rate fran the ICB used is the design basis leak rate of 0.1% vol/ day for the duration of the evaluation.
- 2) The core inventory release fractions listed in (1) above were utilized in aanpliance with specific directicn by the Nuclear Regulatory Otmnission j
(see Reference 2 on page 15.A-9 of PSAR), and are etnsidered by the Appli-cant to be highly conservative.
'Ihe RCB leak rate utilized is based on an R3 gessure of 10 psig, diich is also considered to be a conservative value since no desicy1 basis accidents result in pressures approaching 10 psig.
l l
SE.TV AA-100
- 3) 'Ihe data ard models used to calculate the site boundary 2-hr ard the Icw population zone accident duration cbses were chosen to yield upper
~
bound values or conservative doses.
Itens 1) ard 2) provide additional infonnation concerning the bomding analyses ard conservative set of asstrnptions utilize 3 in the site suitability assessment.
- 4) Develorrnent of the infut parameters w1s done tnder the supervision of L.
E.
Strawbridge, Manager, Nuclear Safety and Licensing, Westinghouse Advanced Peactors Divisjon.
- 5) tbne
- 6) No further research work of Applicant in this area has been identified.
- 7) At the present time, the applicants have not determined the experts, if any, Whcm they intend to have testify cn the subject natter questioned.
ANSWERS II(B)
'Ihe information provided in these answers pertains to the HAA-3B ccxrester code.
- 1) The following input data were used in the Site Suitability Source Tenn dose analysis to ocznpute depleticn factors as described in 15.A.2.2 of tn=
PSAR:
Source Tenn Attenuaticn Within Containment 8
Initial particle nurtber p tration XIN(1) = 1.337 X 10 of Aerosols (particles /cm )
Aerosol voltane Variance SIGAIR = 8.000 3
-3 Aerosol Mass Mean Voltane (m )
VAIR
= 1.000 X 10 3
Density of Aerosol Msterial (g/czn )
RfD
= 10.55 2
-4 Viscosity of Air (dyne sec/cm )
VISC
= 2.264 X 10 2
Tetperature ( K)
TEMP
= 3.940 X 10 SET V AA-101
-5 Diffusional Boundary layer Thickness DELTA = 4.000 X 10
~1 AIPIR = 1.000 X 10 a
c IFF
= 1.000
- 2) 'Ihe initial particle number concentraticn aM the density of aerosol material were based cn the initial airborne mss concentration and the effective density, respectively of the mass associated with the IG non-gaseous source term species. 'Ihe selection of aerosol voltrae variance and mass mean size was based cn the experimental measurernent stmnarized in Reference D-7 on page D-6 of CRBRP-3 Voltane 2.
The air viscosity used is based cn the RCB design temperature, which is conservatively assumed to be the atnespheric tanperature for this analysis.
'Ibe value of a c was calculated by a conservative linear extrapolation measure which was insed on the experimental data reported in Reference D-1 on page D-6 of CRBRP-3 Volume 2.
- 3) The data and nodels used to calculate the clean-up factors which were used to calculate the site bouMary 2 hr. ard the low populaticn zone accident duration doses were chosen to yield conservative cbses.
- 4) All questions regarding the soluticn of parameters or input datum should be referred to L.
E.
Strawbridge, Manager, Nuclear Safety and Licensirg, Westinghouse Advanced Reactors Division.
- 5) References frczn Item 2 above have been or will be made available for insoecticn and copying.
- 6) No further research work of Applicant in this area has been identified.
- 7) At the present time, the Applicants have not detennined the experts, if any, dxzn they intend to have testify cm the subject matter questioned.
I SEP V AA-102
_UESTICH III (GDERAL)
Q Request for the following informe. tion is based cn our cxxrerns with respect to validaticn (iii) and (iv) above.
In the Applicant's answers to the generic gaestions (b) and (c) below, the Applicant is requested to be l
responsive to these concerns.
With respect to each statement, assertion or assumption (fron 15.A in Part II of Appendix F of the PSAR) identified below, please provide the follcu-ing information (tnless noted otherwise). (NorE:
'Ihe following ntrnbered Interrogatories are identified by the page and/or paragraph ntarber fran the PSAR in parentheses.) [Where appropriate, the parts of the question have been restated to reflect the protocol for discovery agreed to by Appli-cants, Staff, and Intervenors NRDC et al.]
a) Identify by name, title ard affiliaticn the primary Applicant em-ployee(s) cr consultant (s) that has the expert knowledge required to support the statement, asserticn, or asstanption.
b) Describe in detail the supporting evidence for the statenent, asser-ticn, or asstrnpticn ard where appropriate the raticmle for the approach taken.
c) Provide any additional infonnaticn requested following each <atanent, assertion, or asstanpticn.
d) 'Ib the extent that any answers to the above questions are based cn reference 3 material, please supply the references.
e) Explain W ether Applicants are presently engaged in or intend to engage in any further research or work which may affect Applicants answer.
'lhis answer need be provided cnly in cases where Applicants intend to rely upon cn going research not included in section 1.5 of the PSAR at the IMA or construction permit hearing cn the CtBR. Failure to provide such an answer means that Applicants do not intend to rely upm the SEr V AA-103
existence of any such research at the IJA or construction permit hearing on the CRBR.
f) Identify the expert (s), if any, Whan Applicants intend to have testify cn the subject matter questione3.
State the qualifications of each such expert.
Wis answer need not be provided until Applicants have identified the expert (s) in question or determined that no expert (s) will testify, as long as such answer provides reasonable retice to Intervenors.
ANSWER III (GENERAL) he questions in this part (III) of this fifth set of interrogatories are based cn statements, assertions and asstanptions in Secticn 15.A of Part II of Appendix F of the PSAR, as it existed in 1976.
Appendix F in its entirety has been deletal fran the PSAR and the Parallel Design is not part of the license application.
Hypothetical (bre Disruptive Accidents are discussed in detail in CRBRP-3, Volumes 1 and 2.
Because of the changes since 1976, Questions 1 through 5 and 6e are no longer applicable.
We responses below are, therefore, limited to questions 6 and 7.
We following responses a, d, e, and f are identical for interrogatories 6 and 7:
(a) The work in this area ms performed tnder the supervision of L.
E.
Strawbridge, Manager, Nuclear Safety and Licensing, Westinghouse Advanced Reactors Division.
(d) We referenced doctanents have been or will be made available for inspecticn and copying.
(e) No further research work of Applicant in this area has been identifia3.
i l
l 1
l SET V AA-104
(f) At the present time, the applicants have not detemined the experts, if any, h they intend to have testify cn the subject matter questioned.
OlJESTICN III-l (15.A-2, par. 2)
Define Irecisely the first containment barrier bcondary which includes the reactor cavity ard the SHAA.
ANSER III-l (b) and (c)
'Ihis question is no longer applicable; refer to the general response (above).
I QLEl!rrION III-2 (15.A-2, Inr. 2)
Define all potential leakage paths fran the first con-talment boundary.
ANSWER III-2 (b) and (c)
'Ihis question is rn longer applicable; refer to the general response (above).
QLESTION III-3 mat are the design leakage rates for each of the leakage paths identified in (3), and what is the design leakage path for the first contairment barrier defined in (2)?
ANSERIII-3(b)and(p
'1his cpestion is no longer applicable; refer to the general zusponse (above).
SET V AA-105
QUESTION III-4 Describe in detail the test program that will demonstrate that the leakage in (4) does not exceal the design leakage as requiral by 10 CFR 50, Appendix J.
ANSWER III-4 (b) and (c) t
'Ihis cpestion is no longer applicable; refer to the general response (above).
QUESTICN III-5 (15.A-7, par. 2) Sme of the volatile fission Iroducts may escape to the RC durirg the transition of the fuel frcm the vessel to the core catcher but this will not affect the results significantly since these will con-tribute to the equilibritm concentraticn which will be established between the isotopes in the sodium and in the atnesphere.
ANSWER III-5 (b) and (c)
'Ihis cpestion is no longer applicable; refer to the general response (above).
QUESTICN III-6 (15.A-9, par. 3) The use of the containment design leak rate (0.1% vol/ Day) for the duraticn of the site suitability source term evaluation is con-servative.
i l
I (c) mat is the basis for asstating the core catcher will work?
SET V AA-106
~
ANSbER III-6 (b) and (c)
'Ihe conservatism in the use of the containment design leak rate for the duration of the site suitability source term is addressed in the response to Question II-2 of this Fifth Set of Interrogatories as it relates to the COMRAIEK Code.
Part (c) of this question is no longer applicable; refer to the general response (above).
QUESTICN III-7 (15.A-lO) Generic answers (a) and (b) not required here.
(c)(i)
Were the F-factors (ren/ curie) utilized in the site suitability source term radiological cbse evaluations the same as those presented in Table 7.1-3 (p. 7.1-47) of the ER?
(ii)
Were these same F-factors utilized as a basis for excluding the effects of certain transuranium isotopes fran the site source term cal-culations? hhich isotopes were excluded?
(iii)
Are we mrrect in asstming that References 5 through 8 cited in Table 7.1-3 of the ER are the sole references used to determine the F-factors? If not, please identify and supply all references used to deter-mine the F-factors.
(iv) In reviewing the literature relative to the F-factors, were not scme references enmuntered that suggested higher F-factors?
If so, explain in detail the basis for rejecting these higher values for the F-factors?
(v)
'Ihere are many fissicn and activatial products in the EDEC radio-nuclide inventory of the CRIR.
Upon dat basis have each and every of
(
these radionuclides bees included cr excluded fran the dose calculations?
Explain this basis in detail by including all relevant information such as SEP V AA-107
inventory, activation cross-section and data related to F-factors.
In answer to this question, please be responsive to (iv) above.
ANSWER III-7 (c)(i) The F-factors utilized in the site suitability source term radio-logical dose analysis are based cn References 2, 3 aM 4 cn page 15.A-9 of the PSAR. 'Ihe F-factors in Table 7.1-3 of the ER are based cn Reference 3 above, which is NURD3-Ol72.
(c)(ii)
The basis for excluding the effects of certain transuranium isotopes, the calculational procedure used, aM the particular transuranitm isotopes excluded were identified in detail during the January 22, 1976 meetirrj between the CRBRP Project arr3 the NIC Staff in Bethesda, Maryland.
'Ihe information presented by the Project showed that excluding the trans-uranitm isotopes resulted in underestimating the potential bone and lung doses by only 4% and 3%, respectively. A detailed smmary of the informa-ticn presented by the Project is provided in "Smmary of Meeting with CRBRP Representatives" Irepared by the NRC Staff, dated February 2,1976.
(c)(iii)
'Ihe references identifed in Iten (i) above are the sole refer-ences used to determine the F-factors.
(c)(iv) !b specific literature search was cxxv3ucted with the objective of finding the highest F-factors reparted in the literature.
'Ihe major F-factor references used are part of docunentaticn provided by the NBC for guidance in cbing dose analyses.
(c)(v)
Two bases are generally usal to excitrie particular isotopes fran dose calculations.
'Ihe first is a consideration of the isotope half-life.
Isotopes with very short half-lives (<1 minute) necessarily undergo rapid radioactive decay and it is physically unrealizable for stxt isotopes to be released frcm contairment barriers and be transported off-site yrior to their decay to insignificant activity levels. 'Ihe second basis consists of a crnparison of a particular isotope's inventory and potential biological hazard (measured by its Maxinun Permissible Cbncentraticn per 10CFR2O or SEP V AA-lO8
alternately by its F-factor for a particular body organ) to the inventory and biological hazard associated with sme controlling radioisotope, i.e.,
the radiciosotope Whose inventory ard biological hazard results in it being a najor cbse contributor to a particular body organ.
Ebr example, the biological characteristics of Icdine coupled with its inventory following irradiation result in it being the nost significant, by a substantial margin, contributor to thyroid exposure.
'Ihus by conparing the biological characteristics and inventory of a par'icular isotope to Iodine, it may be shown that this particular isotope can be neglected when conputing expc>-
sures, without significantly tnderestimating these potential exposures.
'Ihis procedure can be implenented for any isotope ard organ as, for ex-anple, was the case for the transuranium isotopes discussed in Iten (ii) above.
Note that the CDfRAIEX conputer code, used for the evaluation of the site suitability source term, includes the dose contributions of over 100 individual isotopes.
A substantial rainber of these individual isotopes could be eliminated fran the dose analysis without significantly affecting the results.
However, this large runber of isotopes is retained for cunpleteness and only those isotopes, based cn sound engineering judgment in conjunction with the Focedures discussed above, that are clearly i
negligible are eliminated fran the dose analysis.
SEP V AA-lO9
SIXm INTERRCGA'IORY SL'r QUESTICN The sixth interrogatory set requests additional information beyond that supplied by the Applicants in the responses to the secord interroga-tory set.
QUESTICN I(A)(1)
In response to Interrogatory I(A)(1) [of the Seccnd Set of Interrogatories to Applicant], the Applicant indicated SAS3A evolved fran the SAS2A code which evolval frcrn the SASIA code, and these are cbetrnented in References, 1, 22, 29, 30 and 31 cm page F6.2-119, F6.1-120, and F6.2-121 of the PSAR.
Separately, for the main routine, the drive routines and for each sub-routine identified in Figure 2 (p.26) of ANL/ RAS 75-17 (PSAR, Ref.
1, p.F6.2-119) please provide the following information:
(a) Indicate whether the routine appears in (i)
SASlA (ii) SAS2A (b)
Were the routine appears in more than cne code, including SAS3A (e.g., in SAS2A and SAS3A) indicate whether the routine is exactly the same in each code, or whether coding changes have been made but the name of the routine renained unchanged; l
(c) mere coding changes have been made describe fully the changes that were made, why they were made and precisely where these changes are docu-mented in the references identified above.
[tKTIE: Interrogatory I(A)(1) of the Second Set of Interrogatories to Applicant and the tpdated answer are cm pp. AA-3 through AA-6.]
SET VI AA-llO
ANSWER I(A)(1) l Parts (a) - (c) of this interrogatory concerning develogrnent of the SAS3A and SAS3D code are being responded to as a whole in the form of an appro-priately-footnoted table. Before explaining the form and substance of the table, a general ccmnent must be made concernirg the evolution of the SAS3A and SAS3D code. As stated, this interrogatory appears to be based cn the concept that the SAS2A code evolved frcm SASIA in same incremental fashion and that SAS3A evolved frcm SAS2A in the same fashion.
On the cne hand it is true that perhaps a few thousand FORTRAN language statsnents frcm the original SAS1A coding still survive intact in SAS3A.
However, it is neither useful nor instructive to think of SAS3A as havirg evolved frcrn SAS2A and SAS2A frcm SASIA in the sense of SAS3A being SAS2A plus a few model improvanents plus minor ucdel additions.
In fact, only the nest fundamental of nodeling concepts (e.g.,
nodeling a collection of like subassenblies by an average pin, using point kinetics, treatirg only radial heat transfer in the pins, etc.) have survived intact frcm SASIA into SAS3A.
Both when it was decided to develop SAS2A and then to develop SAS3A, the basic approach was to make a fresh start at determining what phencmena relevant to 11FBR accident scenarios would be nodeled ard then what would be required to nodel these phenonena to a sufficient level of detail.
Only after givirg these a great deal of consideraticn did the developers return to the previous code to determine what parts could be lifted reasonably intact and used in the new code.
On the other hand, the SAS3D code was developed directly fran SAS3A using the same physical j
models; the major differences between SAS3A and SAS3D are in the treatment of the data nanagsnent to allow greater gecrnetric nodeling flexibility (i.e., more channels) and in reptwcasnirg to obtain better efficiency.
In the table diich follows, the relationship of the subroutines in SAS3D to SAS3A, SAS2A and SAS1A is explained in a five-coltstn format. Coltran 1 of the table sinply lists the SAS3D subroutines, cne line per subroutine (including the MAIN routine). In Coltrrn 2, an "x" appears cm the sane line if a subroutine of the same name (but consisting of EORrRAN statenents which may or may not be similar to the statenents in the like-named SAS3D subroutine) also was found in SAS3A.
A hyphen appears in Coltrrn 2 on that SEr VI AA-lll
line if a subroutine with that name did not appear in SAS3A. Cb1tran 3 uses the same notaticn to relate SAS3D subroutines to SAS2A. Coltznn 4 uses the same notation to relate SAS3D subroutines to SASlA. 'Ihree exceptions to this notaticn occur in Coltrnn 2 and 3 and need further clarification.
Subroutines INPOrl and INPCTr2, the input and editing subroutines in SAS3A, contain scrne coding similar to that fourd in subroutine INPOUT in SAS2A and an identically-named subroutine in SAS1A.
(bding found in subroutines CDOIJB and FUELEB was lifted to form a part of subroutine FEEDBK in SAS2A, and a like-named routine appears in SAS3A.
Likewise, a part of the coding fran subroutine TSCDOL in SASIA went into subroutine TSCl in SAS2A. When going fran SAS3A to SAS3D, a ntraber of subroutines were split into separate routines, mainly to improve code portability. These splits are irdicated in the table. In Cbitran 5 of the table, the ntrnbers found on the line opposite a particular SAS3D subroutine indicate that the particular footnotes iden-tified by those ntrnbers and found at the end of the table apply to that subroutine.
'Ihere are cnly ten subroutines in SAS3A W11ch are identical to subroutines in SAS2A and/or SASlA.
These subroutines are 01IN, FISGAS, FITZ, PIPFID, PREA, SHAPE, SSPK, '1EERCD, TSCB AND 'ISPK.
Since SAS3A can cnly be con-siderM conceptually as being an entirely new code which " borrowed" col-lections of FORTRAN statenents fron SAS2A, it is not meaningful to discuss
" changes" that were made to SAS2A to obtain SAS3A.
Thus, these " changes" have never been specifically doctrnented as such.
Rather, the SAS3A docu-mentaticn (which currently consists of parts or all of several reports, as indicated in the response to Part (1) of Interrogatory I(A) of the Second Set of Interrogatories) provides a descripticn of the cuerall conceptual framework of the code and its nirnerous nodels, gesents the formulae which form the mathematical basis of each model, aM discussed briefly the algorithms used to solve these equation sets.
A great deal of the al-goritimic detail of the models is left for the interested reader to dis-cover by physically examining the FORTRAN coding ccmprising the nodels.
'Ihe pop shirg changes made in generating SAS3D fran SAS3A were carried out in order to achieve the following cpals: a) to modify the way Iroblem data were stored to allow c;reater flexibility in channel specification, b)
SEP VI AA-ll2 4
to update the FORTRAN code to brirg it into consistency with standard practices, and c) to inprove the performance (speed) of the overall code.
In the process of generatirg SAS3D fran SAS3A, pr%= uirg charges were validated by denanstrating that the intermediate versions of SAS3A and SAS3D oculd produce conputationally identical results.
In addition, sone subroutines were split into multiple subroutines and FORTRlLN source changes were made to pronote consistency with accepted standard programing prac-tices and to allow code exportability.
These charges may be seen in the FORTRAN code. A nunber of subroutines were added to provide the new data managenent stragegy.
i l
SET VI AA-ll3
i I
I Relationship of SAS.3D Subroutines to SAS3A SAS2A and SAS1A Subroutines SAS3D Ebund in Ebund in Ebtnd in Applicable Subroutine Name SAS3A SAS2A SAS1A Footnotes MAIN X
X X
1 ADITIT 13 AXMESH 13 BIDOR 13 BLE20Z 13 BLOWUP X
6 CAVINr X
7 CHIN X
X 1
CIADIT 13 CIAZAS X
5 CNCOW 14 CONFIS X
6 CDOLIT 13 CROEF 17 CRODJL X
10 DA'INOV 14 DATScr 14 DEFCR4 l
DEFINT L DEEDRM DEFORM DEFORM 1
BotND DTFND l
L DrEND DITND DrFND 1,6,7 DrFou l
INYNALL 14 I
i SEr VI AA-114
i SAS3D Pbund in Ebund in Ebund in Applicable Subroutine Name SAS3A SAS2A SASlA Ebotnotes EDITIT 13 EKT 17 EDIAZ 13 EQOOK 13 13 EQGNC EQECIC 13 EDFUEL 13 E3 EAT 13 EOSILM 13 ETBAR 13 EDZFI 13 ERIORS 17 1
EKORN 13 FALUE FALL 2 FALL 3 FALUN 6,16 FAuA FAILS N
s 3
PBKCLZ FBKCDL fBKFCI y
F1EDBK X
- CODIEB, 1,5,6, FBKHET EUELEB 7,16 PBKSIN s FISGAS X
X 2
FIT 4 X
X X
1 l
i l
SEP VI AA-115
SAS3D Fbund in Found in Found in Applicable Subroutine Name SAS3A SAS2A SASIA Footnotes FK CIO CPF PROGRE MDS y
MDL K
K K
1,16 SCRAM YELDPP YOUNG s
EUAREA X
6 FUEILI 13 EUELIT 3
GEDII7T 13 GETRDY X
6 IN WEK 13 INEDIT 13 w
INPOID INPOrl INPorl INPar2 I INPor2 INPOUP INPOUT 5,6,7,16 INPr21 ECIO TBSCAN PtUr IOUT ADUr D M OLJr SETIC J
INTERP X
X X
1 INTIRP X
X 1
INTRP X
10 IGUEL X
10 10EDPY 14 SErr VI AA-ll6
SAS3D Ebund in Fbtni in Ebtn3 in Applicable Subroutine Name SAS3A SAS2A SASlA Ebotnotes LINES 15 ID mEK 13 MAPDRV 13 M NPK2 14 NAPRES X
6 OVERIAY 16 OVLY61 13 OVIX62 13 OVLY63 13 OVLY64 13 OVLY65 13 PIPFID X
X 2
- PNORM7, 13 POINST 14 POWADJ l
SWMPF 6
SIINP2 l
POWREA 13 PREA PREA PREA PREA 1
PROGRE j
PRESS X
6 PRFCI X
7 PRIMAR SSPRIM PRIMAR PRIMAR 2,9 p
PRIM 7f I
REAMC 14 READI 14 READIN X
X X
1 REED 14 REKMAP 13 RESTAR X
X 3
l REZONE X
6 i
SET VI AA-ll7 i
i
SAS3D Pburd in Ebund in Ebund in Applicable Subroutine Name SAS3A SAS2A SASIA Ebotnotes MOL NOIFL 10 MOIFL 1
RHOS RHOSFL 10 MOSFL l
RNGPOS 13 RPOWRE 13 SASFCI INIGL SASFCI 7,16 FCIZ SETITLT 13 SETINP 13 4
SETINS 13 SEITNT 13 SHAPE X
X X
1 SIIJMP2 i
SLUMPF 6,16 T90V46 l
SS000L X
X X
1 SSDRIV X
X X
1,11 SSFUEL X
11 SSifIR X
X X
1,10,11 SSPK X
X X
1,12 SSPRNP X
X X
1,11 STATUS 14 TEMPER 13 TEMPUL X
6
'INERCD X
X 2
TLEFT l
'ITEFT TIEFT TLEFF
'IELAPS I
'IRIGED 13 SET VI AA-ll8
SAS3D Foun3 in Eburd in Eburd in Applicable Subroutine Name SAS3A SAS2A SASIA Ebotnotes TSCA TSCA TSCA 2,9 2
TSCB X
X TSCC TSCC2
(
TSCC 8
TSCr3 8
TST1 M7 TSC2 X
X 2
TSC3 X
X 2
TSC31 HCCN HOCN 2
TSC4A TSC4B TSC41A TSC41B TSC42A TT4 TSC4 3,5,6,7, TSC42B 8,16 TSC43A TSC43B TABFIS TSC5 X
X 2,8 TSO6 X
X 2
TSC7 X
X 2
TSCB TSCB2 TSCB3 TSCB4 TSC8 TSC8 2,7,16 TSCBS TSCB6
'ISCBUB s
2 TSC9 X
X i
SET VI AA-119
SAS3D Fotn! in Ebuni in Ebtn3 in Applicable Subroutine Name SAS3A SAS2A SASIA Footnotes TSDRIV FAIIUR SCRAM TSDRIV TSDRIV TSDRIV 1,2,6,7, TSOV30 9,16 TSOV50 M
s TSIER TSifrR TSilTR TSHTR 1,5,6,7,16 FUELTP TSOV40 TSOV41 TSOV42 TSOV43 TSOV44 TSCl TSCl TSCOOL 1,2,7,16 TSOV45 TSOV47 TSCINT TSCSET TSPK X
X X
1,12 TSPLDP TSPlWP PRINr2 PRINf3 TSPlWP TSPRTP TSPIMP 1,5,6,7,8, PRINT 5 9,16 SSPIUr FORGAS SHORT s
'INOGEO 13 VEU!KN X
6 VFCHAN 13 VFRI'IE 13 NOIDIT 13 J
SElr VI AA-120
SAS3D Pbtznd in Ebund in Ebund in Applicable Subroutine Name SAS3A FAS2A SASIA Pbotnotes i.
NGE006 13 13 i
WRAPIT WRITEE 14 WRITEI 14 XSRITE 13 ZAPPA 14 ZAPPF 14 NN 13 ZCORE 13 VENIN 4
SINP 4
l l
i l
l SEP VI AA-121
Footnotes to Table:
1.
We Maltichannel concept, as described in Sec. I of ANIe8138, "The SAS2A INFBR Accident Analysis C%rnputer Code," by F. E. Dum, et al.,
required that a channel subscript be added to a ntsnber of the EDRTRAN arrays which were lifted frm SASlA.
Subroutines frm SAS2A noted with this footnote contained pieces of coding lifted from SASIA which referenced sone of these arrays.
2.
Wese subroutines were added or were extensively modified in imple-menting the new voidirg model in SAS2A, as described in Sect. I.B. of ANIe8138.
3.
Wese subroutines provide the restart capability for their respective codes, as described in Sect. III.F.9 of ANIc8138 for SAS2A and in Sect. V.F.1 of ANL/ RAS 75-17.
4.
%ese subroutines provide direct ca2pling between SAS2A and the VENCE-II code. his capability is not operational in SAS3A.
5.
Wese subroutines were coded new or were modified frcm their SAS2A form in hnpleentirg the CLAZAS clad moticn model in SAS3A.
6.
Wese subroutines were coded new or were modified frcm their SAS2A form in implmenting the SIIMPY fuel noticn nodel in SAS3A.
7.
Wese subroutines were coded new or were modified frm their SAS2A form in impleentirg the SAS/ECI fuel-coolant interacticn model in SAS3A.
8.
These subroutines were coded new or were modified frcin their SAS2A form in impimenting the soditm film noticn model in SAS3A.
9.
Wese subroutines were coded new or were nodified frcm their SAS2A form in implementirg the PRIMAR-2 primary loop model in SAS3A.
- 10. %ese subroutines were added or nodified frcm their SAS2A form to improve ecmputational rpeed, as described in Appendix A of ANL/ RAS 75-17.
- 11. %ese subroutines were added or modified fran their SAS2A form in implementirg the steady-state fuels categorizaticn model, as described in Sect. III.B of ANL/ RAS 75-17.
- 12. These subroutines were added or modified frcm their SAS2A form in inpleenting the decay heat treatment, as described in Sect. III.A of ANL/ RAS 75-17.
- 13. % ese subroutines were added to SAS3D for the steady-state neutronics coupling.
- 14. New routines added to SAS3D for data managment.
SET VI AA-122
- 15. Ibutines added cr nodified in SAS3D to provide inorcned cx:mputational speed cr additional printout.
- 16. Ibutines added or modified in SAS3D for inproved code portability.
- 17. Additional opticn in SAS3D.
1 r
i f
l f
i i
i StrP VI' AA-123 i
QUESTICH I(A)(2)
In response to Interrogatory I(A)(4) [of the Second Set of Interrogatories to Applicant], the Applicant indicated The entire CAS3A code, including all subroutines, has been checked and rechecked to assure that the nirnerical algorithns which are inplernented in SAS3A to solve the equation sets which constitute the SG3A code as well as with the specific model beirg added to assure that these ntrierical algorithms, both individually and collee tvely, behave in a stable fashicn ard produce accurate solutions to the original equation sets.
'Ihis was carried out by ocmparirg SAS3A results with the output frcrn other codes, with the results of hand calculations, and with what sound engineering judgment desned to be physically reasonable.
Separately for each routine identified in Figure 2 (p. 26) of ANL/ RAS 75-17 (PSAR, Ref.1, p.F6.2-119), please provide the followiry information:
(a)
Mis the routine verified by acrnparison with other codes, or by ccrn-pariscn with the results of hand calculations, or by comparison with what sound engineering jidgment deened to be physically reasonable?
(b)
If the routine was verified by cx2nparison with other codes, hcw was the other code or codes verified? Identify the other code or codes.
(c)
If the answer to (a) or (b) above is that the routine was verified by hand calculaticns, please supply the hand calculations or the appropriate doctznentaticn, i.e.,
(i) the name(s) of the individual (s) who performed the calculations and made the acrnparison; and (ii) the lateratory notebook, mENnorandtrn or other written record that doctanents the acrnparison.
(d) If the answer to (a) cr (b) above is that the routine was verified by otznpariscn with what sourd engineering judgnent deemed to be physically reasonable, please describe in detail the nature of and basis for the engineerirg judgnent.
In addition, supply:
l (i) the name(s) of the individual (s) who rerdered the judg-ment and made the cx2nparison; and SEP VI AA-124
(ii) the lateratory rotebook, manorandtrn or other written recorti that doctrnents the ccrnparison.
(e) Did the author (s) of the nodels actually perform the coding?
If not, identify the progranmer(s).
[NCTTE:
Interrogatory I(A)(4) of the Second Set of Interrogatories to Applicant and the updatal answer are cn pp. AA-3 through AA-6.]
ANSWE:RS I(A)(2)
Parts (a), (b), (c), and (d) of this interrogatory are being responded to as a whole.
Any checkout of new or extensively-modifial codirg does not generally proceed cn a subroutine-by-subroutine basis.
Rather, it is carrial out at the very least en a nodel-by-model basis, where each nodel (fuel notion, clad notion, coolant dynamics, etc., in the case of SAS3A and SAS3D) could consist of a nunber of whole subroutines plus parts of others (where it is coupled to the rest of the code). 'Ihe collecticn of sub-routines ccmprisire one of these nodels is generally referred to in the SAS vernacular as a nodule.
'Ihus, in the case of SAS3A and SAS3D as in the case of many other large-scale codes, the checkout proceeded cn a module-by-nodule basis.
Ocriparisons of the output of SAS3A and SAS3D modules and the entire code with the output of other codes, with simple hand calculations, and with what engineering jtrkynent deemed to be reasonable have been and continue to be carried by the model and code developers.
Ibwever, except as explained i
in the next paragraph, such efforts are not formally or informally doctanented.
'Ihe doctanentation that exists is in the form of the references provided in the response to interrogatory I(A)(1) of the Second set of Interrogatories.
'Ihese ANL reports serve to doctanent the mathenatical bases and provide a broad cuerview of the otznputational algorithms associated with each of the j
I SET VI AA-125
~
models and the code as a Whole.
It is implicit in the publication of these reptsrts that the authors have satisfied thenselves that the FORTRAN prcy-granmirg in the code is correct.
(e)
It is standard practice within the Accident Analysis Section of the Reactcr Analysis ard Safety Divisicn of Argonne National Laboratory that the authors of the SAS3A and SAS3D nodels, as identified by the authors listed in the docunents referenced in the above paragraph, cb their own coding and subsequently actually perform or directly supervise any subse-quent modifications to that coding.
QUEE7 PION I(A)(3)
How does the Applicant continue to assure itself that the cuerall code and its subroutines accurately reproduce the nodels as described in the PSAR and its references?
ANSWER I(A)(3)
'Ihe Applicant continues to assure itself that the overall code and its subroutines accurately reproduce the nodels as described in CRBRP-GEFR-00103 and CRBRP-GEFR-00523 and their references by careful inspection of the output results for every case analyzed and by mnpariscn of the output results for each case analyzed with the results of previous cases which are similar in part or in whole to the particular case analyzed.
In addition, the cunputer system messages are checked to assure that the job was prop-erly executed, without error, by the cunputer systen.
QUEErfICN I(A)(4)
Please identify and povide all Intra-Laboratory Heroranda generated by personnel in the Accident Analysis Secticn, the Ctolant Dynamics Section and other Sections of the ANL Reactor Analysis and Safety Division that l
critique cr otherwise evaluate the models developed by other personnel in l
1 Serf VI AA-126
these respective sections, limited to the development of any and all tredels and subroutines that are used in SAS3A (i.e., routines identified in Figure 2, p.26 of ANL/ RAS 75-17).
Also provide all subsequent mertoranda that are responses to criticisms or evaluations identified above or that represent a continuation of the dialogue related to the model evaluaticn.
ANSWER I(A)(4)
See schedule of doctrnents " Applicants' Response to hPDC Interrogatories" dated Atx3ust 30, 1976.
'Ihe files and documents have been ard will be available for inspection at the Argonne National laboratory and provisions have been ard will be made for copying. 'Ibe schedule of doctments is being updated and the update will be furnished upon acrtpletion.
Doctments referred to in the update will be available for inspecticn ard copying at Argonne National Iaboratory.
QUESTI(N I(A)(5)
Please identify pertinent sections of all ANL policy and procedures tranuals that discuss policies and procedures related to validation of models and codes, incitding those codes that are part of the ANL library.
ANSWER I(A)(5)
'Ihe pertinent sections of ANL Policy and Procedures Manuals are the follow-ing:
(1) Sections II-3.0 and 11-4.0 of the bactor Analysis and Safety Policy and Procedures Manual, Ebr Trail Use Only, dated March,1972 (applic-able to SAS3A, SAS3D, VENUS-II, and PIlJIO 1) and Sections II-3.0 ard II-4.0 of the Reactor Analysis ard Safety Policy and Procedures mnual dated m y, 1979 (applicable to PilTIO 2 only).
(2) 'Ihe Quality Assurance Policy frcm the ANL Policy and Practice Guide.
(3) Sections I and III of the ANL Quality Assurance Policy and' Procedures Manual.
SET VI AA-127
(4) Argonne Code Center 7nstallation Representative Guide.
GASTION I(B)(1)
In response to Interrogatory I(B)(4) [of the Second Set of Interrogatories to Applicant], the Applicant stated:
We entire VENUS-II code has been thoroughly checked to assure that the equaticn sets an3 algorithns given in Ref. 2 cn p.F6.2-119 of the PSAR are accurately progranmed into VENUS-II.
Because these equaticn uts are relatively sinple, this was done by ecoparing output frcm the various subroutines against hand calculations.
Please identify eac:h ard every routine in the entire VI2RJS-II code.
(a)
Separately, for each routine identified above, please supply the hand calculations or the appropriate doctrnentaticn, i.e.,
(i) the name(s) of the individtal(s) who performed the calculations and made the conparison; and (ii) the laboratory notebook, meurandtra or other writeen record that doctanents the ecmparison.
(b) Did the author (s) of the nodels actually perform the coding?
If not, identify the progranmer(s).
[NUPE:
Interrogatory I(B)(4) of the Second Set of Interrogatories to Applicant and the updated answer are cn pp. AA-4 and AA-6.]
1 ANSM!R I(B)(1)
Se table cm the following page lists each of the subroutines in the l
VI!NLE-II code and contains a brief description of each of then.
1 SI!T VI AA-128
(a) As explained in the response to Part I(A)(2) of this interrogatory, no formal or informal doctanentation of the verification activity associated with checkirg out VDE-II exists, nor is it reasonable to expect it to exist.
'Ihe authors of VDE-II carried out these cartparisons prior to their releasirg the code fcr general use (and subsequently to the Argonne Code Center). 'Ihis is implicit in the release of the code and the publica-tion of the topical report describing VEN.5-II, A W 7 31 (Ref.
2, in CRBRP-GEFR-00103).
(b) The authors of the VENUS-II code, as identified in AW7951, performed or directly supervised all of the coding of VDE-II.
l SLT VI AA-129
VENUS-II Subroutine Names and Brief Description SUBICUTINE NAME DESCRIPTION MIN Master routine which calls input routine, hydrodynamics, neutronics feedback and prints edits INPUT Paarh input fran cards and sets constants HYDfD Calls for point kinetics calculation, determines new densities and energies and calls equation-of-state routine HYDRIN Entry point in HYDRO which sets hydrodynamic and thentodynamic initial ecoditions EQlITTA Determines pressure and tanperature frcm density and internal energy INTEUR Detennines material motion feedback contribution to reactivity DT Determines Doppler contribution to reactivity PIGTIS Solves point kinetics equations IN1ERP Calculates constants needed for material noticn feedback (worth gradients) and nonnalizes power as well as worth gradients ITERAT Used to obtain coefficients for quadratic time series used to find material end Ibppler feedback EuCIO Used to determine reactivity when reactivity input in tabular form FDEN Used to allow for a void fraction for non-voided initial conditions FITZ Quadratic interpolation to determine reactivity fran time series DISPLY Writes limited accuracy edit CUIWAY Prints out pictorial view of core un3er investigation pnrJER Used in pictorial PICIURE '
- PIfAI,
)
PIDr3D l
DSChIE Plotting routines for 3-D plots and time history plots
'IPt@1AT M
1 DAXIS DRAW WRITE s
7,
,i SEP y AA-130 n
'jp
QUESTICN I(B)(2)
How does the Applicant continue to assure itself that the overall code end its subroutines accurately reproduce the nodels as described in the PSAR and its references?
ANSWER I(B)(2)
'Ihe Applicant continues to assure itself that the overall code and its subroutines accurately reproduce the models as described in the PSAR and its references by careful inspection of the output results for every case analyzed and by mnparison of the output results for each case analyzed with the results of previous cases which are similar in part or in whole to the particular case analyza3.
In addition, the ccmputer systen messages are checked to assure that the job was properly executed, without error, by the ccznputer systen.
QUESTICN I(B)(2)
Please identify and provide all Intra-Laboratory Menoranda generated by personnel in the Accident Analysis Section, the Ox>lant Dynamics Section and other Sections of the ANL Beactor Analysis and Safety Division that critique or otherwise evaluate the nodels developed by other personnel in these respective sections, limited to the developnent of any and all nodels and subroutines that are used in VENUS-II (i.e.,
routines identified in response to (1) above).
Also provide all subsequent memoranda that are responses to criticisms or evaluations identified above or that represent a continuation of the dialogue related to the model evaluation.
ANSER I(B)(3)
See schedule of doctanents " Applicants' Response to NRDC Interrogatories" dated August 30, 1976.
'Ihe files and doctanents have been and will be available for inspection at the Argcnne Natimal Tahnratory and' provisions have been arrl will be made for copying. 'Ihe schedule of doctanents is being SET VI AA-131
updated and the update will be furnished upon ecmpleticn.
Doctznents referred to in the update will be available for inspection and mpying at Argonne National Laboratory.
QUESTION I(C)(1)
In response to Interrogatory I(C)(4) [of the Second Set of Interrogatories to Applicant], the Applicant stated:
The PIITIO code has been checked and rechecked to assure that the ntrnerical algorithns which are implemented in PLUIO to solve the equation sets have been progranmed correctly.
Furthermore, test calculations were performed to assure that these nirnerical al-gorithms behave in a stable fashion and produce accurate solutions to the original equaticn sets.
'Ihis was carried out by conparing PIUID results with the output fran another code (see, H. U. Wider, J. F. Jacksco, L. L. Snith, and D. T. Eggen, An Improved Analysis of Fuel hion During an Overpwer Excursion, Proc. of the Fast Reactor Safety Meeting, CONF-740401-P3, p.1541, 1974) with the results of hand calculations, and with what sound engineering judcynent deemed to be physically reasonable.
'Ihis reference will be made available for inspection and copying.
Please identify each ard every routine in the PIRIO code.
[ NOTE:
Interrogatory I(C)(4) of the Seccnd Set of Interrogatories to Applicant and the updated answer are cm pp. AA-4 and AA-ll.]
ANSWER I(C)(1)
'Ihe PIITIO 1 code consists of four subroutines:
(1) 'Ihe main driver calls the other three routines and also solves the set of ocznIressible hydrodynamic equations describing the fuel, soditun, and fission-gas moticn in the coolant channels.
Furthermore, this routine calculates the ejection of fuel and fission gas fran the pins.
Srr VI AA-132
{
(2) The subroutine PIREZO has as its main function the rezoning of the two Lagrangian ntsnerical grids in the channel.
This subroutine also performs the nepping of cne of these grids cn the other (e.g.,
for determining the fuel density or the fuel tanperature or the fuel velocity en the sodium grid).
(3) The subroutine SMrIM solves the set of acmpressible hydrodynamic equations which describe the fuel and fission-gas notion inside the pin.
(4) The subroutine PLIO reads the input and produces the output.
mre-over, it calculates the reactivity changes caused by the fuel ard sodium notion.
'Ihe PIITro 2 subroutines are as follows:
PIEDIN - Main PIITIO2 driver; calls all PIITIO 2 routines except PLSAIN, PLINPP, and PISET. Includes autanatic time step calculation, fuel and voiding reactivity calculation, and writes the output.
Called by TSTHIN.
PLSAIN - Picks up data frcm SAS 41ch are necessary to set up the interacticn zone. Called by FAIIIR which is called by DEDIN3.
PLINPT - Initiates nostly channel variables, edits PIDIO 2 input.
Called by FAIIIR which is called by DEDIN3. Is shared with LEVITATE.
PISEF - Initiates r:ostly pin cavity variables, calculates auxiliary terms used in the code.
Also edits PIRIO 2 input.
Called by FAIIER which is called by INDIN3. Is shared with LEVITATE.
PISE1'2 - Called by PIUDRV. Reinitializes tenporary variables whenever castrol is transferred to PIUDRV frcm 'ISIBIN.
PLIF - Calculates sitx3 interface locations ard interface locations of the fuel, fission-gas, and fuel vapor regicns in the channel.
Also calcualtes the claMirg rupture propagaticn. Called by PIDIRV.
PIREZO - Adds or deletes channel cells Wenever the liquid sodium plug interfaces cross mesh-cell boundaries. Called by PIDIRV.
Solves the channel ness conservation equations for the PINACD 2-phase soditan mixture, fuel, fissicn, gas, ard fuel vapor. Called by PIUDRV.
PLNCFR - Calculates fuel, liquid sadium, and gas void' fractions.
Calculates the thickness of the liquid soditan film.
Determines the fuel flow regions for each node. Called by PIUDRV.
SEP VI AA-133
PIMISC - Calculates various channel heat transfer and friction cx>-
efficients.
Deternnnes the frozen fuel gecnetry.
Solves the channel energy equations for nobile and plated out fuel. Called by PLUDRV.
PL'IEG - Calculates cladding and structu*e taperatures.
Called by PIUDRV.
and sirx3 e-gas-I ase energy equations for PINAEN - Solves two-phase 1
h the mixture of sodiun and fissicn gas. Called by PIDDRV.
PLlPIN - Solves nass and energy equations in the pin cavity.
Also calculates the fuel arx3 gas ejecticn rates into the channel as well as the fuel melt-in rates. Called by FIDDRV.
PL2 PIN - Solves the fuel /fissicn gas mcmentun equations inside the pin and produces pin related output. Called by PIUDRV.
PINOCD - Solves the fuel arri sodiun/fissicn gas rxmentun equations in the channel. Also calcualtes the sodium slug velocities.
PLFREZ - Determines the anount of frozen fuel plateout and release if the inderlying clad is melted.
QUESTICN I(C)(2)
Separately, for each routine identified in (1) above, please supply the follcwirr; informaticn:
(a) ms the routine verified by cunparison with other codes, or by ccm-pariscn with the results of hand calculations, or by ocmpariscn with what sound engineering jtrigment deemed to be physically reasonable?
(b)
If the routine was verifia3 by acrnpariscn with other codes, how was the other code or codes verified? Identify the other code or cades.
(c)
If the answer to (a) or (b) above is that the routine was verified by hand calculations, please suIply the hand calculations or the appropriate docunentaticn, i.e.,
(i) the name(s) of the individual (s) who performed the calculations arri made the conparison; and SET VI AA-134
(ii) the laboratory notebook, merrorandun or other written record that docunents the cxxnparison.
(d) If the answer to (a) cr (b) above is that the subroutine was verified by canpariscn with what sourri engineering judgment desned to be physically reasonable, please describe in detail the nature of and basis for the engineerirg judgment. In addition, supply:
(i) the name(s) of the individual (s) who rendered the judg-ment and made the aanparison; ard (ii) the laboratory notebook, menorandun or other written record that doctrnents the canparison.
(e) Did the author (s) of the models actually perform the mding?
If not, identify the pwgcumer(s).
ANSWER I(C)(2)
As with the response to Question I(A)(2) of this interrogatory set, Parts (a), (b), (c), ard (d) are being responded to collectively here. Similarly to SAS3A and VDRE-II, the PIUIO 1 and PIUID 2 codes were extensively checked durirg their developnent and at their coupletion.
Ebr the same reasons given for SAS3A and VENUS-II, these efforts were not docunented in other than the formal PIUID 1 and PIBIO 2 reports and also in several meeting abstracts and proceedings as roted in the paragraph that follows.
Durirg its developnent, the PIUIO 1 ocde consisted of only two subroutines which were strongly cornected anc therefore, tested together.
A rather stringent check cn the two-phase PIUIO 1 hydrodynamics was made by calcu-lating a shock ropagation through a two-phase meditrn (H. U. Wider, J. F.
I Jackscn, and D. T. Bygen, "An Improved Viscous Pressure Ebrmulaticn for
'I%o-Phase Q2npressible Hydrodynamics Oilculations," Trans. Am. Nucl. Soc.,
17, p. 246, 1974).
In additicn, conpariscn calculations with the SAS/ECI model have been performed, Ref. 49 in CRBRP-GEFR-OO103, with certain aspects such as the PCI, Fuel Injecticn into the channel arid the fuel motion in the channel.
SET VI AA-135
Durirg the PIITIO 2 developnent, comparison calculations with the F111IO 1 code were made (see H.
U.
Wider, "PIITIO 2:
A Cunputer (bde for the Analysis of Overpower Accidents in INFBRs," TANSAO 27, p.
533, 1977 and also see H.
U.
Wider, et al., ANL-RDP-63,
- p. 6.8).
PIITIO 2 was also conpared with the EPIC code (see H. U.
Wider, et al., "'Ihe PLITIO 2 Over-power Excursicn Code ard a (bnpariscn with EPIC", Proceedings of the International Meeting cn Fast Reactor Safety Technology, Vol. 1, p. 120, Seattle, 1979).
'Ihe multiphase hydrodynamics model was checked by analyz-ing standard ficw expansicn and contraction Iroblems as well as by attempting to achieve steady-state conditions with this time-dependent conpressible code (see A. M. Tentner and H. U. Wider, " Pressure Drop in Variable Area, Multiphase, Transient Flcw," 2nd Multi-Phase Flcw and Heat Transfer Sympositun - Workshop, p.
1137, Miami Beach, 1979). PLITIO 2 was also conpared to in-pile experiments (see CRBRP-GEFR-00523, References E-3 and E-4).
(e)
The author of the PIITIO 1 and PIITIO 2 models performed or directly supervised all of the codirg contained in the PIITIO 1 and PIIJIO 2 codes.
QUESTICH I(C)(3)
Ibw does the Applicant continue to assure itself that the overall code and its subroutines accurately reproduce the nodels as described in the PSAR and its references?
ANSWER I(C)(3) l I(C)(3) The Applicant continues to assure itself that the overall code and its subroutines accurately reproduce the nodels as described in the PSAR and its references by careful inspection of the output results for every case analyzed ard by conparison of the output results far eadi case analyzed with the results of previous cases Which are similar in part or in whole to the particular case analyzed.
In additicn, the conputer system SET. VI AA-136
messages are checked to assure that the job was properly executed, without error, by the cx:mputer system.
QUESTICN I(C)(4)
Please identify and provide all Intra-Laboratory Memoranda generated by personnel in the Accident Analysis Section, the Cbolant Dyranics Section and other Sections of the ANL Reactor Analysis and Safety Di. vision that critique or otherwise evaluate the ncdels developed by other personnel in these respective sections, limited to the developnent of any and all nodels and subroutines that are used in PIUID (i.e., subroutines identified in (1) above). Also provide all subsequent memoranda that are responses to criticisms or evaluations identified above or that represent a continuation of the dialogue related to the nodel evaluation.
ANSWER I(C)(4)
See schedule of doctanents " Applicants' Response to IEDC Interrogatories" dated August 30, 1976. 'Ihe files ard doctraents have been and will be available for inspection at the Argonne National Laboratory and provisions have been and will be made for copying. 'Ibe schedule of doctrnents is being updated artl the update will be furnished upon atmpletion.
Doctanents referred to in the update will be available for inspection ard copying at Argonne National Laboratory.
QUESTICH II (GENERAL)
Please answer part (e) of questions 1-69 of Part II of the Second Sat of Interrogatories to the Applicant (pp. AA-13 through AA-14).
ANSIER II (GENERAL) r
'Ihe requested information is provided in tim (revised) responses to the secord interrogatory set (see p. AA-14).
SETF VI AA-137
QUESTIONS II-l In response to Interrogatory II 2(b) [of the Second Set of Interrogatories l
to Applicant], the Applicant stated:
i
'Ibe use of point kinetics rrodel with fuel displacement feedback j
obtained by stmming over fuel worth tables is judged adequate, so long as snall, local displacerents are considered.
Gross reloca-ticn of fuel in large segments of the core can be addressed by use of FXVARI or similar diffusion type codes to reccznpute the fuel worth tables when such reosnputaticn is judged necessary.
Section F6.2.1 of the PSAR discusses the approach used.
(a) Wat is the basis for the first sentence of this response? Explain in detail.
(b) Quantify sat is meant by "small" displacenents and " gross" relocation of fuel.
(c) Wat criteria are used to decide when recczoputation of the fuel worth tables is necessary? Describe in detail.
(d)
Wich CDA calculations identified in EE.2 of the PSAR met these criteria? In each case were fuel worth tables recx2nputed usirg FXVARI? If not,
- y not?
[ NOTE:
Interrogatory II-2(b) of the Seccmd Set of Interrogatories to Applicant and the updated answer are cn pp. AA-15 and AA-16.]
ANSMR II-1(a) through (c)
See Section 3.2.13 of CRBRP-GEPR-00103.
SET VI AA-138
ANSWER II-1(d)-
None of the ICIA analyses reported in CRBRP-GEFR-00lO3 and CRBRP-GEFR-00523 met the criteria for reconputation of material worth values.
Therefore, fuel worth values were not recmputed using FX-2.
QUE57 PIONS II-2 In response to Interrogatory II 3(b) [of the Second Set of Interrogatories to Applicant], the Applicant stated:
'Ihe two distinct nodes of failure identified by the terms
'Slunping' ard ' Fuel-Coolant Interaction' are mutually exclusive extranes of a cmtinuous spectrum of failure nodes.
(a) What is meant by " extremes"?
(b)
Describe in detail the basis for the above statenent that the two failure modes are on the extranes.
(c) Is it not lessible to have toth extremes as part of the same accident scenario?
(d)
Describe fully and precisely the nature of, and the application of,
" technical judgments used to ensure that limiting conditions for the available subroutines are applied to bound the resulting energy release."
[ NOTE: Interrogatory II-3(b) of the Secmd Set of Interrogatories to Applicant and the tpdated answer are on p. AA-16.]
ANSWER II-2(a), (b)
'Ihe SIINPY Model was devised to represent fuel notion following disruption of fuel geanetry in a voided coolant chamel.
'Ihe SAS/ECI Model was SEP VI AA-139
~
devised to represent the fuel-coolant interacticm process which by defini-tion requires the Iresence of liquid coolant.
Since the SL1 MPY rnodule requires the absence of coolant to function as interded and the SAS/ECI module requires the Iresence of liquid coolant to function as intended, they are en the opposite ends of the spectrun of all possible scenarios for i
fuel motion with cr without coolant interaction.
ANSWER II-2(c)
Both SLLMPY and FCI analyses may be used in the same accident scenario but not in the same channel at the same time.
ANSWER II-2(d)
If neither nodel can be eliminated by reference to first Irir.ciples of physical reality, the procedure is to perfonn both types of analysis and choose the path leading to higher energetics.
QUESTICN II-3 (PREAMB2)
The Applicant's responses to Interrogatories II-5(b) and (c) in the Second Set of Interrogatories are inadequate.
With regard to Applicant's response to II-5(c), we are requesting a quan-titative rather than a qualitative response which the Applicant should keep in mind when answering (a) through (d) below.
We see very little differ-ence between "significant" and "much more sensitive."
GJESTICH II-3(a)
Please Irovide a detailed and rigorous quantitative response to II-5(b) [of the Second Set of Interrogatories to Applicant] that includes the time histories of (i) clad taperature, (ii) pressure at the fuel pin center i
l l
i SET VI AA-140 l
line and at the fuel cladding interface, (iii) clad stress, and (iv) clad strain, each as functions of the height frm the reflector bottm.
[NCTTE:
Interrogatory II-5(b) of the Second Set of Interrogatories to Applicant and the updated answer are cn pp. M-17 and AA-18.]
ANSWER II-3(a)
A quantitative assessment of the influence of the clad failure character-istics cn the axial location of pin failure can be obtained by cmparison of the SAS calculated B0fr and EDEC 'IOP $0.50/sec pin failure location predictions with failures predicted by the Damage Parameter ernpirical pin failure correlation, described cn pages 6-32 and 6-33 of CRBRP-GEFR-00103.
Ocmparison of the predicted axial failure locations in correspording channels given in Table 6-6, Inge 6-51 of CRBRP-GEFR-00103 shows that both failure nodels predict failure at acrnparable axial locations above the midplane. 'Ihis emparison is applicable to the 'IOP analysis in CRBRPHEFR-l 00523.
'Ihis ocznparison quantitatively confirms that use of the experimental cladding failum strength as a functicn of taperature predicts pin failure toward the upper part of the pin in hypothetical unprotected 'IDP events (2.4 to 504/sec rartp rates).
'Ihe uncertainty is greatest at low ramp rates. Recent information cn the SLSF W-2 test at low ranp rates is being evaluated.
The effect of clad inhcrtogeneities cn cladding failure during a hypotheti-cal transient overpower or loss of ficw transient has not been quantita-tively assessed since no statistical data cn cladding failure due to clad inhcmogeneities is available.
Qualitatively, the effects of cladding inhmogeneities nay cause an individual pin to fail randomly alcng its l
length.
Cladding irhup_neities are not expected to resalt in the otherent failure of large groups of pins.
'Ihe possible failure of a small nmber of pins during the transient due to clad inhcrnageneities is not expected to alter SET VI AA-141
the failure scenarios and energetics resulting fron the failure of large groups of pins as predicted by current pin failure models usirg the best available experimental data for irradiated and unirradiated cladding failure strergth.
'Ibe ternperature dependent clad failure strength curves used for fresh and irradiated claddirg are based cn the best available experimental data, which are given in Refs. 14 and 42 in CRBRP-EER-00lO3.
'Ihe use of the experimental clad failure data with the transient cladding stress and tenperature calculations in the SAS code result in the predicted axial locaticm of cladding failure.
The staternent "The slope of the cladding strength as a function of tanperature significantly influences the degree of bias in pin failure toward the upper part of the pin" is a general statenent, applicable to 'IOP transients in the rarge of approx-imately 2.4 to 504/sec, which is supported by the results of SAS ccxle calculations using the referenced experimental failure data.
The clad tenperature-time history for the BOT and EDT 'IOP 504/see tran-sients is given in Figure 6-66 on page 6-117 and in Figure 6-32 on page 6-83, respectively, in CRBRP-GEFR-00103.
Clad temperature histories denonstrate the trend of the cladding temperature increases during low ramp rates (2.4 to 504/sec) hypothetical 'IDP transients.
'Ihe pressure at the fuel pin center line in channel 10, the first channel to fail in the best-estimate BOEC 'IOP 104/sec transient, is shcwn in Figure 6-69 cn page 6-120 in CRBRP-GEFR-00103.
'Ihis figure illustrates the trend of the pressure increase at the fuel pin center durirg the low ramp rate 'IOP transients.
Pin failure is predicted to occur where the clad ciretsnfer-ential stress and the correspcmding clad midpoint tenperature satisfies the experimental failure strength data. 'Ihe mechanistic burst pressure failure critericm in SAS is based cn clad failure strength rather than cladding strain.
SET VI AA-142
QUESTION II-3(b)
~
Please Irovide the quantitative probability of clad failure with the appropriate uncertainties as a functicn of time ard height above the reflector.
ANSWER II-3(b)
A quantitatiave probability of clad failure with the appropriate uncer-tainties as a functicn of time and axial position does not exist nor is it necessary since for each transient analyzed clad failure was explicitly calculated.
'Ibere is no axial reflector in the CRERP.
QUESTION II-3(c)
Please provide a detailed write-up of the Stuart model as formulated for SAS3A.
Ref.10 cn page F6.2-120 of the PSAR is an abstract and does not provide sufficient detail.
ANSWER II-3(c)
A 'Stuart M:xlel' as such was not formulated for SAS3A and SAS3D.
- Rather, Snith formulated a mathematical roodel of clad loading based cn Stuart's fissicn gas Iressure loading theory.
'Ihe essential elenents of Stuart's theory are stated in paragraphs four and l
five or. page 655 in Ref. 6 in CRBRN-00103. Snith's clad loading model 1
was based cn Stuart's approach that fissicn gas released frcm the fuel during a transient cuerpower event will add to the steady state fission gas mntained in the central cavity to load the cladding.
SEP VI AA-143
A detailed description of Snith's clad loading nodel used in the burst pressure failure criterion in SAS is given in Refs. 7 and 13 in CRBRP-GEFR-00103.
QUEErrION II-3(d) t Does the nodel assume the pressure at the fuel-cladding interface is independent of the height above the reflector? If so, please justify.
ANSWER II-3(d)
'Ihe 'IOP fuel pin failure nodel described in Section 3.2.3 of CRBRP-Gr.FR-00103 does not asstrne that the pressure at the fuel cladding interface is independent of the height above the lower blanket.
t QUEErrION II-3(e)
Please Irovide a detailed description of the curves in Figure M.2-14 Y.P.(1,2) gp(4)
Burst (3)
Burst irradiated (p.F6.1-176).
What are clad (
and Snith-Stevenson? Wiat da these symbols and references nean?
Please cross eference the curves in Figure M.2-14 with the curves in Reference 52 on p.M.2-122 of the PSAR.
l ANSWER II-3(e)
Pointer 50 on page 4-14 of CRBRP-GEER-00103 provides a detailed description of the clad strength curves in Figure 3-3 (p. 3-43) of CRBRP-GEER-00103.
References diich contain the data upon which the curves are based are also cited. 'Ihe curve in Figure 3-3 of CRBRP-GEER-00103 is the same as Figure M.2-14.
'Ihe claMing strength table values for fuel types 1 and 2, ' denoted as I
Y.P.( ' I, are yield strength values for unirradiated 20% CM 316 SS, SET VI AA-144
property code 2102, page revision 1, 6-10-74 in Ref. 42 in CRBRP-GEFR-00103.
h cladding strength table values for fuel type 3, denoted as BURST ( },
are burst strength values for unirradiated 20% CW 316 SS, property code 2203, page revision 2, 4-16-75 in Ref. 42.
'Ihe cladding strength table values for fuel type 4, denoted as Ur
, are ultimate tensile strength values for tnirradiated 20% CW 316 SS, property code 2101, page revisicn 1, 6-10-74 in Ref. 42.
The cladding strength table values for fuel type 5,
denoted as BURST IRRADIATED CIAD(5) are irradiated clad burst strength values for 20% CW 316 SS cladding heated at a transient heating rate of 100 F/sec.
This curve was developed frcm a logarithmic interpolaticn between the 10 F/sec and 200 F/sec curves given in Ref. 14 in CRBRP4ZFR-00103.
As stated in Ref. 7 in CRBRP-GEFR-00103, the SMITH-STEVENSON curve for ultimate tensile strength is given by a 1/T fit to high strain rate data for unirradiated 20% CW 316 SS.
QUESTION II-3(f)
In the above response, please consider the inplications of the nodel and calculations presented by H. G. Bogensbirger ard C. Ranchi (Nuclear Tech-nology, Vol. 29, April 1976, pp. 73-85).
(i) Itw daes this nodel differ fran that used in SAS3A and in the PSAR?
ANSWER II-3(f)
'Ihe Applicant can make no quantitative assessment of the inplications of the model ard calculations presented by H. C. Bogensberger and C. Ibnchi (Nuclear Technology, Vol. 29, April 1976, pp. 73-85) relative to the CRBRP.
~
'Ihe reference applied the fissicn gas behavior nodel in the analysis of a SEP VI AA-145
1 J
$5/sec unprotected 'IOP event in the SNR-300 Mark I core. 1b physical bases for ramp rates of the order of $5/sec have been identified in the CRBRP.
'Ihe implications of the use of the referenced fissicn gas behavior mxlel in the analysis of lower ramp rate 'IDP events is not discussed in the refer-ence, ard the Applicant can make no quantitative assessment of the implica-tions of the nodel cn the lower ranp rate (2.4 to 506/sec) hypothetical 'IOP events reported in CRBRP-GEFR-00103.
QUESPION II-4 With regard to Applicant's response to II.11.(b) [of the Second Set of Interrogatories to Applicant] we note that Figure F6.2-1 cm p.F6.2-ll6 of the PSAR (cited on p.F6.2-9) indicates the SAS/ECI mxlel uses cne dimen-sional Lagrangian cells.
(a) Is this a correct interpretation of the SAS/ECI model?
(b) Has the Applicant rigorously tested whether cne-dimensional Lagrangian cells is an adequate formalism, cr is the Applicant simply assuming it is adequate because of the "lorg length in the axial direction of the coolant channel in ocunparison to the relative srnall distances between adjacent pins?"
[tK7fE:
Interrogatory Il-ll(b) of the Second Set of Interrogatories to Applicant and the updated answer are on pp. AA-20 through AA-21.]
ANSWERS II-4
'Ihe questim nost Irobably refers to Fig. F6.2-4 on p. N.2-166 of the PSAR because Fig. F6.2-1 cn p. F6.2-ll6 did not exist. Figure 3-4 en p. 3-44 of CRBRP-GER-00103 is the same as Figure EE.2-4.
SEP VI AA-146
l (a) These are not Lagrangian cells in a rigorous sense because the veloc-ities of the cell boundaries are not calculated by solving the nrrnentum conservation equation.
Rather the velocity is determmed by linear inter-polation of the interface velocities of the two constraining liquid scrlium loops.
(b) The assumption of the cne-dimensional pseud >Lagrangian cells has been made because the coolant channels are very long in the axial direction ecmpared to the small distances between adjacent pins.
%ese pseudo-Lagrangian cells are used cnly for accounting for fuel notion in the channel for calculating fuel noticn reactivity feedback.
We reactivity feedback effect of any radial fuel moticn in the channel would be totally negligible.
QUESTIN II-5 he Applicant apparently failed to answer Interrogatory II.14(c) [of the Second Set of Interrogatories to Applicant]. %e last sentence under 14 in the Applicant's response is totally inadequate.
Please answer II-14(c) [of the Second Set of Interrogatories to Applicant]
fully, identifyirg each nodel by author (s) and reference and. hen explain fully the basis for rejecting these other models.
[NCTTE:
Interrogatory II-14(c) of the Sectrid Set of Interrogatories to Applicant and the updated answer are on pp. AA-22 through AA-23.]
ANSWER II-5 M:x3els in W11ch a rapid superheating of the liquid sodium and subsequent explosive vaporizaticn is asstrned to occur are not considered to be rele-vant for an 1MFBR envircrrnent with its abundance of nucleation sites. All in-pile tests done to date support the above statement (see Ref. 56 in C N -00103).
SET VI AA-147
l
'Ihe current models which asstrae the sodium to be in thernedynamical egai-libritra are all similar to the co-Wright nedel. 'Ihis nedel can simulate a wide variety of situations by varyirg the fuel particle diameter, the mixing and fragmentation time ccnstant, and the fuel sodium heat transfer coefficient as a functicn of the soditrn void fracticn.
This model is adequate for sinulating the mild interactions which have been observed in experiements, typical of INFBR envircriments, as well as hypothetical more energetic interactions.
A heat transfer nedel Wich takes sodium condensation into account and is otherwise similar to the OMright model was usal for the successful analysis of the fresh fuel H-2 experiment (see Ref. 16 in CRBRP-GEFR-00103).
'Ihe SAS/ECI heat transfer nodel is based cn the 01cH4right nedel and it also takes scxiitrn condensaticn into account. Moreover, the SAS/ECI nedel calculates the rate of fuel injection into the sodium.
The PIRIO 1 and PIRIO 2 heat transfer model contains a space-dependent fuel-coolant interaction calculation in which a SAS/ECI type calculation is beirg performed in many axial nodes in the coolant channel.
Other current heat transfer models are reviewed in the following reference:
H. K. Fauske, "CENI Meetirg cm Fuel-Cbolant Interactions," Nuclear Safety, 16, p. 436-422, 1975.
'Ibe heat transfer nedeling utilized in SAS/ECI and PIRIO 1 and PIUID 2 are considered to be adequate for inclusion in a whole-core analysis code such as SAS3A and SAS3D.
I r
I l
l I
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I SET VI AA-148 l
QUESTION II-6 (PREAE LE)
In response to II-15(b) [of the Seccmd Set of Interrogatories to Appli-cant], the Applicant cited Chapter 6 of Ref. 22, p.F6.2-122 of the PSAR for sodium thermodynamic Iroperties. 'Ihis reference cn p.% states:
Lack of data ard inability to maintain consistency in the superheated vapor region and the regicn above the sodim critical tmperature have led to a neglection of sodim properties in these regions. Since nest calculations with the SAS/ECI model do not require properties in these regions, the equationm f-state model is adequate. Ibwever, extremely high heat transfer rates can lead to tmperatures and pressures abcne the critical values as well as to superheated vapor. A nore crmprehensive egaation of state will be necessary to handle these cases.
and on p.lOl:
Although this method for the subcooled region cbes satisfy the basic thernodynamic relations at the saturation line, it does have scme Iroblems away frcm the saturation line. 'Ibe nest glaring problem is that fcr large pressures Eq. (6.15) may yield a negative isothermal coefficient of bulk cxmpressibility. Similar effects may occur for the thermal expansion coefficient of Eq. (6.17).
Except in severe cases, the pressures and tartperatures predicted by the SAS/ECI model generally do not result in these anmalies.
With respect to each CR analysis Iresented in EE.2 of the PSAR (see I(A) through (D) of NRDC's Eburth Set of Interrogatories), please provide the following information (separately for each QR analysis):
[NtyrE: Interrogatory II-15(b) of the Second Set of Interrogatories to Applicant and the tpdated answer are cm p. AA-23.]
I OUESTION II-6(a)
Does moditm in the core lead to taperatures and Iressures above the critical value?
Is superheated vapor prrxbeed?
Explain fully the basis l
for the answer.
l SET VI AA-149
ANSWER II-6(a) l An examination of the cases in CRBRP-GEFR-00103 indicates that the highest liquid soditrn temperature produced in SAS/ECI was approximately 1000 K below the critical tanperature. Exceeding the critical tenperature was not predicted.
Also, no case appears to have resulted in the Iroduction of superheated vapor.
This is reasonable since the smallest initial interacticn zcne length used was 5 on.
A mininun condition to Iroduce superheated vapor in EAS/ECI is to vaporize all the initial liquid soditan present in the inter-action zone. his requires a voltrae expansion by a factor of 100 to 1000 cr nore, i.e., to an interacticn zone 500 to 5000 on long. We active core is less than 100 on in height.
Heat losses due to condensation will thus arrest the vaporizaticn process before such an expansicn can occur.
QUESTION II-6(b)
If the answer to (a) is yes, describe fully why the Applicant believes the SAS/ECI model is an adequate representaticn of the seditan equation of state.
ANSWER II-6(b) h e answer to II-6(a) is "no".
QUESTION II-6(c)
Do large pressures yield a negative inothermal coefficient of bulk cxan-pressiblity cr thermal expansicn coefficient anywhere in the core? Explain fully the basis for this answer.
SET VI AA-150
i ANSWER II-6(c)
Pressures large maugh to yield these negative values are not calculated anywhere in the core. Such sirgle phase pressures would be very large. Ebr example, at a taperature of 1400 K, a pressure of approximately 4750 atm l
is required to produce negative values of either of these coefficients.
Pressures of this magnitude are not anticipated, since fission gas is generally present in the interaction zone to act as a cushicn, and known heat transfer rates between oxide fuel and sodium cannot produce explosive conditions with expected ICEA phencrnenology.
In any case, the maximum calculated single phase sodium gessures in CRBRP-GEER-00103 and CRBRP-GETR-00523 are generally more than an order of magnitude below the values required for negative coefficients.
QUESTICH II-6(d)
If the answer to (c) is yes, describe fully why the Applicant believes the SAS/EUI nodel is an adequate representation of the scditzn equation of state.
ANSWER II-6(d)
'Ihe answer to II-6(c) is "no".
QUESTICN II-6(e)
If the answer to (a) cr (c) is no, sich are the mininun changes in the more sensitive parameters (e.g.,
reactivity rap rate) that would be necessary before the answer to either (a) or (c) is yes?
ANSWER II-6(e)
No reasonable, i.e., physically meaningful, change to any sensitive param-eters can be made to change the response in sections (a) an3 (c) from no to l
yes.
SET VI AA-151
GJESTION II-7
'Ihe Applicant also stated the thernodynamic properties of cladding and fuel used in SAS/KI are ststrarized in Section 6.2.2.3 of the PSAR.
(a) Precisely where in Section 6.2.2.3 are these properties stmrarized?
(b)
Explain fully (rather than sunmarize) the basis for the choice of these properties.
(c)
Explain fully why the Applicant believes the choice of these Irop-erties is adequate.
ANSWERS II-7 (a)
SAS input for SAS/KI is simnarized cn page 3-9 and 4-16 of CRBRP-G12R-00103.
(b) No special input for clad and fuel thermodynamic properties is needed for SAS/KI. 'Ihe SAS/KI module uses the standard claddirg and fuel prop-erties.
(c)
It is nainly the tncertainties in accident h
t encmenology, not the uncertainties in the thermodynamic properties enployed, that lead to the spectrtan of hypothetical accident scenarios 1 resented in CRBRP-GEFR-00103 arti CRBRP-GEER-00523.
A cxznplete study of the influence of all possible property variations has not been done, but the remaining uncertainties are not expected to lead to a broadening of the spectrtan of hypothetical acx:ident scenario. As a dranatic exartple, Iresent axide vapor Iressure j
uncertainties were shown to have only a snall effect in fast reactor disassembly calculations.
See the following reference: J. F. Jackson et l
al.,
"'Ihe Influence of Equaticm-of-State Uncertainties on Fast Reactor Disassetbly Calculations," Trans. Am. Nuc. Soc., 22, p. 368,1975.
l l
SET VI AA-152
GJESTICN II-8 Document the basis for the Applicant's response to Interrogatory II 16(b)
[of the Second Set of Interrogatories to Applicant].
1
[ NOTE:
Interrogatory II-16(b) of the Seccnd Set of Interrogatories to Applicant ard the updated answer are on pp. AA-23 through AA-24.]
ANSWER II-8 h second sentence in the response to Interrugatory II 16(b) of the Second Set of Interrogatories to Applicant may have been misinterpreted. Whether or not the sadim voiding reactivity associated with a fuel-coolant inter-actim is larger than or smaller than the associated fuel rioticn reactivity is dependent on the particular FCI event being discussed, although the fuel rroticn reactivity is generally the dminant reactivity mee the event passes the first few milliseconds.
h criginal interrogatory asked for i
support of the assmpticn that the sodim voidirg reactivity can be adequately determined in SAS/ECI frm the average, smeared sodium density of the interactim zone.
In support of this aseption, calculations of ECI events with the PIUID 1 and PIUID 2 codes have been done a nmber of times.
In these codes, the sodim voiding reactivity is determined frm the detailed axial distributicn of sodium in the interacticn zone provided by the Lagrangian mesh. 'Ihe PIUID 1 and PIUIO 2 calculations predict sodium voiding reactivities that do not differ significantly frcm those calculated by SAS/ECI if the constrainirs sodim slug velocities predicted by both models are similar (e.g., see page 7-84 in CRBRP-GEER-00103).
h reason for this is that the detailed distribution of the srall anount of scdim in the interaction zone as calculated by PLUID 1 and PIUID 2 is not inportant for the sodim voidirg reactivity feedback calculation.
SEr VI AA-153
QUESTICN II-9 Identify in Appendix F of the PSAR (by page and paragraph) the SASBTK calculations that represent the parametric variation of the loss coeffi-cient (see Applicant's Response to Interrogatory II 44(b) [of the Second Set of Interrogatories to Applicant].
[NCTTE:
Interrogatory II-44(b) of the Second Set of Interrogatories to Applicant and the updated answer are on p. AA-37.]
ANSWER II-9
'Ihe SASBWK calculations that represent the parametric variation of the loss coefficient referred to in the Applicant's response to Interrogatory II 44(b) are identified in CRBRP-GEFR-00103 (by page and paragraph) as follows:
6-4, para. 2 and 3; p. 6-5, para.1 and 2; and p. 6-7, para. 2.
QUESTION II-lO With regard to the Applicant's Answer to Interrogatory II 45(b) [of the Second Set of Interrogatories to Applicant], identify the other locations of the blockages that were considered.
Discuss the sensitivity of the CIA energetics to the locaticn of the blockage.
[NCTTE:
Interrogatory II-45(b) of the Second Set of Interrogatories to Aplicant ard the updated answer is cn pp. AA-37 through AA-38.]
l ANSWER II-10
'Ihe other locations sere agg1cmeration of material could occur are the upper blanket and the core.
l l
'Ihe HCEA energetics are not expected to be significantly influenced by the locaticn of the blockage.
Fuel blockages are progressively more difficult to maintain in a stable coolable configuration as the blockage location l
SET VI AA-154
approaches the core mid-plane. As the reactor power continues to increase due to the continued control red withdrawal, blockages closer to the core are expected to be partially or totally dispersed, depending cn the Inrtion of the blockage material which cannot be cooled below the melting point.
We analysis of Section 10.1.1 of CRBRP-GEFR-00lO3 was performed to assess the pessimistic asstanpticn that fuel blockages could not be sufficiently cooled and would sitzp upon melting.
he results of those calculations showed that sltanping of the melted blockages would not result in recriti-cality. Werefore, the location of the blockage is not expected to have a significant effect cn possible FK2R energetics which might result frcm the altmping of melted blockages.
QUESTICN II-ll Please review the Applicant's Response to Interrogatory II 47(b) [see 47(c) of the Second Set of Interrogatories to Applicant] for correctness.
We Applicant's response here is to refer NRDC to its lbsponse to Interrogatory 27.
Interrogatory 27 addresses the method of estimating fission-gas tanperatures, an inrelated subject.
Perhaps the Applicant meant to refer to another interrogatory.
[ NOTE:
Interrogatory II 47(c) of the Second E.rt of Interrogatories to Applicant and the updated answer is cn Ip. M-38 through M-39.]
ANSWER II-ll he original cross-reference has been corrected. See page identified above and p. M-75 fcr answer.
SEP VI M-155
i QUESTION III
~
In response to Interrogatories I(A)(10), I(B)(10), and I(C)(10) [of the Second Set of Interrogatories to Applicant], the Applicants stated:
"(10) The Applicants are currently analyzing this area and will provide pertinent information as it beccmes available."
We find this response to be inadequate and request the follcwing informa-tion:
(a) Miat is the Irecise nature of the analysis or analyses currently being performed in this area, ard what is the nature of the uncertainty (ies) to be resolved by the analysis or analyses?
(b) Who is performirg the analysis or analyses?
(c) When is the analysis or analyses expected to be ocmpleted?
[tUTE: Interrogatories I(A)l0, I(B)10, and I(C)10 of the Second Set of.
Interrogatories to Applicant and the tpdated answers are cn pp.
M-5 and M-9, pp. M-5 and M-10, and pp. M-5 and M-12 respectively.]
ANSWER III
'Ihe original answers to interrogatories I(A)(10), I(B)(10) and I(C)(10) have been updated. See pages noted above.
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DOCKET 2. B-537 PMWECT MAllAGDIENT CORPORATION
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TlaBIE55EE ELEY Almt0RITY
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4 MFIDAVIT F 005115 M. SWIT!CK Densik N. Smitick, being 61y suorn, deposes and says as follows:
o 1.
That he is emplchl by the General Electric Compaq as Managhe, Safety Analysis, Advanced Reactor Systems Department 310 De Eulgne Drive, Seng vale, California M006.
~ t, That he is 41y authorized to answer the interrogatories numbered I(A)
(that asterial related to the 5458 LOK andule), !!(1-8), !!(28-47) in ME's Second Set of Interrogatortes, !(A), !!(1-4),11(6-29) in letDC's Third Set of Interrogatories, !(1-7), !!(1-16) in lutoC's Fourth Set of Interrogatories, and 11(2.3),!!(9-11), and !!! in letDC's Sixth Set of j
i Interrogatories and that answers to Interrogatories !!(1-8) (Second Set),
and I(1-7) (Fourth Set) include results of work performed and developed under the Reactor Analysis and Safety Division. Argonne National Laboratory and that the answers to !!(6-29) (Third Set) and 1(1-7),11(1-
- 16) (Fourth Set) include results of work performed by Fauske and, Assectates Incorporated.
l 3.
That the above-eentfomed and attached answers are true and correct to the best of his knowledge and belief.
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sehecribed and suern a before se this,zf* de of I, test.
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i WITED STATES F NERICA WCLEAa ama 4 TORY C88415510N l
In the metter of
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WITED STATES EPARllENT F DERGY )
SOCKET 10. 50-531 POWECT MANA4 DENT CORPORATION
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18uur N E VALLEY AUTMotITY
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FFIBAVIT F L. WLTER EITRICH q
L. M1ter Deltrich, being 41y asern, deposes and says as follows:
f 1
1.
That be is employed Iqr the Reactor Analysis and Safety Division of National Laboratory, 9700 50. Cass Avenue Argonne, I)linofs
, as Associate Divisten Director.
l 2.
That he is 61y authorized to answer the Interrogatories unsered I(A),
1(8),1(C) II(9-27), and 11(45-69) fn NRDC's Second set of laterrogatories, excepting unterial related to the SA38 LOK module in the response to Interrogatory !(A), Interrogatory 11(5) of lutDC's Third Set of laterrogatories, and Interrogatories 1(A). I(8), 1(C), !!(1), and
!!(4-4) of IstDC's 5fxth set of Interrogatories.
i 5.
That the above-asntioned and attached answers are true and correct to the I
best of his knowledge and belief.
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Subscribed and sworn to before un this g day of Mes.1982.
1 od v 7ary r ue tw Cminim upires //[rly t-!
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the matter of
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UNITED STATES DEPARTMENT OF ENERGY)
DOCKET NO. 50-537 PROJECT MANAGEMENT CORPORATION
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TENNESSEE VALLEY AUTHORITY
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l AFFIDAVIT OF LEE E. STRAWBRIDGE Lee E. Strawbridge, being duly sworn, deposes and says as follows:
1.
That he is employed by Westinghouse Electric Corporation as Manager Nuclear Safety and Licensing, Westinghouse Advanced Reactors Division, P. O. Box 158, Madison, Pennsylvania 15663.
1 2.
That he is duly authorized to answer the Interrogatories in NRDC's Fifth set of Interrogatories.
3.
That the above-mentioned and attached answers are true and correct to the best of his knowledge and belief.
A (Signature) v Subscribed and sworn to before me this,~),.) day of& c.u.
,1982.
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6% s s t W. /
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Notary Public-V My Comission expires C
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
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In the Matter of
)
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UNITED STATES DEPARTMENT OF ENERGY
)
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PROJECT MANAGEMENT CORPORATION
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Docket No. 50-537
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TENNESSEE VALLEY AUTHORITY
)
)
(Clinch River Breeder Reactor Plant)
)
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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.
20545 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 U. S. Nuclear Regulatory Commission Washington, D. C.
20545
- Daniel Swanson, Esquire
- Stuart Treby, Esquire Office of Executive Legal Director U. S. Nuclear Regulatory Commission Washington, D. C.
20545 (2 copies)
-2_
- Atomic Safety & Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20545
- Atomic Safety & Licensing Board Panel i
U. S. Nuclear Regulatory Commission Washington, D. C.
20545
- Docketing & Service Section Office of the Secretary U. S. Nuclear Regulatory Commissien 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 j
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)
Mt. 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
3-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 l
A Geor.g&'L. Edggb '
Attorney for l
Project Management Corporation DATED:
April 28, 1982
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Denotes hand delivery to 1717 "H" Street, N.W., Washington, D. C.
- / Denotes hand delivery to indicated address.
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