ML20052C505
| ML20052C505 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 04/30/1982 |
| From: | Anderson C, Dickson P, Disney R JOINT APPLICANTS - CLINCH RIVER BREEDER REACTOR |
| To: | National Resources Defense Council, Sierra Club |
| References | |
| NUDOCS 8205050062 | |
| Download: ML20052C505 (152) | |
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, 4/30/82 c rmin s
LEITED STATES T AME:RICA
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NJCLEAR REGUUCDIE COMISSION S m %l ip'
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In the Matter of ) & '
) '
V IMI'IED STATES IEPARINENT T DERGY ) Docket No. 50- "J O ,gkg%
- P!O7ECT MWAGEMENT CORPORATION ) - ;-
b T!!ttMME VAILEY ALTIHORITY )
~~
$ h594I
) 0) , $9 Clindi River Breeder Reactor Plant ) M g3 4 ,s 3 \Q APPLICANTS' UPDATED RESPONSE #2 'ID NATURAL RESOUICES DEFTNSE: OJUtCIL, DC.
AND SIERRA QUB INTERIOGA'IDRIES (SEVENIH, NINIH, TENIH AND 'IHIRTEENIH SETS) Pursuant to 10 CER paragraph 2.740b, and in accordance with the Board's Prehearing Conference Order of February 11, 1982, the United States Department of Energy, Project Management Corporation, ard the Tennessee Valley Authority (the Applicants) hereby update their responses to the Natural Resources Defense (buncil, Inc. and the Sierra Club Seventh, Ninth,
'I%tnth and 'Ihirteenth Sets of Interrogatories to the Applicants, dated April 7, 1976, May 13, 1976, August 13, 1976, and January 14, 1977, respectively.
In these updated responses the following style has been utilized: Pbr endt set of interrogatories the Preanble 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 from the responses initially furnished. Ibwever, for convenience those unchanged responses also have been set forth after the appropriate interrogatories. QSCd 5 I( 8205050062 G20430 PDR ADOCK 05000537 Q PDR SET VII AB-1
5'- , '1he answers contained in this Updated Response #2 supercede all prior answers to the interrogatories as to which they are applicable. In aane instances, interrogatories specifically related to the previous parallel design cxwered in Appendix F to the PSAR. Appendix F was withdrawn fran the application in 1976. Applicants have atternpted in these updated answers to trovide 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 meant a substantial amount 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. Were Applicant believes the interrogatories are related to Appendix F and the previous parallel design and are not appropriately applicable to the current design, Applicant has so noted. I SET VII AB-2 _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ . _ _ i
O O SE:VENIH INTERROGATORY SET GENERAL QUESTIN Each question is instructed to be answered in six parts, as follows [Where appropriate, the parts of the questicn have been restated to reflect the l protocol for discovery agreed to by Applicants, Staff, and Intervenors NRDC i et al.]: ! t A) Provide the direct answer to the question. B) Identify all documents ard studies, and the particular parts thereof, relied upon by Applicants, now or in the past, dich serve as the basis for the answer. In lieu thereof, at Applicants' option, a copy of such docu-ment and study may be attached to the answer. C) Identify principal documents and studies, ard the particular parts thereof, specifically examined but not cited in B). In lieu thereof, at Applicants' opticn, a copy of each such docunerr. ard sttdy may be attached to the answer. D) Describe the methodology of analysis, inclMing all assumptions, and test results of the studies identified in subptrts b) ard c) of each , answer. I E) Explain 4 ether Applicants are presently engaged in or intend to engage in any further research cr work which may affect Applicants' answer. This ; answer need be povided only in cases Were Applicants intend to rely upon cm going research not included in Section 1.5 of the PSAR at the M or , construction penttit hearing on the CRBR. Failure to provide such an answer ! means that Applicants do not intend to rely tpcn the existence of any such research at the M or construction pennit hearing cm the CRBR. SET VII AB-3
1
, F) Identify the expert (s), if any, whcm Applicants intend to have testify cn the subject matter questicned. State the qualifications of each such expert. 'Ihis answer need not be Irovided until Applicants have identified the expert (s) in questicn or determined that no expert (s) will testify, as lcng as such answer Irovides reasonable notice to Intervenors.
GENERAL ANSWER The following responses arc identical for all interrogatories except where supplementary informaticn is provided in the responses below. (A) See ntsnbered responses below. (B) The ch,mts which serve as a basis for the Applicants answer are identified in the responses below arri have been or will be made available l for inspection and copying. (C) The Applicants have examined and evaluated many cbetrnents pertaining to the subject matter questions, however, unless otherwise indicated in the responses below, documents and other studies pertaining to the subject matter have been examined but not relied upcn by the Applicants. This does not imply that the Applicants have examined all cbcunents in existence which could pertain to the subject matter questioned. Of the doctrnents examined by the Applicants which might pertain to the subject matter questions, only that material relied upcn by the Applicants has been retained in retrievable form by the Applicants. 'Ihis material is identified in response to subpart B. (D) 'Ihe nethodology of analysis, including all assumptions, and test results of the studies identified in subparts (B) arti (C) of each answer is described cr referenced in the cbetanent itself. (E) Except where otherwise noted below, the Applicants' program of further research work is described in Section 1.5 of the PSAR and in Appendix A of CRIEP-3, Vol. 1 and Vol. 2. SET VII AB-4
(F) At the present time the Applicants have not detennined the experts, if any, tan they intend to have testi#y cn the subject etter. QUESTICN 1 (PREMELE) Fuel Element Bowing: 'Ihe elimination of an core spacer pads between fuel rod bundle ducts (now placed in the axial blanket) for the CRBR raises the question of a possible autocatalytic power transient as occurred in EBR-I core meltcbwn due to inward fuel bowirg. The PSAR notes that a positive reactivity feedback is Iredicted in the CRBR, but that it will be offset by negative Doppler feedback to make the net reactivity coefficient negative. ;
\
QUESTICN 1(a) Does a negative coefficient depend cn maintaining the reactor power level, coolant flow and fuel and coolant temperatures within prescribed ranges?
'Ihat is, does the thermal hydraulic-mechanical nodel predict a net positive feedback for any cxmbination or set of ciretrnstances?
ANSWER 1(a) CRBRP core restraint analyses generally indicate a negative bowing reactiv- t ity above scme power to flow ratio, (P/F)o, where P and F represent frac-tions of full Iner and flow respectively. (P/F)g will generally range fran O to ~0.8 depending cn the particular constraint asstrnptions employed. l Consequently for P/F > (P/F)g the power coefficient is negative for any l acnbinaticn of the stated parameters, since bowing is the culy conpanent which can contribute positive reactivity. For P/F < (P/F)g a positive bowing reactivity effect is calculated. Under these conditions a negative pawer asefficient would depend cm the Dogpler l faadhrk, coolant flow level, temperature gradient field and the inlet t temperature variation associated with a given power transition. 'Ihis situation is analogous to the EBR-II response (Ref.1) in which positive i SET VII AB-5
, bowing reactivity effects exhibited various degrees of importance dependirq on the power, flow acrrbination.
QUESTI' ION 1(b) Does the PHENIX fast power reactor (France) errploy spacer pads in the core, as distinguished fran beirg located outside the core regicn, vix blanket - (as in GBR)? R. Carle, et al., in ANL-7520 (Part 2). p. 247, indicated that the pads are in fact in the core for Phenix? l ANSWER 1(b) Page 247 of ANL 7520, Part 2, indicates that spacer pads are located within the core regicn for PHENIX. 'Ihe title of the above article (page 243) is ' PHENIX: Status of the Design before Cbnstruction and was presented in November 1968. 9 Personal crmnunication with Argonne National laboratory personnel in 1975 indicates that the spacer pads were renoved fran the active core area and were placed a few inches above the active core. Because of the age of the above report, it is believed that the latter information is representative of the Iresent design. QUESTICN 1(c) How is PHENIX designed to cope with neutron-induced swelling, if the fuel ducts are made of stainless steel? ANSM!:R 1(c) l
'Ihe response to question 1(b) indicates that the present PHENIX design copes with neutrcn induced swelling by locating the load pads above the SET VII AB-6 l
, care regicn. Spacirg between ducts can acuami.date limited swelliry in the core region.
QUEErrION 1(d) Is the exinun fast neutron (> 1 MEV) fluence for PHENIX below the limit Where swellity would cause the ducts to be stuck together at the pads? ANSWER 1(d) I h current / neutron fluence of the peak HENIX fuel duct is estimated to be 1.0 to 1.1 x 10 23 n/an2 , E > .1 EN. '1he fluence at the spacer pad loca-tion (above the active core) is believed to be between 1/10 to 1/15 that of the peak fluence locaticn. Ib swelling contact is expected for the fluence l I experienced at the load pads. QUESTICN 1(e) Does the British EPR use spacer pads in the core? ANSWER 1(e) h information tresented below is fran an article in a Fast Reactor Power Stations Doctanent by the British Nuclear Energy Society; Proceedings of the International Conference Organized by the British Nuclear Ehergy Society held cn 11-14 March,1974, at the Instituticn of Civil Engineers, London. The title of the article en pages 307-318 is Support of PFR Sub-Assenblies and Associated Developnents by J. A. C. Holmes.
'Ihe lower pads are below the lower axial treeder and the typer pads are located at the juncture of the lower axial breeder with the bottan of the care as shown in Figure 2 and 3 of the above reference, attached.
SEP VII AB-7
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.' In the restrained care design the spacer pads are located above the upper breeder ard below the lower breeder as shown in Figure 5 of the above Reference attached.
QUESTION 1(f) mat are the distances of the spacer pads' fran the core midheight level for the CRBRP, Phenix, and FPR (British)? ANSWER 1(f) We distance to the spacer pad centerline fran the core mid height center-line is: 71.1 on CRBRP approximately 55.0 on PHENIX 46.0 on PFR QUEErrIONS 1(g) Please describe the detailed theoretical calculation (results, models and theory) which predict, (i) the anount of fuel duct bowirg (at several elevations of the core), (ii) the initial clearances between ducts (mean, lower and upper limits due to design tolerances), and (iii) the reactivity worth per tnit distance of towing. N l_bd. We description of the analytical techniques and crznputer nodeling splayed in PSAR duct. bowirg analyses is described in Section 4.2.2.4.3. We theoretical treatment is consistent with approaches described in consider-able detail in References 2 and 3. Doctrnentation of the current analytical approaches used in CRBRP core restraint analysis together with their theoretical bases will be available for the PSAR. SET VII AB-11
ii. CRBRP Core Restraint System Gaps (At Uniform Tenperatures) (all dimensions in inches). Top Imd Plane Gaps Interasserably gaps: 0.015 1 004 Peripheral gap: .120 i .012 Above Core Imd Plane Gaps Interassembly gaps: 0.015 1 004 Peripheral gaps: .054 i .012 iii. The radial worth factor for several cases are presenced in the PSAR: Tables 4.3-24 and 25 QUESTICE 1(h) Describe the experiments, if any, which confirm the thermal-mechanical bowing predictions discussed above. ANSWER 1(h) A direct crrnparison of bowing reactivity analytical predictions with experimentally deduced bowing reactivity effects has been performed for the EBR-II reactor (Ref. 3,4). 'Ihese otrnparisons showed that bowing reactivity effects can be predictal with very good accuracy. These results add con-fidence that CRBRP core restraint analyses will be capable of predicting < bowing reactivity effects in the operating reactor.
'Ihe analytical nodel used for CRBRP Cbre Restraint Design is being verified
, using a full core simulation in the Naticmal Core Restraint 'Ibst Facility. i Tests were begm in 1977 and completed in 1981 and the verification will appear in the FSAR. t l SET VII AB-12
. References
- 1. EBR-II System Design Description, Voltzne II Primary System, 01 apter 2 Reactor.
- 2. ANL-8068, NJBOW: A EDRTRAN-IV Prograrr for the Static Elastic Structural Analysis of Ibwed Reactor Cores, G. A. McLennan, April 1974.
- 3. ANL/EBR-Ol4, IDW-V; A CDC-3600 Program to Calculate the Bpilibritra Configurations of a Thermally Ibwed Reactor Core, D. A. Kucera and D.
lehr, January 1970. i
- 4. Prediction and Measurement of Reactivity Effects Due to 'Ihennal Bow 2ng of EBR-II Subassemblies, D. R2tr, A. Gopalakrishnan, ANS Transactions, October 1971.
QUESTICN 2 (a) Discuss fully, including the theoretical and experimental basis for the stated positicn, the extent to which CRBR mixed-oxide fuel will densify upon heating-up to the melting temperature. (b) Has it been established that PuO - UD2 f el will not densify uptm 2 heating up (to the melting temperature) due to any phase changes? ; i ANSWER 2(a) and (b)
'Ihe density of stoichicmetric 75 w/o (weight percent) UO 2
- 25 w/o Pu0 '
2 which is representative of CRIRP fuel pellets, as a function of temperature (up to the melting point) has been derived based span the reommended l relationship fcr thermal expansicn which has been measured. i r
'Ihe derivation is as follows: 'Ihe density at tatperature, P,p, is the ratio of Mass, M, to Voltrne, V,p) f P,7 =h (1)
T SET VII AB-13
, , For isotopic thermal expansion, the relationship between volumetric expansicn and linear expansicn is i l
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=P
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where gP = theoretical density at T = 0 C 3
= 11.00 g/cm for .75 00 .25 PuO 2 2 and h
= ratio of length at temperature to length at C C b (4) o
= [1 + am (T-0)]
The thermal expansion, ag, tp to the melting (solidus) temperature of UO2 ~ Puo2 (2725 C), is
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a ,= 6.8 x 10 + 2.9 x 10 T (5) Finally, cmbinirg equations (3), (4) and (5) gives the final equaticn for density: PT (g/ n ) = 11.08 [3(14 .8 x 10 T + 2.9 x 10
-9 T2 ) 1 (6) 2e basis fcr Equaticn (5) is the thermal expansion data of U0 to 2200 C 2
- fran Reference 2-1 (Section B). Wis data was shown by R. L. Gibbey in Reference 2-2 (Secticn B) to be in excellent agreement with experimentally measured values for 0.75 002 .25 PuO2 and PuO 2 . Wus, minor ocupositional variations between QURP mixed-oxide fuel (0.67 w/o U0 2 - 0.33 w/o Pu02 )
and the representative 0.75 002 - 0.25 PuO2 fuel will n t affect linear l SEP VII AB-14 l i
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i , thermal expansion. 'Ihe behavior of the density with tanperature is l pecwided in Figure 2-1. I 4 1 i In addition to considerations of thermal expansion and ghase change there j is a potential for densificaticn of as-fabricated mixed oxide fuel having I lower density than theoretical. A topical report (CRBRP-ARD-Ol68) by [ Bishop 2-2a considered the impact of fuel densification m CRBRP fuel i performance. 'Ihe Bishop report yrcuides a stunary of existing information : relative to fuel densification and the deformaticn acocmpanying it, applies f this information to predict GBRP fuel dimensional change, and assesses the deformaticn due to densification cn CRIRP fuel performance. j i
'Iherefore, as indicated in Figure 2-1 and as supported by References 2-1 [
and 2-2 Section B: 1 i l (a) 'Ihe extent of the densification (negative linear thermal expansion) of representative CRIRP mixed-oxide fuel upcn heating up to the melting point
- is not significant (less than 10% change). In fact, there is no densifica-ticn but rather a slight decrease in overall density.
l I (b) The impact of fuel densification due to tnrestructured (as-fabricated) ; fuel densification cn GIRP fuel performance has been evaluated in the ! Bishop report. 'Ihe effects of increased stored heat, increased heat l t generaticn rate, decreased heat transfer capability and axial gaps in the [ fuel coltrin were addressed for design basis steady-state and transient ! operation. Any potential degradaticn due to these effects was fourri in l each case to be less than that asstrned in the CRBRP design process de- ; scribed in the PSAR. 'Iherefore, pellet deformaticn due to fuel densifica- r tion is expected to have negligible adverse inpact cm the design capability ! of CRERP fuel.
'Ihare is no evidence of densification of PuO , U0 # "i**O u0 - Puo due j 2 2 2 2 to phase change upcn heating tp to the melting tarperature. i i
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'Ihe Irimary references relied tpon in answering this question are as I follows:
Reference 2-1 > Conway, J. B., R. M. Fincel, Jr., and R. A. Hein, "'Ihe 'Ihermal Expansion and Heat Capacity of UO 2 to 2200 C:, Transactions of the American Nuclear Society Voltane 6, tamber 1,1%3, p.153. Reference 2-2 Gibby, R. L. , " Thermal Expansion of Mixed-Oxide Ebel," HEDIANE-74-3, HEDL Qua terly Technical Report, July, August, Septernber 1974, Voltane 1, Applied Research, pp. A-8 to A-10, A-15, A-22 and A-23. ; i Reference 2-2a Bishop, B. A., "Inpact of Ebel Densification m C'RBRP Ebel Performance," l CRBRP-ARD-0168, June 1977. !
'Ihe secondary references Wiich were examined bt:t were not relied upon in !
answering this questim are indicated below. These references provide collaborating data and/or nodels for thermal expansion of PuO2 (Reference 2-3), .75 002 .25 PuO2 (Reference 2-4 to 2-6) at .00 00 1 ,g
.20 Puoy,94 i (Referere 2-7) as indicated in the attached Figures 7 and 8 (pp. A-22 ard ,
A-23) of Reference 2-2. Reference 2-3 ' Tokar, M., A. W. Nutt and T. K. Keenan, "Idnear 'Ihermal Expansion of PuOh Nuclear Technology, Voltane 17, 1973, pp. 147-152. i i Reference 2-4 Takemura, T., S. Kashima, H. Matsui and M. Koizuni, " Thermal Expansion of l r (Pur-U)O2-x, " Annual Rep rt f 'Ibkai Works, April 1,1968 - March 31, 49, [ t PtCF-AR-68, Power Reactor and Nuclear Ebel Developnent (brporation, 'Ibka.. Japan, Nowmber 1%9. i t l l SET VII AB-17 l - _ , . _ . - - ._. --. . .
. Reference 2-5 IABlanc, J. M. and H. Andriessen, "Research cn Thermal Expansion of UO2 ' 2 and (U, Pu)O2 ," BJRAEC 434, Translaticn of Big-101, Brussels, June PuO 1%2. Reference 2-6 Skavdahl, R. E. and E. L. Zebroski, "High Tenperature Phase Studies," Sodium Cooled Reactors Fast Ceramic Reactor Develognent Program 'Nenty Eight Quarterly Report, August-october, 1968. GEAP-5700, pp. 57, 60-63. Reference 2-7 Both, J. , M. E. Hubert, J. R. Cherry, C. S. Caldwell, "'Ihe Effects of Stoichicmetry on the 'Ihermal Expansion of 20 wt % PuO - U0 Fast Reactor 2 2 Ebel", Transactions of the American Nuclear Society, Volume 10,1%9, pges 457-458.
'Ihe test results fran References 2-1 to 2-7 (other than Reference 2-2a) are shown in the attached Figures 7 and 8 of Reference 2-2. 'Ibe method of deriving the experimental data and in some cases developing the associated mathenatical correlations are described in the indicated pages of the respective references.
Characterization of the out-of-pile and in-reactor behavior of reference QURP mixed-oxide fuel (e.g., dilataneter measurements of pellet sintering behavior) is a continuing effort of IMFBR Mixed Oxide Ebels Developnent Pr@taa arti the IJFIR Reference Fuel Steady State Irradiation Fr@ tan. I f r I i SET VII AB-18
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SEP VII AB-20
CrJESTICN 3 l l 1 G. S. Iellouche in the EBR-I Incident: A Reexamination (Nuc. Sci . & Erg. , 56, 3:303-307) showed that the reactivity changes during the EBR-I power excursion (sub-prmpt) were grossly different than previously believed. (a) 'Ib what extent is the Applicant, or anyone to the Applicant's knul-edge, analyzing the EBR-I incident with SAS3A (or other SAS codes) nodified for metallic fuel, to see if the least-squares-deduced tmperature co-efficient of Iellouche can be predicted accurately? (b) If ro such analysis is being performed, why is this considered unnecessary? ANSWER 3(a) Neither the Applicants nor anytme known to the Applicants is analyzing the EER-I incident with SAS3A (or other SAS codes), to see if the least-squares-deduced temperature coefficient of Iellouche can be predicted accurately. ANSWER 3(b)
'Ihe major difference between earlier analyses of the EBR-1 incident and that of Lellouche lies in his categorization of the fast coefficient as a sm of the fuel tm perature coefficient and the coolant tenperature coefficient. 'Ihe analysis of EBR-I by Lellouche is not relevant for CRBRP safety analysis, because the SAS3A code has terms which account for both effects. 'Ihe sodium worth tables used in SAS account for the dependence of sodim worth cm radial core location in that channel-dependent reactivity will be intrcduced as the coolant density changes. Reactivity changes caused by increases in fuel temperature are accounted for via the Doppler coefficient. 'Ihus, during the early porticm (i.e., before voiding), the SET VII AB-21 l
i P
, approach taken by Iellouche is implicitly used in SAS3A. Coolant density *-
induced reactivity changes are rot nearly as great as those resulting when sodium voiding begins. Once a channel starts to void, it rapidly loses its sodium, i.e., the coolant density drops to nearly zero almost inmediately. QUESTICN 4 r Does there exist a doctanent cr set of h=mts other than References 1, 22,19, 30 and 31 cn pages F6.2-119 to F6.2-120 ard ANL-8131 and ANL-7607,
- which systematically describe the experimental basis for the SAS3A code?
If so, please identify them. ANSWER 4
'Ihe cnly references diich describe the experimental basis for the SAS3A and SAS3D conputer codes are identified in response to question I(A)(6-7) in the Seccnd Set of Interrogatories to Applicants. (See p. AA-5).
QUESTICN 5 QUESTION 5(a) In each of the several core disruptive accidents calculated in the PSAR I (identified under (A) through (D) in Part I of NRDC, et al.'s, Eburth Set of Interrogatories to Applicant), including the nest pessimistic cnes, what fracticn of the com remains in, er is converted to, the vapor state after the fuel expands cbwn to cne atnosphere pressure (assuming no containment as a boundirg calculation)? ANSIGER 5(a) Referring to (A) through (D) in Part I of NRDC g al. 's, Eburth Set of Interrogatories to AIplicant, the fraction of the core in the vapor state i SET VII AB-22 1
. after expansicn to one atmosphere is as follows for selected cases in
. CRBRP-GEER-00103:
Case Ranp Rate, ($/sec) Vapor Fraction
" (a) BOE I N 40 0.16 50 0.13 100 0.20 (b) EDEC I M 40 0.15 (c) EDE 'IOP 50 (partial Na) 0.02 50 (full Na) 0.005 75 (full Na) 0.02 100 (full Na) 0.04 (d) EDE IN 28 0.13 Imnediate reentry 37 0.13 45 0.13 63 0.19 Honogenized Core 30 (21 Po) 0.18 Reentry 30 (2.1 Pg ) 0.20 30 (210 Pg ) 0.10
'1he case in CRBRP-GER-00523 that required a VENW-II calculation inplied an input reactiivity ranp rate of 43S/see and produced a vapor fraction at 1 atm of about .01.
SET VII AB-23
. GJEEFFION 5(b)
Provide the same answer as in (a) above with the assumption that the closure hold down devices failed after absorbirg, say, one half of its design basis mechanical energy, due to faulty equignent. F ANSWER 5(b)
'Ihe closure assenbly is designed to remain in place and essentially leak tight fcr the first 1000 seconds following the SEDB loading. This design assures that sufficient time is available for substantial fuel vapor condensaticn and aerosol plate-out within the reactor vessel. 'Ihe closure hold down device will ret fail if subjected to the sugg, ited value of one-half its design basis mechanical energy. -
I e r QUESTION 5(c) With respect to the fuel Wiich does not vaporize, does a dynamic theory exist which could predict the particle size of the unvaporized fuel? ANSWER 5(c) To the best of the Applicants' knowledge, a dynamic theory does not currently exist which predicts the particle size of the unvaporized fuel. QUESTION 5(d) In the worst QR calculated in the PSAR, Wat fraction of the core would be expected to be made into an aerosol upon the asstanpticn in (b) of a faulty holddown structure, and Wat would be the particle size distribution? ! ANSER 5(d) Section 4.2 and Appendix D of CPBRP-3, Vol. 2, Irovides this informaticn. t SEP VII AB-24
i I 1 l . 1 GJESTION 6 i Se PSAR's analysis of CIns is based cn the assumption that no sodium vapor l explosicn will occur in the event of nelten fuel interacting with coolant. (a) Is it the position of the Applicant that such explosions are (i) l impossible, (ii) highly unlikely, or (iii) unlikely? Quantify your answer if possible. (b) Please supply or identify the scientific justification for the conclusicn in (a) if it is different fran the response to Interrogatory II-16 in the NRDC, g al.'s., hird Set of Interrogatories to the Applicare.. (c) Are any tole core meltdown tests being planned or considered to l investigate the possibility of soditun vapor explosions? If so, describe and doctanent these fully. l ANSWERS 6(a) and 6(b) he Applicants' position has been stated in the Applicants' response to Interrogatory II-16 in the NRDC, et al.'s, Wird Set of Interrogatories to the Applicant (see p. AA-70) and is further explained in Section 8.2.6 of GBRP-GEFR-00523. ANSWER 6(c) No sole core meltdown tests are being planned by the Applicants to inves-tigate the possibility of soditan vapor explosions. SEP VII AB-25
6 I r t l ; j . QUESTICES 7 i l . QJESPICN 7(a) could a coolant pipe rupture (without a pipe sleeve) lead to a core i disruptive accident cr fuel melting even if a SCRAM occurred? ! ANSWER 7(a) , 4 It is possible that the nost severe pipe rupture (i.e., the hypothetical double-ended rupture) at the worst location (i.e., the reactor vessel inlet nozzle) could result in cladding and fuel melting in scrne fuel assenblies ! and radial blanket assentlies. Since the core would be shu due to the j rapid PPS scram, the initial conditions are considerably different than those of the ICR's analyzed in CREP-3 for which the core is at nearly { full power or increasing power. It is not known whether the consequences I of the upper-limit pipe rupture at the worst locaticn could result in full I core involvanent or whether flow would be reestablished with limited local damage. Ibwever, there is no reascn to expect any substantial energetics in either case tecause the core is shut down by a large margin. I I QUESTICN 7(b) ! Ilas the coolant pipe rupture accident without SCRAM been analyzed? If so, please supply the analysis. ANSWER 7(b) i No, the coolant pipe rupture without scram has not been analyzed because the event would require a cabinaticm of a pipe rupture and failure of the ' redundant and diverse plant protection systens, three extranely inprobable , events. 'Ihe probability of pipe rupture is discussed in Reference 7-1 and the Ir,hahility of failure of both shutdown systems is assessed in Refer-ence 7-2. i Ser VII AB-26 l 5
. _ . . . . - - - . .- - . - ~ - . , . - . .
References:
7-1. CRERP-ARD-Ol85, "OtBRP Integrity of Primary and Intermediate Heat Transport Systen Pipirg in Contairrnent," September 1977. 7-2. WARD-D-Oll8, " Reliability Asse'ssment of CRBRP Reactor Shutdown Systen," Novernber 1975. QUESTIONS 8
'Ihe PSAR notes that candidate CDA initiators were eliminated because of a detenninaticn of low probability (p. F3-1).
(a) Please identify all of the candidates that were considered and eliminated, ard in each case supply the supportirg basis for the deter- ' mination of the low Irobability. (b) Please identify those that were not eliminated, namely events (b) of Sec. F3.2. ANSWERS 8(a) and 8(b) IKDA initators are all of low Irobability for CRBRP. 'Ihe CRBRP is designed to prevent the occurrence of conditions under which an HCDA might be initiated as discussed in depth in Section 2 of CRBRP-3, Voltrae 1. An assessment of IEDA initiators is discussed in Section 3 of CRBRP-3, Vol.ane 1. 'Ihe Irocess for selecting and eliminating potential initiators is explained in Section 3.2 of CRIRP-3, Voltane 1. 7 3 PION 9 (PREAMLE) Soditan Void Reactivity 'Ihe PSAR (p. F3-1) states that oore disruption cannot occtr inless a coolable gecraetry is lost. SEP VII AB-27
s
. QUESTICN 9(a)
Does this mean that Iresent analyses cb not predict a superprcunpt-critical ' Iower excursicn caused by rapid sodiun voidirg? ANSWER 9(a) No I i l l QUESTICN 9(b) Please grovide the bnsis for the answer in (a) above. , ANSWER 9(b) In amne hypothetical core disruptive accident scenarios analyzed (loss of flow with failure to scram), the sodiun boiling in the hottest assenblies could Irovide a positive reactivity feedback that could result in a power excursicn and possible core disrupticn. %e Applicants have conservatively considered that coolable core geometry is lost when coolant boiling occurs. Thus, even if the fuel pins are intact at the onset of the power excursion, coolable gecmetry is already considered to be lost based cn the Applicants' I conservative applicaticn of the coolable gecmetry definiticn. QJESTION 10 I he control assembly withdrawal accident at startup was analyzed assuming 600 F inlet tanperature and 1 PW initial power (p. F3-3) . (a) Were these parameters varied, such as assuming 1 watt of power? (b) If so, please provide the detailed analysis. f SET VII AB-28
ANSWER 10 Ib; a parametric study cn initial cxmditions for tnterminated control assanbly withdrawl accidents at starttp has not been done. QUESTICN 11 Please supply the analysis ddch supports the conclusion (p. F3-6) that the fuel will remain intact upcn a SCRAM. ANSWER 11 It is assumed this interrogatory refers to a rod drop without punp trip, an event that is not part of the CRBRP plant design bases since protective functions (flux to delayed flux trip) have been incorporated to assure overall plant arx3 pturp trip. 'Ihe event,oroduces a rapid reactor shutdown with full flow, leading to excessive cooling of the reactor, a conditicn in the thermally safe directicn. 'Ihe rapid decrease in hot leg ternperature, however, adversely affects the service life of plant hot leg cxznpanents. This leads to a requirement for protecticn system action to prevent the rapid Int leg tanperature decrease, hence Irovision is made to assure a full plant trip, includirg ptrnps, fcr this event. For anall roass, thin secticn replaceable cx2nptnents such as fuel, rods, and cladding, rapid reactcr shutdwn with full flow is accep' Ale, because it results in overcooling below termal operating levels. As indicated in PSAR Section 4.1, fuel rod damage which, if excessive, leads to loss of integ-rity occurs Irimarily due to loadings at higher tarperatures; i.e., normal operaticn arx3 abnormal undercooling events. GJE!TfICN 12 Wiat does PPS mean cn p. F3-14? l S!!T VII AB-29
i i
. ANSWER 32 PPS is an acronym for Plant Protection System Which is the system Which !
autanatically shuts the reactor down neutronically if an abnormal condition ! exists. I l QUESPICN 13(a) Have the possibilities of injection of a gas bubble and an oil slug (hydrogenous material) into the reactor and core been considered? : ANSWER 13(a) Yes. '1he passibilities of these occurrences were considered and the design of the plant implemented to preclude than as sunmarized in Reference 13-1. Furthermore, their impact tpcn the plant should they occur has been considered (References 13-1 and 13-2).
References:
13-1 PSAR Section 15.2.3.2; 13-2 Response to NRC Question l 001.338, Amerxhent 15 to the PSAR; I?-3 R. W. Hardie arri W. W. Little, ' Jr . , " PERT-V, A 'I%o-Dimensional Perturbation (bde for Past Reactor Anal-ysis," BtE-1162, Pacific Northwest Laboratory (1%9); 13-4 J. N. Ebx et al., "EDRE-II: A Ctmputational Pres tain for the Analysis of Steady State and i Transient Reactor Performance," GEAP-5273, General Electric Ocnpany (Sept. l 1966); 13-5 D. R. Vissers g al., "A Hydrogen Manitor for Detection of Leaks in IJEER Steam Generators," Nuclear Technology, Vol. 12, p. 218, October 1971; 13-6 " Methods for the Analysis of Sodian and Cover Gas," RDP F3-40, Appendix N, June 1975; 13-7 Applied Physics Divisicn Annual Report, ANL7-190 p. 241, January 1972. Methodology for Gas khble Analysis -
'1his analysis is . reprted in Reference 13-1. 'Ihe sodiun void reactivity worths were calculated using i
PERP-V (Reference 13-3) . Iarge bubbles are assuned to nove at the average i sodium velocity through the core. 'Ihe transient variations in reactivity l insertion were estimated fran the bubble gecnetry, void worth and bubble SEP VII AB-30
O
. velocity. %ese reactivity insertions were used as input to the code FORE-II (Reference 13-4) and the transient changes in power and fuel pin cladding temperature mlculated. Additional cladding tarperature effects due to the insulating effect of the bubble were calculated arx3 the tem-perature increases sunrned. %ese results are Iresented in Reference 13-1.
Small localized bubbles prcduce insignificant reactivity effects ard cnly the insulating effect need be considered. If it is conservatively asstrned that all the heat is absorbed in the claddirg for the duration of bubble ccntact, bubbles of 5, 7.5 and greater than 12 inches length are required in order to raise the cladding temperature 25 F for fuel, radial blanket, and control assemblies respective:.y. Methodology for Oil Slug Analysis - If oil enters the sodiun in the pump tank, it will react to form carbonaceous particles and gases and sone hydroge (Reference 13-2). Since the reaction will be at or near the surface of the sodium, the gases would rise into the pinnp cover gas and be reprocessed by the RAPS. Any hydrogen which does not enter the cover gas would tend to go into solution. Wis requires scrne hydrogen in the cover gas in order to provide a partial pressure to maintain the hydrogen in solution. his sodium fran the ptznp tank will mix with the sodium in the affected PifIS loop and further mix with the sodiun in the two other loops in the reactor vessel inlet plenum. Hydrogen will cx2ne out of solution at the surfaces with other cover gas spaces in order to establish an appropri-ate partial Iressure in all gas spaces. %e saturation level of hydrogen in sodiun at 730 F is 51 parts per billion (References 13-5 and 13-6) corresponding to 24.3 kg hylvge_a distributed throughout the Irimary sodiun. In order to attain this hydrogen level, it would require more than 2 years continuous oil leakage fran cne pianp at 10 c.c./ lour and errone-ously replenishing the oil supply eight times during this period. We estimate also conservatively assunes that all hydrogen is liberated as gas, no gas enters the purp cover gas, the cold trap does not renove any hy&vy.n and no hyi@ enters the other cover gas spaces in cxmtact with the primary soditan. 1 I I SEP VII AB-31 L
Using hydrogen worths based cn ZPPR-2 measured values (Reference 13-7),
, this corresponds to a reactivity ramp rate of 0.8 x 10 cents / hour to a maximtrn inserticn of 2.0 cents based cn unifonn hydrogen distribution of the hydrogen in the core and blankets.
Consideraticn of the relatively insoluble carbonaceous gases formed by the oil / sodium reaction (Reference 13-2) shows that they will gradually build tp with time. The time required for this uniformly distributed gas, asstmting rn depletion, to cause a reactivity insertion of' 1 cent is about 2,900 hours.
'1his conservatively assumes that the temperat*2re and pressure of the gas corresponds to hot leg conditions throughout the core and blankets.
QUESTICN 13(b) Miat absolute assurance is there that such possibilities won't occur? ANSWER 13(b)
'Ihe plant has been designed so that gas bubbles will not acetrulate.
Ibwever, even if a significant size bubble should enter the reactor vessel inlet plenum, the turbulence would result in dispersion of the bubble into anall bubbles before entry to the core arri large coherent bubbles entering the core are inpossible. 'Ihese anall bubbles would have little effect of reactivity. This is substantiated by tests in support of the FFTF. i (Reference J. Miraka, g al., " Gas Bubble Dispersion Test Reactor Inlet ! Madel," HEEL 'DE-71-69, May 1971).
'Ihe bearing / seal assent)1y concept is not required to function normally in order to keg the oil out of the primary soditrn. Separaticn is maintained l by design features Wilch include cuersized drainage reservoirs capable of holding the entire oil inventory even in the event of mechanical failure of the face seals. Mareover, due to the reaction of oil and soditzn, it is l
SEP VII AB-32
inpossible for oil to reach the core. A algnificant buildup of hydrogen would require at least the followirg failures:
.. failure of the pump lubricating oil level indication systems, failure to take action for a period of time depending upon the oil seal leakage rate in spite of oil level alarms,
.. erroneous replenichment of the oil supply.
.. failure of the RAPS to remwe hydrogen frcm the cover gas.
.. failure of the cold trap system to maintain the hydrogen level in the soditrn at 0.1 p.p.m.
'Ihe consequences of oil leakage, should it occur despite the aformentioned protective features, are negligible and " absolute assurance" of no leakage is not required. I QUESTION 14 mat are all of the identifiable er krown possibilities for ramps and steps "beyord the PPS design basis" (see p. F3-15)?
ANSWER 14 No mechanistic sources for ranps or steps beymd the PPS design basis have been identified or are known possibilities. Section 3 of CRIRP-3, Voltrne 1, discussed ramps and steps beyond the PPS design basis and showed that the PPS has a substantial margin for acwm dating reactivity insertions beymd its design basis. Large reactivity insertions that could exceed the PPS mpability could only result fran material motions after coolable ge metry is already lost in a postulated severe accident. his, very large SET VII AB-33
~
reactivity insertico rates (tens of cbilars per second) have been consid-ered fcr OURP in the context of the progression of accidents with postu-lated failure of both shutdown systems and which lead to loss of core coolable ge m etry. NOTE: Question 15 pertains to a Ix>stulated fuel assably flow blockage , l accident. QUESTIN 15(a) Has this accident (with and without SCRAM) been analyzed (calculated) to determine the course it could take (see p. F3-19) l ANSWER 15(a) l i
'Ihis event has not been analyzed since a cxmplete flow blockage is pre-cluded by the design. Ibwever, the consequences are considered to be enveloped by the 'IOP and I& events analyzed in detail in Sections 2.1.1 and 2.1.2 of CRBRP-GE2R-00523.
Since the Fermi blockage incident, major dianges have been introduced in the design of fuel assertlies and the inlet structures to preclude fuel assably inlet blockage. 'Ihe design of the inlet nozzle of the fuel assably arrl the inlet module and module liner are presented in Secticns 4.2.1 and 4.2.2 of the PSAR respectively. QUESTIONS 15(b) If so, please Irovide the analysis. ANSWER 15(b) Not aplicable. SET VII AB-34
. N7FE: Question 16 pertains to a hypothetical 'IOP at starttp without scram.
QUESTION 16 Has this accident been analyzed (calculated) for the course it could take (see p. F3-19)? If so, please identify where in the PSAR this accident is discussed. If not, t y not? ANSER 16 An analysis has rot been performed for the continuous control assembly withdrawal at starttp with shutdcui system failure for the current design. I Section 6.3 of CRBRP-GEER-00103 contains analysis of this event for the hcmogenous design. NOTE: Question 17 pertains to a hypothetical CBE without scram. QUESTIONS 17 (a) Has this been calculated to determine the course it could take? (b) If so, please provide the analysis. ANSWERS 17(a) an8,17(b)
'Ihis event has trE: been analyzed for the current core design. Parametric analysis in Section 8 of CRBRP-GEER-00103 shows the effects of step reactivity canbiru;d with an IN-HOR for the hcznogeneous core.
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. 6
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SCP VII AB-35 i.. .
QUESTICNS 18 QUESTICN 18(a) mat is the basis for the assumption that cnly cne rod will stick in light of the Monticello (IMR) experience where the control rod drive mechanism exhibited internal cracking that would have prevented twenty three rods fran scramning? Why not asstne two, three, or trore stuck rods? ANSWER 18(a)
'Ihis question has been discussed with the Mmticello Plant Superintendent and his etaff. They state that to such occurrence, or anything relatable to it, occurred at Penticello. 'Ibere is no 'asstanption' that only cne rod will stick. Ibwever, there is a basic criterion that requires both the primary and secondary systens to be able to perform their intended design and plant equignent protection functicn even if one of the shutdown systems fails entirely ard one rod sticks in the systen that operates. 'Ihis requirement results in the desigr.
having considerable capability to achuhte nere than one stuck rod while satisfying safety considerations. Additionally, it is pointed out that diversity between the CRBRP primary ard secondary shutdcun systen mechan-ical design minimizes the potential for atmnon failure of both systems. QUESTICN 18(b) Have any CIA or DEA analyses ever been performed assunting nore than cne , t stuck rod? If so, please provide the analyses. - I ANSWER 18(b) Yes. 'Ihe HC[R analyses reported in CRBRP-3 typically asstunes no additional control rods are inserted in either of the two shutdown systens following l the initiation of the transient. 'Ihe IEA analyses reported in Chapter 15 l SEP VII AB-36
of the PSAR typically assume cne of the two shutdown systens is inoperable and one rod is stuck in the operable system. QUESTION 18(c) How many stuck rods can be tolerated for each of the various CDAs or DBAs considered without exceedirg the CRBR design basis - asstruirg the worse set of stuck rods? (See p. F3-22). ANSWER 18(c) See Answer 18(b) above for the failure assumptions used in the analyses.
'Ihe mininun numirr of rods required is dependent on the challenging event and den the event occurs. ('Ihe runber of zods available or required to scram is dependent cn variables such as pwer level, time in fuel cycle, individual rod worths, and the type of transient, e.g., loss of flow or overpower.) No assessment of the mininun ntmber of control rods required to terminate each DBA transient is available. 'Ibe question does not apply in the case of the ICDA where the asstrnption of no rods operating is a prerequisite.
QUESPICN 18(d) Identify, describe, and doctznent all almormal incidences in ocumercial, experimental, military, and prtducticn reactor - danestic and foreign - related to, or associated with control zod or safety rod failure, including failure to scram, accidental withdrawal, stuck rod (s), and errosion design, installation, repair, and operation. (This question is also related to Contention 2.) ANSIER 18(d)
'Ihe Applicants cb not naintain a file Wiich doctanents abnormal occurrences associated with the control rods as delineated in the questirn. However, l
SEP VII AB-37
data sources containing the informaticn are continually and routinely reviewed by design and safety personnel in the course of their work. 'Ihe data sources are available to the public and include the following: a) NRC Quarterly Reports to Congress Canputer Listings of Licensee Events Reports Operating Unit Status Reports Special Topical Reports b) Nuclear Safety Information Center (NSIC/OFNL Unusual Occurrence Reports Special Topical Reports c) Institute for Nuclear Power Operations (INPO) Additionally, a cmprehensive test Irogr m has been initiated to verify the design, performance, and safety aspects of the CRBRP control rod systen design. 'Ihis Irogram is described in the PSAR, Appendix C. h program will provide data which is prototypic to CRBRP. Information obtained from the review of the data sources listed above supplanents the data obtained fran this test program. QUESTICN 19 Does the parallel design involve any core design changes, such as the Doppler mefficient, relative to the reference design? ANSWER 19 Not applicable. h " parallel design" has been deleted fran the license application. I i l i l SET VII AB-38 l l
. QUESTICH 20 Will the sealed EDW be tested to withstand its design basis CEE explosion using simulant explosions? (See p. F4-1). If not, why not? ANSWER 20 Not applicable. 'Ihe license application cbes not cxmtain provisions for sealirg the head access area. QUESTION 21 Identify cr supply the analysis which dertonstrates that criteria (c) of Sec. F5.3.2 is satisfied (p. F5-4). ANSWER 21 Criterion (c) of F5.3.2 is no longer applied to in-vessel equipnent. 'Ihe current reactor ocmponent design requirenents are specified in Section 5.2 of CRBRP-3, Volume 1. 'Ihe requirements for in-vessel equipment are in Section 5.2.3 of CRBRP-3, Voltune 1. QUESTICN 22
'Ihe Vessel Closure Head and other canponents and systens of Section ES.3.3 and FS.3.4 are designed to withstand the structural loads of the design basis CIRs. Has any analysis ever been performed of the consegaences of failure of any, several, cr all of these ocznponents under such loading? If so, please Irovide the analyses.
ANSWER 22 As discussed in CRBRP-3, Voltane 1, the ocmponents are required to withstand the hypothetical core disruptive accident dynamic loads without the SEP VII AB-39
l formation of missiles or excessive leakage. Designing to the CRBRP-3, Voltme 1 (Appendix B) requirements Irovides assurance that the ecznponents will maintain their integrity under HCDA dynamic loadirg conditions. The CRBRP Safety Study (CRBRP-1) includes analysis of consequences of hypo-thetical accidents, scme of which imply failure of the vessel closure head or other ocmponents. QUESTICN 23 Has the Beginning-of-Life (new core; zero burnup throughout) condition ever been analyzed with regard to CDAs usire the same qualitative degree of pessimism used in Appendix F to explore the worst possible course the 1DP and Is initiator accidents might take? The concern here is the lack of fission gases Wiich the PSAR expects would likely cause self-shutdown for the cases of BOEC and EDEC. ANSWER 23
'Ihe analysis of a Beginning-of-Life core for the current design is in CRBRP-GEER-00523. Tnis analysis accounts for lack of fissicn gases.
QUESTICN 24 With respect to the report by H. K. Fauske, identified as Reference 5 on p. F6.2-119, and the conclusicn of Secticn F6.2.1 that no vapor explosion would occur in any core melting event, do the authors of References 5 and 6 en p. F6.2-Il9 agree with this interpretation? ANSWER 24
'Ihe authors of References 20 and 63 in CRBRN-00103 agree with the positicn that the occurrence of coherent energetic vapor explosion in the CRBRP envircrvnent with cuide fuel can be considered highly unlikely. Ebr further details see Section 8.2.6 of CRBRP-GEER-00523.
SET VII AB-40
QUESTICN 25 (PREAELE)
'Ihe PSAR states that point kinetics is used in SAS3A (p. F6.2-6) and VENUS (p. F6.2-105-106). Point kinetics is an approximation to the method of calculating the reactivity from eigenvalue differences using static diffusicn theory, which is an approximation of the FX-2 type factorization method (quasi-static method), tich is an approximation of the time-dependent diffusicn, which in turn is an approximation of time-dependent 1
neutron transport theory. i l QUESTION 25(a) l mat specific allowances for the error of each of these approximations have been made, if any, ard what is the mathenatical basis for the magnitude of these allowances? ANSWER 25(a) No specific allowances for the errors which might exist in the approxima-tions mentioned were made, since such errors were judged to be small for CRBRP. The mathematical tases for the magnitudes of such errors are as follows: (1) With respect to the differences between diffusion theory and transport theory, the estimates of reactivity changes will not be significantly different miess significant void spaces exist, such that the diffusion coefficient be.wes infinite over large areas of the core. Such differ-ences do not exist in any of the cases considered in CRBRP-GEFR-00103 or CREP-GEFR-00523 so that diffusion theory is adequate. 'Ihe mathematical basis can be found in discussion en the relationships between neutron diffusion theory and neutron transport theory such as in Reactor Handbook, H. &vvhk (Editor) Interscience, John Wiley and Sons, second editico, p. 140. SET VII AB-41
. (2) With respect to the differences among approximate methods of solving the time-dependent diffusion equation, namely, point kinetics, eigenvalue differences using static diffusion theory (" adiabatic" methods), and factorization methods, the mathematical basis can be found in Reference 1 below.
It is shown therein that all three methods are factorization methods, differing only in their methcds of testirg the time dependence of the flux shape. QUESTION 25(b) In Reference 75 (PSAR, p. M.2-124), the error between the static eigen-value methoc3 and the FX method was found to be 33% non-conservative in the energy yield of the power excursion; and yet a stronger Dogpler coefficient was used than the values used in the CRBR PSAR. What is the error between these two methods for the GBR cases ocnsidered, i.e., using CRBR Ibppler coefficients ard the CRBR CDAs considered. ANSWER 25(b) No otrnparison was mde between (i) FX2-V"WS space-time kinetics (ii) FX2-VEN W with point kinetics (" static-eigenvalue method"), and (iii) VENUS-II (point kinetics) for GBRP. QUESTIWS 25(c) and 25(d) (c) 14mt is the basis for not performing the CA analyses (SAS and VENW) with at least the quasistatic method? (d) 'Ihe PSAR (p. M.2-lO6) asserts that the inaccuracies incurred by not using the quasi-static method are not serious, and cites Sha, et al. (Reference 75) for the 1mais. Yet, Reference 75 concludes that **Ihe cases SEP VII AB-42
. described here do not constitute an adequate test of these approximations . in any sense" (o.147) . Specifically, how is it judged frcrn Reference 75 !
that the error of SAS/VDRE relative to the quasi-static neutron kinetics approximation would not be serious? ANSWERS 25(c) and 25(d) { r h Applicants response to Interrogatory II-1(a-c) of the Sixth Set of - Interrogatories to Applicant (see p. AA-138) presents an analysis which + shows that the quasistatic method is not significantly nore accurate than ! point kinetics with worth tables generated usirg the linearized leakage ; treatment for the initiating t ase h (SAS) for CRBRP analyses. Regarding the j disassenbly phase, the cmparisons provided in Reference 68 in CRBRP-GEFR-00103 for the high tenperature initial conditicns Irovide the basis for concludirg that the results predicted when not using the quasistatic method [ do not differ significantly. QUESTICN 25(e) i mat is the basis for the PSAR's conclusion that the plant design can t safely contain the consequences of a CDA, in view of the fact that Sha, et_ j al., (Reference 75) qualified the quasistatic method (known theoretically to be at least less approximate than the sinple neutronic model used for : the PSAR) in regard to accuracy, stating, "Cbviously, no firm statement can ; be made concerning the errors introduced by these models (used in the quasistatic method)? In this regard, it has been noted by Meneley, g ; al., that the quasistatic method is an approximation of neutron diffusion theory, W11ch is in itself an approximation of neutron transport theory. t (Sgt Reference 74, p. 486). l i ANSWER 25(e) l
?
h quote, fran page 151 of Reference 68 in CRBRP43EPR-00103 was not [ intended to be a qualification of the quasistatic method but rather referred to methods used by the " analysts" do " apply a point kinetics t i l I ! i SET VII AB-43 i
i I l
. 1 nodel with coefficients mlculated for conditions which are representative of the conditions expected durirg the transient." In other words, the anount of error introduced by using point kinetics is really case-dependent. 'Ihe quasistatic methcd itself is a ntanerical method of solving the time-dependent diffusion equations, tich has been shown (2,3) to !
converge to the soluticn as calculated using finite difference methods on a very fine nesh, as the runber of shape function calculations is increased. a
'Ihe point kinetics fornpla is one in which only one shape function is used to describe the space dependence Wether this shape is selected to be for the state at the beginning of the transient, or foc a state expected at sone time during the transient. If the transient tmder consideration does not include gross material motions, such a use of point kinetics may be adequate. Ebr the CRBRP cases analyzed and 1 resented in CRBRP-CEER-00103 and CRBRP-GEFR-00523 this was indeed true, as is demonstrated in the answer i to Interrogatory II-1 in the Sixth Set of Interrogatories (see p. AA-138). l QUESTICNS 25(f) and 25(g)
(f) Furthermore, it is noted that even though the SAS/VINUS calculations are ultimately based cn diffusicn theory, Boudreau, et al., (A Proposal for Computer Investigation of LMFBR Core Meltdown Accidents, for Alamor, IA-UR-74-243 (undated), pp. 14, 28), have concluded that time dependent transport theory is necessary, at least in order to investigate the possibility and severity of secondary criticality (secondary power excur-sions). In fact, in cne calculation, the reactivity was tmderpredicted by diffusicn theory by about 3% K reactivity units (p.16), which would be disastrous if a substantial amount of this reactivity could be " inserted" - in an explosive recompacticn event. Is Boudreau, g al.'s analysis correct? (Here, we have reference to a private cmmunicaticn between Wetb and D. Ferguson, ANL). If not, why not? If it is correct, in answering (e) above, also be responsible to Boudreau , g al.'s analysis and conclusions cited here. ' l l i SET VII AB44
, (g) If Boudreau, et al.'s analysis is incorrect, discuss in detail why the
, Applicants believe diffusion theory is still adequate for all CRBR CDA analyses of interest.
ANSWERS 25(f) and 25(g) Boudreau's analysis is not correct. An inconsistency was present between the two treatments of the diffusion terns. When this inconsistency is renoved, nuch closer agreement results. 'Ihe inportant point to be derived fran Figure 3, cm p. 16 of IA-UR-74-243, is that, even though the magni-tudes differ, the slopes of the curves of k,ff vs separation distance are nearly the sane. The slope is what determines the reactivity insertion rate; the mergy release fran a disassernbly calculation is dependent on the ranp rate chosen. When the two curves are brought together by movirg the
'INC7fRAN-II curve (bwnward, it will be seen that the slope at protpt criti-cal (Ak units /en) will be nearly the same for each curve. with a knowledge of the slug velocity at this time, crie can then derive a ranp rate (Ak/
sec), which can then be used to drive the hydrodynamic disassembly. GESTICN 25(h) Miat (even if considered outside the area of interest) would have to be postulated before diffusion theory would be considered inadequate? Explain in detail. ANSWER 25(h) Cnly a recriticality event in Wiict very large voids were Iresent at the time of pronpt criticality would require consideration of the use of transport theory. Such a conditica has not been identified in any of the OURP analyses performed to date. SET VII AB-45
. QUESTION 25(i) Please supply Reference 25 in the Ibudreau Ircposal, namely, IA-4432 - l Theory and Use of 'IWOTRAN. ANSWER 25(i) i
'Ihis cbetraent has been or will be made available for inspection and copying.
QUESPICN 25(i) As noted by Boudreau, g al., transport theory predicted a secondary power excursicn, whereas diffusion theaty did not, fte the case examined (p.14). In light of this result is the CRBR project planning to reanalyze the CRs , using time-dependent transport the<ry; or is it not practical to do so? If , not,
- y not? If so, please discuss the details of the proposed program.
ANSWER 25(j)
'Ihere is no plan to reanalyze ICRs using tMependent transport theory. 'Ihe reasons are found in the responses to parts (a)-(h) above.
QUESTICN 25(k) On *at grounds can the Applicut justify the safety of the CRBR relative to the CDAs analyza3 in light of the apparent need for time-dependent transport theory? 'Ihe answer to this question can be included in (c) above (See (f) above.) ANSER 25(k) As indicated in parts (a)-(j), diffusion theory is adequate for - ICA analyses for the CRBRP. SEP VII AB-46
QUEETPION 25(1) In the Ibudreau 'IWCTIRAN transport mlculation, which used the SN methcd, diat was the order of quadrature (N = ?)? Prestm.bly, Boudreau used a (rest of question missing) ANSWER 25(1)
'Ibe Ibudreau 'IWOPRAN " transport" calculation was agP calculation. 'Ihe Sn methcd was not used.
QUEErrION 25(m) Has this order used been established to be sufficiently accurate within the limits of the S method, n by ocmpariscn with higher order calculations? ANSWER 25(m) APgcalculation is no more accurate than diffusion theory calculatior.s. QUESTICN 25(n) In several of the CIRs a cavity is predicted or postulated to develop. What order quadrature would be necessary to accurately calculate the reactivity? (Present the mthenatical analysis in support of the answer). ANSE R 25(n) Such an analysis has not been mrried out. Diffusion theory is suffi-ciently accurate to analyze all cases of interest. Therefore, S is also i 2 sufficiently accurate. 1 i SEr VII AB-47 l
i l
. QUESTION 25(o)
It should be noted that 'IWOTRAN is a two-dimensional Sn . e re any plans to calculate any of the more severe CIA unirg three-dimensional S , say with N = 4? n ANSWER 25(o)
'Ihere are no plans to calculate any FEIAs using three-dimensional Sn methods.
CUESTICN 25(p) e Is there any analysis Which shcws that the additional accuracy afforded by a 3-D, 4S meM is n Wed? If so, please provide. ANSWER 25(p) As indicated in parts (a) - (j), the methods Iresently utilized for CRBRP analyses of ICIRs are adequate. N3 analysis exists which would show the benefits, if any, of utilizing a 3-D, S method 4 of analysis. QUESTICN 25(q) Has time-dependent transport theory ever been acrnpared with an adiabatic-type transport calculation, similar to the classic space-time calculaticms of Yasinsky and Henry (N. S. and E., 22: 171-181)? Ebr the CRBR core desicy1? ' ANSWER 25(q) ' No. An indirect ocznparison exists in a paper by E. L. Fuller in Reference :
- 4. In this work static diffusicn theory, S , and hwA,t. diffusion 4
SET VII AB-48
, thecry (FX2) are ocmpared in an analysis in which void spaces were present.
, Agreenent was very close arrong these three methods. Since the major differences betwem " adiabatic" methcds and " exact" space-time methods lie in the treatment of the variation of the Irecursor concentrations, and since the transients resulting fran ms are very rapid, it is concluded that time-dependent transport theory would give essentially the same results as eigenvalue differencing using transport theory for the cases of interest.
QUESTI m 25(r)
'Ib use time-dependent transport theory, there is the <pestion of the " stability" of the S n finite difference calculaticnal methcd. In 1964 Clark and Hansen noted that no proof of runerical crmputations stability '
exists (Nanerical Methods of Reactor Analysis, p. 223) . (i) Has a proof now been made? Where? (ii) What is the significance of this? (iii) Should a mlculational result be mstable, can it be shown that it will always be obvious? (iv) Could it be that a result will be grossly in error without any irdicaticn of a calculaticra being on the verge of being unstable cr slightly unstable? ANSWER 25(r) (i) 'Ihe applicant is not mere of such a proof. (ii) 'Ihe fact that a mathanatical proof for the unconditional stability of the Sn finite difference calculational method does not currently exist is of little significance in regard to the practical application of the method, and is of no significance to CRBRP M evaluations. (iii) It cannot be mathematically shown that, should a calculational result be metable, it will always be obvious. Ibwever, should a :nnerical SET VII AB-49
. instability occur, it will usually be discovered by a careful inspection of
. the calculaticnal results.
(iv) To the best of the Applicant's knowledge it has not been shown that a result could be grossly in errx without any indicaticn of a calculation t being cn the verge of being instable x slightly stable. QUESTICN 25(s) Wiat is the status of the cnnputer investigation proposed by Ibxireau, g al., at Ios Alamos? Have any results been issued in report (including draft) form? If so, please p ovide these. r ANSWER 25(s)
'Ihe SI!@ER code developnent by Boudreau g al. is still in progress.
QUESTICN 25(t) Please supply; the cneMJroup and tse-group diffusion constants
- for the 1 CRBR, for va#ious fuel zcnes at various depleticn states for each of the cases, BOL, IYRC, and EDIr.
*Namely, D, k, f ,, M, 2. f, 32 , h2, etc.
ANSWER 25(t) OrwHJroup and twt >-group constants do not exist for the GBRP. 'Ihe sets that were used for the GBRP-GEFR-00103 calculations were the 27-group set described cm page 5-1 and the nine-group set, mentioned cn pages 10-16 of CRBRP-GER-00103. A similar set was used in CRBRP-GEFR-00523. l SET VII AB-50
. QUEErrICN 25(u) In the SAS calculation the use of neutron diffusion theory may be consid-f ered by the Applicant as having been shown to be an adequate approximation I of transport theory by Ferguson, g al. (Reference 3, PSAR, p. E6.2-119), citing Fert3uscn, et al.'s, fair agreement usirg an S4calculaticn. However, l Ferguson, et al., emphasized that they assumed that the S calculation is 4 sufficiently accurate (Referenca 3, p. IX-68). What plans, if any, does the Applicant have to verify the validity of this asstanption? ANSWER 25(u) l i
'1he Applicants have no plans to verify the validity of the assumption mentioned.
QUESPICN 25(v) , 1 Do any of the CDAs analyzed in the PSAR involve significantly nore fuel sitznpirg and/cr voiding of core materials which may shoe a greater trans-port effect than the fuel sltanping situation calculated by Ferguson, g _al.? ANSWER 25(v) l l l
'Ihe fuel sitmping situation mentioned involves about the same amount of fuel as is involved in the innediate-reentry analyses in CRBRIH3EFR-00103.
In addition, the relative degrees of sodium voiding are similar. QUEErrICE 25 (References)
- 1. K. O. Ott and D. A. Meneley, " Accuracy of the Quasistatic Treatment of Spatial Reactot Kinetics," Nucl. Sci. Eng. 36, 402-411 (1969).
- 2. E. L. Fuller, "One-Dimensional Space-Time Kinetics Berd n=rk Calcula-tions," Argonne National Laboratory, Applied Physics Division Annual SEP VII AB-51
. Report, July 1, 1970 to June 30, 1971, ANL-7910, pp. 497-502 (Jan. 1972).
- 3. H. L. Dodds, Jr., " Accuracy of the Quasistatic Method for 'Iko-Dimen-sional '1hermal Reactor Transients with Feedback," Nucl. Sci. Eng. 59, 271-281 (197?)
- 4. E. L. Fuller, " Reactivity Effects of Cbre Sltrping in Fast Reactors: A Case Study," Argonne National Laboratory, Applied Physics Division Annual Report, July 1, 1971 to June 30, 1972, ANIe8010, Ip. 583-587 (1976).
QUESTION 26 Is there any consideration Whatsoever being given to core destruct experi-ments, includirg partial core destruct experiments? If so, please provide a detailed description of what is plamed in this area, including when results are expected. Please provide all doctments related to any consid-eraticms given to core disruptive experiments. ANSWER 26
'1he Applicants have given no consideration to core denruct experiments and has produced no doctments relative to such experiments.
QUESTION 27
'1he SWMPY fuel motion rrodel contains a pseudo-viscous Iressure for the purpose of providing ntnerical stability.
(a) mat is the physical basis for this pseudo pressure? (b) Is this the von Netmann term which is founded on the Hugoniot shock relation? SET VII AB-52
l ANSWER 27 h SWMPY pseudo-viscous Iressure Irovides an autcznatic treatment of shocks that can arise due to the low sonic velocities that can occur in the two-phase system that SWMPY models. 'Ihe algorithm is based cn the two-phase fornulaticn fcr two-phase calculaticms without a significant increase in the widths of the shock frcnts (H. U. Wider et al., "An Improved Viscous Pressure Fornulaticn for '1%o-Phase Ctmpressible Hydrodynamics Calcula-tions," Trans. Am. Nucl. Soc., 17, p. 246, 1973). In a sirgle-phase l situaticn, this formulation indeed reduces to the von Netsnann term founded on the Hugoniot shock relation (J. von Neumann and R. D. Richtmyer, "A Method fcr the Numerical Calculaticn of Hydrodynamic Shocks," J. of Applied r Physics, 21, pp. 232-237, 1950), with the inprovernent that the pseudo-viscous pressure is zero when the material is undergoing an expansion (as of a free surface) (R. D. Richtmyer, Difference Methods for Initial-Value [ Problens, Interscience Publishers, Inc., New York, pp. 210-211, 1957). i QUESTICN 28(a) Can the Doppler coefficient be readily varied (reduced) in the core design change without significantly affectirg thermal, mechanical, and hydraulic design? ANSWER 28(a) h Doppler coefficient i.1 CRBRP cannot be readily varied by core design changes without affecting thermal, mechanical, and hydraulic design ecmditions. Any fuel cr core design changes diich would be considered to increase the breeding capability are constrained by ptmp head (coolant Iressure drop), fuel lifetime, and other thermal / hydraulic cr mechanical P limits. '1he Doppler effect in CRBRP is primarily a function of the U-238 ness (fertile-to-fissile ratio) and the fraction of the neutron flux in the , r resonance range (neutron moderation). i I L SEP VII AB-53 i
QUESTION 28(b) Please supply a curve of breeding " doubling time" versus the Doppler coefficient (sodim in/out). ANSWER 28(b)
'Ihe CRBRP breeding ratio is a performance parameter and, as such, it is only calculated fcr naninal (sodim in) reactor conditions. Ibwever, a unique curve of treeding ratio versus Doppler coefficient is not meaningful due to the dependency of, for exanple, Doppler coefficient en U-238 ccmtent, fuel cxznposition, neutron nederation (spectrm) frcm steel and sodim content, etc., ard the variety of core design changes which could be proposed to result in a particular treedire ratio change. 'Ihe change in the Doppler coefficient is assessed explicitly for any particular fuel or core design change considered for CRBRP.
QUESTICE 29 In each of the SAS3A Iredictions of self-shutdown by fuel ejection that were considered in the PSAR, what percentage of the fuel in the core is ejected frcm the core? ANSWER 29 CRBRP-EFR-00523 describes the analysis for the current core design. GJEErrICN 30 In the various CDAs analyzed, dat are core average and (local) maxim.ra fractions of fissicn gases that were originally in the core, that bubble out of the nolten fuel Irior to any recriticality event, including power excursicn event? SEP VII AB-54
ANSWER 30
'1he analysis of recriticality events in GBRP-GEFR-00103 did not consider the presence of fissicn gas as a dispersant. Therefore,100 percent of any fission gas ); resent was assumed to bubble out of the molten fuel glor to any recriticality event, includirg power excursicn event. t QUESTICES 31 (PREAMBLE) ;
i In view of the four-fold difference in the neutron mean free path between i LD2 anS sodium, one might expect a streamirg effect of neutrons along the ! coolant channels (sodium in). Een fuel sitznping occurs in which there is a substantial loss of core gecznetry, this streaming effect might conceiv-ably add a significant reactivity effect of core ecznpaction or expansion. QUESTICN 31(a) Has it been shown theoretically cr experimentally that such a streaming effect is negligible? ANSWER 31(a) Neutron streaming has a negligible reactivity effect in CRBRP. Even though ' fuel (UO2 ) and c lant (soditan or partially scxiitzn void condition) have substantially different neutron mean free paths, neutron streaming is negligible when these materials are relatively hcmageneously mixed because the meditan is then essentially irotropic. Cbnsideration of the CRBRP fuel ascenbly design shows the fuel and coolant are relatively hcznogeneously , mixed, i.e., 0.23 inch diameter fuel rods with a pitch-to-diameter ratio of ; 1.256. Neutron streaming mn cnly have a non-negligible effect if the fuel and i coolant are arranged in a very heterogeneous manner such that (1) the SEP VII AB-55
radial fuel dimensicn represents a substantial fraction of the neutrcn mean free path and/or (2) a two-dimensional streaming path exists. Neither of ' these conditions exists in CRBRP.
'Iherefore, geonetric aansiderations alone are sufficient to denenstrate that neutrcn streaming in CRBRP is negligible.
QUESTION 31(b) In regard to neutron streaming in the case of sodiun out, such as the loss of ficw without SCRAM, which is predicted by SAS to have much or nest of the core voided of coolant, have any theoretical estimates been made as to the reactivity effect of the streaming alone? I ANSWER 31(b)
'Iheoretical estimates have been made of the reactivity effects of neutron streaming in CRBRP with sodiun out. These results were presented in Reference 1. From the results of three-dimensional Mante-Carlo Calcula-tions (Reference 1) heterogeneous neutron streaming effects are not expected to be inportant for the sodium void worth and cladding worth in CRBRP.
1 i QUESTICNS 31(c)
'Ihe NRCs Reactor Safety Research Program (NURB3 75/059, pp. 26-27) notes that Monte Carlo calculations of streaming are being attanpted.
(i) 'Ib Wat extent is the Applicant, its consultants, or other researchers performing these particular calculations? Identify *ere and by whan this work is being performed. (ii) Are there any results as to reactivity estimates? (iii) W at is the purpose of these calculations? SEP VII AB-56
(iv) Does (and if so, how cbes) the Applicant, or its consultants , expect to use the Monte Carlo calculations in assessing the safety of the INFBR? (v) Please supply all documents relating to research work identified above, including the research proposal, CRDA approval and comnenting memoranda. ANSWER 31(c) _
'Ihe Applicants have not- depended cm the doctrnent referenced in the inter-rogatcry. 'As stated in response to parts a) and b) above, the Applicants has considered the potential inpacts of neutron streaming and the inpact is not expected to be significant.
s GEStrICN 31(d) Kohler and Ligow Irelicted a reactivity effect due to neutron streaming in the Gas-Cooled Fast Faactor of about 1% dK (Nuc. Sci. & Eng., 54:357-60). Discuss any and all atmsiderations given by the Applicant cr its con-sultants or experts known to the Applicant to the possibility that during a
" hydrodynamic disassetly," e.g. , severe power excursion in a core voided substantially of coolant, as the fuel rods swell during the early phase of the excursion that the reduction of neutron streaming may cause an auto-catalytic reactivity fearnwk effect?
ANSWER 31(d) . l
'Ihe cnly analysis applicable to OtBRP known to the Applicants is that presented in Reference 1. Frcm this analysis it is judged that such ccmsideratirms are not inportant for CRBRP.
l l 1 l l l SET VII + AB-57
QUESTICN 31(e) Itiat theoretical effort, if any, has or is being undertaken to include this process (changes in neutrcn strenming) in the SAS/ VENUS CIA analyses? Discuss fully any results of this effort. ANSWER 31(e)
'Ihe Applicants have reviewed reference 2 for its applicability to CRBRP HCIA analysis.
QUESTIG7 31 (References)
- 1. F. E. Dunn and R. Iell, " Heterogeneous Neutron Streaming Effects in the Clindi River Breeder Reactcr," Trans. Am. Nucl. Soc., 22, 373 (1975).
- 2. Gerald Lee Goldsmith and Richard B. Nicholson, " Reactivity Due to Neutrcn Streaming in the Voids of A Bubbly Pool Core, Design Basis Accident Studies, Final Report, Richard B. Nicholson, editor, Chio State University, CDO-2286-3, pp.104-171 (1974) .
QUESTION 32 Professor R. B. Nicholson was an AEC consultant in INFBR safety research after March 1972. Please supply all of the doctanents, writings, papers, articles, letters to the AEC, by Dr. Nicholson in which the results of his research are presented and discussed. ANSWER 32
'Ihe Applicants are not famil iar with Dr. Idenolson's Irecise role and activities as an ABC consultant cm INETR safety research. The Applicants believe that he may have been a consultant to A Q Regulatory. He has not been a DCE consultant cr involved in CREP safety analyses. 'Ihe Applicants do not have the cbetanents, writings, papers, etc., by Dr. Nicholson in which the results of his research are presented and discussed.
SEP VII AB-58
e , QUIETION 33 In the various loss of flow accidents, dat otmsideration has been given to the possibility of om or more coolant peps restarting, which could ocnceivably lead to rapid cxrnpaction of a pliable cnre - made pliable by overheating - due to the onrush of returniry coolant flow? Please supply all relevant analyses and supporting cbetanents. ANSWER 33 If one acmsiders the extreely low Irebability case of loss of off-site , power ard the resulting main ptstp loss coupled with the failure of the primary and secondary cxmtrol assenblies to insert, cne could attempt to restart the primary heat transport system punps if off-site electrical power were restored. Ibwever, flow would not return autanatically on reestablishirg off-site power. However, restartirg the Irimary pumps even if electric Iower were available is not a simple one-step operaticn. 'Ib satisfy installed interlocks, the operator would have to reestablish lube oil flow to the ptrnp power supplies (motor-generator sets) before reenergizing the pump power supplies. Procedures would then require establishing intermediate flow before [ attsptire to establish primary coolant flow. Even if the procedures are violated, the Irimary heat transport system ptrnp breakers are inter 1ccked with the shutdown systes to prevent establishiry primary ficw whenever any scram treaker of the Irimary electrical subsystem cr solenoid valve of the seccrdary electrical subsystem is open. This interlock with primary planp breakers would Irevent reestablishing flow in this postulated extree low probability case. With this plant design, it is rot reasonable to postu- ! late that the Im ICIR could occur, offsite electrical power is restored, and a primary planp is restarted. i l SEP VII AB-59
QUESTICN 34 l hat ocxisideration, if any, has been given to whole cnre destruct tests with zero burntp and fueled only by U-235 (no plutonium)? Such tests would I avoid a serious radioactivity hazard. ANSWER 34 No consideration has been given by the Applicants to whole core destruct test of any type and therefore not to the test supOsed QUESTICN 35 Please list all of the physical differences between each of the TREAT experiments and CRBRP design, including but not limited to differences in fuel rod and lattice dimensions, coolant flow rates and tanperatures, fuel burntp levels, fuel canpositicn, fuel red height, axial blanket lengths, power output, fuel temperatures (initial), reactor transient period. (Simply citing references to TREAT and CRBR data would not be an adequate response to this question.) ANSWER 35
'Ihe TREAT reactor has been used for many experiments, a considerable ntrnber of which are not applicable to the CRBRP m analysis and which, there-fore, have not been used to support CRBRP m evaluations. Within the context of the subject matter questicned, those physical differences believed to be significant and associated cnly with those TREAT tests that have been utilized to support the first principles modeling and engineering jtukynents etployed in the CRBRP m analyses can be identified. 'Ihe TREAT reactor is a transient test facility having a thermal neutron flux with no significant cooling system. It operates en the p-inciple of utilizing the heat capacity in the reactcr fuel to absorb the energy generated during the transient. 'Ib con 3uct fuel-in-sodium tests the TREAT SEP VII AB-60
O facility requires that the test be conducted in test capsules which contain the sodiun ard fuel test specimens. A variety of these sodium filled capsules have been designed for use in the TREMP facility. Three major categories of capsules can be identified: static capsules, circulating sodium capsules (Mark II loop capsules), and transient modiun flow capsules (R-loop capsules). Each of these types of capsules introduces physical differences into the experiments. In addition to the test vehicle introducirg its own differences, the experimental capsules may contain a single fuel pin, multiple fuel pins, or actual fuel pins and sinulated fuel pins.
'Ihe experimenter's choice of capsule and fuel pin test assembly design is a function of the physical phenanena to be investigated. Atternpts are made to produce the best approximation to the physical phencrnena to be inves-tigated within the constraints of capsule type and test fuel pin design.
Another extremely inportant variable is the instnanentation capability associated with the capsule. In the interest of obtainirg greater amounts of measured data, the experimenter may accept additional differences in the I test and reactor conditicn. It should, therefore, be understood that several physical differences nay exist between a TREAT experiment and a particular reactcr design. mny of these differences cb not have an inportant effect cm the principal physical phenonena that the TRFAT test may be designad to investigate. Table I sumnarizes tat the Applicants cxmsider to be the significant physical differences that exist between the TREAT experiJnents that have been most directly utilized to support the validity of the CRBRP IE M analysis ard the CRIRP design. l SET VII AB-61 l
1NER I. Wristm of TIEftf 1tset Parameters and OEEt Desip Parameters sM III
.!1tERP N M OtIEL Typimi Valise R4,R5, or Core Wide ILv.ge RS,R7, Doei p Parameter 1 hits at Power (975 Pett) BB L2 L3 1A 15 04A C48 d A.CSB S11,S12 IOP-3C Active Mael Height m 91 (4) 34 34 34 86 36 36 61 15 34 Ugper Blanket Height (2) on 36 17 17 1 30 9 12 12 36 1 Mael Pin Diameter au 0.53 (4) (4) (4) (4) (4) 0.64 0.64 0.64 (4) (4)
Fuel Sremer Density As t 85 (4) 87 87 (4) (4) 90 90 90 (4) 88 Percent of Theoretimi Mael apar-ific Power I3I att/gn 50.-180 (4) (4) (4) (4) 190 988(6) 1475(6) 1057(6) 32,000(6) 214 fuel Mael Burnup OfD/T O.6-110 Zero Zero (4) (4) (4) Zero Zero (4) Zero (4) Mael Amentaly coolant Cb 333 (4) 400 400 460 400 310 425 s650 s150 s425 N Inlet Temperature Coolant Preneure Drty Epnocal 70-110 (4) (5) (5) (5) (5) (7) (7) (7) (7) (7) Onr Active Fuel Imngth molant Average leise gq/sec 80-112 130 (4) (4) (4) 49 (7) (7) (7) (7) (7) Flow Rate Per Pin Coolant and Spacer wire - 0.75 (4) (4) (4) (4) (5) .69 .69 (5) 2.5,2.0 1.11 Area to Fuel Pin Ares , lah of Mael Pins - 217 7 7 7 7 3 1 1 1 1 1 Per Assent >1y 10t115: 1. Values reported in er calculated fran IEIL ard Alt, Reports.
- 2. Reflector and/or blanket netterial atzgrisirvj thermal heat sink ard inertial restraint.
- 3. Flattcp value durirg 11 TEMP controlled transists.
- 4. Value is within range of, or approximates, OtBRP design value.
- 5. Detailed value not reported cr readily available.
- 6. 11 TEAT natural transient, half niaminima p21se height of transient.
- 7. Static capsule desip, parameter not applicable.
1 NEE I. (Ckmtirmed) thymristm of 11Eftr itset Pariumsters ard OEut Desicpi Parameters s H III
'litEXT EXPERIPENr VAIDE cent Typimi Value or Cbre Wide Range 16, R9, Design Parismeter thits at Ibwer (975 PWt) HJr5-3A H3 H4 H5 B6 E7 L7 IB H6 ES J1 F1 F2 R12 Active Fuel Height m 91 34 34 34 34 34 34 86 86 34 34 34 34 34 (4)
Upper Blarhet Height (2) on 36 30 1 35 1 17 30 9 9 40 40 27 30 14 17 Fuel Pin Dimieter an 0.53 (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) Fuel Samar Density As t 85 (4) 88 (4) 88 (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) Percent of 1heoretical Fuel Specific Power I watt /cyn 50-180 246 243 383 (4) 272 231 188 408 158 168 315 198 206 230 fuel Fuel Burnup GID/T O.6-110 (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) 0.34 Zero W Fuel M1y Ozolant CD 333 s250 377 368 354 393 407 400 400 470 407 476 25 25 (4) Inlet Temperature Ocx21 ant Pressure p op Epascal 70-110 (7) 22 22 22 22 22 (5) (5) 44 (5) 47 (7) (7) (4) Over Active Fuel Length Cbolant Jwerage thes cyn/sec 80-112 (7) (4) (4) (4) (4) (4) 50 50 (4) (4) (4) (7) (7) 136 Flow Rate Per Pin Ozolant and spacer Wire - 0.75 1.12 .% .% .% .97 (4) .85 .85 (4) .95 .89 .82 .32 (4) Area to Fuel Pin Area NLsiter of Fuel Pins - 217 1 7 7 7 7 7 3 3 7 7 7 1 1 7 Per Asseibly te: PRES: 1. Values reported in er calculated fran EETL ard Art Reports.
- 2. IWflector ard/or blanket interial apprising thermal heat sirk ard inertial restraint.
- 3. Flattg value durirg 11tERP contro11a1 transients.
- 4. Value is within range of, cr g5roximates, QUMtP design value.
- 5. Detatisi value not reported cr readily available.
- 6. 11trAT natural transient, half imxirasn gulse height of transient.
- 7. Static capsule design, parameter not applicable.
l QUESTICN 36
'Ihe PSAR assumed reactivity insertion rates starting the TOP as high as 10$/sec (p. F6.2-86) . What possible 'IOP initiator would yield a 10$/sec rate? l l
ANSWER 36
'Ihere is no TOP initiator that would result in a reactivity insertion rate anywhere near 10$/sec. The 10$/sec ranp included in CRBRP-GEFR-00103 was part of parametric calculations to determine if any new phencrnena became important at higher than realizable ranp rates.
IUTE: Question 37 pertains to the consequence of assaning midplane failures. - QUESTICN 37
'Ihe PSAR considers the possibility of midplane fuel failure.
(a) What fraction of the fuel rods were assumed to fail at the midplane, and what was the negnitude of the reactivity feedback frcm the midplane failure fcr the various CIAs considered in Appendix F7 (b) What percentage of the core was voided of coolant when the positive reactivity feedback occurred? (c) The PSAR (p. FE.2-86) analyses 0.5$/see to 10$/sec ranp rates. Have 0.01 to 0.lS/sec ramp rates been considered when assuming midplane failures? (d) Would the sodium be nore likely to be boiled out of the cxare for ramp rates lower than 0.5$/see? , SEP VII AB-64
(e) Has the BOIC for 'IUP been analyzed asstaning midplane fuel rod failures for 0.01 to 10$/sec (P6.2.4.3.2 is not clear cn this point)? If not, why not? If so, where is this discussed in the PSAR? ANSWER 37(a-e)
'Ihis question appears specific to the iuuy.neous core sich is not the current design. Accordingly, it requires no answer. Section 6.1 of CRBRP-GEFR-00523 describes the 'IOP BOC-1 analysir. for the current oore design.
QUESTICE 38 (a) In the transition puise frczn IN events, Wat fractions of the initial fissicn gas inventory would renain in the molten core for the disassembly phase? (See p. EE.2-93). (b) Supply the basis fcr these estimates. ANSWER 38
'Ihe question implies that the transition phase is followed by a disassenbly phase. It stould be reenphasized that a rapid recriticality laading to hydrodynamic disasserobly frcm the transition puise is highly inlikely because of the inherent dispersal characteristics of a nelten core with entrained steel.
'Ihe literal answer to part (a) is that no quantitative estimate of the anount of fissicm gas renaining can currently be given. Qualitatively, the fission gas concentration should be low, as is suggested by analysis of the resolidified molten fuel in TREWP tests arx3 other out-of-pile heating experiments cm irradiated fuel. Because ;recise quantitative estimates of fissicn gas availability were not available, the calculations of specula-tive disasserrbly transients in CRBRP-CEER-00103 were done in a conservative fashicn, asstaning no fissicm gas was present. '1he fact that the dis-assenbly calculations were conservative is an adequate basis for the SEP VII AB-65
. estimate. It is well known that the tresence of only a few percent of the . fission products could significantly reduce disassembly energetics, although delays in pressurization due to these fission Iroducts of only a few milliseconds will substanHally reduce the effect (J. F. Jackscn and A. M. Enton, " Pressurization Ihte Effects in Irradiated Cbre Disassembly Calculations." Trans. AM. Nucl. Soc. 22, p. 370, (1975)). NCTTE: Question 39 pertains to extended fuel notion. QUESTION 39 QUESTION 39(a) Wen noiten fuel (and molten steel) noves into the blanket region, could it encounter sodim there? (See p. F6.2.5.2.2). ANSWER 39(a)
'Ihe possibility that the nelten fuel-steel mixture entering the blanket regicn will encounter liquid sodiun cannot be ruled out.
QUESTICN 39(b) If so, how was this treated in the analyses of CRBR CIRs? ANSWER 39(b)
'Ihis subject is evaluated in Sections 8.2 and 8.3.6 of CRBRP-GEFR-00523.
QUESTION 40 Please supply all program proposals and test reports related to the " Upper Plerun Injection" experiments mentioned on p. EE.2-101 SEP VII AB-66
l
. ANSWER 40 ANL/ RAS 76-4, Upper Plentra Injection Tests tb. 1 and No. 2, Ibbert E.
Henry, 3 al., February 1976 is the only doctrnentation available on conpleted tests. 'Ihe Applicants have not Iroposed any programs for addi-tional tests of this type at this time. GJESTION 41
*Ib explore reactivity effects in a disruptive core, the PSAR asstrned an IEEE core:
(a) Shouldn't the BOL and IDEC cases be examined too, since they would contain more excess reactivity? If not, why not? If so, why hasn't this been examined in the PSAR? ANSWER 41(a) Reactivity effects for both the IOC-1 and the EOC-4 disrupted cores have been evaluated ard are reported in Appendix F of CRBRP-GEFR-00523. QJESTION 42 (a) Have any core destruct tests ever been considered which could provide the needed experimental confirmation that CDAs will not likely lead into a superprompt critical power excursion and hydrodynamic disassembly (explosion)? (b) If so, diat are the cost estimates of such tests and how many experi-monts (and of what kind) were or are being considered? l (c) Please supply all dcx:tunents W11ch describe the core destruct tests l 1 ' l (including internal motoranda, etc.) that may have been considered. SI!T VII AB-67 l
. ANSWERS 42(a)-(c) (a) 'Ihe Applicants have not identified a need for such tests and is not considerirg such tests for support of CRBRP. (b) No such information has been developed by the Applicants. (c) lb such doceents have been developed by the Applicants. QUESTICN 43 Elaborate cm the pessimistic assmptions which nust be made regarding fissicn gas effects, to generate disassenbly. (See P.'RR, p. F6.2-105) . ANSWER 43
'Ihis question appears specific to the imugeneous core Which is not the current design.
Accordingly, it requires no answer. Section 7.2.3 in CRBRN-00523 describes the consequences of pessimistic assmptioins regardirg fissicn gas effects in IN-HCDA of the current design. QUESTICN 44 (PREMELE)
'Ihe PSAR states that modim is still largely in the core in 'IOP cases Irior :
to disassenbly (p. EE.2-105). l l l SET VII AIHi8
\
l QUErrION 44(a) Are there any possible situations, such as lower initial ranp rates, where this may not be true? l ANSWER 44(a)
'Ihe Applicants have not been able to identify a situation in the analysis of a 'IOP disassembly event in which soditn would not be expected to still I
be largely in the core. It nust be enphasized that the failure of each channel must be forced to occur at the core midplane in order to satisfy the ocriditions necessary to begin a hydrodynamic disassernbly calculation. 1 QUESTICE 44(b) l l Explain in detail the insis for the answer in Part (a). ANSWER 44(b) In the 'IOP event sodian voiding results fran fuel-coolant interaction in the failed channels. 'Ibe failed channels still contain a large fraction of liquid sMium since the coolant pumps continue to operate during the transient. This is in contrast to the loss-of-floar event, in which an operatire ptznp head is not available to maintain liquid sodium in the channels. In additicm, the channels which have not failed in the 'IOP event are ocznpletely filled with sodium, dereas in an IM event sodium voiding may occur prior to pin failure. Therefore, in 'IOP events which satisfy the conditions for a hydrodynamic disassembly calculation sodiun is still largely in the core. QUESTICN 45 sh dreau, g al. , are concerned about the possibility of rapid, regional core conpaction during disassably which may lead to secondary criticality SET VII AB-69
l t
. with high ranp rates. Identify and supply the studies, if any, that have . been rede to explore this possibility. (See Boudreau and Erdmann's Ch Autocatalysis, Nuc. Sci. & Eng., 51:206-22; and the Proposal for Chnputer Investigation . . ., mentioned in question 25 above.)
ANSWER 45 It is the Applicants' understanding that Boudreau, et al., were not concerned with rapid regional core ecmpacticn during disassembly, but rather with fuel rempetion following an initial disassembly to cause a recriticality. 'Ihe reentry cases presented in Secticn 11.3 of CRBRP--GEFR-00103 address such a situation as it is hypothesized to occur in CRBRP. Ref. 65 in CRBRP-GEER-00103 by J. F. Jackson et al., addresses the situa-tion as it is hypothesized to occur in FETF. Ebrther analyses are provided in the paper by J. E. Boudreau and J. F. Jackscn, entitled "Recriticiality Cbnsiderations in IMFBR Accidents" Iresented at the Fast Reactor Safety meeting at Beverly Hills, California in 1974. 'Ihe Applicants are not aware of any other relevant analyses addressing the ocncern of Boudreau, except-ing for those referred to by the author in the questicn. QUESTICN 46_ ht effects do the fission gases have cn (possibly) mitigating the power excursicn of a $100 per seccni I& accident (BOEE and ECEC)? l ANSWER 46 Since the CRBRP-GER-00523 and CRBRP-GFR-00103 analysis supports the conclusion that a power excursicm of 100$/sec in the CRERP (BOEC or EDEC iuiupmous design) cr in the current design core does not appear to be . 1 physically realizable, the situation hypothesized is not very sigaificant ' to CRERP. 'Ihe inportance of fission gas in the theoretical Iroblem postulated is not answerable in the general case. '1he initial part (about 60% of the pulse width) of the power pulse frcm a 100$/sec excursicn is SEP VII AB-70
, ccmtrolled solely by the Doppler effect and is unaffected by the gas or , void distribution in the core. 'Ihe energy in the pulse tail is affected by gas, soditan and void distribution. Ibwever, to decide to what extent fissicn gas is a mitigating effect requires knowledge of the specific distributions of soditan, void (gas) and fuel.
QUESTICN 47
'Ihe PSAR asserts that the Irebable course of CRs is self-shutdcun without explosicn due to fuel ejecticn. Please supply for each CDA considered in the PSAR numerical figures that would show how much margin there is between the nest pecbab1e course of a CR and the nest pessimistic course considercd. 'Ihat is, identify those parameters ard variables which control the course of fuel and coolant notion and associated reactivity effects, ard indicate the extent that each parameter and variable would need to be varied for the CR to take the nost pessimistic course considered.
ANSWER 47 Table 2-2 en pages 2-4 through 2-8 of CRBRP-GEFR-00103 and Table 2-1 on pages 2-0 through 2-10 of CRBRP--GEER-00523 surplies the ntanerical figures that show the trargin that exists between the best-estimate course of 'IDP and IN ICDAs and the pessimistic cases considered. The extent to which the particular variable cr variables were varied for the HCDAs to take the pessimistic courses considered is discussed in the sections of CRBRP-GEFR-00103 and OtBRP-GEFR-00523 which Iresent the results of the cases stra-marized in the tables. GJESTICES 48(a) and 48(b) (a) Is sodium hanmer (analogous to water hanmer) possible in an INFBR accident situaticm? (b) If so, has it been evaluated as a cause of fuel failure and fuel noticn in the CRBR? l SET VII AB-71 l
l ANSWERS 48(a) and 48(b) (a) A sodium hanmer Iressure transient may result in the CRBR frm the closure of the check valve in a primary sodim coolant loop following the hypothetical seizure of a sodim coolant pump. 1 I 1 (b) 'Ihe effect of a sodium hanmer pressure transient was analyzed using ; the DEMO code. 'Ihe results are presented in Section 15.3.2.1.2 of the PSAR. 'Ihe analyses irdicated that the most severe situaticn postulated, that of the instantaneous closure of the check valve, would result in a pressure change of less than 1 psi in each of the c,ther primary sodium coolant loops. 'Ihe effect of such a small hydraulic perturbation cn the operaticn of the remaining peps and cn the coolant flow in the reactor is considered inconsequential. 'Iherefore, a sodim hanmer pressure transient resulting frm a check valve closure is predicted to cause no failure of fuel rods in the OBR reactor. QUESTICN 49 Is the possibility of sodim vapor explosion-driven core reccmpaction for autocatalysis ruled out? (See PSAR, p. F6.2-101). (b) Wouldn't core destruct tests be necessary before one could make any firm conclusicn that sodim vapor explosions strong enough to drive a core back to criticality are not possible in CI:As? (c) If not, explain fully on what basis are they unnecessary? (d) Wat are the irdivliual opinions of the various ANL experts (and other experts whose opinions are known by the Applicant) on fuel-coolant inter-actions alx32t the need for core destruct tests? SET VII AB-72
ANSWERS 49(b)-(d)
'Ihis subject is evaluated in Sections 8.2 and 8.3.6 of CRBRP-GEER-00523.
Answers to Parts (b), (c), and (d) are as follows: (b) No. (c) Definitive c:enclusions regarding the possibility for vapor explosion drive fuel cmpacticn can be obtained by carrying out out-of-pile lab-oratory experiments and analyses to determine in a general way the require-ments for vapor explosive events. (d) To the Applicants' best knowledge, none of the ANL experts (and other I experts) believe that oore destruct tests would be necessary. QUESTION 50 Should the value of the gradient at the top of p. M.2-112 (.15254/cm) be
.1525S/cm?
ANSWER 50 See CRBRP-GEFR-00103 for the corrected value. QUESTION 51
'Ihe PSAR notes that there is a positive reactivity feedback due to fuel inploding into a cavity upcn fuel reentry in an EDEC IDF (see Section M .2.6.4.1 of p. M .1-ll2).
(a) Please elaborate by Iroviding a ocmplete analysis of these disasserrbly calculations. (b) Identify how nuch feedback was estinnted. SEP VII AB-73
l (c) Wat was the rate of reactivity feedback? (d) Provide drawing of the changing configuration of the core (elevation and plan view) during this process. (e) Wat fraction of the core or zones was involved in this inplosion, e.g, fraction of core planar area which the imploiing fuel crossed f (horizontal plane through the core)? l t (f) Since transport theory was not used, is there any plan to recalculate this CDA using transport theory? ANSWER 51
'Ihis question appears specific to the h2rogeneous core which is not the current design. Accordingly, it requires no ansver. GEFR00523 describes the analysis for the current core design.
QUEFf1CNS 52 i i QUESTION 52(a) i i Does the design basis of 102 Mj of sodium slug energy for the head bolts include a safety factcr? ANSWER 52(a) No, the~102 M7 is the value calculated explicitly fran the Structural Margin Beyond the Design Base energetics. It has not been arbitrarily I increased to provide an additional safety factor. Sl!T VII AB-74
QUESTION 52(b) mat is the value of this safety factor? That is, at 4at slug energy would the head bolts be expected to fail? ANSWER 52(b) h design of the head restraint meets the requirenents of Section 5.2.2 of CREP-3, Voltrae 1. It is not known at what slug energy level the head restraint would be expected to fail. QUESTIONS 53 (Soditrn Slug Rebound) QUESTICE 53(a) After a violent oore disassenbly event in which the sodium slug slams under the closure head, what consideraticn has bem given to the possibility of the soditun slug rebounding and blasting fuel back into the core region to cause a secondary power excursion by reassembly of enough fuel? ANSWER 53(a) h scenario proposed in this question has not been specifically identified for analysis. Ebwever, Secticn 11.3 in CRBRP-GEFR-00103 discusses the potential for tennination of the initiating phase. 'Ihese recriticality analyses provide results characteristic of those that would be generated by the troposed scenario. 'Ihe results of the VDRE-II analyses for various parameter possibilities is presented in Secticn 11.3 of CRBRP-GEFR-00103. Evaluations in CRBRP-GFR-00523 do not identify any mechanisms for this type of accident. SET VII AB-75
I
~
l QUEErrICN 53(b) t t . Could fuel noving away radically frm the core center rebound and return to meet the down-cming fuel mass? ANSWER 53(b)
'1here is substantially nore fuel dispersal predicted to occur axially than radially in the CRBRP core disruptive analyses ard therefore the propensity for recriticality to occur frm axial reempaction is greater than for radial reempacticn. Since the analyses in Section 11.3 of CRBRP-GEFR-00103 are done with ramp rate as a parameter it is possible to equate these ranp rates to reempaction frm cmbined directions. We CRBRP-GEFR-00103 evaluation was empared to axial recompaction only because it is judged that if significant reempacticn is to occur it would be far nore likely to cme from the axial direction.
QUmrICN 53(c) Identify and supply any analyses that have been done to explore the rapid fuel moticn after soidtzn slug hnpact. ANSWER 53(c) he only analysis that the Applicants have done that is appropriate to explore rapid fuel moticn after soditan slug impact is presented in Section 11.3 of CRBRP-GSFR-00103 and in Section 9 of CRBRP-GEFR-00523. QUESTION 54 Wat are the planned " Safety Test Facilities" for fast reactor transient testing mentioned in the NIC's Reactor Safety Research F%umu (NUREG-75/058, p. 41), and how do they fit into the decision-making process for the CRBR? SET VII AB-76
t ANSWER 54
'lhe only Iroject fitting the description of " Safety Test Facilities" for
" Fast Reactor, transient in-reactor tests" known by the Applicant to be !
planned is the TREAT Upgrade, now in3er way at ANL. '1he Applicant does rot ! believe that sudt facilities are necessary to reach a decision on CRBRP.
'Ihe Applicants tnSerstand that there were studies conducted by NRC relating ,
to " Safety Test Facilities." However, the Applicants are not aware that there are any finn plans by NRC to construct such facilities. 'Ihe Ap- [ Plicants cb not believe that such facilities are necessary to reach a i decision cm CRBRP. i
+
P l I l l l i r I 6 r l l l SET VII AB-77
NIN1H INTERROGATOIU SET QUESTICNS (GENERAL) Each of the following questions is to be answered in 6 parts, as follows [Where appropriate, the parts of the question have been restated to reflect the Irotocol for discovery agreed to by Applicants, Staff, and Intervenors NRDC et al.]: (A) Provide the direct answer to the question. (B) Identify all doctrnents ard studies, and the particular parts thereof, relied upon by Applicants, now or in the past, 41ch serve as the basis for the answer. In lieu thereof, at Applicants' option, a copy of such doctrnent and study may be attached to the answer. (C) Identify principal doctrnents and studies, and the particular parts thereof, specifically examined but not cited in B) . In lieu thereof, at Applicants' opticn, a copy of each such doctrnent ard study may be attached to the answer. (D) Identify by name, title and affiliation the primary Applicant em-
' playee(s) or consultant (s) wto provided the answer to the question.
l l l (E) Explain Wether Applicants are Iresently engaged in or intend to engage l in any further research or work which may affect Applicants' answer. This 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 la or l 1 construction pennit hearing cn the CRBR. Failure to provide such an answer means that Applicants cb not intend to rely upcn the existence of any such research at the im or construction permit hearing cn the CRBR. (F) Identify the expert (s), if any, whczn Applicants intend to have testify on the subject natter questioned. State the qualifications of each such
O
~
expert. 'Ihis answer need not be provided until Applicants have identified the expert (s) in questicn or determined that no e.xpert(s) will testify, as lcng as such answer Irovides reasonable notice to Intervenors. ANSWERS (GENERAL)
'Ihe following answers are identical for all interrogatories except where supplementary informaticn is provided in the answers which follow.
(A) See direct answers below tnder heading " ANSWER". (B) The documents which serve as a basis for the Applicants' answer are identified in the responses below. 1 (C) Unless otherwise indicated below in regard to the answers under heading I
" ANSWER (REFERENCES)"; none.
(D) See the attached affidavits. I
\
(E) Except W ere otherwise noted below, the Applicants' program of further research work is described in Secticn 1.5 of the PSAR. l i (F) At the present time the Applicants have not determined the experts, if any, whan they intend to have testify cm the subject matter questioned. PART A: IhriuuCGA'IORIES RIIATED 'IO (ORIGINAL) CINIHTTICN NO. 8 [NOW CINTENTICE NO.113 l l QUESTICE I On page 12.1-2 of the PSAR, the Applicant states that personnel exposure in routinely occupied restricted areas will be limited to approximately 1/10 of the limits of 10 CFR 20. Ch page 12.1-3, the Applicant indicates that [ ..
there will be other zones in the plant where dose rates may range fran 2 mren/hr to 100 mren/hr. Etw has it been determined for each of these zones that the radiation level is AIARA? ANSWER I
'Ihe zoning criteria referred to in this question is pert of the overall CRBRP radiaticn protecticn and shieldirg design which will fully meet the intent of Regulatory Guide 8.8, Ikwision 1. It is recognized in this Regula'ory Guide that IWR occupational doses have been below the applicable limits of 10 CFR 20. '1herefore, the intent of Regulatory Guide 8.8 is to "prcmote a more formal approach to keepire doses AIARA, to identify and prcrote continuance of good Iractices, and to Ircmote further improvements where practicable."
'Ihe CRBRP radiation protection design limits the exposure df the individual to 10 CFR 20 occupational limits while keepirg the total man-ren dose to the total staff AIARA. As specifically noted in Regulatory Guide 8.8, "It would be inappropriate to hold the irdividual doses to a fraction of the applicable limit if this resulted in the irradiation of nere people and increased the total man-ren dose."
'Ihe radiation zone for a given cell is determined by its access require-ments. As noted in Secticn 12.1.5 of the PSAR, the anticipated fractional time spent in Zones II and III is 25% and 5%, respectively. 'Ihe expected man-hours of occupancy y area and radiaticn zone is shown on pages Q331.17-1 and 17-2 of the PSAR (Amendnent 6). 'Ihe rmmancy requirements of Zones II ard III represent approximately 24% and 2.5%, respectively, of the total access requirenents, and are therefore consistent with the original design basis.
'Ihe nn-ren dose for required activities within radiation Zones I, II, and III are discussed cm PSAR pages Q331.19-7 (Amen &nent 8) and Q331.21-1 (Amendnent 14). 'Ihese en-rem doses for operations, maintenance within accessible cells, rad ste operations and refueling / fuel handling opera-tions result in an estimated exposure of 58 man-ren per year, or an average SET IX AB-80 t . ..
l i l l
~
I of c. bout 0.4 rern/ year to an individual cm the CRBRP plant staff. Based on data provided in Reference 1, the dose fran these activities is consistent i with that in IMR experience. Wis data shows the overall radiation exposure due to these operations to be V. ARA. 'Ihe radiation exposure in the various radiation zones are not disproportionate nor has any sirgle ! r activity been identified as having an undue fracticn of the allowed dose.
'Ihe CRBRP AIARA review Irogram provides a basis for continual review of the r activities during design and operaticn of the facility. This program is -
discussed in PSAR Q331.1 (Amendment 1) and 0331.3 (Amencinent 20) of the PSAR and 0331.2 and 0331.4 (per NRC response). ANSWER I (REFERENCES) ;
- 1. L. A. Johnson, " Occupational Radiation Exposure at Light Water Cboled '
Power Reactors," NUREG-0323, March 1978. Doctanents used as reference naterial in developing this reply: [
- 1. L. A. Johnscn, " Occupational Radiaticn Exposure at Light Water Cooled f Power Reactors," NUREG-0323, March 1978. ,
I Doctrnents examined during preparaticm of this reply: i
- 1. Pelletier, 01arles A., et al . , "Carpilation and Analysis of Data on f Occupational Radiaticn Exposure Examined at Operating Nuclear Power -
Plants," Atanic Industrial Ebrun, Inc. I i I QUESTICE II f Page 8.8-2 and 8.8-3 of Regulatory Guide 8.8 tabulates specific information [ that should be provided (iterns a through r) at the constructicm permit i stage to ensure that Irovisions have been included to achieve AIARA. With i rwi. to endt itan (a) through (r) separately, precisely how is com-pliance with this guide being inplemented? i i J
- ,< g, ,.--w- - - - , , - , . . , - - - - - , , - - - - n - _- ..~,m -, - - , , - -,w-:, --
ANSWER II
'Ihe rnethod of inplementing the Irovisions of items (a) through (r) of Regulatory Guide 8.8 (Revisicn 1) are gival below:
Item (a) General service and access design criteria are included in overall plant design requirenents. Features specific to irdividual systems are l included in the system design requirernents. Examples of overall plant service and access design criteria are as follows:
- 1. All cx2nponents shall be made readily accessible and rnaintainable with a logical rerreval path defined ard doctanented. Provisions shall be included where Iracticable, for isolating cnnponents to permit continued operation of the plant. Pad-eyes shall be strategically located in radioactive cells for installation of portable shielding or for nounting pipe restraints or ;
tooling. '
- 2. 'Ihe plant design shall be such that ronintenance can be performed with adequate maintenan access for personnel ard for required tools, and with minimization of scaffolding, rigging, and portable shielding required to facilitate the work for both scheduled ard unscheduled events.
- 3. mintenance access for servicing and/or reieval or replacement shall be provided for each canponent that is to be maintained. All systen acrn-ponents shall be designed for raioval and replacement.
- 4. Clearance shall be providal between adjacent ccmponents and structures for personnel access, installation, and operation of tooling, and instal-lation of temporary shieldirx3 Overhead roon shall be provided for equipnent rutoval and replacanent. 'Ibe following represents specific l maintenance envelope requirements: l l
- a. A nmtinal 3'-0" naintensnee clearance space shall be Irovided for all majcr canponents ard piping 24" ard lart3er.
SET IX AB-62
em
- b. For in-service maintenance requirirg cuttirg and rewelding of pipe, access space m2st be Irovided for manual and/or autanatic cutting and weldirg equipnent. Se most restrictive clearance is expected to la for cutting the pipe. Specific access requirements as a function of pipe ram al and axial dimensions have been developed for project use.
l
- 5. A mininun of 7'-0" clearance fran the floor to rny overhead obstruction shall be provided cn all stairs, walkways, and other personnel access ways.
Item (b) The general service and access design criteria for the overall plant requires that all electrical juncticn boxes and instrunent and instrunent junctions shall be external to all normally inaccessible areas cr cells. This criteria excludes the location of these components in high radiation areas which are inaccessible during operation. Ebr example, the flux monitoring instrunentaticn and calibraticn equipnent will be located in the HAA. Provisions have been made to remove nuclear detectors as required fran the reactor cavity through the HAA. Rus, entry to the reactor cavity would not be required for this activity. Itsu (c) Wherever possible nonradioactive plant conpanents are locate in accessible areas as a part of the overall plant design criteria. 'IN maintenance systen design criteria provides for the removal and cleanirg of several najor cagonents such as the PifrS grimary punp, check valves, et al. Se response to PSAR Q331.4 (Amencinent 20) discusses in detail the accessibility and retuvability of the following systems:
- 1. Liquid, Gaseous and Solid Radwaste
- 2. Closure Head Operations
- 3. Refueling and Fuel Handling Systens
- 4. (bntrol Ibd Drive Removal Operations
- 5. Maintenance Work on Large Equipnent l
l Item (d) The overall plant design criteria recognize the inportance of "best" grade conpanents to minimize radiation exposure. Specifically, design-dictated naintenance will be reduced through application of fail-safe features, designating conponents which require little or no preventive SET IX AB-83
maintenance and assigning tolerances which allow for use and wear through-out the equipnent's useful life.
'the specific groble related to valves is well recognized by all systens and discussed in the response to PSAR Question 331.4.
Item (e) The design requirenents for penetrations are discussed in PSAR p. 12.1-5 through 12.1-6. Specific design criteria are part of the overall plant design criteria as follows: Allowable Dose Rate Iccation of Penetration Description Range at Penetration Restricted Area, and Penetration <200 mrem /hr. The cbse Radiaticm Area, Cell radiation peak rate at accessible loca-Accessible, Penetration 9 feet or nere tions frcm normally Nonnally Inacessible above normal inacessible penetrations working surface. shall also be limited Access requires to not exceed the re-special platform requirement stated below, cr lacMer. Restricted Area, and Penetration Factor of 3 greater than Radiaticn Area, Cell radiaticn peak cell or area design dose Accessible, Penetration less than 9 rate (general area dose Normally Accessible feet above rate increase at work normal working location limited to 1.2 surface. times the level without penetrations). Unrestricted Area,
<2 mrun/hr (10 CER 20 Penetration Accessible limit).
or Inaccessible Iten (f) Radioactive sources in accessible areas are controlled as required to meet the PSAR radiaticn zoning criteria. 'Ihe radiaticn zoning criteria is a part of the overall plant design criteria. Were transport of a substantial source through an accessible zone is required, radiation exposure will be controlled by shielding, access restrictions or both. '1he dose rate due to a transient source is limited by the overall plant design criteria to 200 mrem /hr. Any exceptions to this will require specific design and operaticmal features to provide positive protection to per-sonnel. R1rther discussion of inplementing this AIARA feature can be found SET IX AB-84
in response to PSAR Ouestions 0331.6 (Amendment 1) and 0331.4 (Amendnent 20). Item (g) The overall plant design criteria sets the following general requirements:
- 1. Facilities shall be provided for convenient inspection, remwal, and repair or replacernent of reactor internal ecmponents. Provisicri for interim storage of cxrponents to be replaced shall be considered.
l
- 2. Adequate local lay-dcwn space shall be: provided for all equipnent such as shield plugs required for maintenance operations. On-site storage space shall be provided for maintenance equipnent and tooling, spare parts, tenporary shielding, etc.
- 3. Cell liners, drip and splash pans, ard/or other devices shall be provided as ranquired to limit damage caused by scditan leaks and to facili-tate cleanup. Call finishing should provide stooth nonporous surfaces and eliminate hard-to-reach corners ard pockets so as to ease decontamination.
- 4. 'Ihe ventilation system shall be designed to facilitate the flow of potentially contaminated air frcm the less contaminated to the more ccntaminated area, thus minimizing the spread of contamination.
Additional design features being implemented by individual systems design I criteria are discussed in response to PSAR Q331.4 (Amendmnt 20). Itan _(h) 'Dx follovity overall plant design requirernent has been included to -limit the nisnber of tMrainable locations: Liquid containing ' systems and/or ocanpanents shall be designed to facilitate cxmplete drairnge. Ibr conpanents that cannot be ccznpletely drained by normal means, provisions shall be included in the design of the ocanponent
'to perreit use of other liquid removal methods utilizing naintenance equipnent.
1< , SEP IX AB-65
The design of soditan-containing equipnent and/or acrnponents shall minimize crevices and pockets sich make atmplete scditrn rerzwal difficult. This requirenent to limit moditan-containing pockets also serves to limit potential crevices for solids. Iten (i) The design requirenent discussed under itern (h) also serves to permit flushirg. A couplete discussion of the large arnponent cleaning and decontamination systen can be found in response to PSAR 0331.4 (Amendnent 20). These facilities are provided as part of the CRBRP Maintenance Systen. Iten (i) The overall plant design requires that area configurations, gas purges, and differential pressures should be established to assure leakage occurs fran less contaminated to nore contaminated areas. The Nuclear Island HV7C systern has the additional requirenent to limit the spread of airborne radioactive materials, where they may be present within the NI building. Itan (k) A ccrnplete radiation and airborne contamination ntnitoring system with both fixed and remote readouts / alarms has been inplanented in the design. 'Ihis systen is discussed in detail in PSAR Section 12.2.4. Iten (1) CRBRP cells which contain significant heat and radiaticn sources are inerted with Argon or Nitrogen and are cooled by the recirculating gas cooling systen (ROCS). 'Ihe ccznponents requiring maintenance for the cooling systen are located outside the radioactive cells in radiation Zone III as described in PSAR 12.1. The radiation level in these cells is designed to give ALARA radiation exposures under the following conditions:
- 1. 'Ihe RGCS shall be designed for contact maintenance during plant full pcwer operation and normal fuel handling operations.
- 2. Provisions shall be made for fan, blower, and cooler replacanent during plant full-power operation.
SET IX AB-06
. 3. Ventilation is not grovided except as required to acquire cell access during maintenance periods after the cell radiaticn has sufficiently decayed.
Iten (m) The scope and extent of the CRBRP shield is discussed in PSAR Secticn 12.1. The doses have been shcwn to be AIARA (See response to I-1) . ! Men (n) The overall plant design criteria require the following provisions for tenporary shielding:
- 1. Radiation fran soure.es within the cell shall be limited either by retoval (e.g., primary soditra), or by permanent or tenporary local shield-ing as required. !
- 2. All cartponents shall be made readily accessible and maintainable with a logical renoval path defined and doctanented. Provisions shall ba included, where gracticable, for isolating amponents to permit continued operation of the plant. Pad-eyes shall be strategically located in radioactive cells for installation of portable shielding or for nounting pipe restraints or tooling.
- 3. Clearance shall be grovided between adjacent aanponents and structures !
for personnel access, installation and operation of toolirg, and instal- ; lation of tanporary shielding. Overhead rocm shall be provided for equipnent retoval ard replacement. Iten (o) Detailed information cn the radioactive waste shielding require-ments are discussed in PSAR Chapter 11 and 12. Detailed information on the AIARA aspects of the radioactive waste disposal system are given in l response to PSAR Q331.4 (Air.E:rdir d. 20). Item (p) The overall plant design requires that each systen consider the following:
'Ihe logistics of all naintenance operations shall be considered, including the paths all equipnent must follow; the availability, capacity, lift, and l l
l l l l SET IX AB-87
area coverage of handling devices; port and hatch size ard locations, rotating and other special handling requirements; operator stationing with respect to safety and visibility; and requirements for pits or other tarporary storage or transfer areas. Special equipnent shall be identified as required by eadi systen. Examples of renote handling equipnent are given in response to PSAR 0331.4 (Amen &nent 20). Itan (q) The plant radiation Irotection and shielding design source terms are based cn maxinzn expectal failures of fuel elenents and " conservative" analysis. A conplete discussion of these design requirements are provided in Sections 11.1 and 12.1 of the PSAR. Itan (r) The manned access points for sanpling sites are in radiation Zone II. The radiaticn zoning has best found to be AIARA in general (see response to I-1) and included required sampling locations. Additional information cn sanple handling equignent is discussed in response to PSAR Q331.5. Source terms at sanpling stations are discussed in PSAR Section 12.1.
Pages AB-89--AB-93 Deliberately Left Blank AB-89--AB-93
PARP B: INIERIOGMORIES REIATED 'IO (ORIGINM,) CINITNTION NO.14 [NEW CINITNTICN NO. O'l, QUESTICN I
'Ihe Draft EIS on the CRBR (NUREG-0024, p. 10-4) identifies 9 civilian nuclear power facilities that were or are in the process of beirg deocrn-missioned. With respect to the CRBR and each of these 9 facilities, please identify and cmpare at the time of deccrimissioning:
- a. cmpositicn of the reactor vessel, vessel intemals and the concrete shieldirg. In responding to this question, we request that you designate the rmterials for the crrrponents of the reactor, e.g., plate, course and flarge forgings, nozzle forgings, boltings, nuts, support, pipe, weld roi, cladding, etc. By designate we mean list current AS'IM designations, or ASME Section III specifications, includirg chemical cmposition in weight percent of C, Ni, Fe, and Mn.
- b. the Irincipal activation Iroducts (with particular attention to Fe-55, Ni-59, OcHK), and Ni-63) within the reactor vessel arri the vessel inter-nals;
- c. the principal activation Iroducts in the inmediate concrete shielding;
- d. the reactcr shielding exposure in W-days or scme other convenient unit; and >
l
- e. the flux density, the neutron fluence, rwt, to which the reactor vessel and concrete shieldirs has been exposed, with particular attention to the vessel inner surface at the beltline.
SEP IX AB-94
. ANSWER I
'this response deals cnly with those atmponents of the Clinch Rive.r Breeder Reactor Plant (GBRP) which are not designed to be removable. Any activat-ed remwable reactor intemals would be transported to facilities for re-processirg cr storage of radioactive materials at the time of deccm-missioning.
i
'Ihe Applicants are not in possession of the information sought which pertains to the other nine nuclear facilities. Information relating to CRBRP is available to the Applicants and such information will form the l basis fcr the following response.
(a) Table 1 lists the ccm p nents, and the RDT standards (1,2,3) and material types required for the current design of CRBRP. Table 2 gives chemical ocuposition of the steels as specified in the ASME references (4,5,6), along with any additional constraints imposed by RDT standards or specific design requirements. 'Ihe assuned atmposition of the primary concrete shield is given in Table 3. l (b) Table 4 lists the rincipal activation reactions which occur in the GBRP permanent steel conponents (8,9). 'Ihe activity of the nest highly r activated otnponent, the Fixed Radial Shield, and of the Reactor Vessel wall is givet fcr the time of deocmmissioning. 'Ihe buildup of radio-isotopes was calculated based cn the conservative asstmption that the , reactcr will operate for 30 effective full pcwer years (EFPY).
'Ihe activity in pCi/CM 3was calculated based cn an average neutron flux !
spectrun at the inner surfaces of the fixed radial shield and reactor vessel over a two foot high surface caea centered at the radial midplane of the reactor core (i.e., the locaticn of the maximan neutrcn flux level). (c) Table 5 lists the Irincipal activation reacticms Milch occur in the CRMP Primary Carm.id.e Shield (9). 'the activity in pCi/CM 3was calculated based cm a neutron flux spectnzn incident cm the concrete surface at the core midplane. 'Ihis is the locaticn of maxinzn neutron flux arti thus ! l i SET IX AB-95 l I
, . maximtsn neutrcn activaticn in the concrete shield. %e conservative
. asstsrption was nede that the reactor will cperate for 30 effective full power years (EFPY).
(d) The design pwer rating of the GBRP is 975W . We design reactor lifetime is 30 years, with a 75 percent plant capacity. This corresponds 6 to a total reactor power generation of about 8.0 x 10 & days over the in-service life of the CRBRP. (e) The total neutron flux, neutzen energy spectrun and neutron fluence for the inner surface of the CRBRP Reactor Vessel at the core midplane eleva-tion are given in Table 6. %e total neutron flux, neutron energy spec-trun, and neutrcn fluen at the surface of the Primary Concrete Shield at oore midplane elevation are given in Table 7. Fluence levels are based cn a 30 effective full power years of operaticn. QUESirION II Harwood, g al. , in " Activation Products in a Nuclear Reactor," state on page 6: he length of time the reactor vessel nust renain isolated fran the enviumment. depends cn the criteria for safe radiation levels. %e criteria sich has been adopted in several deocmnissionings is contained in 10 CFR Part 20.105(b) (1):
' Radiation levels Wich, if an individual were continuously Iresent in the area, could result in his receiving a dose in excess of two millirens in any cne hour. . . '
Do you agree cr disagree with this assertion? If you disagree, please l amplain the basis fcr the disagreement. SErr IX AB-%
ANSWER II
'the Applicants tnderstand that the quoted criteria have geviously been applied as a guideline for deccrimissioning. It should be noted that pursuant to 10 C.F.R. 50.82, decatmissioning can be undertaken upon Cbmtissicn approval of the Applicants' deccmnissioning plan, and that the criteria for decomtissioning would be established pursuant to review and approval of a given deccanissioning plan.
( I' SElP IX AB-97
. 1 I
TABLE 1 MPCERIAL REQUIRDefrS FOR CRBR PERMANENT SIEEL COMPONENTS Otmpcment Product Ebrm RDP Standard
- Grade or Type _
Reactor Vessel Plate M 5-1 304 Plate M 5-1 316 Ebrging M 2-2 F-304 Suppressor Plate Plate (ASME SA-240) 316 Guard Vessel Plate (As4E SA-240) 304 Core Former Plate M 5-1 304 Structure Plate M 5-1 316 Ebrgire M 2-2 F-316 Cbre Support Plate M 5-1 304 Structure Ebrging M 2-2 F-304 Pbrging M 2-4 F-8 Fixed Radial Plate M 5-1 316 Shield Horizontal Baffle Plate M 5-1 316 Bypass Flow Ebrgirg M 2-2 F-304 Module Ebrging M 2-4 F-8
*RDF M 5-1 is A9E SA-240 with additicmal requirenents I3)
RDF M 2-2 is A9E SA-182 with additional requirenents(1) RDF M 2-4 is A9E SA-336 with additional requirenents(2) l ser IX AB-98
l l TABLE 2 OB4ICAL COWOSITICN OF MNTERIAIS REQUIRED EDR CRBR PEIMWENT STEEL COMPONENTS Capositicn (weight %) SA-240(6) SA-240(6) SA-182 I4) SA-182 I4) SA-336(5) , Element Type 304 Type 316 Type F-304 Type F-316 Type F-8 Carbon 0.08 0.08 0.08 0.08 0.08 (max) Manganese 2.00 2.00 2.00 2.00 2.00 (max) Phosphorus 0.045 0.045 0.040 0.040 0.040 (max) Sulfur 0.030 0.030 0.030 0.030 0.030 (max) Silicxm 1.00 1.00 1.00 1.00 1.00 (max) Nickel 8.00-10.50 10.00-1400 8.00-11.00 10.00-1400 8.00-11.00 Otrcmium 18.00-20.00 16.00-18.00 18.00-20.00 16.00-18.00 18.00-20.00 Malybdenn 2.00-3.00 2.00-3.00 1),2) n umbim (max) 0.02 0.02 0.05 o.os o.05 Tantalm (Max) Titanium2 ) 0.05 o.os o.os 0.05 0.05 (max) Cobalt3 ) 0.10 0.10 0.10 0.10 0.10 (max)
- 1) Colunbium and Tantalm are cxmtrolled together
- 2) Agplicable to welded austenitic stainless steel items Which may be '
subjected to service at temperatures over 800 ?.
- 3) Applicable to core former structure, core support structure, and bypass flow module steels in contact with the soditun pool.
SEP IX AB-99
TABLE 3 CHIMICAL 00MPOSITIGJ OF 'IHE CRBRP PRIMMU SHIELD Camposition (weight %) Element Ordinary Concrete (Portland) Iron 1.22 Hydrogen 0.56 Oxygen . 49.83 Magnesim 0.24 1 Calcium 8.26 Solim 1.71 Silicon 31.58 Altunintan 4.56 Sulfur O.12 Potassiun 1.92 SEP IX AB-LOO
, 'IABII 4 PRINCIPAL ACTIVATION PRODUCTS IN 'IHE CRBRP PERMANENT SIEEL COMPCNDCS I
Activity
- Activity
- InPixedRadiag)
(pci/m Shield In Reactor yessel Activaticn Reaction (pci/m ) Cr O(n,y) Cr51 9.8 x 10 4 8.7 x 10 3 , fen (n,a) Cr51 5.1 x 10 0 1.5 x 10-3 Fe (n,p) MnN 1.2 x 10 2 3.2 x 10 -2 255(n,2n) MnN 1.3 x 10
-1 -5 2.2 x 10 ;
Fe (n,N)Fe59 3.7 x 10 3 2.5 x 10 3 , Cb59(n,p) Fe 9 1.5 x 10 -2 2.6 x 10 Ni (n,a) Fe59 1.1 x 10 -3 1.3 x 10 ~7 l Ni (n,p) 00 5.1 x 10 2 9.2 x 10 -2 Co (n,2n) CoS8 5.2 x 10-3
~
- 8. x 10 CoS9(n,y) 0060 2.1 x 10 5 8 9 x 10 3 ;
Ni60(n,p) Co60 4.4 x 10 0 7.7 x 10 Ta181(n,y) Ta182 7.4 x 10 4 5.0 x 10 3 Fe (n,y) Fe 5.1 x 10 5 1.2 x 10 4 > NiS8(n,y) NiS9 3.1 x 10 0 1 5.7 x 10 Nib 2(n,y) Ni6 5.9 x 10 3 4.2 x 10 2 , 293(n,y) Nb 0 4.5 x 10 7.3 x 10 -2
*After 30 effective full power years (EFPY) of reactor operation and averaged over a two foot hicjh section of the inner surface of each u, weit at radial midplane.
t l l SEP IX AB-101
. 'mHLE 5 PRINCIPAL ACTIVATION PRODUCTS IN '1HE CRBRP PRIMARY 3NCRET. E SHIELD r Maxinun Activity
- rete Shield Activaticn Reaction In 'Ihe p)
( Ci/an 01 (n,a) C14 2.0 x 10
-3 Na23(n,y) Na24 7.3 x 10 1 '
Mg26(n,y) Mg27 -2 9.3 x 10 A127(n,y) A128 1 6.9 x 10 Si30(n,y) Si31 6.5 x 10 0 S 33(n,p) P 8.8 x 10 -5 S (n,y) S35 ,9.5 x 10
-2 S 36(n, y) S37 4.9 x 10 K (n,y) K* 3.0 x 10 g41(n,y) K42 0 6.4 x 10 ,
Ca*(n,y) Ca41 2.9 x 10 -2 Ca*(n,a) Ar37 8.1 x 10 ~1 Ca (n,y) Ca4 6.6 x 10 0 Ca (n,y) Ca47 5.0 x 10 -3 Ca (n,y) Ca4 7.6 x 10 ~1 Fe (n,y) Fe ' l.1 x 10 1 ' Fe (n,y) Fe ~1 2.1 x 10
*After 30 effective full power years (EFPY) of reactor operation at core radial midplane. j SET IX AB-102 '
l i r TABLE 6
. L NEL7f!ON M.1DC AT 'IHE QURP IUCIOR VESSEL DEER SURFACE AT CORE MIDPIANE EIJNATION Flux Fluence
- 2 2 (n/an sec) (n/an)
'Ibtal 4.3 x 10 11 3.1 x 10 20 9 18 Thermal 1.9 x 10 1.3 x 10 .
D0.1 MeV 5.9 x 10 9 4.2 x 10 18 16 E>l.0 MeV 2.2 x 10 1.6 x 10 t
- Fluence is based cn 30 effective full power years (EFPY) of reactor ,
operaticn. i i f I SET IX AB-103
TABLE 7 NEUI10N MIDC AT 'HE CRBRP PRIMARY 00tCRETE SHIELD AT CORE MIDPIANE ELEVATION Flux Fluence * (n/an2 sec) (n/an)2 10
'Ibtal 1.1 x 10 7.8 x 10 18
'Ihermal 9 18 2.5 x 10 1.8 x 10 E>0.1 MeV 1.3 x 10 0 9.4 x 10 11 E>l.0 MeV 4 13 8.6 x 10 6.1 x 10
- Fluence is based cm 30 effective full power years (EFPY) of reactor operaticn.
I 'l l SET IX AB-104
ANSWERS (PARP B) (REFERENCES) nn,, m ts used as reference material in developing this reply:
- 1. RDP M2-2T, " Stainless and Iow Alloy Steel Forgings," Decerter,1974.
- 2. RDF M2-4T, " Alloy Steel Ebrgings," Noverter,1974.
- 3. RDF MS-lT, " Stainless Steel Plate, Sheet, and Strip," Noverter,1974.
- 4. A94E SA-182, " Specification for Ebrged or Ib11ed Alloy-Steel Pipe Flanges, Forged Fittings, and Valves and Parts for High-Temperature Service."
- 5. ASME SA-336, " Specification fbr Alloy Steel Ebrgings for Seamless Drun, Heads, and Other Pressure Vessels."
- 6. ASME SA-249 " Specification for Heat-Resistin3 Chronitzn and Onunitra-Nickel Stainless Steel Plate, Sheet, and Strip for Fusicn - Welded Unfired Pressure Vessels."
- 7. Jaeger, R. G., et al. , " Engineering Ccrnpendium cn Radiation Shielding, Vol. 1," Springer-Verlag, New York, 1968, p. 177.
- 8. HEDEMME-72-135, "Maltigroup Reactor Cross Sections for ETR Applica-ticn," R. B. Kidman.
- 9. Nuclear Data Tables, Vol. All, No. 8-9, " Neutron Activation Cross Sections, Measural and Seniempirical," July,1973.
SET IX AB-105 l l
I . ! TEN 1H INTERROGATORY SET GENERAL QUESTICN Each questian is instructed to be answered in 6 parts, as follows [Where appropriata, the parts of the questicn have been restated to reflect the protocol for discovery agreed to by Applicants, Staff, and Intervenors NRDC et al.]: (A) Provide the direct answer to the question. (B) Identify all documents and studies, and the particular parts thereof, relied upon by Applicants, rru or in the past, which serve as the basis for the anseer. In lieu thereof, at Applicants' optim, a copy of each such document and study may be attached to the answer. (C) Identify principal doctrnents and studies, and the particular parts thereof, examined but not relied upon by Applicants, which pertain to the ' subject matter questiened. In lieu thereof, at Applicants' option, a copy , of each such doctanent and study may be attached to the answer. (D) Identify by name, title and affiliaticn the primary Applicant em- l playee(s) cr consultant (s) who Irovided the answer to the question. (E) 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 only in cases where Applicants intend to rely upon cn cping research not included 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 cb not intend to rely upon the existence of any such research at the IMA or constructicn permit hearing cn the CRBR. (F) Identify the expert (s), if any, dxrn Applicants intend to have testify cm the subject matter questicned. State the qualifications of each such SET X AB-lO6
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 ana e provides reasonable notice to Intervenors. GENERAL ANSWE'RS
'Ihe following responses are identical for all interrogatories with the exception of those instances where additional or supplementary information is provided in the responses to the interrogatories th eselves:
(A) See numbered responses below. (B) The cbetunents which serve as the basis for the Applicants' answer are identified in the appropriate nunbered response and have been or will be made available for inspection and copying. (C) The Applicants have examined an3 evaluated ntanerous doctrnents pertain-ing to the subject atter questioned, however, unless otherwise indicated in the responses below, doctanents and other studies pertaining to the - subject atter have been examined but not relied upon by the Applicants. This does not imply that the Applicants have examined all doctanents in existence drich could pertain to the subject etter questioned. Of the doctanents examined by the Applicants which might pertain to the subject matter questioned, crily that sterial relied upon by the Applicants has bem retained in retrievable form by the Applicants. '1his material is identified in the response to Subpart B. (D) See the attached affidavits. (E) Except Were otherwise noted below, the Applicants' further research work is described in Section 1.5 of the PSAR. (F) At the gesent time, the Applicants have not determined the experts, if any, whan they intend to have testify on the subject matter questioned. i SEP X AB-107
QUESTION I Attached to the letter Van Nort to Boyd (April 30, 1976) was the PT Smmary of the reliability program meeting between the CRBR Project and NRC. On page 2 of the strrmary the following appears which is attributed to R. Denise: (b) A cpal of 1 X 10 events / year for events leading to conse-- quences exceeding 10 GR 100 guidelines is considered acceptable by NBC, In the CRBRP Reliability Program report, dated January 1976, the following appears cm page 7: 1.2.1 Rationale Regulations for licensing of power reactors do not establish proba-bility guidelines which can be used directly in this reliability program, although regulatory doctanents do provide guidance. 'Ihis guidance has been considere$ in establishirg the following ntunerical i reliability criterion: The probability of exceeding 10CFR100 guidelines shall be less , than one chance in one millicn per reactor year. With respect to the cpal of 1 X 10 events / year or the ntynerical reliabil-ity critericn of one chance in one millicn per reactor year, please answer the following questions:
- 1. Wat factual information was utilized to support the inplied contention that this specific goal or critericn is in fact achievable?
(a) In answering this question, please refrain frcun theoretical discussions. We are gnly interested here in factual informaticn - that SET X AB-108
is, documentaticm that a ntanerically amparable reliability goal was proposed and achieved in a project omnparable in cxrnplexity to the CRBRP. (b) In answering this questicn, please cite only factual doctrnented information that is relevant to a Iroject canparable in aanplexity, in being essentially first-of-a-kind and where the criterien was introduced at the design stage and was required to be inplemented during the construction phase and to be effective in the subsequent operaticnal phase. (c) In answering the above question, please specify those portions of the doctanentation that specifically treat the matters of htrnan error and comen node failure during the design, construction and operational phases.
- 2. If the answer to the above question is that there is no such factual information, is factual infonnaticn available relative to a higher prob-ability (10-5, 10 , 10-3) goal or critericm?
(a) In answering this question please be responsive to (a), (b), and (c) of Interzugatory I.l.
- 3. Since cne can often learn fran past mistakes, are there cbetrnented
~ ~
cases wherein a goal or critericn of 10 , 10 , 10 , or 10 was estab-lished but was not realized? , (a) In answering this question be responsive to (a), (b), and (c) of Interrogatory I.l. (b) In answering this question please indicate Irecisely What was learned frun the failure and how it can be utilized in the CRBR project.
- 4. In the event that the above questions are conceived as being too '
restrictive, what other factual information is available to demonstrate that a goal cr criterion of 10 , 10-5, 10 or 10-3 is achievable? l l SET X AB-109
l
. ANSWER I (1-4)
Se Applicant agrees that in the NRC letter to the Applicant dated My 6, 1976, the probability of one chance in a million per year resulting in consequences exceeding 10 CFR 100 Guide Lines is the " safety objective almirg point" rather than a fixed nurrber which must be desnonstrated for a given plant. We factual information which supports the Applicants' position that the CRBRP reliability goal is achievable is the experience that has been gained with nuclear reactor systens and the NRC conclusions regarding achievement of systen reliabilities presented in WASH-1270. Since this goal is only one aspect of the over-all design safety approach for the CRBRP, it is not intended nor necessary that the achievenent of this goal will be confirmed in a rigorous statistical fashion. Werefore, doctanentation that a ntrner-ically atmparable goal had been proposed and achieved in a project ccm-parable in ocuplexity to CRBRP would culy be of acadanic interest unless an overall design safety approach utilizing ocmparable regulations, regulatory guides, and other design constraints, performance criteria, and successful precedents, had also been imposed cn the hypothetical project in question. he operating experience that has been gained on current IMR plants which are ocmparable to the CRBRP in ccrnplexity do provide an indication of the achievability of the criterion for the purpose for which it was intended, i.e., as one aspect of a balanced design safety approach. We CRBRP utilizes concepts and equipnent very similar to those arployed in IMRs. Where differences cb exist by necessity (e.g., scxlitan vs. water coolant) specific attention has been given to those differences to assure they are well understood and accui ulated in the CRBRP design ard factored into the CRBRP reliability program. Experience to date for shutdown systens for IMRs supports the attairunent of an unreliability of less than 10 failures per year for single systens. Een redundant and diverse systems are siployed to accanplish a particular functicm, it is reasonable to expect to be able to achieve net tmreliabilities in the range of 10 to 10 fail-ures per year. , SET X AB-llO
Further, since there are no known instances where a cmmercial reactor incident has led to significa:it inpacts tpon the health and safety of the public, confirmaticn as to whether or not reactor systes can achieve such a pal is nore appropriately derived fr a actual cperating experience rather than by hypothetical data whidt might support or contradict attain-ment. With regard to dwther cr rot a mznerically canparable reliability program has been proposed and achieved in a project cmparable in emplex-ity to the CPIRP, the Applicants agree with NRDC's suggestion that the questions posed by NRDC are somewhat restrictive. 'Ihe Applicants believe that the nore appropriate acrnparisons and analogies should be made with the ocmmercial IMR experience. Ocmnercial IMRs involve projects of similar conplexity to the GBRP and also have dernonstrated a high degree of reli-ability in protectirg the public. 'Ihe ccmnercial reactor experience includes all generations of reactors including those which were at the time "first of a kird." It also includes all types and aspects of htsnan error and ocmmon cause failures. 'Ihe fact that a runnerically amparable reli-ability goal was not set prior to the achievement of their demonstrated high reliability in Irotecting the public (an achievement of an unprec-edented public safety record) shoold not really be the pertinent ccmsideration. Although not directly relied tpon, the United Kingdczn has reached similar conclusions and actions relative to the use of reliability techniques in reactor design. 'Ihe United Kingdczn began work in the area of reliability for reactor plant design and development during tlw early 1960's. Dedicated reliability programs have been utilized in the United Kingdom and methods for inplanentaticn of such prograns have been described in reasonable detail by A. E. Green (the Reliability Assessment of Bnergency Electrical Supplies: Prc-: - -Wgs 1975 Annual Reliability ard Maintainability. Sym-posium - Washington, D. C. Jan.1975), M. C. Pugh ('Ihe Use of Probability Techniques in a Reactor Desip Office - SRS/GR/5-LEAEA) , and others. Although not specifically stated in these references, such Irocedures and techniques have been anployed by the LEAEA in nuclear projects.
'Ihe Prototype Past Reactor (PFR), Iresently operating in the United Kingom., utilized a reliability program developed for the various systerns SI!T X AB-lll
essential to safe operation of that plant, and reliability program elements (similar to those applied cn the GBRP) were applied throughout the design stage. This program involved implementation of reliability requirements in design, assessments of system reliability, and data gathering and applica-ticn. This progrart was designai to carry on through long-term operaticn of the PFR. Ibwever, detailed descriptions of this program are not available l at the present time and no documentation is expected in the near future.
. 'Ihe brief description provided here was obtained by a personal comunica-tion with Mr. John Bourne of the LEAEA, Directorate of Safety and Reli-ability. Both the British and the CRBRP project have recognized the effectiveness of a reliability program ard have integrated similar programs into design.
Reliability methodology and confirmation techniques have evolved over a long period of time cn many and varied programs. Intensive application of reliability techniques in a formal manner began over two decades ago and has matured through applicaticn and developnent in large numbers of mili-tary, space, transportation and industrial programs during this time. 'Ihis has resulted in a vast " storehouse" of tried and proven approaches and methods Wich can be applied to the Iroject with confidence. 'Ihus, this project is drawing cri the vast experience and varied techniques developed on these Irograms to Ircnide additional neans of assuring a safe and reliable power generator. There is more than enough history of application and successful results to provide gnWnce for a well-balanced Irogram of goal-setting and confirmatory analyses and tests. By taking maximum aAantage of this extensive history and the experience gained frcm IMRs, the reliability p vp an cn this project will be an additional factor in assuring that the CRBRP will indeed be safe and reliable. 'Ihe project has and is making a concerted effort to judiciously apply the ordered discipl-ine of Reliability Assurance to CRBRP to assure that these objectives are met.
'Ihe reliability technology referred to above has been applied with varying Qc of discipline to a variety of systems and programs which range frcm simple to ocmplex. Information frce these programs has a direct bearing in advancing the state-of-the-art of the reliability discipline and enhancing SEr X AB-ll2
1 the reliability data base. Itwever, since the CRBRP Project is unaware of any Iroject dtich has ambined the use of a reliability goal, emparable regulations, guides, codes and standards, and successful precedents, direct ntanerical results fran and conclusions reached by reliability Irograms for any other projects would have to be qualified prior to canpariscn with the CRBRP Program to such an extent that they might be of doubtful significance in any evaluaticn of the CRIRP reliability program. GJESTION II (PREAMBLE) On Inge 3 of the Reliability Meeting Sturmary the following appears and is attributed to Dr. Ian Wall: As hunnn errors, test and maintenance activities and omnon node failures are major contributors to systen unavailability, and these are difficult if not inpossible to accurately quantify, attainnent of the goals cannot be demonstrated by analysis. ' Neither can they be denonstrated by test because of the nature of the equipnent arri such rare events would reqaire an unrealistic amount of time to test. With respect to this statement, please answer the following gaestions: OJESTIONS II-l, II-2 ;
- 1. Ibes the Applicant agree with this staterent? '
(a) If not, why not? (b) In answering this question, please reconcile your answer with your answers to Interrogatories I-1 through I-4 above.
- 2. If the answer to II.1 above is yes, then explain how it will be pos-sible to detonstrate that the CRBR design is acceptable.
SEP X AB-ll3
e 4
. l ANSWERS II-1, II-2
_ i I Se Applicants agree that hman errors, test and maintenance activities, ard cormm mode failures are major mntributors to system unavailability in a systs which has been engineered for high availability, such as the CRERP shutdown ard shutdcwn heat renoval systems. %e Applicants agree that mathmatically rigorous quantification of the unavailability contributed by [ these causes is difficult. We Applicants agree that systen testing to : demonstrate achievement of very high availability goals reauires an un- ; realistic amount of time ard equipnent. However, the Applicants do not agree that the attainment of such goals cannot be demonstratal. Attairrnent of such goals for high availability with respect to oczmon mode type failures can be assessed by cmparison with operatirg systen data ard evaluations to determine the degree of design inmunity to ocztman node or htsnan failures. Ctmparison to operating systens shows tha a single system can achieve a given availability in practice when subjected to test, maintenance, or other htsnan errors. Wese data for single systens can be applied to diverse systems by appropriate : cmbinations of these single systens and by taking into account the extent to which the unavailability of the diverse systens deperds upon estimated single system unavailabilities. Careful evaluation of certman rtode failure potential (includirg htstan, maintenance, ard test) provides the basis for determining the degree of indeperdence, for eliminating or minimizing the influence of cmmtn causes through design or procedural changes, ard for determining the relative retoteness of postulated failures. Werefore, i through the use of operating data, evaluation of the design regarding ccmton node failure, careful consideration of htsnan errors, and test and maintenance systens, and assessments based on cunponent failure rates, a high availability can be estimated. Given the use of reliability tecimiques and goals as one aspect of a balanced design safety approach, this attain-ment by qualitative and quantitative argtunents is consistent with the answer to I above. SEP X AB-ll4
l
. i I
! QUESTION II-3 l Is it possible to quantify the probability of deliberate htsnan acts such as sabotage? (a) Is it possible that the Irobability of a deliberate act of sabotage is as large as 10 , 10-3, or 10-2 per year? (b) Is it possible that a deliberate act could Iroduce a CIR ard/or a situaticn wherein 10 CER 100 criteria could be exceeded? (c) If deliberate acts such as sabotage were considered as accident ini-tiating events, would it not be possible to include design ard operational features that could significantly reduce the residual risk of and fran such acts?
- 1. Are such acts being considered with respect to the CRBR and if not, why not?
ii. If the answer to (i) above is yes, explain in detail those specific design and operational features that have been included for this purpose. ANSWER II-3 (a) The Applicants have not fcund any technically credible way to quantify , the probability that an act of sabotage would be atteropted. Ebr the reasons stated in their response to Subparts (b) and (c) below, the Appli- I cants believe that it is highly inyrebeble that an act of sabotage would be successful. Wreover, specific plant design features and the CRBRP physi-cal security givpcuu will be implemented to further reduce the probability of such acts. (b) As described in detail in the Applicants' response to a prior inter-rogatory propounded by NRDC in NRDC's Eighth Set of Interrogatories to the SEP X AB-ll5
. Applicants, it is possible, but highly improbable, that a deliberate act 1 1
could produce a CIR. ' (c) As described in the Applicants' response to the interrogatory refer-enced above, multiple layers of controls and safeguards to preclude such acts have already been inwrporated in the plant design and physical security systems. GJESTICN Iy Cbnsidering your answers to the above interrogatories (I and II) and considering that we are concernM with determining the precisicn (quanti-tatively) with Which the residual risks fran the operation of the CRBR can be determinM, how can it be safely concluded that CDAs can be excluded as DBAs?
- 1. In answering this question please be responsive to the remark attrib-uted to R. Denise cm page 2 of the Sunmary of the Reliability Program meeting:
'Ihe doctmentation Wich NIC has received is basically not ac-ceptable for their audit and review due to the extensive use of engineering judgment.
- 2. In answering this question please be responsive to the following Which appears cn page 6 of the CRBRP Reliability Pispaii report:
'Ihe overall design of the CRBRP is based cn the natural three levels of design which Regulatory uses to evaluate the adequacy of Iroposed nuclear power plants . . . 'Ihe third level provides assurance that the public is protected even in the event of extranely miikely ciretanstances of failures or malfunctions.
I 1 S!!T X AB-116 i
A '
.u i
l - L l With respect to this quotatirai, please also' answer the' followm3 questions: ~ l :. (a) What specific design fectures are included in this third level? w
" ~
-(b) 'Ib what estrenely tnlikely circtetances are they Cirect.ed?
~i. In answering this, please consider year answers to all the above interrtgatories and indicate how unlikely are these circumstances.
ANSWER III It.may be safely concluded thst the hypothetical coro di'ruptive s accidents (FKI%s) can be excluded as a Deslgn Basis Accident. 'Ibe preventive design features included in the GBRP and aLKJaented by reliability Irogram acti-vities renders lEDAs to hypotethical events. As described in the PSAR, the p: eventive , features ' tha'. meet ~ the GBRP Design Criteria and appropriate Federal Regulations ax1 Criteria are: , Provided in accordance with the three levels of safety approach
- Cbnparable to IIR preventive features which have bem shown effective in practice Under continuing scrutiny thrcugh the interact on of reliability and .
engineering to further reduce the likelihood of failure.
'Ihe engineering design (and subsequent steps throucih to final cperation) of these deterministic criteria are augnented in the CRBRP thnxtgh the reli- i 1
ability Irogram to pecreide further assurance of the' sufficiently low l probability of the initiation ,of an ICDA. I It is important to recognize that exclusion of IOh frcm the list of DBAs is not syr. suas with exclusicn fran mnsideration altogether. On the contrary, as is explained several times in the information contained in the Public Record (see e.g. , References 1-2), events beyond the . design base _ have been examined by the Applicants to determine the inpacts- asyriated l m. SEP X AB-ll7 -
l
. l with then and, more particularly, the inpacts associated with HCDAs have ,
- i been evaluated in the CRBRP Aoplicaticm. }
he treatment applied to HCDAs is described in References 1 and 2, sich fom a part of the CRBRP Application. As shown in Reference 2, the postu- : lated off-site doses for HCDAs are not excessive considering the highly [ imprr*mble occurrence of ICDAs. This pinrides the necessary information to ascertain the residual risks associated with events beyond the design f basis. j i
%e Interrogatory requests that the Applicants' answer be responsive to a renark attributed to R. Denise on page 2 of the Stmnary of the Reliability Program meeting. %e renark as quoted in the interrogatory has been taken out of the context in which it was presented. The stunary of Mr. Denise's f renarks is given in Reference 3. he Applicants' understanding of Mr. l J
Denise's position, as derived fran the totality of his renarks, is that the
- Reliability Prc3 ram should be re-oriented so as to place less enphasis on numerical allocation of reliability and more ertphasis an conformance to other elenents of licensing practice, such as design criteria. Quantita-tive reliability studies cn the GBRP are directed toward effecting design ,
improvenents. In crder to correctly identify any areas in the design where ' i improvements may be necessary, realistic reliability models and calcula- ; tions utilizing sensitivity studies are necessary. Very conservative models, calculations, and data can distort the realistic nature of these studies and result in erroneous conclusions. Developing realistic nodels, data, ard asstanptions requires the utilization of sound engineering judg- ; ment. Berefore, the use of qualified engineering judcynent is consistent I with and appropriate for the intents and purposes of the CRBRP Reliability ! Program. i he Interrogatory also quotes statenents fran page 6 of the GBRP Reli- l ability Fispc.u, and poses certain other questions relating thereto. We statements originate fran Section 1.1.2.1 of the PSAR, Were essentially l identical statements are made, and are anplified into a full descripticn of : the design safety approach for GBRP. [ SEP X AB-ll8
Table 1.1-2 of the PSAR illustrates the classification of events into the three various levels of design. 'Ibe detailed classificaticn of a much broader spectrtrn of events is Iresented in "able 15.1.3-2 of the PSAR. Since the PSAR was docketed, the evoluticn of the GBRP design has resulted in the aucynentation of the design by the addition of Margin Beyond the Design Base features. These are reported in References 1 and 2.
'Ihe Margin Beyond the Design Base features are directed at mitigating the consequences of a postulated accident resulting in the melt-through of the reactor vessel. 'Ihe scenario used for purposes of this evaluation is described in Section 3 of Reference 2.
It is the judcynent of the Applicants, concurred in by the Nuclear Regu-latory Conmissim (Reference 4), that events leading to a scenario of this kind are sufficiently inprobable that they need not be included in the list of design basis events for the CRBRP. References for III
- 1. CRBRP-3, Voltrne 1: Ehergetics and Strtdural mrgin Beyond the Design Base
- 2. GBRP-3, Voltrne 2: Assessment of 'Ihermal Margin Beyond the Design Base
- 3. Attachnent to letter, P. S. Van Nort to R. S. Ibyd, "Stranary of Meeting Held Between CRBRP Project and NBC to Discuss GBRP Reliability Program and Related Doctrnentation," April 30, 1976, Docket No. 50-537.
- 4. Ietter, R. P. Denise to L. W. Caffey, May 6,1976, Ibcket No. 50-537.
QUESTION IV If one of the design features specified in III-2(a) above is not a core catcher, precisely how was it exclude 3 as a design feature? SET X AB-119 ; 1
- 1. In answering this, please consider all of the above interrogatories (I, II and III) and indicate how unlikely are the circumstances that would require a core catcher.
ANSWER IV A core catcher is not included as a design feature for CRBRP. 7t is not necessary to include such a device since the anticipated consequences of the postulated event *ich it might otherwise have been argued would require a core catcher, have been shown to be acceptable without the ' inclusion of this devicce. See CRBRP-3, Voltrne 2, Assessment of 'Ihermal Margin Beyord the Design Base. GJESTICN V (PREAMBLE) Questions V relate to July 14, 1976 letter to Ibger Boyd frcm Inchlin Caffey. QUESTICE V(1) Are Applicants avare of the NRC Iractice of establishing the source term for site suitability which could only occur if an accident substantially nore severe than the DBA occurred? ANSWER V(1) Yes. t GESTICN V(2) l If the answer is yes, dat is A;plicants trderstanding of the basis for l that approach? l l SEP X AB-120 l l
ANSWER V(2)
'1he basis for that approach is the requirement for ccupliance with 10 CFR 100.11(a), using the guidance given in the footnote thereto.
QUESTION V(3) M1y do you believe such an approach is not appropriate for the CRBR? In your answer, take into account the relative lack of information ard exper-ience with INFBRs as ampared to IWRs. ANSWER V(3)
'Ihe assumption that GBRP is not in coupliance with 10 CFR 100.11(a) is invalid. '1he CRBRP has been ccmmitted to meet the Federal Regulations, including the requirenents of 10 CFR 100.11(a).
QUESTICN V(4) Provide a copy of the Project's evaluation of the experimental data base for the alternation!1 rocesses p for IMR, HIGR, and INFBR facilities. If no such evaluation exists, describe in detail how the evaluation was done, what data was analyzed, who conducted the analysis, how long it took, and when it es coupleted. I
!A nI the responses that follow, the Applicants asstzne that NRDC meant l
" attenuation" instead of the term "alternaticx)"
l l l l SEP X AB-121
It is presuned that this question relates to the following sentence, which is contained in Mr. Caffey's letter of July 14, 1976:
"Regarding the halogen source term, the Project has evaluated the experimental data base and further empared the attenuation Irocesses for IMR, HIGR and INFBR facilities."
'Ihis is to be interpreted as meaning that the Project has evaluated the experimental data base relative to the halogen source term for INFBRs, and has, in addition, empared the attenuation Irocesses for IMR, HIGR, and INFER facilities. Based cn the precedent established in the HIGR, the provision of a technical case with supporting experimental data should suffice to permit credit for physical attenuation processes. 'Ihe Appli-cants have not and cb not intend to evaluate the data inse for either HIGR or IMR since the detailed experimental support for attenuation in these reactors is not relevant to an INFBR. 'Ihe relevant factor is the precedent for credit where experknentally supported argunents are presented.
'Ihe Project evaluation of the experimental data base relative to the halogen source term for INFBRs is contained in Mr. Caffey's letter to Mr.
Boyd dated March 12, 1976. 'Ihis is now in the Public Record and attention is directed to Sections 2.1.1, 2.1.2, and 2.1.3 of the attachnent to that letter. Further information relating to experimental determination of halogen attenuation in scdiun under ccmditions in which large bubbles of fission gas are released is contained in References V-1 and V-2. QUESI'ICN V(5) Provide a copy of the Project's further emparison of the alternationb processes for IMR, HITR and INFIR facilities. If to sudi ocmparison i exists, describe in detail how the cxmparison was cbne, dat data was analyzed, who ocnducted the empariscm, how long it took, and when it was ocmpleted. SET X AB-122 l
- 1. __-_ _ _ __- _- _ - ___._ _ _.- --
1 I . ANSWER V(5) Q:mparison of IMR, HIGR and UFIR attenuation processes was ccniucted as follows: 1 Ebr IMRs, the guidance given in Reference 3 is that 50% of the i 1 I halogens are to be assumed to be released into the reactor building, and, of this fraction, 50% is to be assumed to be absorbed cnto internal sur-faces of the reactor building or adhere to internal ccmponents. These asstmptions are Iredicated cn a Instulation of IOCA with degraded per-formance of engineered safety features which will result in a substantial reduction in water level within the reactor vessel during the period in which fissicn products are being released. This is in etntrast to the situation for DFBRs, in which there would be no change in soditzn level in the reactor vessel. Pbr HIUR, the attenuation mechanism consists of hold-up of fission products within the fuel particle coatings during adiabatic heat-up (Reference 4) .
'Ihus, it is established that credit may be given to such attenuaticn mechanisms as may exist due to the unique characteristics of a given reactor design.
QUESTICN V(6) Miat is the Project's firm reason to believe that the source term is overly ccnservative and that insufficient credit for halogen alternationb is given? ANSWER V(6)
'!he Project's firm reasons for belief that the source term is overly conservative are stated in Mr. Caffey's letter to Mr. Boyd of March 12, 1976.
SEP X AB-123
l I i
= 1 QUEErrION V(7)
Describe in detail the alternation /1 mechanisms which the Project claims exist and fcr which warrant further credit should be given. ANSWE:R V(7)
'Ihe answer to this question is contained in Sections 2.1.1, 2.1.2, 2.1.3 of Mr. Caffey's letter to Mr. Ik:ryd dated March 12, 1976.
QUESTPION V(8) Provide copies of all the research data upon which the claim for greater alternation /1- mechanisns is based. ANSWE:R V(8)
'Ihe data is that contained in References V-1 and V-2 quoted above, together with that frcm References 4, 6, 7, 9,10, and 12 of Mr. Caffey's March 12 letter.
SET X AB-124
O e e e SET X AB-125
i i QUESTICN V(ll) Explain the Project's reliance upon the general approach and result of the Reactor Safety Studies for Release Categories 1-7 for IMRs as a basis for a ' licensing decision en contairrnent venting in light of the following Ccm-mission Interim General Statement of Policy (39 Fed. Reg. 30964) (August 27, 1974): I Accordingly, it is the interim position of the (%mnission that, pending ocznpleticn and detailed evaluation of the final (Reactor Safety Study) study, including public ccrrment thereon, (1) no changes < in the Ccrimission's safety or environmental regulations pertaining to nuclear power plants are now warranted, (2) the Carmission's exist-ing requirenents should not be relaxed, and (3) the contents of the draft study are not an wuslate basis for licensing decisions. ! ANSWER V(ll) As stated in the letter frczn Mr. Caffey to Mr. R. S. Boyd dated July 14, 1976, "the Project is using the general approach (realistic assessments) and results of the Reactor Safety Studies for release categories 1-7 for IMRs to gauge the consequences associated with core disruptive and core melt accidents." 'Ihe Project (bes not rely cm this as the sole basis for a licensing decisicn. j l l t i SEP X AB-126
. The cited statsnent in the interrogatory must be considered in the light of L
- the remarks of then NRC Olairman W. A. Anders, at the time of the release of the final versicn of the Reactor Safety Sttdy.
"'Ihe Omnission believes that the Reactor Safety Study report provides an objective ard meaningful estimate of the public risks associated with the cperation of present-day light water power reactors in the United States. 'Ihe final report is a soundly based and impressive work. Its overall ocnclusicn is that the risk attached to the operation of nuclear power plants is very low canpared with other natural and man-made risks.
The report reinforces the Omnission's belief that a nuclear power plant designed, constructed and operated in accordance with NBC's canprehensive regulatory requiranents govides adequate protection to public health and safety and the environment. of course, such regulatory requirements must ' be continually reviewed in the light of new knowledge, including that 7 derived fran a vigorous regulatory research program." r
'Ihus, NBC is not using the results of the Reactor Safety Study directly in the licensing of IMRs sin one of the major corx:lusions of NASH-1400 is that the current NRC licensing procedures govide adequate gotection for the general public. Ebever, the Reactor Safety Sttdy results are being utilized indirectly in IMR licensing in areas not previously addressed by licensing procedures and regulations (See Reference V-6). Similarly, it is appropriate that the general approach and results contained in the Reactor Safety Study be utilized as one measure of the conparability of the CRBRP to IMRs. l GJESPIQI V (References)
- 1. 'IC-537 IMEBR Source Term Attenuation by D. R. Dickinson and F. H.
Nantsnaker, Decenber 1975.
- 2. AI-ERIR-13172 Quarterly Technical Progress Report, January Mach 1976, pp. 7-17.
I SE:P X AB-127
. 3. TID 14844 Calculaticm of Distance Factors for Power and Test Reactor . Sites by J. J. DiNunno et al., Nrch 23,1%2.
- 4. NUREG 75/004 Safety Evaluaticn of the Stmnit Power Station, Docket No.
50-450, January 1975, pp. 15-19 through 15-22.
- 5. WMH-1400, " Reactor Safety Sttriy - An Assessment of Accident Risks in U. S. Chnmercial Nuclear Power Plants," Final Report, October 1975.
- 6. Letter frcIn B. C. Rusche, Director, Office of Nuclear Reactor Regula-tion, to J. E. Wrd, Atcznic Industrial Ebrtan, Inc., dated August 13, 1976.
QUESTION VI (GDERAL) Considering your answers to the previous interrogatories in this set, how is it possible to justify the statenent cn page 11 of the CRBRP Reliability Program report: Based on the Ireceding discussion, it is ocncitx3ed that consistent with Nuclear Regulatory requirenents and IMR precedents, a spectrtru of even's (which include floods and earthquakes in excess of the stipulations of the Regulations and Regulatory Guides and aircraft impacts) is not appropriate for inclusion in the design bases of the CRBRP. ANSWER VI (GENERAL) As was stated in the response to the questions in Section III of these interrogatories, exclusion of events fran the CRBRP Desian Base is not the same as exclusion fran consideration altogether. Events of the type noted in the questicm (floods, earthquakes and aircraft inpacts) have been considered in the design. Based cm the principle of ocriparability with light water reactors, the approach adopted has been to establish design bases in precisely the same manner as for light water reactors, and then to test the capability of the design to a ---.. shte some larger event. SET X AB-128
i . Rr exanple, in the case of earthquakes beyond the design base, a consid-erable amount of relevant data was 1 resented at the June 23, 1976, meeting of the Advisory Ocmmittee cn Reactor Safeguards, CRBRP SLWittee. See { e.g. pp. 167 - 186 of the transcript of that meeting. j
'Ihus, material in the public record has already demonstrated the large l margin of reserve capability which exists to acu mulate earthquakes nere severe than those within the design base, and more severe than required for siting at the overwhelming najority of locations within the continental United States.
i With regard to f1 coding, the Irotection provided is by purely conventional nwans, e.g., assurance of flood barriers, placernent of building access above maxinun flood levels, etc. See Section 2.4, PSAR. Cbnsideration of aircraft proximity is given in Section 2.2.2 of the Preliminary Safety Analysis Report, in which it is shown that there are no ecumercial flight i paths within 10 miles of the Clinch River site, and that the nearest airport (Meadowlake, 10 miles away) cnly handles anall sport-type civil aircraft. Cbnsideration of other events beyond the design base is included in the response to questions in Section III. , GJESTICN VI(l) In answering this interrogatory, be responsive to the consideration that this is essentially a first-of-a-kind facility and answer separately for [ each of the cbjectives of the CRBR relative to the INFBR Program as stated cm page B-2 of NURB3-0024:
'Ihe identified role of the CRBRP was stated in the PEES as follows: It is a key element in both the engineering and manufacuring phase and the SET X AB-129 !
l l l utility ocmnitment phase. In its role as an IMFBR denonstraticn plant, the . principal cbjectives of the CRBRP are to: (a) Denonstrate safe, clean and reliable operaticn with high availability in a utility envirorrnent.
)
(b) Ebcus the develognent of systems and crmponents. (c) Develop industrial and utility capabilities to design, c3ntrol, operate and maintain an INFBR. (d) Derrenstrate the licensability of INFBRs. ANSWE:R VI(l) (a) As has been dertenstratM above, there exists considerable capability to protect the public fran the consequences of a range of such events beyond the design base. (b) One element of focusing the developnent of systems and otrnponents is the establishnent of appropriate licensing-related requirements on these systems and crmponents. It is clearly desirable, if possible, to cb this by reference to existing licensing requirements. Ebr that reason, the process adopted has been that of establishing design bases in these areas on the same prenise as is done for light water reactors, and then to evaluate the capability to acocamodate events beyond the design base. 'Ihe denonstraticn that use of design base requirenents in this manner provides a significant level of a$ditional capability is a major step in establish-ing that certain traditional light water reactor licensing requirements can be directly translated into IMFBR licensing requirenents. (c) Apart frcm the design element, which is addressed in (b), such ocn-siderations are not really directly related to end-of-spectrun events. (d) 'Ihe response to this question is contained in (b), above. SEP X AB-130
QUEErfICN VI(2) In answering this interrogatory, explain how, without land use or other regulations, an airport or an FAA airway can be excluded fran the CRBR vicinity or that of future INFBRs. (a) Also explain how sone other ptentially mdesirable neighbor can be excluded. ANSWER VI(2) As indicated above, the nearest crmnercial flight path is 10 miles away fran the CRBRP and the nearest airport is 10 miles away. In additicn, the question of land use planning in the CRBRP has been addressed in question response 310.34 (See Section 2.2.3 of the PSAR). QUESTICE VI(3) In answering this interrogatory, consider the implications relative to the above-quoted objectives of the events recently revealed at the North Anna seismic fault site, the recently revealed magnitude of the fault and seismic events near Diablo Canycn, and of the size of the SSE and OBE at sites such as those along the Pacific (bast. (a) We are concerned here not cnly with the licensability of the CRBRP but also with the demonstration of meeting the overall INFBR Program ciojective of the CRBR. ANSWER VI(3) (bnsistent with the Iractice utilized in the licensing of IMRs, the CRBRP plant desicp will be shown to be conpatible, with appropriate margins, for the site location stipulated in the license application. As discussed in response (1) of this interrogatory, there is significant reserve seismic 1 1 SET X AB-131
capability within the current CRIRP design. 'Ihe existing margin is con-sidered nore than adegaate for siting at the overwhelming najority of locations within the continental United States. QUEErfICN VII (GMERAL)
'Ihe following appears cm page 6 of the CRBRP Reliability Program report:
Due to the lack of precedents for INFBR plants, the CRBRP design approach utilizes reliability techniques extensively to provide a systematic determination of events to be included in the plant design basis. On page 7 of the same report, the following appears:
'Ihe probability objective of one clance in a millicn was selected after a revig)of 6ppuriate nucleargfety orientgliterature
, WASH-1400 , WNSH-1250 g
as NASEI-Ig Cha q q tarr's work , NASH-1285 , WASH-1318 and others. Since the first quotaticn above describes the purpose of the reliability program and the se<:orx1 cites cnly ame of the reliability or safety oriented literature reviewed, please supply us with a more emplete listing of the pertinent literature reviewed by the Applicants. ANSWER VII (GMERAL) The literature determined to be nere pertinent to the stated Project objectives are those referenced in the intrcxiuction of the interrogatory. It should be noted that it is not reasonable to cite all references which ! provide a source of information used to establish a design approach, l program objectives, cr assessment techniques. 'Ibe reference having the j greatest influence cn the specific value chosen was WASH-1270. 'Ihe re- l mainder of the quoted references were supportive in nature, in that they provided ad3itional evidence that the objective chosen was (1) reasonable SEP X AB-132
in relation to assuring adequate protection of the public and (2) achiev-able using proven design concepts. In addition to those references already noted, the following references are typical of the types of literature which were reviewed ard factored into the CRBRP decision-making process as appropriate:
- Farmer, F. R., " Reactor Safety and Sitirx3: A Proposed Risk Criterion,"
Nuclear Safety, Vol. 8, No. 6, Nw.-Dec.1%7.
- Otway, H. S. and Erdmann, R. C. , " Reactor Siting and Design fran a Risk Criterion," Nuclear Engineering Design, 1970. - Godbout, P., " Appendix III, Probabilistic Safety Analysis of a Hypothet-ioal 1000 We Liquid Metal Past Breeder Reactor," Public Health Risks of 'Ihermal Power Plants, Report No. UCIA-E2G-7242, School of Engineering and Applied S:lence, UCIA, my 1972.
U.S. Atanic Energy Cimnissicn, " Nuclear Power Plants: Seismic and Geologic Siting Criteria," No. 10 CPR 100, Federal Register, Vol. 36, No. 228, 1971. '
- Salvatori, R., "Systenatic Approach to Safety Design and Evaluation,"
IEEE Transactions on Nuclear Science, Vol. NS-18, February 1971.
- IEEE, " Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Protection Systen," IEEE Std. 352-1975, 1975. ' - Marcos, A., "'Ihe Ible of Probability in Nuclear Plant Design," (knsulting Engineer, Decernber 1973. - Rasmussen, N., " Reactor Safety: Real Probabilities," Ctnbustion, June '
l t 1974.
- Boeovsky, I., " Reliability T ory and Practice," Prentice Hall, Ihglewood Cliffs, New Jersey,1%1.
1 SET X AB-133
- Green, A. E. ard Bourne, A. J., " Reliability Technology," Wiley Inter-science, New York, New York, 1972.
- Epler, E. P., "Ocrmon 2de Failure Cbnsiderations in the Design of Systens for Protecticn and Control," Nuclear Safety, Vo.10, No.1, Jan.-
Feb. , 1%9. During the five year period since 1977, considerable attention has been given to formalizing ProbabilisLic Risk Assessment (PRA) methods and goals.
'Ihe Applicant has nonitored these efforts and reviewed the doctrnentation produced. 'Ihe objective of the monitoring and review was to ascertain that the Applicants' Reliability Prog 1.m cbjectives and cpals mntinued to be
- current with the evolving industrial practices.
'Ihe following list of doctanents identifies the Irincipal cbetrnentation reviewed durirrj the last five years:
- 1. U.S. Nuclear Regulatory C%2mtission, Safety Goals for Nuclear Power Plants: A Discussion Paper (NUREX3-0000) (Feb 1982);
- 2. Itaclear Regulatory Ckmnission Statement of Risk Assessment in Light of the Risk Assessment Review Grotp Report (Jan. 18, 1979);
- 3. lac Iroposed rule requiring inprovements in reactor design to reduce the risks frtzn anticipated trnasients without scram ( "A' INS") events (46 Fed. _ Reg. 57521) (Nov. 24, 1981);
- 4. Risk Maessment Review Group Report to the U.S. Nuclear Regulatory Ccranissim (NURE!3/CR-0400);
- 5. Ssain, A.D. , A.G. Guttmann, "Har t v4 of lisnan Reliability Analysis with Bnphasis cn 1&aclear Power Plant Applications," Draft Report (NJREI3/
CR-1278) (Oct. 1980); SET X AB-134
i
- 6. PrWing of the International ANS/ ENS 'Ibpical Meeting cn Probabilis-tic Risk Analysis (Sept. 20-24, 1981) Port Chester, NY; !
- 7. U.S. Nuclear Regulatory Ommission, NRC Action Plan Developed as a i Result of the 'IMI-2 Accident (NURED-0660) (Aug 1980); l
- 8. " Licensing Requirements for Pending Applications for Cbnstruction l Pemits and Manufacturing License. (Mar 1981) (NURED-0718) (Rev. July 10, i 1981); j r
s
- 9. PRA Procedures Guide, A guide to the performance of Probabilistic Risk Assessments for Nuclear Power Plants (Review Draft) (1982) (NUREn-2300, l Rev. 1). ;
QUESTION VII(l) I i i Has the Applicant reviewed any literature that is skeptical or critical of reliability assessments or estimates such as those presented in NASH-14007 ANSWE:R VII(l) i a
'Ihe Applicants are aware of viewpoints that differ frcm the cxmsensus of opinicn regarding the applicaticn of reliability technology in general and the assessments Iresented in WASH-1400 in particular. .
I i With regarti to WASH-1400, the Applicants are aware of specific skepticisms ; or criticisms Which have been published. Typical of these are cxmnents by . l the Union of Concerned Scientists, National Intervenors, Californians for Safe Nuclear Energy, Joel Yellin and William Bryan. 'Ihese omments were h considered in light of IRR experience, reliability experience with other programs, state-of-the-art of reliability technology, a detailed review of WASH-1400 arri its inplications, and a mntinuing interacticn with personnel I involved in the Ireparation of that report. In those cases Were the concerns reflected constructive criticisms, they have been appropriately ' SET X AB-135
considered in their application to the CRBRP Reliability Pr@ tan. Many of the criticisms, however, were found to be without merit.
%e more recent literature which the Applicants have reviewed is overwhelm-ingly in support of the application of reliability methods.
QUESTION VIII l Mr. William M. Bryan, former manager of reliability studies in the NERVA and Apollo pr@tas, has made the following criticism of WASH-1400 (a copy of his entire report is attached; the section quoted below is found on page 4): As an example, the Apello 4th stage rocket engine had an assigned reliability of .999 (o- 1 failure allowed per 10,000 missions). This assigned value was derived in the early 1%0's quite similarly to the quantitative values calculated in the WASH 1400 report. '1he highest reliability estimate achieved by this engire after thousands of actual tests was approximately 0.% (or 4 failures per 100 missions). hus, in this case an error er uncertainty factor of 400 existed between the predicted and actual reliability. Use of this type of ' reliability assigment (predicticn) techniques consistently led to such an overstatment of reliability, which was one of the main reasons these techniques were abandoned by NASA. By otrnparison to a nuclear power plant, this engine was a very simple system which t represented an off-the-shelf technolocjy. Many of the failures that occurred were due to htrnan errors during manufacture and essernbly, errors diich got through tmnoticed despite the nest sophisticated quality control procedures ever utilized to discover ard prevent them (quality control procedures and funding considerably superior to those currently used in nuclear power plant construction). Many others were due to design mistakes or lack of anwledge. None were planned. Each time the systs was tested, the engineers ard analysts expected it to fully succeed, much the same as the technical optimism that is evident in WASH 1400. With respect to this statement by Mr. Bryan, please answer the following questions:
- 1. Ibes the Applicant agree with the statement? If not, why not?
j SET X AB-136
l
- 2. (bnsidering that one of the major objectives of the Apollo program was the safety of the astronauts, does the Applicant consider that the incin-eration of cnly 3 astronauts cn the pad at Chpe Kennedy is representative of an adequate program? If not, why not?
ANSER VIII
'Ihe cantnents referred to in this interrogatory are part of an overall set of ccmnents cn WASH-1400 by Mr. Bryan. 'Ihe principal point of these ccmnents is that the analytical techniques reed in the study lead to optimistic conclusions. It was further asserted that the National Aero-nautics and Space Administration (NASA) and the Aerospace Industry abandoned the use of these techniques for this reason.
In arriving at a decision as to the validity of these ocmnents, it is instructive to review NASA's own ecmnents on the stuiy's methodology.
'Iheir ocmnents are contained in a letter of June 16, 1975, frcm the Admin-istrator of NASA to the Chairman of the U.S. Nuclear Regulatory Ccmnission (a copy of which is grovided as Attachnent 1). Briefly, the NASA letter states that the techniques used in the study are effective and are capable of producing runerical assessments of value if the data base fran which failure probabilities are determined has sufficient accuracy and content in light of the quantification being performed.
Additional insight into this question is provided by A. E. Green, Manager of the Systems Reliability Service (SRS) operated by the United Kingdom Atanic Energy Authority. A copy of a letter fran Mr. Green to the Project Staff Director of the WMiH-1400 study is also included as Attachment 2. Mr. Green states that where they have applied quantitative reliability techniques for Irediction, there has been reasonable agreement with field experience when it became known, generally within a factor of two. In support of this realistic grediction capability a graph fran Reliability
'I%chnology is provided as Attachnent 3, which shows the close agreement the SRS group has experienced between gredicted and observed systen failure rates fcr scme 50 systen elenents.
l SEP X AB-137 l l
An objective review of these ard other ccmnents leads to the conviction
~
that the analytical techniques used in WPSI-1400 were current, applied appropriately, and can serve to provide realistic system failure predicti.ons.
'Ibe foregoing is rot intended as a justification of WPGI-1400, rather it is an indication of the awareness of the Applicants of the state-of-the-art of reliability technology and the omviction that when appropriately applied it forms an effective part of the CRBRP developnent process.
i I l l SET X AB-138 l
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'1HIRFEENIH INTERROGA'IORY SET PRENHE 'IO QUEETTICES
'Ihere are several INBR safety systes known to us that are currently being discussed in the literature at react r safety meetings, etc., but that are not Iresently incorporated in the OtBR design. 'Ihese include (a) the use of a third shut-down syste that is self-actuated and that does not rely on any instrinnentation, e.g., using curie point magnet as a release mechanism, l
(b) the " parfait" core arrangment, i.e., interspacing the core and blanket materials; (c) use of heavy flywheels on the Irimary soditan ptanps; (d) designirx3 the reactor to operate at a lower soditan tmperature, thereby sacrificing efficiency for safety, and (e) the use of a core catcher arx1 other desigri alternatives presently being considered but not incorporated in the Iresent CRBR design.
'Ihe third self-actuated shutdown systen was discussed by Deetrious L.
Basdekas of the NBC Staff at the Decenber 3, 1976, ACRS meeting (Trans-cripts, pp. 233-241 and attachments). 'Ihe " parfait" core was discussed at the Internatic.nal Meeting on Fast Reactor Safety and Related Physics in Chicago, October 3-8,1976. 'Ihe use of flywheels en the soditan prnps is utilized in the Phenix to delay onset of a CIR in the IM scenario. 'Ihe use of a lower sodiun operating terrperature has been sq M to ERIA personnel by Hans Bethe. 'Ihe core catcher is well known to the Applicants / Staff and is incorporated in the German IMIR demo design. We are interested in i identifying all such safety systems that are not presently ira.wpurated in the CRBR design, obtaining complete documentation on these systems, determining the effects and safety advantages that these systes would have if incorporated into the OtBR design, incitzif ng the inplication on criteria related to site suitability, CIA probability and energetics and, finally, the effects m cost and schedule if these systens were h to the CRBR. SEP XIII AB-139
Each of the following questions is to be answered in six parts, as follows [Where appropriate, the parts of the question have been restated to reflect the grotocol for discovery agreed to by Applicants, Staff, and Intervenors NRDC et al.]: (a) Provide the direct answer to the question. (b) Identify all doctrnents and studies, and the particular parts thereof, relied upon by the Applicants, now or in the past, which serve as the basis for the answer. In lieu thereof, at Applicants' option, a copy of such doctrnent and study may be attached to the answer. (c) Identify principal doctrnents ard studies, and the particular parts thereof, specifically examined but not cited in (b) . In lieu thereof, at Applicants' opticm, a copy of each such doctrnent and sttdy may be attached to the answer. (d) Identify by name, title and affiliation the primary Applicant anployee(s) or ccnsultant(s) who provided the answer to the question. (e) Explain diether the Applicants are presently engaged in or intend to engage in any further research or work which may affect the Applicants' answer. 'Ihis answer need be provided cnly in cases where Applicants intend to rely upon cm going research not included in Secticn 1.5 of the PSAR at the DR or construction 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 permit hearing on the CRBR. l (*) Identify the expert (s), if any, whcm the Applicants / Staff interd to have testify cn the stibject 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. i SEI' XIII AB-140
QUESPION 1 Please identify all safety systes, materials, concepts and significant design alternatives, including those discussed above that are not presently incorporated in the GBR design but that could ocmceivably inpact on the probabi}ity or energetics of a GBR CDA. - ANSWER 1 he safety systes, materials, and design concepts incorporated into the CRBRP design that have an impact on the probability of CDAs are discussed in the following PSAR Sections: 1.6 (Reference 10), 3.1, 3.2, 3.8, 4.1, 4.2, 4.3, 4.4, 5.2, 5.3, 5.4, 5.5, 5.6, 7.1, 7.2, 7.3, 7.4, 7.5, 7.6, 8.2, 10.4, 15.1, 15.2, 15.4. (a) % ird Shutdown System P We Applicants have incorporated two shutdown systems in the reactor . I shutdown system to ensure that all postulated off-normal occurrences are terminated without initiation of an HCDA. Any new shutdown systems would be less certain ard less tested. There is no reliability data for them, as there are for the existing Irimary and secondary shutdown systems. All identified failure modes are addressed by the primary and secondary shutdown systems; a new syste does not address any other failure nodes. Based cm the guidance of WASH-1270 (A'IWS), the May 6,1976, letter fran R. , Denise to L. W. Caffey, and Applicants' assessments, the Applicants do not believe that the additicn of a third shutdown syste would have a signifi-cant inpact an further reducing the Irobability of an HCDA, and therefore, the third system was not included in the GIRP design. (b) Parfait Core Concept he Applicants have not conducted a detailed analysis of the Imrticular
" parfait" core vacwL as studied at MIT ard discussed at the ANS Inter-national Meeting cm Fast Reactor Safety and Related Physics. Many " Parfait" SEP XIII AB-141
and " Radial Parfait" or heterogeneous concepts have been postulated to enhance breeding characteristics of IJEBRs. Heterogeneous core concepts were evaluated by the Applicants. Che suited to the CRBRP design objective was adopted and is described in the PSAR as the current reference design. l Se preliminary indications are that the energetics potential for the initiatirr3 phase of the ICDA may be even less with the heterogeneous design than with the earlier lu+eous fuel enagement schema. We heterogeneous core is described in PSAR Secticn 4. %e reference fuel management schane has been extensively analyzed by the Applicants and has been shown to have negligible potential for energetics resulting fran a postulated HCDA (See CRBRP-GEER-00523 ) . Since the design of the systems which grevent initiation of HCDAs did ret materially change for the heterogeneous core design, there is a negligible difference in the Ircbability of initiation of an HCDA occurring with either fuel management scheme. %e previous high degree of redundancy and diversity in the CRBRP design is retained with the heterogeneous fuel management concept. (c) Heavy Flywheels he CRBRP primary flow coastdown cinaracteristics have been selected by balancing two requirements, which are:
- 1. 'Ib provide adequate coolant flow to the core and radial blanket for all design basis events including postulated loss of power to all the three primary pianps.
- 2. 'Ib minimize the thermal transients associated with reactor and plant trips.
Se required flow coastdown characteristics will be provided for the CR4e by building sufficient inertia directly into the pt:1p drive motor rotor such that the nonentum of the pianp-drive notor asserrbly will be available for these purposes. This inertia satisfies both of the abcne requirenents SEP XIII AB-142
i
]
i and obviates the need for the addition of a separate flywheel for these ! purposes. 6 Iower probability events which are beyond the design base have also been i cmsidered. We probability or the resultant scenario of the postulated transient overpywer ('IDP) events, which assume failure of both reactor , shutdown systens, would not be changed by rrodified primary ptrip inertia since no ptunp trip is involved. %e effect of increased Irimary ptrnp ; inertia for the postulated loss of flow (IM) events, which also asstrne failure of both reactor shutdown systems, would be to slightly change the i thne scale for the events but not the overall conclusions. We time for i initiation of boiling might increase, but cnce boiling is initiated, the sequence of events would be controlled by the phentrnencn related to boiling, such as void reactivity insertion, cladding dryout, fuel failure i and post-failure fuel and cladding moticn. Rese phencunena lead to a prediction of event termination in a non-energetic rranner. Increased pinnp inertia would not change the probability of soditan boiling ard the re- ! sultant consequences. Sus, increased planp inertia would be ineffective in significantly impacting the probability or consequences of a IM m, and,
- therefore, was not included in the CRBRP design.
(d) Operation at Iower Tenperature [ he operating terrperatures of the plant have been selected based upon [ detailed plant performance analyses which evaluated a variety of factors, including operation of the plant at high thermal efficiency consistent with : fuel breeding goals and material capabilities. Secticn 15 of the PSAR,
" Accident Analysis," clearly derronstrates that the selected plant operating }
conditions are acceptable. t i
%e effect of choosing a lower plant operating tertperatures would not i significantly change the ' IIP M consequences because the current 'IOP scenario results in molten fuel release fran the pin before coolant boiling l' occurs. Thus, the overall conclusions regarding the 'IOP m would nc,t be '
influenced by a choice of lower cperating terrperature. SEF XIII AB-143
Se effect of lower operating taperatures cn the Irobability and con-sequences of a IN HCIA is sintilar to that described in (c) above. We time to initiate boiling would be slightly increased, but the p,ubability or consequences of soditrn boilirg would not change. R us, a lower operat-ing tanperature would not significantly inpact the Irobability or con-sequences of a IN ICIA. ! (e) Core Catcher l l We core catcher would not have any inpact cn the probability or energetics of an ICIA; rather the presence or absence of a core catcher would impact upan the level of consequences to occur in the unlikely event of an HCDA with core melt. he Applicants' current assessment entitled Wird Invel hennal Margin (TUIM) indicates that the current plant design, without the core catcher, provides acceptable protection to the public health and safety. QUESTION 2 (PREMELE) Questions 2(a) through 2(f) pertain to each of the systems, etc. Identi-fled in the response to Question 1. QUESTION 2(a) Please identify and supply emplete and current doctanentation of these systans, materials, cucw and design alternatives. ANSER 2(a) Refer to the PSAR sections and documentation entunerated in the response to Questian 1 above. SET XIII AB-144
QUESTICE 2(b) Describe to the fullest extent possible and quantify where possible the Impact ea6 would have cn the probability and/or energetics of CIms if each was inc3rporated in the CRBR design. ANSWER 2(b) See the response to Question 1. QUESTION 2(c) M1at effect in terms of CRBR cost and scheduling would be felt if it was incorporated in the CRBR design? ANSWER 2(c)
'Ibere is to inpact cn CRBRP cost and scheduling for those safety systems, materials, and desicp1 concepts already incorporated into the design of the CRBkP as discussed in the PSAR. 'Ihe Applicants have not performed detailed cost and schedule analyses to assess the impact of those additional design features identified in the background provided with this interrogatory.
OLESTION 2(d) ht was the basis for the Applicants determination that it should be excluded frcm the C3tBR design?
. ANSWER 2(d) h additional design features Iroposed by this interrogatory would not provi& significant additicnal protection for the health and safety of the l public nor are they necessary to meet NIC criteria inclniing those set SEP XIII AB-145
forth in the NRC my 6,1976, letter. hose features that are necessary
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are discussed in the PSAR sections identified above. QJESTION 2(e)
'Ib diat extent do the Applicants' conclusions regarding (i) CDA energetics, (ii) probability of CDA, (iii) adequacy of the current design to cope with CDAs with respect to their Irevention or mitigation, or (iv) the adequacy of the NIC criteria set forth in the May 6,1976, letter by Denise impact on whether it should be included or excluded fram the CRBR design?
ANSWER 2(e) he Applicants' conclusions regarding CDA energetics, the probability of a CDA, and the adequacy of the current design to cope with CDAs with respect to their Irevention or mitigation and the inpact diich these features had cn the AIplicants' decision to exclude these additional design features is presented in 2(a) through 2(d) above. %e CRBRP design features specified in the PSAR are in conpliance with the NRC criteria set forth in the my 6, 1976 letter by Denise. OJEStrION 2(f) : If the Applicants' conclusions regarding items (i) through (iv) in Question 2(e) above were substantially found to be in error and non-conservative, how large an error or What change in criteria would be required before the Applicants would likely cocritar incorporating it in the CRBR design? ANS6ER 2(f) Se Applicants have Irovided adequate nargins in the QURP design to provide protecticn for the health and safety of the public, as discussed in the PSAR and the hird Invel hermal mrgin Report. %ese margins also SEP XIII AB-146
provide sufficient grotection against tocertainties Weh might exist in the data inse. Em irmpretion of the additional features mentioned above into the QURP design would not Irovide significant additional protection to the public health and safety for the raamerus stated in response to Question 1. I 1 l l ( l SEP XIII AB-147
UNITED STATES OF AMERICA 4 WUCLEAR REGULATORY COMMISSION In the Matter of .
) .
DEPARTMENT OF ENERGY ) - DOCKET NO. 50-537 ' PROJECT MANAGEMENT CORPORATION. -
) -
TENNESSEE VALLEY AUTHORITY -
) .
4 AFFIDAVIT OF - PAUL W. DTCKnOM. JR . _ _ baing duly sworn, deposes and says as follows:
- 1. That he is employed by Westin4 house Electric Corporhtion Cs Technical Director, _ Clinch River B_reeder Reactor Project, Westinghouse Advanced Reactors Division, Post Office Box W, Oak Ridge, 'renne s s e.e _
37830 2 That he 1s duly authorized to answer the Interrogatories numbered 5,2,38-21,33_,48,52, and 54 - in NRDC's Seven th set of Interrogatories, and all Interrogatories except V (9.30) -- - in NRDC's Tenth set of .
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Interrogatories. .
- 3. That the above-mentioned and. attached answers are trde,and l correct to the best of his knowledge and belief. -
1 . f J V&? , A ~ SIGNA'ERE - . SUBSCRIBE 0 and SWORN to before ne this f.8 day of l#/ , 1982 .
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g' . fyp) y My Commiss!on Expires April 28.1984 . Notary 1Public 9019-9E9 Sid 3DOIM XWO dM8M3 1ND 11:03 083/Wb0 My c ==iision muniram 18
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, l UNITED STATES OF AERICA NUCLEAR REGULATORY ComISSION In the matter of )
INIITED STATES DEPARTENT OF EMERGY) Docket No. 50-537 PROJECT MANAGEMENT CORPORATION )
. TENNESSEE VALLEY AUTHORITY )
AFFIDAVIT OF RICHARD KEPPLE DISNEY Richard K. Disney, being duly sworn, deposes and says as follows:
- 1. That he is employed by, Westinghouse Electric Corporation Advanced Reactors Division P.O. Box 158 Madison, Pennsylvania 15663 as Manager, Shielding Analysis.
- 2. That he is duly authorized to answer the Interrogatories nunt>ered I and 11 for (Old) Contention 8 in NRDC's 9th set of Interrogatories, and I and II of (Old) Contention 14 in NRDC's 9th set of Interrogatories.
- 3. That the above mentioned and attached answers are true and correct to the best of his knowledge and belief.
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( fignature[ Subscribed and sworn to before me this M M day of #N/,1982. Y %., l3 7}<yb, . ()sotaryPub1N
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;i; ) DOCKET NO. 50-537 26 WITED STATES DEPARetT OF ENERGY
,M.b . ) f {ijje .... .MMMECT M4llA8EMEIIT ) CORPORATION
.i. TElBESSEE VALLET AlmWRITY h) f^. AFFIDAVIT OF DR. CARL ALBERT ANDERSON. JR.
v< , [; Dr. Carl A. Anderson, being duly sworn, deposes and says as follows:
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sg";. 1. That ;T is suplayed by idestinghouse Electric Corporation s, Advanced Reactors Division jN, - Post Office Box 158 Madison, Pennsylvania 15663 l D; F.. as Manager.. Reactor Projects. f 6 2, The he is duly authorized to answer the Interrogatories numb [.2:
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2 (a) thru (f) in futDC's 13 set of Interrogatories. f.' ., 1.v. ,.
~ 3. That the above< mentioned and attached answers a
'~ best of his:hessledge and belief.
(Signature) gIl.,]. _,1982.
'i;" . Sescribed and suern to before Me this Mbay of ,
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WITED STATES IEPARTMDiT F DERGY ) SOCIET W . 5-537 l l PRlWECT MAnMiceli CORPetATION ) , TDAES$EE MLLEYp!TY 3 )
- MF10AVIT F LEE E. STRAW 8 RIDGE
. y ' . Lee E. Straubridge. befag dely suern, deposes and says as fellows: l i - - l
- 1. That he is empicyed by the Weitfnghouse Advanced Reactors Division as
, Hamager, unclear Safety and Licensing, P. O. Sex 158, Madison, j Poemay1vania 15663. j
- 2. That he is 41y authorized to answer the Interngatories numbered 7-16, Its 26, 28, 34, 36, and 42 of MDC's Seventh Set of Interrogatories.
- 3. That the above-eentioned and attached answers are true and corriet to the best of his knowledge and belief.
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subscribed and morn to before me this M day of / . 1982. j i
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, T. i-WITED STATES E NERICA l ELEAR EGULATORY COMISSION ] - la the entter of ) 1 4 f WITH STATES DEPAKTIST W DERGY ) SOCKET W. 50-537 f peWECT sumspui coerceAT!on 1 [ i 1suurtw r ilALLEY Alm 10RITY )4 t 4 j NFISAVIT W IEleIIS N. SWITICK r i i Beasts N. Switick, being daly morn, deposes and says as folleus: I !
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- 1. Inst he is supittyed by the General Electric Compatty as Manager, Safety i' .
Analysis, Advanced Reactor $ystems Department, 310 De Guigme Drive, Soutrvale, Callfornia 94006.
! 2. That he is esly authorized to mswer the interrogatories embered 3-6, ,.
17, 23, 24, 25, 29-32, 37-41, 43-47, 49-51, and 53 of IulDC's Seventh Set of Interrogatories. !
- 3. That the above-eentioned and attached answers are true and correct tol best of his knowledge and belief.
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_-r M. _ k l r (f.1gnaturin j I Subscribed and suorn to before se this g my of _M ,1982. t 1 i 7 7tary runne - - i i sv cemetssim aptres FIs4V . l' i l
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W ITED STATES F MERICA WELEAR REGEATORY CGet!5SION In the aratter of . 1 mITEn STATES SEPARMENT Ch DERGY )' BOCKET W. B-5U N&1ECT fumAGOENT CORPORATION )
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1marttrr igLLET AUTHORITY ) 5
. NFIDAVIT F L. WLTER DEITRICH
, 9 L. hiter hitrich, being 41y suorn, deposes and says as follows:
- 1. That he is ampladed ly the Reactor Analysis and safety Otvision of Hetional Laboratory, WOO So. Cass Avenue, Argonne, Illinois
, as Associated Divisten Director.
- 2. 1het he is 41y authorized to answer the Interrogatories numbered 27 and
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N of M DC's Seventh $st of Interrogatories.
- 3. That the above-eentioned and attached answers are true and correct to the best of his kassiedge and belief.
b I-151gnature) Subscribed and suorn to before ne this g day of 1982.
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, UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION i
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In the Matter of )
)
UNITED STATES DEPARTMINT OF ENERGY )
)
PROJECT MANAGEMENT CORPORATION ) Docket No. 50-537
)
TENNESSEE VALLEY AUTHORITY )
) ,
(Clinch River Breeder Reactor Plant) ) l l ) 1 CERTIFICATE OF SERVICE l 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 Com=ission 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_ !
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- Atomic Safety & Licensing Appeal Board i
. U. S. Nuclear Regulatory Commission Washington, D. C. 20545 -
- Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission :
Washington, D. C. 20545
- Docketing & Service Section l Office of the Secretary ;
U. S. Nuclear Regulatory Co= mission - l Washington, D. C. 20545 (3 copies) William M. Leech, Jr., Attorney General William B. Hubbard, Chief i Deputy Attorney General ! Lee Breckenridge, Assistant l Attorney General State of Tennessee ! Office of the Attorney General . 450 James Robertson Parkway ! Nashville, Tennessee 37219 l Oak Ridge Public Library . Civic Center ! Oak Ridge, Tennessee 37820 l Herbert S. Sanger, Jr., Esquire ! Lewis E. Wallace, Esquire i W. Walter LaRoche, Esquire . James F. Burger, Esquire Edward J. Vigluicci, Esquire i Office of the General Counsel i Tennessee Valley Authority i 400 Commerce Avenue : Knoxville, Tennessee 37902 (2 copies) I i
**Dr. Thomas Cochran i Barbara A. Finamore, Esquire :
Natural Resources Defense Council > 1725 Eye Street, N. W., Suite 600 l Washington, D. C. 20006 (2 copies) Mr. Joe H. Walker 401 Roane Street Harriman, Tennessee 37748 , Ellyn R. Weiss Harmon & Weiss i 1725 Eye Street, N. W., Suite 506 Washington, D. C. 20006 l l
3-Lawson McGhee Public Library 500 West Church Street
. Knoxville, Tennessee 37902 :
William E. Lantrip, Esq. i 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) i
**Eldon V. C. Greenberg l Tuttle & Taylor 1901 L Street, N. W., Suite 805 Washington, D. C. 20036 ,
Commissioner James Cotham Tennessee Department of Economic and Community Development Andrew Jackson Building, Suite 1007 Nashville, Tennessee 37219 A Geor M . Edg W Attorney for l Proj ect Management Corporation DATED: April 30, 1982 l t
*/ Denotes hand delivery to 1717 "H" Street, N.W., Washington, D. C.
**/ Denotes hand delivery to indicated address.
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