ML13196A421

NRC Staff Answer to Motion for Summary Disposition of Contention 4B - Attachment 4B-E

ML13196A421
Person / Time
Site:	Seabrook
Issue date:	07/15/2013
From:	Atomic Safety and Licensing Board Panel
To:
SECY RAS
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50-443-LR, ASLBP 10-906-02-LR-BD01, RAS 24821
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NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1)

License Renewal Application NRC Staff Answer to Motion for Summary Disposition of Contention 4B ATTACHMENT 4B-E

A DIRECT COMPARISON OF MELCOR 1.8.3 AND MAAP4 RESULTS FOR SEVERAL PWR & BWR ACCIDENT SEQUENCES M.T. Leonard, S.G.Ashbaugh R.K. Cole, K.D. Bergeron K. Nagashma ITS Corporation Sandia National Laboratories- NUPEC 8015 Mountain Rd. Place, Ste 210 P.O. 5800 17-1,3-chome Toranomon Albuquerque, NM 87110 (USA) Albuquerque, NM 87185 (USA) . Minato-ku, Tokyo 105 (JkpAN)

(505) 254-1005 (505) 844-2507 (+8 1) 3-5470-55 16 ABSTRACT_.-- - nuclear reactor vessel, supporting coolant piping, steam

.. -. - generators,. and containment structures are represented via This paper presents a comparison of calculations of user input to either code can cause major differences in severe accident progression for several postulated calculated results. To m e r complicate matters, the accident sequences for representative PWR and BWR format and nomenclature used to present results to the nuclear power plants performed with the MELCOR 1.8.3 code user differs between the two codes for some and the MAAP4 computer codes. The PWR system parameters. As a result, it can be difficult to determine examined in this study is a 1100 MWe system similar in whether the two codes calculate different values for the design to a Westinghouse 3-loop plant with a large dry same parameter because, in fact, the same parameter containment; the BWR is a 1100 MWe system similar in can represent slightly different quantities within each design to General Electric BWR14 with a Mark I code.

containment. A total of nine accident sequences were studied with both codes. Results of these calculations are With these challenges in mind, a systematic effort compared to (a) identify major differences in the timing of was made to compare results of calculations performed key events in the calculated accident progression or other with both computer codes for five severe accident important aspects of severe accident behavior, and (b) to sequences in a representative BWRl4 - Mark I identify specific sources of the observed differences. containment system and four accident sequence in a representative 3-loop (Westinghouse) PWR with a large I. INTRODUCTION dry containment. The calculations were performed with the MAAP4 (version 4.0.2) and MELCOR 1.8.3 computer The MELCOR and MAAP2 computer codes are used codes. This paper summarizes the findings of this by many organizations world-wide to calculate the. comparison effort.

response of commercial nuclear power plants to postulated accidents that invoIve substantia1 damage to 11.

SUMMARY

OF WORK PERFORMED reactor fuel (i.e., severe accidents). Although both codes are designed to address the same general problem (i.e., the MAAP4 and MELCOR 1.8.3 calculations were transient response of nuclear reactor systems to severe performed for the accident sequences shown in Table 1.

accidents), the modeling approach used to represent some The scope of the current comparison was limited to important phenomena and the level of detail with which calculated results related to severe accident behavior.

certain models are developed differ substantially between Particular emphasis was placed on early thermal hydraulic the two codes. As a result, differences in calculated behavior (specifically, factors governing the depletion of results are often observed. the primary coolant system inventory), in-vessel and ex-vessel core melt progression, and resulting containment However, differences in results of MAAP and response. Calculated results regarding fission product MELCOR calculations are also frequently observed due to release from fuel, deposition in various factors unrelated to the inherent differences in their reactor/containment systems and ultimate release to the modeling approaches. For example, seemingly slight environment were @ examined. A-differences in the way a large complex system such as a m8TP1BUT1ON OF TwfS DOCUMENT IS U N L f M m * - -

U P 1 This work was supoorfed by the United-States Deparfrnent of Energy under

. Contract DE-ACO4 - 94AL85OQO.

-- _ 1 - ---- - - - - - ---_ ---

DISCLAlMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States.Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied. or assumes any legal liability or responsibility for the accuracy, completeness, or use-fulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any spe-cific commercial product, process, or service by trade name, trademark, manufac-,

turer, or otherwise does not necessarily constitute or imply its endorsement, recom-mendation, or favoring by the United States Government or any agency thereof.

The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

$1 I"

Table 1. Accident Sequences Examined with MELCOR 1.8.3 and MAAP4 I

PWR: BWR.

0 Station blackout with an induced reactor coolant 0 Station blackout in which steam-driven coolant pump seal LOCA (TMLB); injection systems operate until battery (dc) power is 0 Large break LOCA with failure of emergency coolant exhausted (TB);

injection and containment sprays in the recirculation 0 Transient with failure of decay heat removal, high-mode (AHF); pressure injection and automatic depressurization 0 Large break LOCA with failure of emergency coolant (TQW; injection and containment sprays (ADC) Transient with failure of decay heat removal, high-0 Small break LOCA with failure of emergency coolant and low-pressure injection (TQUV);

injection and containment sprays in the recirculation 0 Transient with failure of containment heat removal mode (&E); (Tw);

0 Large break LOCA with failure of all coolant injection (AE).

I III. COMPARISON OF PWR CALCULATIONS The larger differences in the time to core Results of the MAAP4/MELCOR comparison for the . uncovery shown in Table 2 for the two large-break LOCA four PWR accident sequences are described below. For sequences appear to be the result of at least two significant brevity, a detailed discussion of the comparison is differences in the way in which emergency coolant provided for one representative sequence, station injection flow into the primary coolant system was blackout. However, this discussion is preceded by a modeled. These differences are controlled primarily by summary of major findings of the comparisons for all four user input, and do not appear to be the result of code sequences. models.

A. Summary of Comparisons for All Sequences 2. The time required for vessel breach to occur after large quantities of debris relocate into the lower head A summary of the calculated timing of key events for (i.e., after lower core support structure failure) is each of the four PWR accident sequences is given in calculated to be much shorter in MELCOR than in Table 2. This information is presented both in terms of MAAP. This result arises from significant differences the calculated time between major events as well as between the two codes' models for debris heat transfer differences in the cumulative time to an event from the within the reactor vessel lower head, and for structural start of the accident. failure of the lower head. In MELCOR 1.8.3, vessel failure is calculated based on a relatively simple thermal penetration model; i.e., failure is assumed to occur when The following general observations can be made from this the temperature of a penetration (if modeled) or the inner information and from a review of the individual surface of the lower head reaches a user-specified calculations by means of pIotted variables. temperature. In W 4 , vessel failure is calculated based on a cumulative damage (ie., Larson-Miller creep

1. With the exception of the large-break LOCA rupture) failure model. As described below (Section B), a sequences (AHF & ADC), the early hydrodynamic similar model has recently been installed in MELCOR, response of the system (i.e., the time required for the which, ifactivated, produces a time to vessel breach much primary coolant inventory to be depleted to the point that closer to the MAAP result.

the top of active fuel is uncovered) is shown to be in good agreement between the two codes. This suggests that the factors influencing the overall mass and energy balances of the primary coolant system prior to the onset of core damage are calculated in a similar manner by the two codes. This was confirmed by a closer examination of plots for several calculated parameters.

Table 2. Summary of Calculated Timing of Major Accident Events - PWR Sequences Accident 1 Time to Core Time to Failure of Time to Vessel Time to Sequence All times in seconds i Uncovery Lower Core Breach Containment Support Structure Failure TMLB Time between MAAPj 8,501. 5,396. 3,396. 83,923.

events MELCOR i......................................................................................................................

7,726. 4,956. 64. .......................................

126,238.

Cumulative MAAPf 8,501. 17,225. 101 ,148.

time from start MELCOR i 7,726. 12,746. 138,984.

AHF Timebetween MAAP: 4,221. 3,846. 10,259. 58,651.

events MELCOR 3,211. -.................................................................................

................................................................................................................ 9,168. 67. .......................................

136,809.

Cumulative MAAP: 4,22 1. 18,326. 76,977.

time from start MELCOR 3,211. 12,446. 149,255.

ADC Timebetween MAAPf 1,699. 3,114. 8,80 1. 102,441.

events MELCOR i 128. 3 ........

359 54. 203,094.

............................................................................................................................................... 2 :...............................................................................

Cumulative MAAPi 1,699. 13,614. 116,055.

time fiom start MELCOR 128. 3,541, 206,635.

SzHF Time between MAAP; 13,378. 10,747. 11,860. 47,620.

events MELCOR i 14,416. 22,901. 60. 143,875.

Cumulative MAAPf 13,378. 35,985. 83,605.

time from start MELCOR 1 14,416. 37,377. 181,252.

An additional contributor to the observed MAAP4 calculations predict the formation of a quenched differences in the calculated time to lower head failure is debris bed immediately after vessel breach; this result the way in which heat transfer between core debris and occurs independent of whether water exists in the cavity residual coolant in the reactor vessel lower head is prior to vessel breach (as in sequences SZHFor AHF) or modeled. The MAAP4 model operates on a conceptual arrives coincident with debris ejection (as in TMLB).

• picture of core relocation (from above the lower core Subsequent increases in containment pressure are support structure into the lower head) that is based on a governed primarily by ensuing steam generation in the contiguous pour (or jet) of molten material through a pool cavity. In contrast, the MELCOR calculation does not of water. Jet breakup and material fragmentation provide predict a quenched debris bed after vessel breach for any significant cooling of debris. MELCOR 1.8.3 provides an of the sequences. The rate at which containment pressure optional model for transient "eat transfer (i.e., during increases after vessel breach in the MELCOR calculation relocation into the lower head), however, this model was is governed primarily by gas generation resulting fiom not active in the present calculations. The default corium-concrete interactions.

(operating) model only accounts for debris heat transfer after material has settled onto the inner surface of the B. Specific Results for Sequence TMLB lower head and formed a stable debris bed. The net result of this difference is a significantly lower average debris The following provides more detailed information on temperature in the MAAP4 calculations than in the calculated results for one representative sequence, station MELCOR 1.8.3 calculations when lower head dryout blackout (TMLB).

occurs; this causes a delay in lower head failure in the MAAP4 caIcuIations because the debris must first 1. Early Thermal Hydraulic Response increase in temperature before it can challenge the lower head structure. MAAP4 and MELCOR calculate a very similar thermal hydraulic response of the primary coolant system

3. Finally, a large difference in the time at which prior to vessel breach. In particular, the primary coolant containment over-pressure failure occurs is indicated for pressure history and coolant inventory depletion all of the sequences in Table 2. This is due primarily to characteristics are in very good agreement. As shown in differences in models in the two codes for heat transfer Figure 1, both codes predict a sharp, but temporary, behveen core debris that emerges from the reactor vessel decrease in primary system pressure at 2700 seconds and water in the reactor cavity. Specifically, each of the when the reactor coolant pump (RCP) seal LOCA occurs.

System pressure subsequently increases to the pressurizer the Henry-Fauske critical flow model to calculate mass relief valve setpoint and remains at that level until vessel loss. MELCOR also calculates local fluid void fraction failure occurs, The only significant discrepancy in the by comparing the elevation of the opening in the primary calculated pressure response is the time at which the rapid system to the swelled level in the local control volume.

depressurization accompanying vessel breach occurs. The MELCOR then applies analytic fits to the Moody critical reasons for this difference were discussed in the summary flow tables to calculate mass loss.

section above.

The difference in critical flow models between the two codes is not likely to be responsible for the observed difference in coolant flow rates through the ruptured RCP seal or the relief valve. Rather, the different ways in which the elevation of the openings in the primary system are defined and the calculations of Pmurucr relief valve cycles swelled coolant level are calculated produces different estimates of local fluid conditions.

Rcactorvase1 bruch

2. In-vessel Core Damage Behavior Many similarities are also observed in the initial stages of in-vessel core melt progression. Initial fie1 heat-up rates are similar and subsequent core melting and 0 20 40 60 80 100 120 140 material relocation produces a similar level of cladding (Zircaloy) oxidation prior to large-scale debris relocation into the lower head. After lower core support structure failure, however, a relatively large difference in the cumulative mass of hydrogen generated is observed in the Figure 1. Reactor Vessel Pressure (PWR - TMLB) two calculations. This difference is created when rapid metal oxidation occurs in the MELCOR calculation as debris relocates into residual water in the lower head The rate at which the primary coolant system following failure of the lower core support structure.

inventory is depleted i s also in good agreement between This increment is not observed in the W 4 calculation.

the two codes. This results in reasonably close agreement The extent of clad oxidation calculated by each code can in the time at which the reactor vessel water level be inferred from the cumulative mass of hydrogen decreases to the top of the active fuel, and the onset of generated in-vessel as shown in Figure 2.

core damage. Some discrepancies in the details of the primary system inventory, depletion characteristics are obseryed, however. For example, substantial differences are observed in the calculated flow rates of coolant through the two leak paths from the primary system, i.e.,

the RCP seal LOCA and the pressurizer relief valve.

MELCOR calculates more coolant mass discharged through the RCP seal LOCA than W 4 ; however, this difference is balanced by MELCOR calculating a smaller 400 loss of coolant through the pressurizer relief valves than 550 W 4 . Given the good agreement in important 300 boundary conditions for these calculated parameters (e.g.,

C 250 the two codes calculate a very similar pressurizer water 200 level), the most likely explanation for these differences is 150 I1 100 that they are caused by differences in the way the two 50

--C Y A W 4 codes calculate fluid (donor) conditions at a break (or YClCOR 0

relief valve) location. MAAP4 first compares the 0 20 40 60 80 100 120 140 Time (1OSsc) specified elevation ofthe opening in the primary system to the swelled-up water level in the portion of the system containing the opening. This establishes the local fluid void fraction. It then applies a correlation (curve-fit) for Figure 2. Total Hydrogen Generation (PWR - TMLB) f

The modeling options used in the MAAP4 head are higher in the MELCOR calculation than in the calculation related to debris formation and transport into MAAP4 calculation.

the lower plenum are not known and, therefore, we can only speculate on the specific cause of the observed Second, the models used to calculate when difference in in-vessel oxidation. However, there are structural failure of the lower head occurs are different in several hndamental differences between MAAP4 and the two codes. In MELCOR 1.8.3, lower head failure is MELCOR related to late-phase material relocation which calculated based on a thermal penetration model; Le.,

would produce such an effect. In particular, the MAAP4 failure is assumed to occur when the temperature of a models are based on a conceptual picture of late-phase penetration (if modeled) or the inner surface of the lower material relocation that emphasizes the formation of a head reaches a user-specified temperature, typically molten pool above the lower core support structure 1273K. In W 4 , vessel failure is calculated based on a (similar to the one which formed in the TMI-2 accident). creep rupture (i.e., Larson-Miller) failure model. The This molten material subsequently relocates in the form of combination of higher calculated debris temperatures at a jet, of molten material into the lower head. The the inner surface of the lower head and the different lower process of debris bed formation within the lower head head failure model in MELCOR resulted in the shorter involves the breakup of this molten jet, and the relocation time to vessel breach.

of collapsing solid materials into several separated layers of particulate debris, molten metallic components and A Larson-Miller creep rupture model has

partially-frozen ceramic debris. The resulting material recently been implemented in MELCOR, although it was geometry limits the extent to which unoxidized metallic not active in the present calculations. A sensitivity components are exposed to steam generated as a result of calculation for the TMLB accident sequence performed debris relocation and cooling in the lower head, thereby with this new model resulted in a delay in the time to limiting hydrogen generation. In contrast, the geometric lower head failure of nearly 3000 seconds, bringing the picture represented by MELCOR can be thought of as a MELCOR result within approximately 400 seconds of the relatively open lattice of particulate and conglomerate MAAP prediction. That is, much closer agreement debris within which unoxidized metals may exist. When behveen the two codes can be achieved when similar the core water level decreases to very low elevations in modeling approaches are used.

the core, the rate of metal oxidation is limited mostly by the rate at which either (downward-directed) radiation heat transfer from the core or the relocation of hot debris 3. Ex-vessel Debris Behavior into the water pool in the lower head generates sufficient steam to oxidize exposed metals. Thus, when lower Among the more important differences in the hvo support structure failure occurs, the large amount of steam calculations is the containment pressure history. With the produced can result in a brief, but significant, increase in exception of the rime at which the prompt rise in oxidation. containment pressure accompanying vessel breach occurs, the very early containment pressure response (described Failure of the reactor vessel lower head occurs later) is quite similar in the two codes. However, the shortly following failure of lower core support structures long-term response is quite different. MAAP predicts in the MELCOR calculation; in contrast, W 4 does over-pressure failure to occur 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> after the start of not predict vessel breach to occur until approximately one the accident; MELCOR predicts failure to occur nearly 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after lower support structure failure. There are two hours later. The cause of this large difference in timing reasons for this substantial difference in time. First, the can be traced to fundamental differencesin models for differences in the way the two codes model late-phase heat transfer between debris that emerges from the reactor material relocation (described above) result in vessel following lower head failure and water in the significantly different debris temperatures within the reactor cavity. This difference is observed in the lower head. The W 4 model represents the formation calculations for three of the four P W R accident sequences of a debris crust against the inner wall of the lower head, (sequence ADC being the only exception.)

which partially insulates this structure from the molten ceramic material. The particulate debris bed that forms In both TMLB calculations, the cavity is dry above the molten pool is at least partially quenched by when debris first emerges from the reactor vessel.

residual water. While MELCOR also calculates debris Therefore, the initial mass of debris that arrives on the cooling at upper elevations of the debris bed, it does not cavity floor is very hot (Le., approximately 2600K in explicitly represent the formation of an insulating crust on MELCOR and 2100K in MAAP4). However, nearly the surface of the lower head. As a result, debris 100,000 kg of water enters the cavity very soon after temperatures at the bottom of the reactor vessel lower vessel breach. This water is discharged from the

accumulators as the primary coolant system depressurizes 4. Containment Pressure Response following vessel failure. The coincident release of debris and water complicates a direct comparison of models Differences in the thermal state of core debris related to debrislwater heat transfer between the two codes released to the containment in the two calculations allows because different heat transfer models are exercised when different processes to control the calculated containment debris falls into water versus water falling onto an existing pressure response. In the MAAP4 calculation, increases debris bed. in containment pressure immediately following vessel breach (shown in Figure 4) are totally governed by steam Nevertheless, the following observation can be generation in the cavity and the lower compartment (i.e.,

made. The MELCOR model for the TMLB accident cooling of two quenched debris beds). Over this time sequence did not contain any specific input for the FDI period, containment pressure is calculated to increase at a Package. Therefore, heat transfer between debris rate of approximately 40 kPa/hr. Containment pressure discharged from the reactor vessel and water in the response over the same time period in the MELCOR containment is governed exclusively by models that focus calculation is governed only partially by the evaporation on boiling heat transfer at the surface of a stable debris of water in the cavity and containment; a significant bed b; a quenched debris bed can only be attained if the portion of the debris internal energy and decay heat are coolant can penetrate the surface of the debris bed (a consumed in corium-concrete interactions in the cavity.

process that is subject to debris bed flooding limitations - Therefore, containment pressure increases at a lower rate i,e., the Lipinski correlation). As shown in Figure 3, of 25 kPa/br in the MELCOR calculation. When the suffkient debris cooling to prevent aggressive corium- cavity water is eventually boiled away (at approximately concrete interactions (CCI) to begin promptly after vessel 40,000 seconds in the W 4 calculation and 55,000 breach did not occur in the MELCOR calculation. In seconds in the MELCOR calculation), the containment contrast, the MAAP4 calculation allows debris pressurization rate decreases in both calculations. A fragmentation and cooling to occur (at least for the debris second decrease in the rate of containment pressurization mass that is ejected after some water is discharged to the is observed in the MAAP4 calculation (at approximately cavity). The result is significant debris cooling and a 73,000 seconds) when water in the lower compartment is delay in the onset of CCI until the water on the cavity boiled dry; this effect is not observed in the MELCOR floor is completely evaporated. calculation.

2.75 1.1 I

1 .o 2.25 0.9

=?

5 2.00 0.8 g 1.75 n

H t I- 0.7 1.50 n

.-25 n 0.6 1.25 0.5 1.00 5 0.75 0.4 u

0.50 0.3 0.25 0.2 0.00 0.1 0 20 40 60 80 100 120 140 Tim- (rossc) Time (IOSas)

Figure 3. Debris Temperature in Cavity (PWR - TMLB) Figure 4. Containment Pressure (PWR - TMLB)

FDI is the portion of MELCOR that calculates debris-coolant heat transfer relocation from the reactor vessel lower head to the containmentkavity floor.

These models operate in the CAV (cavity) Package.

IV. COMPARISON OF BWR CALCULATIONS when possible, identify the cause(s) of these differences.

The process of identifying differences in the calculated A comparison of the calculated timing of key events results as well as identifying their causes was based for each of the five BWR accident sequences is given in primarily on direct comparisons of code output (in the Table 3; as with the PWR results, this information is form of plot variables).

presented both in terms of the calculated time between major events as well as differences in the cumulative time Several differences were observed in the calculations.

to an event from the start of the accident. Details of the In many cases, the cause of these differences can be traced BWR calculations are not presented here because the to known differences in the mathematical models used by major findings are very similar to those identified from the each code to simulate complex physical phenomena or P W R calculations. However, several of the observations other aspects of severe accident behavior. However, in are worth noting, particularly as they reinforce the some cases differences in calculated results are caused by conclusions one would draw from the PWR results. subtle differences in the specification (via code input) of plant system characteristics.

' 1. The early hydrodynamic response of the reactor pressure vessel (i.e., the time required for the Several of the differences in calculated results coolant inventory to be depleted to the point that the top involved relatively isolated aspects of severe accident of active fuel is uncovered) is shown to be in good progression or were observed only- in particular types of agreement between the two codes. This suggests that the accident sequence simulations. Examples include factors influencing the overall mass and energy balances differences in the calculated distribution of coolant within of the reactor pressure vessel prior to the onset of core the primary system during the blowdown period of large damage are calculated in a similar manner by the two break LOCAs (explained by code hydrodynamic modeling codes. This was confirmed by a closer examination of differences) and differences in the total mass of coolant plots for several calculated parameters. injected to the primary system from accumulators (unique to the PWR ADC and AJ3F accident simulation).

2. The time required for vessel breach to occur after large quantities of debris relocate into the lower head Two signifcant differences in the calculated results for all (i.e., after core support plate failure) is calculated to be of the accident sequences are observed which can be much shorter in MELCOR than in MAAP. This is due to attributed directly to code modeling differences. These the significant differences in the models for debris heat are:

transfer within the lower head and for reactor vessel lower head failure described above. 0 The time required for reactor vessel failure to occur after a substantial mass of core debris

3. In contrast to the calculations of P W R relocates into the lower head is greater in accident sequences, the general characteristics of ex- I MAAP4 than in MELCOR 1.8.3, and vessel behavior of core debris are calculated to be in 0 The time at which containment pressure is reasonably good agreement between the two codes. That calculated to exceed the failure criterion differs is, the calculated temperature histories of debris within the between the two codes in virtually all the reactor pedestal are similar. This results from the minimal accident sequences. In general, containment coolant mass that can accumulate in the drywell pedestal failure occurs earlier in the MAAP4 calculations area to provide debris cooling; core concrete interactions of the PWR accident sequences than in the are calcuIated to begin very soon after vessel breach by corresponding MELCOR calculations; the both codes. However, significant differences are observed reverse is observed for the BWR simulations.

in the calculated temperature of the drywell atmosphere after the onset of CCI; the MELCOR 1.8.3 results being The cause of the first difference (i.e., time to vessel significantlyhigher than the MAAp4 results. breach) is the same for the PWR and BWR calculations.

Namely, the two codes use different models for debris V. CONCLUSIONS heat transfer within the lower head, and for the structural response and failure of the reactor vessel lower head.

Calculations were performed for a wide spectrum of Sensitivity calculations performed with the developmental severe accident sequences in representative PWR and version of MELCOR (post-1.8.3) in which a new creep BWR plant configurations using the MAAP4 and rupture (lower head structural failure) model is invoked MELCOR 1.8.3 computer codes. The primary objectives suggest a significantly closer prediction of time to reactor of the current evaluation were to identify major vessel breach results when similar structural response differences in calculated results from the two codes and, models are used.

I I Table 3. Summary of Calculated Timing of Major Accident Events - BWR Sequences Accident i Time to Core Time to Support Time to Vessel Time to Sequence All times in seconds 3 i Uncovery Plate Failure Breach Containment Failure TB Time between MAApi 36,501. 10,089. 7,144. 52.

events MELCOR i 30,641. - 8,107. 32. 3,328.

Cumulative MAApj 36,501. 53,734. 53,786.

time from start MELCOR i 30,641. 38,780. 42,108.

TQUX Time between MAAPi 1,897. 5,118. 3,269. 23,498.

events MELCOR i 1 902 5 201 34. 17,220.

................................................................................................. ?........:................................ ?........:.................................................................................

Cumulative MAApi 1,897. 10,284. 33,782.

time from start MELCOR i 1,902. 7,137. 24,357.

TQW Time between MAAPj 1,171. 3,837. 8,013. 22,773.

events MELCOR f

..................................................................................................... 1?........

647:..-.......................................................................................................................

3,269. 56. 19,571.

Cumulative MAApj 1,145. 13,021. 35,794.

time from start-- MELCOR t 1,647. 4,972. 24,543.

TW Timebetween MAAPI

• 14,975.- 10,377.

• events MELCOR i

• 11 268 36. *

........................................................................... *.................................................................. ?........:.................................................................................

Cumulative MAAP; 119,516. 144,868. 109,153.

time from start MELCOR i 115,309. 126,613. 108,357.

AE Time between MAAPj 35. 2,861. 10,123. 28,041.

events MELCOR i 29. 1,458. 57. 21

............................................................................................................................ .................................................. ............................................. ?........ 724:..

Cumulative MAAPj 35. 13,019. 41,060.

time from start MELCOR f 29. 1,544. 23,268.

• Containmentfailure precedes onset of core damage.

The major source of differences in the calculated time to steam generation in the cavity. In contrast, the MELCOR containment failure between MAAP and MELCOR in all calculation does not predict a quenched debris bed after of the calculations can be traced to differences in debris- vessel breach for any of the sequences. The rate at which coolant heat transfer after late-phase (large-scale) material containment pressure increases after vessel breach in the relocation. However, these differences manifest MELCOR calculation is governed to a lessor extent by themselves at different times in the PWR versus B W R steam generation in the cavity as a significant portion of calculations. In the case of the BWR calculations, the energy transferred from core debris is involved in corium-differences appear during the period of in-vessel melt concrete interactions.

progression. In particular, significantly lower temperatures of debris within the lower head are observed ACKNOWLEDGMENTS in the MAAP calculations than in the corresponding MELCOR calculations due to more efficient cooling. The work presented in this paper was sponsored by This difference has a significant impact on the time at Nuclear Power Engineering Corporation (NUPEC) of which reactor vessel breach is predicted by the two codes, Tokyo, Japan.

and on the mass of water that remains in the lower head at the time vessel breach occurs. These differences affect REFERENCES containment response by changing the timing and relative amounts of steam and non-condensable gas generation 1. R.M. Summers, et al., MELCOR Computer Code during the early and late periods of accident progression. Manuals, NcTREG/CR-6119, SAND93-2185, Sandia National Laboratories, Albuquerque, NM (1994).

In the PWR calculations, MAAP4 predicts the formation of a quenched debris bed immediately after vessel breach 2. Electric Power Research Institute (EPRI), MAAP4 -

in every sequence. This result occurs independent of Modular Accident Analysis Program for LWR Power Plants, Vols 1-4, RP 3131-02, Palo Alto, CA (1994).

whether water exists in the cavity prior to vessel breach (as in sequences S2HF or AHF)or anives coincident with debris ejection (as in TMLB). Subsequent increases in containment pressure are governed primarily by ensuing