LD-87-067, Forwards Proposed Advanced Reactor Severe Accident Program Resolutions for Four Remaining Nrc/Idcor Severe Accident Issues.Six Valid Doe/Idcor Resolutions Will Be Adopted in Development of Sys 80 Design.Concurrence Requested

Forwards Proposed Advanced Reactor Severe Accident Program Resolutions for Four Remaining Nrc/Idcor Severe Accident Issues.Six Valid Doe/Idcor Resolutions Will Be Adopted in Development of Sys 80 Design.Concurrence Requested

ML20236T350
Person / Time
Site:	05000470
Issue date:	11/24/1987
From:	Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	Miraglia F NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
LD-87-067, LD-87-67, NUDOCS 8712010165
Download: ML20236T350 (27)

Text

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SOISBUSTION ENGINEERING November 24, 1987 LD-87-067 Docket No. STN-50-470F Mr. Frank J. Miraglia Associate Director for Projects Office of Nuclear Reactor Regulation Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Advanced Reactor Severe Accident Program - Topic Paper Submittal

Dear Mr. Miraglia:

On August 3,1987, contractors for the Department of Energy met with Office of Nuclear Reactor Regulation Staff to outline their Advanced Reactor Severe Accident Program (ARSAP). In that meeting, the ARSAP approach to resolution of severe accident issues for Advanced Light Water Reactors

( ALWRs) was described. Basically, we understand that ARSAP will review and, to the extent possible, adopt resolutions which have resulted from interactions between the Nuclear Regulatory Commission (NRC) and the Industry Degraded Core Rulemaking (IDCOR) program. For those issues where modifications are recommended for ALWR's, ARSAP will produce Topic Papers i describing the proposed alternatives, j We understand that the NRC and IDCOR Staff have identified nineteen (19) severe accident issues and have documented agreed-upon resolutions to eleven (11) of them. (These are referred to as Topic Set 1.) Of these l eleven (11), ten (10) are applicable to PWRs. ARSAP has reviewed these ten (10) issues and we are advised that they have concluded that the NRC/IDCOR resolutions to six (6) of them are valid for ALWRs.

Those six issues are:

o Fission Product Release Prior to Vessel Failure - IDCOR Issue 1, o Release Model for Control Rod Materials - IDCOR Issue 3, o In-vessel Steam Explosion and Alpha Mode of Containment Failure -

IDCOR Issue 7, o Ex-vessel Heat Transfer Models from Molten Core to Containment -

IDCOR Issue 10, o Modeling of Emergency Response - IDCOR Issue 14, and o Secondary Containment Performance - IDCOR Issue 16. I off Power Systems 1000 Prospect Hill Road (203) 688-1911 Combustion Engineering. Inc. Post Office Box 500 Telex: 99297 Windsor, Connecticut 06095-0500 8712010165 071124 PDR C

TOPRP ENVC-E PDR _ _ _

Mr. Frank J. Miraglia LD-87-067 November 24, 1987 Page 2 We are futher advised that ARSAP has determined that the four (4) remaining NRC/iDCOR resolutions should be modified before being adopted for ALWRs.

Proposed ARSAP resolutions are attached to this letter for the following issues: ,

l o Reactor Coolant System Natural Circulation - IDCOR Issue 2, 4 o Fission Product and Aerosol Deposition - IDCOR Issues 4 and 12, o Ex-Vessel Fission Product Release - IDCOR Issue 9, and o Revaporization of Fission Products - IDCOR Issue 11.

Combustion Engineering plans to adopt, in the development of the System 80+ M design, the six (6) NRC/IDCOR resolutions identified above and the four (4) resolutions proposed by ARSAP and attached to this letter. We request your early concurrence on their acceptability.

If you have any questions or comments on the attached material, please feel free to call me or Dr. Michael D. Green of my staff at (203) 285-5204.

Very truly yours, COMBUSTION ENGINEERING, INC.

m -

A. E Scherer Director Nuclear Licensing AES:ss Attachment i

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a DOE Advanced Reactor Severe Accident Prostram*

ARSAP Proposed Resolutions for Severe Accident Issues - TOPIC SET 1

• The material in this attachment was developed by the DOE ARSAP in support of C-E's Design Certification Program

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TOPIC SET 1: IDCOR/NRC Resolved Issues l

Many of the issues related to severe accidents in existing LWRs have been l identified previously by the IDCOR Program. During the IDCOR/NRC interaction process,I 19 issues were identified as topics for resolution. These issues were clearly defined, and an acceptable approach to resolution was proposed for present LWRs in order to meet the NRC Severe Accident Policy Statement.2 Eleven of the 19 issues were agreed upon by NRC and IDCOR to have been resolved. Ten of those eleven issues are either generic or PWR specific.

Because ARSAP's interaction with NRC is through the CE System 80+

certification effort, those issues, or aspects of issues, related exclusively to BWRs are not addressed here. The ten PWR-relevant S/A issues considered to be resolved are:

1. Fission Product Release Prior to Vessel Failure (IDCOR Issue 1)

2. Reactor Coolant System Natural Circulation (IDCOR Issue 2)

3. Release Hadel for Control Rod Materials (IDCOR Issue 3)

4. Fission Product and Aerosol Deposition in RCS and Containment (IDCOR Issues 4 and 12)

5. In-vessel Steam Explosion and Alpha Mode of Containment Failure (IDCOR Issue 7)

6. Ex-vessel Fission Product Release (IDCOR Issue 9)

7. Ex-vessel Heat Transfer Models from Molten Core to Containment (IDCOR Issue 10) 1 1

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8. Revaporization of Fission Products (IDCOR Issue 11)

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9. Modeling of Emergency Response (IDCOR Issue 14) l

10. Secondary Containment Performance (IDCOR Issue 16) w ARSAP has reviewed the. documented IDCOR/NRC resolution of these issues and j concludes that, for the majority of issues, the suggested resolutions j contained therein are applicable to both existing plant designs and those l l proposed for ALWRs. In general, the issues have to do with generic I I

l' phenomenological behavior.

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Specifically:

1. For six of these issues in Set 1, resolution should be attained in precisely the manner agreed upon and documented in the NRC/IDCOR interaction process. Those six issues are:

(1) Fission Product Release Prior to Vessel Failure (IDCOR Issue 1)

(3) Release Model for Control Rod Materials (IDCOR Issue 3)

(5) In-Vessel Steam Explosion and Alpha Mode of Containment Failure IDCOR (Issue 7).

(7) Ex-Vessel Heat Transfer Models from Molten Core to Containment (IDCORIssue10). i (9) Modeling of Emergency Response (IDCOR Issue 14).

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(10) Secondary Containment Performance (IDCOR Issue 16).

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2. For two of the issues, planned improvements in the ALWR design provide a resolution mechanism which is described later in this  ;

document. These issues are:

(2) Reactor Coolant System Natural Circulation (IDCOR Issue 2)

(6) Ex-vessel Fission Product Release (IDCOR Issue 9)

3. For the remaining two issues additional analysis has been performed and additional information has been developed which has impacted the ALWR approach to resolution.

The issues effected in this manner for Topic Set 1 are:

(4) Fission Product and Aerosol Deposition in RCS and Containment (IDCOR Issues 4 and 12)

(8) Revaporization of Fission Products (IDCOR Issue 11)

The proposed resolutions of the above four issues that are somewhat modified from the IDCOR/NRC positions are presented in the following sections. Included is a brief writeup containing a statement of the issue, a summary historical perspective, a statement of the IDCOR/NRC resolution and the proposed ARSAP resolution.

References

1. Mitchell, H. A., et al., " Review of IDCOR/NRC Issues", DOE /ID-10162, Advanced Reactor Severe Accident Program, March 1987.

2. Policy Statement on Severe Reactor Accidents regarding Future Designs and Existing Plants, 50FR 32138, August 8,1987.

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1.2 SEVERE ACCIDENT ISSUE TOPIC PAPER REACTOR COOLANT SYSTEM NATURAL CIRCULATION (IDCOR ISSUE 2) l l l I. Issue Definition l

Natural circulation flows can occur between the uncovered core region of PWRs and the upper plenum during a severe reactor accident. These flows are induced by the relevant buoyancy of gases heated in the core versus that of the gases cooled by the upper plenum internal structure. Flows can also occur between the upper plenum region of a PWR and the steam generators via the hot legs. These flows are induced by the relevant buoyancy of gases in the upper plenum versus the gases in the steam generator tubes. The uncertainty in the direction and magnitudes of the flow rates comprise the issue.

Such natural circulation flow is an issue because heat is transported by the flow to regions outside the core. Thus, the magnitude of heat transfer determines the heatup and possible degradation of structures outside the core. For example, melting of upper plenum internals, structural failure of vessel penetrations such as a PWR hot leg nozzle junction, or steam generator tube failure have been postulated as consequences of strong natural recirculation flows.

Since natural circulation carries gases to and from the core region, it is possible that it is an additional mechanism for introducing steam to the core where it could then be reduced by overheated cladding to hydrogen. The exothermic oxidation of cladding serves as an additional driving force for flows.

Natural circulat.on flows also carry fission products throughout the primary system; consequently the amount and location of fission product deposition is dependent upon this phenomena. The overall holdup of fission products in the primary system and the potential for revaporization are directly related to these flows. Deposition and revaporization of fission products are dependent upon natural circulation flows. These are addressed in the resolution of IDCOR Issues 4 and 11.

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II. Hjstorical Perspective

a. Industry Actions to Address the Issue As part of the the IDCOR Program on issue resolution,I mechanistic models of natural circulation were incorporated into the MAAP code 2 for severe accident analysis. These models were inspired by and benchmarked against EPRI/ Westinghouse PWR in-vessel natural circulation experiments 3 and EPRI-sponsored. codes.4 The MAAP code was further upgraded with mechanistic models to account for phenomena anticipated during core degradation but not observable in the experiments, such as cled ballooning, the possibility of inter-channel flow redistribution, radial radiation heat transfer, and heat transfer through the core barrel-baffle assembly.2 Applications of the MAAP models to the EPRI/ Westinghouse tests have been highly successful.I These tests have been performed with high-pressure simulants in a scaled manner and are directly applicable to prototypical conditions. In these experiments, circulation between the core and upper plenum and between the upper

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plenum and steam generators was observed. It is ::oncluded that the MAAP model is reliable and provides a mechanistic description of the phenomenon.

b. NRC Actions to Address the Issue NRC-sponsored researchers at LANL and SNL have performed two-dimensional natural circulation calculations using the MELPROG code,5,6 These calculations have predicted higher temperatures than the MAAP calculations. However, despite the complexity of the MELPROG code, the controlling phenomena of counter-current flow in the hot leg and multi-directional flows iii steam generator tubing have not been incorporated into MELPROG.

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.c. Outstanding Questiq01

'While the upgraded model in MAAP has shown that steam generator tube rupture can be ruled out for certain plants, this has not been.shown-for all plants, nor has the NRC developed a modeling capability of-similar flexibility and complexity to independently verify the IDCOR work. Failure of the hot' leg appears possible, but not certain.

, Therefore, the possibility of temperature-induced failure during high-pressure ' sequences is still an outstanding question.

d. The NRC Position The following position has been presented by NRC:7 The Staff believes that. existing analytical models for natural circulation flow phenomenon need to be improved to realistically-predict the thermal hydraulic coupling with core-melt progression, hydrogen generation, and~ fission product release and deposition.

Based on available information it is the Staff's judgement that the uncertainties in the prediction of temperatures within the RCS are sufficiently large that RCS depressurization via temperature-induced failure in the' primary piping system can be neither assured nor ruled out. Accordingly, the Staff indicated that the industry

-should consider both possibilities in estimating risk for high pressure sequences in PWRs and in the development of the individual plant examination methodology.

Available analysis also suggests that natural circulation may strongly influence the issue of fission product revaporization by altering the spatial distribution of fission product initially deposited in the RCS, as well as the temperature distribution within the RCS. Uncertainties'in predicted RCS distribution temperatures are large; for example, estimates of upper plenum 6

structure temperatures range from 1300 to 2100 F. Since tts effect of temperature differences on this order would substantially affect fission product deposition and vaporization, the industry stould address this matter in its sensitivity study.

III. Technical Anoroach to Resolve the Issue for ALWRs The EPRI Advanced Light Water Reactor Utilities Requirements Document 8 requires that ALWR designs provide an AC independent safety depressurization system. This system provides a means for diverting core heat from the steam generators to the containment. This diversion results in a reduced reactor coolant system heat up. The ARSAP position is that this issue is resolved for ALWRs provided the ALWR designs incorporate the depressurization system called for in the~EPRI Requirements Document. Confirmatory analysis, by the designer, must demonstrate that heat transfer from the core to the steam generators, caused by natural circulation, is interrupted.

IV. References

1. IDCOR Technical Report 85.2, Technical Report Support for Itsue Resolution.

2. MAAP 3.0 User's Manual, Fauske & Associates, Inc., Burr Ridse, Illinois.

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3. Review of EPRI/ Westinghouse Natural Circulation Experiments, Pittsburgh, communication with David Squarer and William Stt art of Westinghouse, April, 1985.

4. Personal communication, T. Wassel, SAI, to M. Kenton, FAI, June, 1987.

5. R. J. Henninger, J. E. Kelly, and J. F.

Dearing,

" Preliminary 2-D MELPROG Calculation for the TMLB' Accident in Surry", Los Alamos and Sandia National Laboratories, March 17, 1986.

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,, r, g 6. J. T. Han, "A Summary of March 17, 1987 Meeting on RCS Natural Circulation Studies", May,1986.

7. " Summary Paper 'for the Resolution of NRC/IDCOR Is!.ue 2: Reactor Coolant System Natural Circulation", attachment to letter, T. Speis, NRC, to A. Buhl, IT, dated September 26, 1986.

8. Advanced Licht WA1.er Reactor Requirements Document, Chapter 5:

Engineered Safeguards. Systems, Electric Power Research Institute, Palo Alto, CA; Draft dated September 1987 (yet to be published).

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a I.4 SEVERE ACCIDENT' ISSUE TOPIC PAPER FISSION PRODUCT AND AEROSOL DEPOSITION'IN RCS AND CONTAINMENT C 11DCOR ISSUES 4 AND 12)

I. ' Issue Definition.

Severe accidents in nuclear power plants release fission product vapors and aerosols into the primary system and containment. These fission products'are removed from the gas phase by various' natural phenomena and engineered safeguards. The deposition of aerosols is controlled by the agglomeration of' submicron source particles. Typically the NRC staff assessments of aerosol agglomeration and deposition rely on detailed computer simulations of the

agglomeration process.= These calculations solve directly the~ aerosol agglomeration equationI ,2.for each accident sequence investigated.

I The aerosol equation has two asymptotic solutions that exist for each.  ;

deposition mechanism of importance to reactor safety.3,4,5,6,7 Unique i nondimensional solutions can be obtained from computer simulations of the  ;

equation and from experimental data. These solutions have correlated in terms of the nondimensional parameters and these correlations provide a '

l simplified, yet accur' ate means for calculating aerosol behavior. . l The technical _ argument pertaining to this issue is whether the aerosol correlation is realistic and accurate over the full range of severe accident  ;

conditions in the RCS and containment. Specifically, NRC contractors concerns over the use of the aerosol correlations for severe accident studies- j were:  ;

I. Mass transfer coefficients for the correlations are limited to a l power law dependence on particle volure. 1 1

2. The correlations cannot be applied to depletions in which more than j two mechanisms are acting simultaneously. l 1

3. The correlations apply only for mature aerosols.

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.4. There is no physical significance to the form of the curve' fit to the computer solutions.

'i . The correlations lack direct dependence between volumes and gas

. flows and the variables used 'in the curve fits.

II. Historical Perspectlyg; p a.- Industry Actions to Address the Issue The industry has made extensive comparisons of the correlations to-the NRC aerosol codes and the NRC-sponsored aerosol deposition and transport experiments. The. industry has worked with the NRC's contractors in explaining the derivation and the nature of the aerosol correlations used in severe accident studies.5

b. NRC Actions to Address the Issue

, The NRC has extensively reviewed the correlations and the outstanding issues and fourd the differences not to be significant-enough to preclude the use of the correlations for severe accident studies.8

c. Outstanding Ducstions The major outstanding issue with the use of the correlations is the third NRC contractor concern listed above. The NRC position on the use of the correlations is that they are adequate to use only for those sequences involving long periods of aerosol depletion. The NRC staff feels that the correlations can not be relied upon for those sequences in which aerosols are rapidly released to the environment. ,

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The short time frame for which the correlations are not considered acceptable by the NRC has not been defined precisely by their contractors.9

d. The NRC Position The NRC's position on the use of the aerosol correlations is that they are sufficient for the analysis of severe accidents, given sufficient time for the aerosol behavior to approach that of.the correlation curve.

IV. Technical Anoroach to Resolution for ALWRs DOE's Advanced Reactor Severe Accident Program developed a sectional version of the MAAP code to asses:1 those sequences that may fall out of bounds of the regime of applicability o# the aerosol correlations.10 Comparisons were made between the two models for sequences that have short residence times for the aerosol in the primary system. For Small Break LOCA sequences that lead to core melt, the amount of fission product material deposited in the primary system agreed to within 13 between the two different models.10 In fact, the sectional model predi:ted more deposition than the correlational model.

In any event, such differences are considered small compared to other uncertainties in the problem. ARSAP believes that the correlational model included in MAAP is adequite for use in ALWRs and intents to analyze the Combustion Engineering design with the correlational models. NRC should review the ARSAP technical report describing the comparisons noted above and provide concurrence on the ARSAP position.

V. References

1. F. Gelbard, Y. Tambour and J. H. Seinfeld, " Sectional Representations for Simulating Aerosol Dynamics, Journal of Colloid .

and Interface Science, 21,1980.

2. N. A. Fuchs, The Mechanics of Aerosols, Pergamon Press, Oxford, 1964.

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3. M.1Epstein, P. G. Ellison, and R. E. Henry, " Correlation of Aerosol Sedimentation", Journal of Collold and Interface Science, Vol.113,

,- No. 2,. October,.1986.

4. . M. Epstein and P. G. Ellison, "A Principle of Similarity for

. Describing Aerosol' Size Distributions", Journal of Colloid 'and

. Interface Science, to be published.

5. "IDC0F. Technical Report for Issue Resolution, T85.2", Atomic Industrial Forum, July,1985.

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6. M. Epstein and P. G. Ellison, " Correlations of the Removal Rate Constant for Coagulating and Depositing Aerosols and Their Application to Nuclear Reactor Safety Problems", Nuclear Enaineerina and De11gn, ts be' published.

7. P. G. Ellison and M. Epstein, " Experimental Evidence of Nuclear Aerost1- Similitude", Nuclear Science and Enaineerina, to be published.

8. T. Krt ss', " Review o'f the FAI Aeroso'1 Correlation" . informal letter repori.to Lisa Chan, USNRC, March 20,~ 1986.

9. Lettet from T. Spets, USNRC, to A. Buhl, IT, March 11, 1987.

10. P.' G. Ellison and M. Epstein, " Nuclear Fission Product Aerosol Trans; ort and Deposition" , DOE Advanced Reactor Severe Accident Progrtm Report, to be published.

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. 1.6 SEVERE ACCIDENT ISSUE TOPIC PAPER' EX. VESSEL' FISSION PRODUCT RELEASE (IDCOR ISSUE 9)

I. . Issue Definiti g Fission products can be released from core debris to the containment atmosphere during core debris-concrete interactions. The technical argument

'of the original'IDCOR/NRC issue concerns the modeling differences between IDCOR and NRC on the fission product release chemistry. A related issue which impacts predictions of the release model is the concrete erosion rate and debris temperature, covered by NRC/IDCOR Issue 10; "Ex-vessel Heat Transfer Models From Molten Core to Concrete / Containment". ,

The primary concern expressed during the NRC/IDCOR dialog was the degree of sophistication of the model and number of species assumed by IDCOR in the i MAAP 2.0 code.1 T11s model does not account for all the species or phenomena considered by the NRC model, VANESA.2 The IDCOR model produced lower overall releases than the NRC model, in some cases by orders of magnitude. Another part of the issue is the chemical state of fission products assumed bI the NRC model, which may neglect species or non-ideal effects.which subs:antially lowcr the total vapor pressures of fission products'or increase the vapor pressures of radiologically inert elements.  !

'This issue is of minor importance if the debris is. covered with water, because overlying yater will scrub fission products released from the debris. Further, if water cools the debris and terminates core concrete interaction, the phenomena does not occur. However, the issue'of debris coolability must be addressed because in addition to cuenching of debris, establishing an adequate heat transport path to remove decay heat is necessary to terminate the accident. The issue of debris coolability is scheduled to be addressed in a subsequent topic paper along with other issues related to conditions for safe stable states.

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• II. Historical Perspective

a. Industry Actions to Address the Issue j

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As part of the NRC/IDCOR issue resolution, new models were developed for ex-vessel fission product release.3 The first such model, EQUUS, contained a large number of species and a chemical equilibrium formulation.4 While this was satisfactory at the time, it revealed that the assumption of overall debris chemical equilibrium versus separate oxide and metal equilibrium5 yielded larger releases for certain species. Furthermore, the EQUUS-model could not be directly incorporated into MAAP 3.0;6 instead, a simplified set of equilibria were, again, under the assumption of separate cxide and metal equilibrium.

Under EPRI sponsorship, a neis model called MET 0XA 7 ,8 was created and integrated with the MAAP 3B code. This version of MAAP is being prepared for imminent release and will be the baseline version of the code for ARSAP. The MET 0XA model considers equilibrium of the entire debris, non-ideal behavior of certain species, and an i extended list of condensed phase material, including control material, beyond that of the NRC model.2 Gas phase species were selected from an extended species list by quantitative ordering of important species using EQUUS. Kinetic effects used within the NRC mode 12 are neglected because it has been shown that equilibrium is attained9,10 within a short time frame.

Results of the MET 0XA model indicate that key areas of uncertainty exist in the predictions related to nonideality of the debris chemistry. However, the model can be upgraded to handle these effects when appropriate data becomes available. This is an important point relative to the NRC conclusions as to the state of issue resolution. The resolution is viewed as conditional on the ability of the industry model to reproduce experimental data.

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, I Experiments are being funded by the industry through EPRI and carried out at ANL.11,12 While these were not created'as part of '

the NRC/IDCOR issue resolution process, information obtainable from this program will be directly applicable to resolution through MAAP {

model upgrading and benchmarking.

b. NRC Action to Address the Issue The NRC is continuing a large-scale integral core-concrete experimental program at SNL, and currently the SURC (sustained urania interacting with concrete) series is being pursued.13 Prior to SURC there have been several series of experiments known as TURC (transient urania interacting with concrete),14 SWISS (sustained water interactions with stainless steel),15 and HSS (hot solid series).16 As part of issue resolution, NRC is continuing the large scale SURC series. The SURC matrix consists of experiments with either stainless steel or uranium / zirconium oxides sustained at high temperatures through induction heating and allowed to penetrate either downward or both downward and radially into concrete. The scale is approximately 40 cm diameter for the initial cavity.

Fission product simulants are employed. Thermal and aerosol measurements are made.

Since the NRC/IDCOR issue resolution process was completed, NRC has begun a new set of experiments known as the WITCH and GHOST series.17 These are intended to provide better information for fission product chemistry and aerosol formation. {

c. Outstanding Questions The technical differences between NRC and IDCOR have been resolved through upgrading of the IDCOR medels. There are currently in progress both industry and NRC sponsored experimental programs which l

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will be used for development of. a database. This' data is intended

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to be available for both industry and NRC codes to use for modeling-of ex-vessel fission product release. Issue resolution, according  ;

I to NRC, .is contingent upon further confirmatory results from the experimental program and subsequent benchmarking of codes against experimental results.

d. The NRC Position i

The NRC has stated the following position 18 which related both to the subject issue and to NRC/IDCOR Issue.10. Ex-vessel Heat Transfer Models from Molten Core to Concrete / Containment:'

"From our review, we conclude that the IDCOR model revisions to include more chemical species are acceptable. The overall heat transfer predictions appear to be in reasonable agreement with the available data. If the SURC data and the comparisons of the IDCOR codes with that data indicate that the IDCOR treatment of core debris coolability is not supported, we vill require modifications to the codes."

"This resolution of Issues 9 and 10 must be viewed as condit'ional and is based on our current understanding of core-concrete phenomenology. This approval is contingent on comparisons being made.with new experimental data when such data becomes available".

"It is our understanding that IDCOR has agreed to this approach. If significant discrepancies are identified through these comparisons, we will require that further model revisions be made to address

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these discrepancies. A significant step forward in acquiring this  !

data is expected to be made with the completion of the SURC tests later this year and in 1987. In addition, there is a strong international effort involved in research and code development in the core-concrete interaction area attesting to the recognized 16 i;

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' importance of this subjact.14 It is expected that this

-international interest'will ensure that.the ongoing improvements;in our capability:to predict the risk contribution from core-concrete interactions will continue."

III.. Technical Accroach to Resolution As' stated in the issue definition, phenomena of ex-vessel fission product release is of_ minor importance if it can be assured that the debris is covered with water because fission products would e effectively be scrubbed..

The EPRI Advanced Light' Water. Reactor Requirements Document requires that the- j containment design and containment water inventory be such that water would naturally be present in-the cavity prior to vessel failure.19 The ARSAP' position.is that the designs which will be generated, using the.EPRI' requirements document as criteria, will prevent the phenomena in question from being a risk'significant issue. ARSAP will continue to benchmark the MAAP codes-as new data, relevant to this issue, is obtained.

V. References

1. MAAP 2.0 User's Manual, Fauske & Associates, Burr Ridge, Illinois.

2. - ' D Powers, et al . , "VANESA: A Mechanistic Model of Radionuclides Release and Aerosol Generation During Core Debris Interactions with Concrete",NUREG/CR-4308, July,1986.

3. IDCOR Technical Report 85.2, Technical Support for Issue Resolution.

4. M. G. Plys and R. E. Henry, "Ex-Vessel Fission Product Release Modeling", Trans. Am. Nucl . Soc., Vol . 50,- November,1985.

5. D. Cubicciotti," Thermodynamics of Vaporization of Fissions Products Under Severe Accident Conditions", Pure and Aoolied Chemistry, Vol. 57, No. 1, 1985.

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6. MAAP 3.0 User's Manual, Fauske & Associates, Burr Ridge, Illinois.

7. " MET 0XA: An Equilibrium Model for Fission Product Release During Core-Concrete. Interactions",FA!/87-20, draft, April,1987.

8. M. G. Plys, M. A. Kenton, and R. E. Henry, Ex-Vessel Fission Product' Release Modeling Within Integrated Accident Analysis, Proc. of the CSNI Specialists' Meetino on Core Debris-Concrete Interactions, EPRI NP-5054 SR, February, 1987.

9. G. A. Greene, Brookhaven National Laboratories, personal communication, 1987.

10. M. G. Plys, P. G. Ellison, and R. E. Henry, " Influence of Containment Thermal-Hydraulics on Source Term Compositions", Proc.

of the Third International Tooical Meetina on Reactor Thermal Hydraulics, October 15-18, 1985.

11. B. W. Spencer, et al., "EPRI/ANL Investigations of MCCI Phenomena and Aerosol Release", Proc. of the CSNI Specialists' Meetina on Core.

Debris-Concrete Interactions, EPRI NP-5054 SR, February, 1987.

12. M. Tetenbaum, et al., " Fission Product Release from Core-Concrete Melts", Proc. of the CSNI Specialists' Meetina on Core Debris-Concrete Interactions, EPRI NP-5054 SR, February, 1987.

13. E. R. Copus and R. Blose, " Sustained Uranium-Concrete Interactions:

The SURC Experiments", Proc. of the CSNI Specialists' Meetino on Core Debris-Concrete Interactions, EPRI NP-5054 SR, February,1987.

14. J. E. Gronager, et al., TURC 2 and 3: "Large Scale U02 /Xt02 /Zr Melt-Concrete Interaction Experiments and Analysis", NUREG/CR-4521, June, 1986.

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. 15. E. R. Blose, et al., " SWISS: Sustained Heated Metallic Melt Concrete

. Interaction with' an Overlying Water Pool Experiment and Analysis,"

SAND-1546, September, 1986.

16. E. R. .Copus and D. R.' Bradley, " Interaction of Hot Solid Cc re Debris -

with Concrete, NUREG/CR-4558", June, 1986.

'17. S. 8. Burson, "The USNRC Research Program on Core Debris /Cencrete

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Interactions and Ex-Vessel Fission Product Release", Proc. of the

' CSNI Specialists' Meetina on Core Debris-Concrete Interactiaal, EPRI NP-5054 SR, February, 1987.

18. " Summary Paper for the- Resolution of IDCOR/NRC Issue 9: Ex-Vessel Fission Product Release and Issue 10: Ex-Vessel Heat Transfer Models from Molten Core to Concrete / Containment", attachment to

. letter, T. Speis, NRC, to A. Buhl, IT, November 26, 1987.

19. Advanced Liaht Water Reactor Requirements Document, Chapter 5:

Engineered Safeguards Systems; Electric Power Research Institute, Palo Alto, CA; September,1987.

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, l '. 8 SEVERE--ACCIDENT ISSUE TOPIC PAPER REVAPORIZATION OF FISSION PRODUCTS (IDCOR' ISSUE 11).

.I. Issue Definition.

For accidents involving long reactor coolant system (RCS). residence times, a large fr' action of volatile. fission products can be retained in the primary system and deposited on structures. Revaporization of.those fission products

-such as Cs0H and.Csl can occur due to increasing RCS surface temperatures that result from decay heat associated with deposited-fission' products.

.Revaporization could have an important effect on source term for accident sequences in which the containment has failed prior to revaporization.

IDCOR and NRC both agree that revaporization can occur. The technical difference pertains to whether present models are sufficiently complete to assess.the rate and timing of the revaporization of deposited liquid and solid fission products. The timing and process of revaporization depend on vapor pressure, which is a function of chemical species and temperature, and thermal hydraulic conditions such as reactor vessel natural circulation and primary system beat loss. The higher the vapor pressure, the earlier revaporization occurs. The' deposited species may react with.the RCS surfaces to form low vapor pressure compounds. Thus, fission product-RCS surface reaction may reduce and delay revaporization. However, the detailed surface reaction chemistry is not well understood.

This issue impacts, to an' extent, all plants and sequences. Significant retention in the RCS is required for revaporization to inpact the source term. Sequences which are susceptible to this impact are generally those at which the primary system remains at pressure and in which cooling (i.e., AFW) is not available after vessel failure.

20 c=1-_--_=--________ _ - _ _ _ -- _ _

. ~II. Historical' Perspective I

a. , Industry Actions to Address the Issue 1

IDCOR/MAAP calculations included a treatment of revaporization using invariant chemical species assumptions (i.e., no credit is taken for j chemical reactions with the primary system steel surfaces) and i

progressively more elaborate RCS decay heating and natural circulation modeling. The revaporization model in MAAP uses i straightforward and conventional mass transfer relations from the literature. The model was compared tc EPRI-sponsored experiments at Argonne National Laboratory (ANL), which investigated revaporization behavior of Cs0H, CsI, and Te02 . The model predictions compare well with ANL Cs0H and Csl revaporization data at 750 C0 and 1000 CD which are the temperature range anc chemical forms of interest in the primary system.I When a large fraction of volatile fisfion products are retained in the primary system, there is potential for long-term revaporization which largely depends on the ability (f the primary system to reject heat. MAAP models primary system heat losses by considering separately the through-insulation and no-through-insulation terms.

It was found that recent surveys of asallable plant data show larger RCS system heat losses than those con:idered in reference plant analyses.1 This would result in lower RCS inner surface temperature and substantially lower revaporization.

To estimate the effect of surface chenical reaction on revaporization, IDCOR, using MAAP, ran two cases:

1. A base case with the pressure correlation in the code appropriate for no surface reaction and

2. a case with vapor pressure of 0.01 of that of the base case.I 21

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, n

.. j i

.o  !

The reduced vapor pressure decreases the revaporization rate and limits the total release of Cs! into the containment at vessel  !

failure. The total release to the environment is reduced, because it is largely controlled by the long-term revaporization rate in the primary system.

b. NRC Actions to Address the Issue At the present time, the Source Term Code Package (STCP),2 which is the tool the NRC uses to calculate source term, does not have a model to account for revaporization after vessel failure. However, both MELCOR and MELPROG include a revaporization model which has yet to undergo validation and benchmarking,

c. Outstanding Questions As stated above, the issue relates to the influence of the surface reaction chemistry and reactor thermal-hydraulics, such as natural convection flow patterns in RCS and heat losses from RCS, on the rate and timing of revaporization. The key question remaining is the modelling of the effect of the surface reaction chemistry.

d. The NRC Position The following are NRC positions:3 "In an effort to resolve this issue, IDCOR conducted sensitivity studies on vapor pressure using MAAP. Reducing vapor pressure to

.01 of the original values, IDCOR recalculated source terms for selected plants and sequences. Revised IDCOR calculations using reduced vapor pressure procuced a modest amount of revaporization; however, this approach, i.e., varying vapor pressure to simulate extent of revaporization, results in release of fission products early in the accident. NRC believes that the use of fission product surface chemistry models in lieu of the vapor pressure approach used 22

. n e.

) i by IDCOR would tend to predict slower but continuous revaporization.

Because delayed revaporization is not mitigated as effectively as early revaporization, the IDCOR model is considered to have l- non-conservative effect on source terms."

"It is the belief of the NRC that neither the STCP nor MAAP adequately accotnt for the impact of chemistry on revaporization, and that both NFC and IDCOR reference plant assessments should address the uncertainties associated with this issue via uncertainty analyses. Recognizing the limitations of the STCP and the lack of a thorough understanding of revaporization chemistry and thermal-hydraulics during and after vessel failure, this issue had been included in the NUREG-1150 uncertainty analysis. Similarly IDCOR should also consider revaporization in their uncertainty studies. Furthermore, the IPEM (Individual Plant Evaluation Methodology) shculd contain screening criteria that will highlight any plant-specific deviations from generic assumptions on revaporization, such as the magnitude of heat losses from the RCS."

III. Technical Acoroach to Resolution for ALWRs The issue is considered resolved, not because the phenomena are completely understood, but because the uncertainty calculations prescribed by the NRC will highlight any desigt deficiencies is ALWRs. Recent reassessment of heat losses through primary system insulation have shown that RCS inner surface temperatures are lower ttan previously thought. ARSAP's approach to resolution of this issue is to assure that the uncertainty calculations are performed and included ir. risk assessments. This assurance will be implemented by including a requirement for the sensitivity assessment in the PRA Assumptions and Grourdrules Section of the EPRI Requirements Document.4 1

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• - V. References.

1. "IDCOR Technical Report 85.2, Technical Support'of the Issue Resolution", July,1985, Atomic Industrial Forum.

2. Silberberg, M., Mitchell, J. A., et al., " Reassessment of the Technical Basis for Estimating Source Terms", NUREG-0956, United 1

States Nuclear Regulatory Commissions, July,1985.

1

3. H. A. Mitchell, J. C. Carter, and M. H. Fontana, " Review of IDCOR/NRC Issues", DOE /ID-10162, March, 1987.

4. Advanced Licht Water Reactor Requirements Document: Probabilistic BigjL ssessment it Key Assumptions and Groundrules Document, Draft.

Electric Power Research Institute, Palo Alto, CA; July 1987.

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