ML20128N529

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Audit Calculations for Station Blackout Sequence in Surry Facility
ML20128N529
Person / Time
Site: Surry, 05000000
Issue date: 01/10/1985
From:
BROOKHAVEN NATIONAL LABORATORY
To:
NRC
Shared Package
ML20127A894 List: ... further results
References
FOIA-85-110, REF-GTECI-A-44, REF-GTECI-EL, TASK-A-44, TASK-OR NUDOCS 8507130115
Download: ML20128N529 (49)
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'g 'N AUDIT CALCULATION 3 FOR A STATION BLACK 0UT SEQUENCE

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IN THE SURRY FACILITY 3 -

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.g s Accident Analysis Group ei and .,

,, E>oerimental Modeling Group Departmeq'. of Nucleai Energy Brookhaven National Laboratory Upton, New York 11973 i

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January 10, 1985 Prepared for U.S. Nuclear Regulatory Commission Washington, D.C. 20555 .

Under Contract No. OE-AC02-76CH00016 FIN A-8507130115 FOIA 850415 PDR PDR ALVAREZ85-110 2A?.

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ABSTRACT Calculations have been performed for a station blackout sequence in the l Surry Plant using the suite of severe accident phenomenology codes developed a

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for the Accident Source Term -Program Office (ASTP0), RES/NRC. This calcula- /

tional effort has demonstrated that the ASTP0 suite of codes can be exported i to an independent organization. A station blackout sequence at Surry was ori-

- ginally calculated by staff at BCL and SNL using the ASTP0 codes and reported in Volume V of BMI-2104. The purpose of the present BNL calculations is to-provide an independent audit of the BMI-2104 results. Hence, the full suite of ASTP0 codes have been obtained from BCL and SNL and made operational on the BNL computing system. No modifications were made to the codes other than those necessary to make them operational at BNL. Code input and output para-meters used in BMI-2104 were reviewed in detail and several inconsistencies related to data transfer between codes were found. Some difficulties were ex-perienced in the use of certain codes. However, a self-consistent station blackout sequence was calculated, and good agreement was obtained between BMI-2104 and the audit calculation.

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-1v-ACKNOWLEDGEMENTS The quality assurance team is indebted to R. A. Bari (BNL), J. Rosenthal (NRC), J. Mitchell (NRC), and J. Read (NRC) for many helpful discussions and guidance while carrying out this work and preparing the document. The ex- ..

pertise and help of Y. Sanborn (BNL)' and W. Bornstein (BNL) in carrying out the computer calculations is gratefully, acknowledged.- Thanks are due to T. .

Rowland for her preparation of the document on short notice.

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TABLE OF CONTENTS P a oe ABSTRACT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii ACKNOWLEDGEMENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . iv LIST OF TARLES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi -

LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii -

1. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . 1

> 1. 2 Ob.iective and Scope of Audit Calculations. . . . . . . . . . . 2 1.3 Calcul ational Methods. . . . . . . . . . . . . . . . . . . . . 2 1.4 Calculational Procedure and Team . . . . . . . . . . . . . . . 5

2. AUDIT CALCULATIONS. . . . . . . . . . . . . . . . . . . . . . . . . 7 2.1 MARCH 2 . . . ... . . . . . . . . . . . . . . . . . . . . . . . 9 2.2 MERGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2.3 CORSOR , . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 2.4 TRAP-MELT. . . . . . . . . . . . . . . . . . . . . . . . . . . 18 2.5 CORCON . . . . . . . . . . . . . . . , . . . . . . . . . . . . 21
2. 6 VANESA . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 2.7 NAUA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
3.

SUMMARY

OF RESULTS AND COMMENTS . . . . . . . . . . . . . . . . . . 35 e

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-vi-LIST OF TABLES Table Title Pace 1.1 BNL staff participating in audit calculations . . . . . . . . 6 2.1 Accident event times (minutes). . . . . . . . . . . . . . . . 12 2.2 Primary system response . . . . . . . . . . . . . . . . . . . 13 ,

2.3 Containment system response . . . . . . . . . . . . . . . . . 14 2.4 In-vessel mass bal ance (CORSOR) . . . . . . . . . . . . . . . 17 2.5 In-vessel mass balance during fission product transport (TRAP-MELT) . . . . . . . . . . . . . . . . . . . . . . . . . 20 2.6 Output from the VANESA model (ORNL/TM-8842) . . . . . . . . . 30 2.7 Breakdown of Cs 2 0 grouping: Su r ry TMLB ' . . . . . . . . . . . 31 2.8 Input to VANESA (kg) . . . . . . . . . . . . . . . . . . . . . 33 2.9 Ex-vessel mass bal ance (NAUA) . . . . . . . . . . . . . . . . 35 3.1 Overall mass balance (kg) . . . . . . . . . . . . . . . . . . 41 3 3.2 Distribution of species after accident (BNL fraction -

based on total mass from Table 3.1) . . . . . . . . . . . . . 42 I

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LIST OF FIGURES Figure Title Page 1.1 Flow diagram of ASTP0 codes for application to Surry . . . . . 4 2.1 Melt temperature history vs. time: Surry TMLB' sequence . . . 23 Integrated gas generated dates (kg) vs. time: Surry' 2.2 TMLB' sequence . . . . . . . . . . . . . . . . . . . . . . . . 24 2.3 Integrated gas generation rates (gm-moles) vs. time:

Surry TMLB' sequence . . . . . . . . . . . . . . . . . . . . . 25 2.4 Concrete erosion depths vs. time: Surry TMLB' sequence. . . . 26 3.1 Airborne and leaked masses (Audit calculations). . . . . . . . 39 3.2 Airborne and leaked masses (BMI-2104 Volume V) . . . . . . . . 40 4

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1. INTRODUCTION 1.1 Backoround During the last several years the NRC has sponsored research related to severe accidents in Light Water Reactors (LWRs). In particular, the Accident-Source Term Program Office (ASTP0) RES/NRC sponsored the development of a

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suite of severe accident phenomenology codes at RCL and SNL. These codes are  ;

intended to describe how a nuclear reactor core might degrade without adequate cooling and hence release radioactive fission products. The codes also follow the subsequent transport of the fission products from the damaged core to the environment if the containment fails or is bypassed. These codes therefore focus on the release' and transport of fission products and were applied to model selected severe accidents for six representative reactor designs. The results of this code application effort is reported in BMI-2104.1 Note that other activities sponsored by the NRC related to determining containment loads and containment performance during severe accidents are reported sepa-rately 2,3 and were not taken into account in Reference (1).

The methodology described in Reference- (1) has received extensive peer

.: review over the last several months and is also under review by the American Physical Society (APS). As a result of questions raised during the APS review a meeting" was held at NRC to develop an appropriate response to the ques-tions. Part of this response was to demonstrate that the ASTP0 suite of severe accident codes could be exported to an independent organization and that (by using similar input parameters and intercode data transfer) similar results to those reported in BMI-2104 could be obtained. BNL was selected by

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the NRC to be the independent organization and this informal report documents the results of this effort.

7 1.2 O_bpctive and Scope of Audit Calculation This effort was performed under a very severe time constraint. The meet- -

ing which . initiated this effort was held on November 20, 1984 A station blackout sequence at Surry was analyzed using the full suite of ASTP0 codes.

The objective of the effort is limited to demonstrating that each of the codes used in BMI-2104 can be made operational at RNL and that the results in --

BMI-2104 can be reproduced. However, we did have sufficient time to verify the appropriateness of code input parameters and intercode data transfer.

This is therefore a limited quality assurance (QA) audit of the results in BMI-2104 We have not h.ad sufficient time to 0A the codes and in fact we were specifically reouested to make no modifications to them and to set the in-ternal options (related to alternative models) as in BMI-2104 Validation ef-forts are described elsewhere. A number of inconsistencies related to data transfer between codes were found and these are discussed in detail later in this informal report.

1.3 Calculational Methods r

The codes used to analyze the station blackout sequence are identified in Figure 1.1, which indicates the relationship between the codes and also pro-u vides a brief description of the purpose of each code. The MARCH code calcu-L L lates in-vessel core degradation (BOIL subroutine), vessel failure (HEAD sub-routine), ex-vessel core debris / water interactions (INTER subroutine). All of these subroutines feed the containment building response model (MACE subrou-l tine). The core temperature history, as predicted by BOIL, is fed to the CORSOR code to calculate in-vessel fission product release. Core outlet gas -

and steam flow rates, as predict' ed by BOIL, are fed to the MERGE code to i

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calculate primary system thermal hydraulics. Output from CORSOR and MERGE are used in the TRAP-MELT code to predict primary system transport of fission pro-ducts. Fission product. output from TRAP-MELT feeds the NAUA code, which cal-culates fission product transport in the containment building. Note that NAUA gets thermal hydraulic data from the MACE subroutine of MARCH. Inconsistency

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between these codes was noted as part of the initial peer review of RMI-2104. .

l The flow rate predicted by PRIMP of MARCH is not consistent with the flow rate predicted by MERGE.

After head failure (HEAD subroutine) MARCH can model core debris / water (HOTDROP) or core debris / concrete (INTER) interactions depending on conditions in the reactor cavity beneath the reactor vessel. The thermal / hydraulic con-ditions in containment depend strongly on the made of ex-vessel core debris interactions. If extensive core / concrete interactions occur then the asso-ciated fission product release must also be calculated, which is done by using the CORCON code (core debris / concrete interactions) together with the VANESA code (fission product release from ex-vessel interactions). The user of the codes should ' explicitly ensure that the start of core / concrete interactions predicted in the MARCH code is consistent with start of interactions assumed in the CORCON/VANESA codes. In addition, the user must be mindful of the in-consistency of' using INTER to calculate the containment response (MACE) and CORCON/VANESA to calculate fission product release. This was also noted as part of the peer review of BMI-2104.

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coAE[cwdETE GOS FLc6J tifTER ACTIONS ' k CoAcon) code CofE/CONcAGE INTERACT)oNi Figure 1.1 Flow diagram of ASTP0 codes for application to Surry. .

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1.4 Calculational Procedure and Team The audit calculations were performed using the suite of codes described in Section 1.3. BNL staff involved in the audit calculation are identified in Table 1.1. The audit calculation was performed by three BNL staff members (refer to " Analyst" column) familiar with running the respective ASTP0 codes.

These calculations were then in turn checked by other BNL staff members (see -'

"QA" column) also familiar with the .various codes. Integration of the codes is an extremely important process and this was ensured by a BNL staff member separate from the analyst team and this process was also subjected to a OA review.

A station blackout sequence (TMLB') in the Surry plant was selected at the November 20 meeting as the basis of the audit calculation. It was de-

.cided3 a'fter the meeting to analyze a basemat (c) failure mode at BNL to en-sure comparison with a calculation in Volume V of BMI-2104 that used a consis-tent set of in-vessel and ex-vessel input parameters. The overpressurization failure mode (6) in BMI-2104 used primary system behavior (MARCH, CORSOR, MERGE, and TRAP-MELT) consistent with an c failure mode rather than the appro-priate 6 failure mode. Differences between the crimary system behavior for the basemat and overpressurization failure modes in BMI-2104 relate primarily to the amount of cladding that oxidizes in-vessel (59% clad reaction for the e failure mode vs. 93% reaction modeled for the 6 failure mode). The amount of

clad reaction principally influences the Te release. The fraction of clad reacted is controlled by MARCH input parameters and a 93% in-vessel reaction is certainly an upper estimate. Consequently, the overpressurization failure mode is based on an-upper bound in-vessel clad oxidation but uses fission pro-duct releases based on 59% clad reaction assumed for the e failure mode. The

influence of this inconsistency on the quantities of fission products released to the environment' is not great. However. . it was decided to base the audit calculations on the basemat failure mode because this was calculated in a con-sistent manner in BMI-2104 In Se'ction 2 we describe the audit calculation and each individual code is briefly discussed. Finally in Section 3 the results of the audit calculation are compared with the BMI-2104 results and some comments are provided.

Table 1.1. BNL staff participating in audit calculation Code Analyst 0A MARCH 2 R. Jaung J. Yang CORS0R R. Jaung H. Ludewig MERGE R. Jaung K. Perkins l _ TRAP-MELT R. Jaung H. Ludewig/W. Yu NAUA W. Yu H. Ludewig CORCON . G..Greene M. Khatib-Rahbar L VANESA G. Greene W. Yu l

l Code Integration H. Ludewig W. T. Pratt l

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2. AUDIT CALCULATIONS A station' blackout sequence resulting in failure of containment via base-mat penetration (TMLB'c) was selected as the basis of the audit calculation.

This sequence results in failure of all active emergency core cooling systems and containment heat removal systems. Initially, the secondary side heat sink (steam generators) boils dry. After the ultimate heat sink is lost the pri- ,

mary system begins to boil at the set point of the relief valves and the pri-mary systen water inventory 'is lost. Eventually, the reactor core is un-covered and it begins to heat up and degrade. Du' ring this process the in-vessel release of fission products begins. Without coolant injection the core will melt and slump into the bottom of the reactor vessel where it will thermally attack the lower vessel head. When the reactor vessel fails any residual primary system water is released to containment while the core mate-rials are released from the vessel. For this sequence, as the primary system depressurizes, the accumulators will inject water. The subsequent accident progression depends on the amount of water in the reactor cavity and on the mode of contact between the core debris, water and concrete.

The codes described in Section 1.3 were applied to the above secuence in a manner similar to that assumed in Volume V of BMI-2104. Input parameters and intercode data transfer have been carefully checked and an independent audit calculation performed. Each code used to model the above accident se-quence is described in the following sections. The sections also give a dis-cussion on the input model assumptions used in BMI-2104 and in some cases sug-gest alternative assumptions. During the audit calculation an inconsistency was found. for the TMLB'-c sequence as reported in BMI-2104. That analysis ,

(Section 2.1) assumes cuenching of the core debris immediately after vessel

2 8-failure and does not predict core / concrete interactions to begin until the cavity boils dry (297 minutes after reactor scram). However, fission product release due to core / concrete interactions is modeled separately from MARCH 4 using the CORCON/VANESA codes and this process was started at vessel failure

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] (157 minutes after reactor scram). Th'e airborne fission product masses in

. containment are modeled using NAUA and the~ fission products generated by -

core / concrete interactions were input at 297 minutes, which is consistent with i

3e MARCH calculation (but inconsistent with the CORCON/VANESA calculation).

This inconsistency was corrected in our audit calculation and its impact on the predicted released of fission products is not great for the late contain-ment failure ' ode (refer to Section 3).

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2.1 MARCH 2 The version of the MARCH code used in BMI-2104 Volume V for the station blackout sequence was 1.98. This version of the code is not available at BNL. At BNL version 2.111 was initially used. This was the most up-to-date version of MARCH operational at BNL at the beginning of the effort and also judged by BCL staff to be the most appropriate. However, difficulties were e experienced with this version of the code since it became clear that the core / concrete interaction models used in versions 1.98 and 2.111 were quite di fferent. A later version of MARCH was delivered to BNL- (namely, 2.151) which (when used with the same input) gave results compatible with MARCH 1.98. All the calculations reported in this document are therefore based on MARCH 2.151.

The input to the MARCH 2.151 calculation is identical to that used in BMI-2104 Volume V. A comparison of the results is shown on Tables 2.1 through 2.3. It is seen that the event times (Table 2.1) and most of the primary sys-tem and containment response parameters agree well. Minor disagreement occurs for the- core temperature, time at temperature and the fraction of zircaloy oxidized. These differences (although small) play a role in determining the amount of tellurium (Te) released during the melting phase of the accident.

The implication of this will be pointed out in the CORSOR discussion in Sec-tion 2.3 The input parameters to MARCH were reviewed and found to be consistent t

with plant construction details. In addition, the input modeling assumptions are reasonable given our current understanding of the progression of a core meltdown accident. The assumption that approximately 60% of the clad will oxidize in-vessel is rather higher than the current " industry" estimates but

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is certainly feasible. The rapid head failure time after core slump is also to be expected given the high primary system pressure for this accident se-quence. It should also be noted that mechanisms exist to depressurize the primary system (due to high temperature degradation of the primary system pressure boundary) prior to direct corium attack on the bottom of the vessel.

Such a possibility can be investigated using di f ferent MARCH inout _

assumptions.

The subsequent ex-vessel interactions of the core debris depends on the configuration of the region below the reactor vessel. In Surry, for a station blackout sequence, the reactor cavity would be relatively dry and hence core debris released from the bottom of the reactor vessel would not immediately contact water. However, as the primary system depressurizes (below 665 psig) the accumulators will inject water, which will eventually reach the core de-bris in the reactor cavity. At this point MARCH can model two possible modes of core debris / water / concrete interactions.

The HOTDROP subroutine in MARCH assumes that the core materials, on con-tact with the water, form fine particles which mix homogenously with the

. water. This configuration allows rapid heat transfer between the core debris f

and water, which rapidly cools the core debris and generates significant quan-tities c' steam. This. rapid steam generation causes a rapid pressure rise in containment. Eventually, if water is not supplied to the reactor cavity, the fission product decay heat will dry 'out the core debris. The core debris is then assumed to reheat (adiabatically in HOTDROP) until it melts, which allows transfer (in MARCH) from the HOTDROP to the INTER (core / concrete interaction)

- subroutine. After INTER begins to calculate core / concrete interactions fis- .

sion product release must also be calculated (refer to Sections 2.5 and 2.6).

Note that this adiabatic reheating of the core debris is a conservative as-sumption and if heat losses were modeled during this period the core debris may not reach its melting point. This assumption is built into the MARCH code and it is beyond the scope of the audit calculation to alter internal code structure. However, the above is the mode of core debris / water / concrete in-teractions assumed in BMI-2104 and some of the conservatisms inherent in the $

calculations should be noted. The second mode of ex-vessel core debris in-teraction is discussed below.

There is some experimental evidence that if water is poured onto molten core materials that a porous crust will form and thus prevent mixing of the water and core materials. The stability of such a crust-across the large sur-face area of a reactor cavity has been questioned, however, thc possibility can be modeled in MARCH by byoassing HOTDROP and assuming core / concrete inter-actions being immediately after vessel failure with an overlying water cool.

INTER is used to model core / concrete interactions and heat transfer ~from the core materials to the water is limited by film boiling. This alternative mode of core debris / water / concrete interactions was not assumed in BMI-2104 but is feasible and it would give different containment loads and ex-vessel fission product release characteristics than given by the HOTDROP assumption.

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Table 2.1 Accident event times (minutes)

Audit -

, (MARCH 2.151) BMI-2104 -

.i Steam Generator Dry 69.0 67.5 Core Uncover 97.25 95.5 Start Melt 118.5 118.3 Core Slump 143.5 146.3 Bottom Head Fail 155.0 157.3 Cavity Dry 213.3 214.9 Start Concrete Attack 287.3 289.9 Containment Fail 738.0 738.2

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Table 2.2 Primary system response Average Core Peak Core Frection Fraction Time (min) Pressure (psla) Temp. (*F) Temp. (F) Core Metted Clad Reacted E vent A* R** A B A B A 8 A 8 A B Core Uncover 97.2  %.5 2369 2369 669 669 674 675 0. 0 0.0 0.0 0.0 5 tart Melt 118.5 118.3 2366 2366 1926 1990 4130 4130 0. 0 - 0.0 .05 06 Core Slump 143.5 146.3 2363 2362 3762 3709 4130 4147 .580 .55 .38 .33 Botton Head Failure 155.0 157.3 2366 2368 3509 3820 - - .87 -

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    • 8MI-2104 Results a

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Table 2.3 Containment system response Compartment Compartment Sump Water Sump Water Reactor Cavity Steam Cond. on Event Time (min) ' Pressure (psla)- Temo.(*F) (Ibe) Temp. (*F) Water (Iba) Walls (Ib/ min)

A* 8" A .8 A 8 A '8 A 8 A 8 A- 8 Steam Generator 69 67.5 13.1 13.0 137 136' 3.43(4)"* 3.39(4) 138 138 0.0 0.0 1070 1051 Ory.

I Core Uncovery 97.25  %.5 28.7 28.8 219 219 2.07(5) 2.07(5) 1% 197 0.0 0.0 2446 2498 Start' Melt 118.5 118.3 26.2 25.7 211 209 2.44(5) 2.47(5) 200 200 0.0 0.0 1039 994

! Core Slump 143.5 146.3 23.0 22.5 199 197 2.65(5) 2.69(5) 198 198 0. 0 0. 0 754 684 Bottom Head 155.0 157.3 46.0 45.9 '253 253 2.80(5) 2.82(5) 199 198 171(5) 171(5) 11649 11810 Fatfure Cavity Dry 213.31 214.9 58.9 58.6 273 272 4.01(5) 4.06(5) 272 ' 221 0.0 0.0 1407 1405 g t

Start Core /

Concrete 281.31 289.9 46.2 46.0 253 253 4.52(5) 4.54(5) 226 225 0.0 0.0 634 653- 7 Interaction Containment 738.03 738.2 52.9 53.7 248 249. 4.86(5) 4.85(5) 228 227 . 0.0 0.0 74 0.0 l

Failure

  • Audit Calculations
    • BMI-2104 Results
  • "3.48(4) = 3.48 x 10*

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e 2.2 MERGE A revised version of the MERGE code was received at BNL in September 1984 and this version was used in this analysis. The advantage of using this ver-sion of the code over earlier versions is primsrily in the capability of rep-resenting the primary system by seven control volumes. In the interests of consistency these same seven control volumes were used in the primary system

, j fission product transport calculation to be discussed in Section 2.4 The I

seven volumes are connected in series and represent the following structures.

Volume 1 - Core Volume 2 - Core plate Volume 3 - Thin metal structure ' upper vessel internals Volume 4 - Thick metal structure upper vessel internals Volume 5 - Piping (pressure vessel to pressurizer)

Volume 6 - Pressurizer Volume 7 - Containment (sink)

MERGE -input consists of volume dimensions and connections to adjacent vol umes. The thermal / hydraulic input data was obtained fran the MARCH 2.151 I

calculation. Direct comparisons with BMI-2104 calculations become difficult at this stage because some of the above volumes were combined for the MERGE step. The results of the four-volume MERGE calculation were extrapolated in BMI-2104 to the seven-volume model used in TRAP-MELT (refer to Section 2.4).

We simply used the same seven control volumes in both MERGE and TRAP-MELT.

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2.3 CORSOR CORSOR input in the form of a core temperature history was obtained from the BOIL subroutine in MARCH, and initial fi ssion product inventories and power peaking factors were taken from BMI-2104 Volume V.' Table 2.4 shows the results of the CORSOR calculation. Fce Cesium (Cs) and Iodine (I) it is seen that essentially all the material is released during the melting phase of the _

accident. In the case of Tellurium (Te) 12'.8 kg are released out of an ini-tial inventory of 25 kg. In comparison, the values quoted in BMI-2104 show the same result for Cs and I. However, the BMI-2104 calculation released 9 kg of Te. This difference is to be expected since there are differences in the temperature history of the core and the fraction of zircaloy oxidized between our Audit calculations and BMI-2104 This implies that different quantities of Te will participate in the next steps of the calculation, i.e., transport

. within the primary system and during ex-vessel interactions of the corium with concrete following vessel failure.

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Table 2.4 In-vessel mass balance (CORSOR)

Element Initial Inventory (kg) Mass Released (kg)

During Core Melt ,_

Cs 146 144,0 I 12.15 12.0 Xe 260 256.21 HKr 13 13.20 Te 25.4 12.80

, Ba ,

61 10.73 Sn 262 98.25 Ru 215 1.53 Zr 179 .03

, Mo 155 '12.89 Sr 48 3.44 Aq 2750 1275.90 l Cd 173 134.24 l

l In 505 71.71 00 2 .79650 18.01 Zr (clad) 16454 1.88 Fe 6486. 73.47 f

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e 2.4 TRAP-MELT The version of TRAP-MELT used in this analysis was implemented at BNL in April 1984. It is written in FORTRAN 5 and allows for five states. The same seven-volume representation of the reactor coolant system used in MERGE and as described in Section 2.2 was used in TRAP-MELT. Thermal / hydraulic data deter-mined by MERGE was used as input for the various volumes and flows between vol umes. The fission product release rate, which acts as the source in this calculation, was determined by CORSOR.

This step in the calculation proved to be quite frustrating, in that the TRAP-MELT solution was unstable and diverged toward the end of the transient.

Different choices of time steps for thermal / hydraulic input and volume nodali-zation did not solve this problem. This numerical instability remains as an area of_ concern. It was decided to use only the stable part of the solution, since the instability only occurred at the end (approximately five minutes from the end) of the time frame of interest. The TRAP-MELT solution spans the time frame from beginning of core melt to bottom head failure. In order to obtain input for NAVA from TRAP-MELT, it was necessary to stop the calculation

[

prior to vessel failure and release all the remaining fission products, which had been calculated to be released during the last five minutes by CORSOR, as a puff. at the end of the TRAP-MELT calculation. In BMI-2104 this problem was l

handled by extrapolating the TRAP-MELT calculation to the time of vessel fail-ure. In this way mass conservation was assured, but the quantity of fission products deposited in the primary system will be slightly underestimated. It was felt that for the particular sequence beino considered, this approximation

. would have a small effect on the mass of fission products leaked to the en- -

vironment, since the containment failure time is almost 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after vessel

v failure. This allows a large amount of time for agglomeration and settling in the containment, and the details of the release at vessel failure are less important. In other accident sequences where the containment may fail ear-lier, this approximation would be even less valid.

By inspecting the results presented on Table 2.5 it is seen that this

~

- shortcoming in the calculation does not affect Csl and Cs0H, since their en- ,

tire mass is emitted by the time the calculation diverges. In the case of Te and aerosols approximately 2 kg will be emitted to the containment at the time of bottom head failure. The slight difference between the releases calculated by CORSOR and those calculated by TRAP-MELT for Csl and Cs0H are due to dif-ferent time nodalizations used in representing the release rate and subse-ouently carrying out a sum to determine the total released mass.

i.

l-l l

l' i

I l

l l

\ -

Table 2.5 In-vessel mass balance during fission product transport (TRAP-MELT)

Fission Released in Mass Mass **

Product Released During Transport Code Retained Retained Species . Melting (CORSOR) (TRAP-MELT) (Audit) (BMI-2104)

Cs1 24.61 24.67 21 21.8 Cs0H 147.78 148.22 126 128 Te 12.80 10.70* 9.50 7.6 Other. 1702.06 1700.03 1605 1612 (Aerosol)

  • 2.1 Kg of Te release as a puff at vessel failure.
    • Reproduced from Table 7.3 of BMI-2104 l

4 T

e

M 2.5 CORCON The CORCON code (Muir,6 et al.,1981) was used to calculate the ex-vessel attack of molten core debris on the basemat and this calculation drives the ex-vessel aerosol and fission product source term calculation. The CCRCON t

code is the state-of-the-art model for the analysis of the interaction between molten fuel and structural materials with concrete. Typical output of CORCON calculations are the concr'ete erosion rate, generation rates of concrete de-composition gases, and the core melt / temperature history. These three quanti-ties are necessary input to the VANESA code for the calculation of the ex-ves-sel release of fission products -and aerosols into the containment during a core melt accident.

The version of the CORCON code used in the all BMI-2104 accident sequence source term calculations was CORCON/M001 with two official Fortran update packages documenting changes made to the original code by Sandia National Lab-o rato ry. This code is heretofore referred to as CORCON/M001-C2. A third up-date which has not been widely distributed to date was identified during this audit calculation and was used in the BNL calculations, j The input to the CORCON code consists primarily of concrete composition data (user input), core melt composition (MARCH output), cavity geometry l (plant specifications /FSAR), surroundings temperature history (user input or 1

MARCH), initial melt temperature and time after SCRAM at start of core / con-crete interaction (MARCH), and melt / concrete surroundings radiative emissivity

'vs. time (user input).

~

Those inout variables which are " user input" are left to the discretion and scientific judgement of the user. In addition, there are no internal parametric model variations possible through input. The input -

for the Surry TMLB' CORCON/M001-C2 calculation is identical to that used in

BMI-2104 Volume V, with the following exceptions. The SNL CORCON calculation was started at 157 minutes after SCRAM at an initial core debris temperature of 1807K and a time-invariant surroundings temperature of 500K. The BNL core / concrete interaction was started at 287 minutes after SCRAM with a time-invariant surrounding temperature of 1373K and initial melt temperature of 1777K. This was done to be consistent with the BNL MARCH calculation results ,_

of initial melt temperature, surroundings temperature and start of core / con-crete interactions at 287 minutes after SCRAM (corresponding to the time INTER allowed core / concrete interaction to occur after adiabatic reheating from ouenching). Both of the above core / concrete ~ interaction calculations were performed at BNL and compared for differences. It was found that only minor diffe.rences were calculated in erosion rate, gas generation rates, and melt temperature vs. time.

The results of the CORCON/M001-C2 calculations are shown in Figures 2.1-2.4 as melt temperature history vs. time, integrated gas generation rates

in kg and moles vs. time, and vertical and lateral erosion depths vs. time, respectiv::ly.

i A more complete assessment of this computer code may be found in ORNL/TM-8842, Chapter V.7 l

l s

O r- -- ,. , .- ._

~

LAYER TEMPERATURES a.004 aanno. l LECESD o = HEAVY OXIDE LAYER

= METALLIC LAYER a LIGHT OXIDE LAYER e

siano-E acano-y O

E-isono- ,

secco -

7ano -

I tecoe

, iso :co aso ana zo 400 *so sao wo TIM EXSir) *ltf Figure.2.1 Melt temperature history vs. time: Surry TMLB' sequence.

O

t CAS GENERATION l

s 30000 LECEND o . co o=cO2 accom-l [0 200a0 -

2 1 sono-2 ianao-sono-l.,

p S ;;;_::

i. ,, 7 E: - _. /

iso zoo aso ano aso eao eso soo sso Truc(scc) to' Figure 2.2 Integrated gas generation dates (kg/) vs. time:

Surry TMI.B' sequence.

l l .-

l

l GAS GENERATION Y*

LEGEND D . CO O = CO2

=0 - . = n2

  • =H2O ase-g =0 -

d 2

$ 110-o 30 4 -

50-

!. __' '_~[ m

_ er: :___ _ : : - -

l LSo ELO 254 30.0 50 400 410 'JAO SS O I

Ts uinEC) *16 i

l I

l l..

Figure 2.3 Integrated gas generation rates (gm-moles) vs. time: Surry TMLB' sequence. .

,.ar. - - , . - -r--m- r v - - - - - - -

EROSION RATE -

30 43- - ; ; ;;2 : - -

,i 4D-3.5 - ,

2 LEGEND

  • o = AX1 AL EROSION o = RADI AL EROSION

{ so. .

u.

SD-g.

[

I. O IS0 ' 200 25 0 304 35 0 480 450 50 0 SS O TlMC(SEC) *10' 4

l Figure 2.4 Concrete erosion depths vs. time:

l Surry TMLB' sequence.

I f

2.6 VANESA In order to account for the ex-vessel release of aerosols and fission products into the containment during the thermal interaction between molten core debris and structural concrete, the VANESA model was used. The VANESA model is a mechanistic description of the aerosol generation and fission pro-duct release during ex-vessel core-concrete interactions. A more complete -

description of the VANESA model may be found in ORNL/TM-8842, Chapter VI.8 The VANESA model accepts as inp'ut the output of the CORSOR and CORCON computer codes.

From CORSOR, VANESA will get the mass of all species in the core melt in-vento ry. At the present state of development, VANESA can accept explicitly 32 species; other species are surrogates of one of the 32 species. Examples of this are as follows:

Gd, Eu, Pm + La on molar basis Np + Ce on molar basis Cd + Cs on molar basis In + Ag on mass basis 1.

In addition, it is required to input the inventory of Nb as the equivalent mass of Nb2 0s directly into VANESA in Subroutine ASSEMB, line 41. These vari-

~

ations on melt species input refl ect the ongoing development of the VANESA model as new species were identified as important during the Accident Source Tenn Reassessment Study.

From CORCON, VANESA receives the following information in time step in- -

tervals as speci fied:

P

oxide melt temperature (K) integrated gas release rates (kg)

H2 HO 2

C0 CO 2 maximum radial erosion radius (m)

SiO2 content of melt (kg) -,

The arrays of these seven input variables were transferred automatically be-tween CORCON and VANESA.

A detailed description of the output from VANESA has been documented

~

elsewhere (Powers,8 1983) and will not be repeated here (see Table 2.6). The most important of the output quantities for subsequent use in the calculation of the ex-vessel source term are:

1. aerosol' mass generation rate,
2. chemical composition of aerosolized mass,

-s

3. aerosol mean particle size and,
4. aerosol density i

,' The version of VANESA used in this audit calculation of the Surry TMLB'

! j-l; sequence as reported in BMI-2104 Volume V was the first exportable Fortran l:

version. This version was adapted from the original non-Fortran version writ-ten at SNL and used in the BMI-2104 study. The code was first received at RNL on November 15, 1984 It was successfully compiled and a sample problem exe-cuted on November 19, 1984. In the process of familiarization with the code, several minor transcription errors were identified in the Fortiran version by -

l the SNL and BNL staff and were corrected.

l

The Surry TMLB' audit calculation was then run with the validated input from MARCH, CORSOR, and CORCON as previously specified. The results of this calculation were tabulated for each time step, listing the species in the aerosol, the source rate, oxide melt temperature, aerosol density, and aerosol mean particle size. One exception that should be explicitly mentioned is the Cs20 group. Recall that this species actually consists of Cs20 and Cd. A ',

breakdown of this group for the first four time steps is given in Table 2.7 on the basis that Cd actually comprises 98.7% of this category on a mass basis.

Also indicated in Table 2.7 are the source rates in gm/s for the times during which a water pool existed over the core melt. In these cases, SNL calculated a decontamination . factor . (DF) by which the source rate was diminished.

BMI-2104 Volume V reports the source rates with the DF included; Table 2.7 lists both sets of results for the BNL audit calculations. Comparison of the results obtained in this calculation to those reported in BMI-2104 Volume V demonstrates excellent agreement on a quantitative basis.

t O

l m m ,7 - ---vy----- - - + - y p % v ,--- S-- w,wy --w s

Table 2.6 Output from the VANESA model (ORNL/TM-8842)

, krosol Properties  !

1. Density of aerosol material (g/cm 3)
2. Maan aerosol particia size (um)
3. Mass flux of aerosols (g/s)
4. Aerosol concentration at STP (g/m3 ) 3 -
5. Aerosol concentration in cavity (3/m )

. Aarosol Composition

1. Fission Products (mass percent Cs , I, Te , Ru , Sb , Mo , Sr , Ba ,

U, Pu, Ce, La, Nb) '

2. Concrete Constituents (mass percent Na , K, Al, Si, Ca , Fe)
3. Fuel and Structural Materials (mass percent Fe, Ni, Cr, Ma, Sn, Zr, U) ,

, Kinetics Data

1. Source of Release (sparging, evaporation, mechanical)
2. Rata limitation (surface area, time, mass transport, or chemical kinetics)
3. Vapor phase speciation Melt Composition
1. Change caused by aerosol formation
2. Change-caused by metal oxidation
3. Change caused by concreta melting Permanent Gas Characteristics
1. Composition (volume percent CO, CO2, H2, H20, OH, O. H)
2. Flux (moles /s)
3. Superficial velocity (m/s)

Y I

t e -

4

_ - . . . , , , , - - . ~ - -

Table 2.7 Breakdown of Cs20 grouping: Surry TMLB' t (sec) 0 1200 2400 3600 i

Cs20 (%) 1.26 .52 41 .36 Cd (%) 95.56 39.34 30.92 27.21 Source -rate (gm/s) w/DF* 1. 5 11.2 40.2 66.5 Source rate w/o DF 4. 4 21.3 53.6 72.4 DF 2.947 1.903 1.334 1.088 l

  • DF - Decontamination Factor i

5 k

j, s

L. .

l

s 2.7 NAUA The NAVA code used in this calculation was transmitted to RNL in December 1984, and includes models for diffusio-phoresis and homogenous riacleation.

This final step of the calculation takes input from TRAP-MELT and VANESA in the form of fission product aerosol source, and from MARCH 2.151 for the con-tainment building temperature, steam addition rate and leak rate. The in-ves- -

sel release is supplied by TRAP-MELT and it begins soon after the start of core melting and continues through to the time of bottom head failure,_ which consists of a puff release of Te and aerosol as described in Section 2.4. At the time of bottom head failure a 1-arge amount nf water and steam is also added to the containment (Table 2.3). The core debris, which is on the cavity floor is assumed to be cooled to the water temperature at this time. During the next phase of the accident the water in the cavity boils away and the core debris eventually drys out. No fission products or aerosols are released dur-inq this re-heating phase. The core debris is assumed to become sufficiently hot to attack the concrete basemat at 287 minutes (approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fol-lowing bottom head failure). This attack is computed to continue until the i

core debris penetrates the concrete basemat and containment failure occurs.

The aerosol source term description is represented by two modes. The j first1 mode represents the source from the primary system into the containment

?

!u~' and the second mode represents the source from core / concrete interaction. The first mode stops after 8945 seconds, which corresponds to the time when the TRAP-MELT calculation dive: ged, and is six minutes before the bottom head fails. The second release mode at that time represents the puff release cor-responding to the unreleased fission products during the melting phase. There -

is 'no aerosol source from the time of bottom head failure until the cavity 6

9 I

dries out (17200 seconds). At this time the core / concrete interaction source starts and continues up to basemat failure. The release rate (gm/sec), frac-tion of -aerosol composition and aerosol size description are determined by CORCON/VANESA. The fission product input to VANESA corresponds to tti mass 1

leaving the Reactor Pressure Vessel following bottom head failure. In the 8

~

calculation described in Section 2.6 this mass was based on fission product ,

masses reported in BMI-2104 Volume V. The masses released in the Audit calcu-lation are different (see Table 2.8) and this difference was accounted for by adjusting the release fraction in the calculatior, described in Section 2.6.

r These adjusted release fractions, together with the release rates and aerosol size description from Section 2.6, form the input to NAUA for the core / con-crete interaction.

Table 2.8 Input to VANESA -(kg)

Species Audit BMI-2104 I

I .15 .1 l- .Cs 1. 8 .7 l'

l Te 12.6 16.4 f

l.>

Table 2.9 shows a mass balance for the containment calculation. It is seen that the bulk of the material is deposited on horizontal surfaces (sedi- -

L mented deposition) with 'a smaller fraction on vertical surfaces (diffusion y ..-m, -_y i m,- . - - . . - . , , , , - . _ . , , . , , . , , . , , , . _ _ , . . , _ , - ..-.._,,-...___-.-_.__--_m

J s

)

- 34-deposition) . "

Other" in this table refers to aerosols leaking out of the primary system and those generated in the core / concrete interaction, where as in Table 2.5 "other" refers to aerosols generated only during the in-vessel release phase.

I h

J I

f f

' :j i

e m - ----- e e'T mF--"'py--r-m-m -v-w 4-9^ r,w w e w ^n+ 1 --

+w v- --ye?--w- - - - ----vv-9 y, .* v_ - --~ =-WmTy

~

Table 2.9 Ex-vessel mass balance (NAUA)'

4 Fission Sedimented Diffusive Airborne '

Total in Leaked Total Product Deposit Deposit Containment Species (kg) (kg) (kg) (kg) (kg) (kg)

Csl 3.397 0.628 0.0019 4.027 0.0403 4.067 Cs0H 17.735 3.264 0.0012 21.0 0.0261 21.026 Te 4.971 0.541' O.389 5.901 1.790 7.691 Other 663.526- 67.719 43.415 774.660 275.0 1049.660 e

S I

e 4

1 e

3.

SUMMARY

Shown on Tables 3.1 and 3.2 is a summary of the results of the audit cal-culations. The tables show thi final distribution of fission products after the accident. A direct comparison with values taken from BMI-2104 is made, and for Cs! and Cs0H the agreement is good. In both calculations approxi-1 mately 15% of the Csl and Cs0H remain in the containmen't, 85% in the reactor .-

coolant system and a small fraction leaks to the environment. The estimated

fractions leaked to the environment agree closely.

In the case of Te the distribution of fission products is somewhat dif-ferent between the two calculations. The audit calculations predict a larger fraction of fission products retained in the primary system that is calculated in BMI-2104 This is due to the higher Te release in the audit calculation during the in-vessel core degradation and thus more Te is available for reten-tion in the primary system. During ex-vessel corium/ concrete interactions the split between sedimented retention on containment surfaces and retention in

- the melt is also different between the two calculations. A larger fraction is predicted to be retained on surfaces in the audit calculation and a smaller

, fraction is retained in the melt than in BMI-2104 These dif ferences are ons-sibly due to differences in the CORCON/VANESA calculation. It has been pointed out in earlier. sections. of this report that there are inconsistencies between the NAUA and. CORCON/VANESA calculations in BMI-2104 These inconsis-tencies were eliminated .in.our audit calculations resulting in the differences noted below:

O

Audit. BMI-2104 CORCON/VANESA NAUA CORCON/VANESA NAUA Start of core / concrete 287. 287 157.3 282 interaction (min)

Initial melt temperature 1777 1777 1807 1807 4

(*K)

Temperature of surroundings 1373 1373 500 500

(*K)

Initial Te mass (kg) 12.6 12.6 16.4 '6.4 i

From the above table it is clear that the CORCON/VANESA calculations in BMI-2104 were started at the point of vessel failure rather than at the point of debris dryout. In the audit calculation, the CORCON/VANESA calculations was started at the time of debris dryout. However, both calcula61ons predict that essentially the same fraction of Te leaks to the environment. Thi s agreement is due primarily to the long ' time before containment failure for this sequence. For a sequence in which the containment function is compro-

!3 mised at an earlier time the agreement may not be as close.

A final check between these two calculations is a comparison between the airborne and leaked mass as a function of time. Figures 1 and 2 illustrate

this comparison. The audit calculation (Figure 1) and the BMI-2104 calcula-tion (Fiqure 2) show the same trends and to a large extent the same magni-t tudes. The airborne mass peaks following vessel failure due to steam in the atmosphere. The peak values agree well between the calculations. A valley develops as the steam condanses, and serosols due to core / concrete interac- -

l tions have not been generated yet. The valley in the audit calculation is

- 1 deeper than in the BMI-2104 calculation. However, the~ remainder of the curve starting at the beginning of core / concrete interactions. agrees well with the BMI-2104 curve. Following containment failure the BNL curve predicts a sharp decrease.

The audit calculation has further shown that some inconsistencies exist in BMI-2104 regarding inter code data transfer. For this particular suite of codes this is a significant difficulty and great care must be taken to ensure

.i

') code compatibility. However s for the particular sequence and failure mode considered in the audit calculation the inconsistencies did not strongly in-fluence the predicted release of fission products. Obviously for any future use of the ASTP0 codes to a specific reactor plant code compatibility should be ensured and also the other input model assumptions noted in this report (but not exerci' sed in BMI-2104) should also be factored into a comprehensive analysis.

In ' summary, this short-term audit calculation has demonstrated that the

)

suite of codes developed under ASTP0 sponsorship can be successfully exported to an independent organization and that the results in BMI-2104 can be reproduced.

u I

t e

e

-39 .

O

=

_i . . . . . . i i e .

LEGEND.  :

O - RLrborne -

b O - Leaked -

'O _ l

-= - _=

g  :

- 1 -

"O'- '

i

_=

O m 2- co -

y cF' ~ -

C

  • a':

._.- O g

- 1 1 g -

g -

- O . -

v1 o O- _.

-E ~ j j a .-

~

~

.., ,O -

O -

' E- ,,

O_ _

~5 -

0 5

' ~

O_

_O_

Start Bottom 2 Core. Head Start core / concrete -

interaction Containment Building Melt Failure Failure e

C 1 1

-2 > > > > >

> .. i 10 10' TLme (MLn)

Figure 3.1 Airborne and leaked masses (Audit calculations).

40 "o .

n. RIRBORN LERKED - -

t

  • a _

i .

"o _

cm :_

s _

en .

(n ~

C .

r

_fo_

C -:

H -

o -

H  :

"o _

"o _

~

5! , . . . . . . . . i '

2 10 10' 2*10' .

TIME, MIN Figure 3.2 Airborne and leaked masses (BMI_2104 Volume V)

(Reproduced from Figure 7.34 of Volume V of BMI_2104)

~

Table 3.1 Overall mass balance (kg)

Species Total in Leaked Mass Retained Mass Retained Total Mass Containment Mass in Primary in Melt System Csl 4.027 0.0403 20.72 - 24.79 .

Cs0H 21.00 -0.0261 125.99 -

147.02 Te 5.901 1.79 9.51 8.2 25.4 i

J

Table 3.2 Distribution of species after accident (BNL fractions

, based on total mass from Table 3.1)

! u. .

Fraction of Core Inventory f

I In Containment In Melt In Primary Leaked 1 Species Ex-Vessel System Audit BMI-2104 Audit BMI-2104 Audit BMI-2104 Audit BMI-2104 l

Cs! .16 .15 - -

.84 .85 1.6(-3)* 2.8(-3)

.) Cs0H .14 .14 - -

.86 .86 1.8(-4) 1.7(-4) i i *

Te .24 .19 .31 43 .38 .30 7.2(-2) 8.1(-2)
  • 1.63(-3) = 1.63 x 10-3

+

6

,, . .____ ,___ - - - - - -+------r - - - - +

4. REFERENCES
1. J. A. Gieseke, et al., "Radionuclide Release Under Specific LWR Accident Conditions ," BMI-2104, Vol umes I-VI,1984
2. " Containment load Working Group Estimates of Early Pressure and Tempera-ture Loads Resulting from Steam Spikes and Other' Accident Phenomena Within First.Few Hours After RPV Failure," NUREG-1079, Draft,1984.
3. NUREG-1037.  ;

4 Silberberg, M. to D. Ross, "Results from Meeting with Contractors," NRC Memorandum, November 29, 1984.

5. 'nning, R. (BCL) to M. Silberberg (NRC), Letter dated November 30, 1984
6. iui r, J. F. , Col e, R. K. , Corradini , M. L. , and Ellis, M. A. , "CORCON-M001: An Improved Model for Molten Core / Concrete Interactions," SAND 80-2415, 1981.
7. Greene, G. A., " Status of Validation of the C0RCON Computer Code," Chap-ter V in " Review of the Status of Validation of the Computer Codes Used in the NRC Accident Source Term Reassessment Study," 0RNL/TM-8842, November 1983.
8. Powers, D. A. and Brockman, J. E., " Status of VANESA Validation ," Chapter VI in " Review of the Status of Validation of the Computer Codes Used in the NRC Accident Source Term Reassessment Study," ORNL/TM-8842, November 1983.

l l

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