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Advanced BWR Core Component Designs and the

Implications for SFD Analysis

L. J. Ott Oak Ridge National Laboratory

P. 0. Box 2009 Oak Ridge, Tennessee 37831-8057

(423) 574-0324

ABSTRACT

Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the “lessons learned” fiom the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes.

The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8x8 fuel assembly with two water rods and a cruciform control blade constructed of B,C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced our current SFD experimental data base and SFD code capabilities. For example, modem BWR control blade design includes hakium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities.

1. INTRODUCTION

After the accident at Three Mile Island (TMI) Unit 2, the United States Nuclear Regulatory Commission (NRC) initiated a severe accident research program, for the general purpose of developing a basis for evaluating reactor core melt progression and ultimately, for assessing the release of fission products from the plant site and the ensuing threat to the public health. An important element of this program has been the development of state-of-the-art computer codes that model the physical and chemical processes occurring in a reactor core that has lost cooling capability. A second important element of this program has been the generation of a data base by which such codes could be assessed and validated. This data base consists of experimental evidence obtained from several in-pile and out-of-pile phenomenologically oriented programs supported by the NRC and its foreign Cooperative Severe Accident Research Program (CSARP) partners.

The roles of the experimental programs have been first to provide data to understand the important in- vessel melt progression processes and, second, to provide data for model (code) development and validation. The analytical tasks focused on

employing the experimental information base to develop computational models and codes that provide the mechanism for translating the data base that has been generated to the analysis of nuclear power plants (NPPs). Computational models serve as the principal mechanism for performing NPP severe accident analyses.

In 199 1 the Committee on the Safety of Nuclear Installations (CSNI) of the Organization for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) issued a state-of-the-art report’ on in-vessel core degradation in light water reactor (LWR) severe accidents. A recommendation of this report was that a “Validation Matrix,” essentially a summary of the severe fuel damage (SFD) experimental data base, be produced. The objective of the SFD Validation Matrix (VM) was to define a basic set of experiments, for which comparison of the measured and calculated parameters forms a basis for establishing the accuracy of test predictions, covering the full range of in-vessel core degradation phenomena expected in LWR (both boiling water and pressurized water reactors) severe accident transients. The emphasis of the VM reportz was on integral experiments, and the review covered LWR designs of Western origin; the coverage of phenomena extends

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from the initial core heat-up through to the introduction of melt into the pressure vessel lower plenum.

Currently only two integral SFD experimental programs are being conducted. The French PHEBUS- FP program2 is investigating core degradation and the resultant fission product (FP) release during unrecovered transients. Second, the German QUENCH program is studying the recovery of degraded or partially degraded cores by water quenching or by rapid cooling with steam injection. The impetus behind the QUENCH program is that - all SFD experiments with reflood, conducted to date, have demonstrated (before cooldown) a renewed rapid temperature escalation and increased hydrogen generation rates, which cannot currently be predicted by the systems level SFD codes.

From the early 1980s to 1993, sixty-one SFD integral experiments have been conducted worldwide with the majority (40) of these tests being conducted at the Forschungszentrum Karlsruhe (FZK), Germany, in the NIELS and CORA facilities. However, only seven experiments [six in CORA and one (DF-4) conducted at Sandia National Laboratories (SNL)] specifically addressed the unique in-core structural features of the Boiling Water Reactor (BWR). These seven experiments were based on the vintage BWR core design (8x8 fuel assemblies with two water rods enclosed in Zircaloy channels with cruciform control blades constructed of B4C-filled tubelets).

BWR severe accident technology programs were initiated by the NRC at the Oak Ridge National Laboratory (ORNL) in 1980. These Programs have studied a series of postulated severe accidents in BWRs and have developed a set of model^^".^ specific to boiling water reactor response under severe accident conditions. Many of these models have been incorporated in NRC’s whole-core accident analysis codes, MELCOR6 and SCDAPRELAP5’. These models (in an experiment-specific f0rmaP3~) have also been successfully applied to pretest and posttest analyses of the DF-4 and CORA BWR integral experiments with excellent agreement between model prediction and experimental data. To date, these models have provided far more precise prediction^^*'^-^^ of the conditions in the BWR experiments than have previously been available. This has provided a basis for more accurate and more complete interpretation of the phenomena for which the experiments were performed.

This paper will present, in Section 2, the “lessons learned” from the DF-4 and CORA BWR experiments and the impact on core models in the

current generation of SFD codes, that is, MELCOR and SCDAP/RELAP5.

As stated above, the DF-4 and CORA BWR test assemblies were modeled on the BWR core component designs circa 1985; that is, the 8x8 fuel assembly with two water rods and a cruciform control blade constructed of B4C-filled tubelets (see Figure 1). Within the past ten years, the state-of-the- art with respect to BWR core component development has out-distanced our current SFD experimental data base and SFD code capabilities. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities in Section 3.

2. LESSONS LEARNED FROM THE DF-4 AND CORA BWR TESTS

The DF-4 BWR experiment was conducted in the Annular Core Research Reactor (ACRR) at SNL on November 2 1,1986; and the CORA BWR tests were performed at FZK from November 1988 to October 1992. A synopsis of the test and operating conditions are given in Table 1.

The CORA facility at FZK was specifically designed for SFD experiments and allowed a systematic assessment of the differences in core damage progression behavior due to variations in test parameters, such as: 1) bundle design (PWR or BWR) and size, 2) system pressure / fuel rod internal pressure, 3) bundle heatup rate, 4) bundle cooling rate (slow cooling in argon or

rapid quenching in water), 5 ) bundle gaseous inlet flow rates and temperatures.

Table 1 clearly illustrates the systematic approach to assessing the differences in core damage progression behavior due to variations in test conditions; that is, only the condition of interest is varied, then the test results can be compared to determine the effect of the change in the test condition.

The comments (“lessons learned”) in the following sections are summarized fiom the experimental posttest analyses of these integral BWR SFD tests (references 8, 10-14).

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Table 1. Synopsis of BWR SFD integral test conditions

CORA-18 2lJune 90 4 8 0.210.6 16 4 1 .o <2300 Electrical ' Slow(&) Large Bundle

CORA-31 25July 91 18 0.210.4 8 2 0.3 2300 Electrical Slow(Ar) Slow heatup

CORA-28 ' 25Feb. 92 18 0.210.4 8 - 2 1 .o 2300 Electrical Slow(&) he-oxidized

CORA-33 1Oct. 92 18 0.210.5 3 Trace 0.3 2300 Electrical Slow(&) ''Y core conditions

2.1 Zirealoy/Steam Oxidation

The posttest analyses performed at ORNL employ a comprehensive kinetic correlation package covering the full temperature range (<900"C, cubic kinetics; 9OO-105OoC, transition between cubic and parabolic kinetics; 105Oo-158O0C, parabolic kinetics; >1580"C, rapid oxidation due to Zircaloy microstructure change) encountered in SFD experiments and postulated accidents. These Zircaloy/steam reaction models give comprehensive coverage of this phenomenon fkom 650°C through Zircaloy melt with no Correlation being applied outside its experimental range.

The systems level codes (MELCOR and SCDAPAWLAPS) currently use the Urbanicrneidrick correlati~n'~ (range of 1580"- 1 850°C) at temperatures greater than 1580°C. SCDAPRELAPS employs the Cathcart/PawelI6 correlation between 1000" and 1580°C and MELCOR utilizes the Urbanic/Heidrick correlation (range of 1050"- 158OOC) between 827" and 158OOC. The lower temperature limit for applicability and even the correlations themselves may be changed in MELCOR via user input sensitivity coefficients; whereas, the SCDAPAWLAPS lower limit and correlations cannot be modified by user input. Thus in the systems level codes, the oxidation kinetics correlations may be extrapolated beyond their applicable temperature ranges, or oxidation may be stopped completely below an artificial lower limit.

The ORNL posttest analyses of DF-4 and CORA BWR experiments have made the following observations with respect to Zkcaloy/steam oxidation kinetic models:

low temperature (<1050"C) oxidation is significant, and recommendations for the

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cubic and transition regimes should be used. In fact, CORA-28 was pre-oxidized (in-situ) at -800°C. The cubic and transition regimes are not represented in systems level codes. parabolic kinetic models (temperature range of 105Oo-158O0C) should not be extrapolated outside their experimental data base range (as is done in systems level codes). This can result in underpredicting the low temperature (<1O5O0C) oxidation. a comprehensive kinetic correlation package (such as used in the ORNL experimental posttest analyses) should be used. available kinetic correlations (applied over their appropriate temperature ranges) are adequate for SFD analyses.

Cladding may undergo simultaneous deformation (ballooning) and oxidation during a severe accident. The deformation increases the available surface for oxidation and thereby increases the volume of oxide (and hydrogen) that is subsequently formed. Deformation generally occurs at lower temperatures than those significant for oxidation; research, during the 1970's and early 1 9 8 0 ' ~ ~ was conducted on cladding burst (deformation and rupture) for typical LOCA conditions. That is, the cladding was subjected to temperature escalations and high differential pressures as would be expected in a depressurized, loss of cooling capability accident scenario. For these conditions, clad burst would be anticipated for the a and a + p Zircaloy phase regimes; that is, at temperatures <975"C, which is below the threshold of the parabolic oxidation kinetics (the p Zircaloy phase regime). In other words, concurrent high cladding oxidation and strain would not be expected in a commercial LWR accident scenario (the cladding would have burst

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before significant oxidation had begun). The differential pressure across the cladding in the CORA BWR experiments is small (see Table 1) but cladding strain in these tests has occurred and has been confirmed by Haste1*. The ORNL analyses use the Erbacher modePg (based on the German clad burst experiment^'^ in the REBEKA facility) for cladding strain determination. Excellent agreement has been obtained with cladding thermal excursions (which is not possible without the strain model) and with predicted cladding failure times. Although it may not be prototypic of commercial LWR accident scenarios, concurrent strain and oxidation in the p Zircaloy phase regime must be considered in the experimental analysis-specially, if useful information is to be gained ftom the early phase structural melting and relocation portion of the experiment.

CORA- 16,- 17, and -28 are similar experiments (see Table 1) with respect to test section steam flows and initial heatup rates. The major difference between these tests (regarding the oxidation induced temperature excursion) is that CORA-28 was preoxidized (which is more prototypical of actual LWR operating conditions where BWR Zircaloy structures will oxidize at the rate of -1 0 mp per cycle) and CORA- 16 and -17 were not. The observed (and calculated) excursions occurred significantly later (100-275 s) in CORA-28 as compared to CORA-I6 and - 17; also the excursion rates of the three tests are the same but the temperature range at which the excursions occur in CORA-28 is significantly higher (by 100’-200OC). Both the delay in the excursion and higher structural temperatures at initiation of the escalation should be expected in a preoxidized bundle. Given the appropriate initial conditions (temperature and oxide thicknesses), current heatup and oxidation models are capable of predicting the observed cladding thermal response.

A time lag between the calculated and experimentally determined CORA hydrogen generation rates has been observed in the ORNL analyses and in the International Standard Problem (ISP) 3 1 exercise20 (based on CORA- 13). At the ISP-3 1 workshop, a theory was proposed that the Zircaloy cladding in the upper portion of the bundle had absorbed the hydrogen and then released it at high temperature, thus causing the observed time lag. Recently, a paper by 0lander2l lends credence to this argument. Olander performed his analyses by integral diffusion theory, and he states that “this methodology provides a better description of the state of the cladding than conventional kinetic rate laws.” This may be true, and this approach does give a credible explanation for the observed time lag; but it should be noted that conventional use of kinetic rate laws (as in the ORNL analyses) does allow excellent

predictions of the chemical energy production and the thermal response of the test section structures.

Another method of studying the hydrogen generation is to break down the evolution into two event intervals; this may appear arbitrary, but there are two logical time periods: first, the period of intact cladding (through incipient clad melting) and second, the time period after initial clad melting (melting, relocation and formation of U-Zr-0 blockages). As demonstrated in the CORA-28l4 and CORA-33l3 analyses, only -25 percent of the hydrogen generated occurs during the “intact cladding” phase . Over 75 percent of the total hydrogen evolved occurs during the meltinglrelocationhlockage period of these experiments. All integrated codes compute hydrogen generation in intact geometry reasonably well, but handle the later phase (post intact cladding) of an experiment (or accident scenario) much less robustly. Obviously, these codes must be capable of accurately estimating the Zircaloy-steam reaction rate during the period of clad degradation and blockage formation since the majority of the hydrogen generation has been demonstrated to occur during this time m e .

Finally, it should be noted that in the initial QUENCH facility experiments22 at FZK the hydrogen generation could not be determined by available Zircaloy/steam oxidation correlations.

2.2 Material Interactions

The components of a BWR core consist of materials that basically are not stable together thermodynamically; that is, chemical interactions have to be expected. The material interactions of interest (in a BWR core) are:

B4C and stainless steel (also Zircaloy),

stainless steel and Zircaloy, Zircaloy and fuel (UOz).

Under normal reactor operating conditions, these chemical interactions are limited and can be tolerated. However, at elevated temperatures, the chemical interaction rates (i.e., kinetics) can increase dramatically. The interaction of Zircaloy and fuel has long been recognized and is included in MATF’R07 routines of SCDAP/RELAPS, but the importance of the other material pairs interactions was not recognized as recently as 1985 (the DF-4 experiment was conducted in November 1986). In fact, the SNL staff requested assistance fi-om the FZK staff on the B4C chemistry issue prior to the conduct of the DF-4 test. This request resulted in the hag en'^^^,^^ preliminary findings that there was limited chemical interaction between B4C and steam, and that there was strong chemical attack of the stainless steel by B4C at -1200°C with complete liquefaction by

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125OOC. This contrasts with the expected failure of the BWR control blade by melting at 1375"-1425°C.

H ~ f m a n n ' s ~ ~ - ~ ~ separate-effects tests (at FZK) in support of LWR severe accident investigations include the determination of the following:

high temperature oxidation kinetics of Zircaloy, extent of the U02 fuel/Zircaloy cladding interactions, dissolution of U02 and Zr02 by molten Zircaloy, reaction kinetics of absorber materials [(Ag, In, Cd) alloy, B4C] with their cladding material, reaction kinetics between A1203/ B4C and Zircaloy (burnable poison rods), onset of melting of core components due to chemical interactions, phase diagrams considerations.

The objective of these studies was to describe these processes by models that can be applied in codes. A summary of the chemical interactions and formation of liquid phases in an LWR fuel rod bundle as a function of temperature is presented in Figure 2.

The integral BWR experiments (DF-4 and CORA) have demonstrated that the initiation of structural liquefaction occurs due to the interaction of B4C and stainless steel in the absorber blade. This melt then attacks and destroys the Zircaloy channel wall with subsequent distribution of the B4C/SS/Zr melt radially and axially within the fuel bundle and interstitial regions. This then results in the liquefaction of the Zircaloy cladding and U02 fuel at temperatures far below their melting points. Metallic- rich blockages form within the interstitial region and within the fuel bundles low in the core.

2.3 Melt Relocation Flow

The CORA experiments have covered much of the range of early-phase melt progression phenomena and have provided basic information on the processes of metallic melt relocation. From the video records (Le., visual observation), melt relocation was found to occur mostly by non-coherent, non-coplanar rivulet flow. Hering29,30 performed an extensive review of the video records for the CORA tests and concluded that melt progression occurred via rivulet flow and droplets with 4 0 percent of the relocation being by droplets.

Current melt relocation modeling employed in the system level codes assumes coherent, coplanar film flow (work is underway to develop a flow regime map for application in SCDAPRELAPS).

Proper modeling of the relocation flow is important in that core structures (especially the BWR channel wall and control blade) below core mid-plane respond dramatically to relocating flow in a very short time frame (before the thermal escalation due to the Zircaloy/steam reaction). To properly predict the temporal and thermal response of lower core structures, the melt flow must be properly modeled.

All degraded core BWR experiments (DF-4 and CORA) have demonstrated varying degrees of melt intrusion (B4C/stainless steel/Zircaloy) from the interstitial region of the core into the fuel rod assemblies within the individual channel boxes. ,

Models to allow melt intrusion into the channel box region and to represent the subsequent freezing/remelting/relocation of these melts within the fuel bundle are needed to adequately predict the melt flow within the core.

In CORA-33, the ''dry" core test, more melt relocated lower into the test section (even below the active heated length) than had been observed in any previous CORA BWR test. This test employed a near-prototypic axial thermal gradient at the time of incipient control blade liquefaction. In the CORA-28 preoxidized test, more melt relocated into the quench tank water (which cools the lower electrical connections of the CORA bundle) than had been observed in the other BWR experiments. In CORA- 28, the level of oxidation is prototypic, but the axial thermal gradient (at incipient control blade liquefaction) was substantially colder (and less steep) than prototypic.

The combined results of CORA-33 (near-prototypic axial thermal gradient) and CORA-28 (more prototypic surface conditions) indicate that substantial melt relocation onto the BWR core plate should be expected in the unlikely event of an actual severe accident, and that integrated code models of the core plate should be capable of predicting the response and possible failure of the core plate due to core melt thermal and mass loading.

The ex-reactor (XR) experiments31-37 at SNL (three tests conducted from July 1993 to October 1995) were intended to resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a dry BWR core as a result of a core-melt accident. In all three XR tests, significant amounts of the melt delivered to the test assemblies traversed the assembly structures to and through the simulated core plate. In XR2- 1, with full scale prototypic lower core BWR structures, more than 50 percent of the material was found below the core plate (having flowed through prototypic bypass flow paths).

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No systems level code has the capability of predicting the observed phenomena in the XR experiments with respect to the core plate behavior. These core plate m0dels ,3~-~~ in the systems level codes, are simplistic or non-existent and are not based on experimental data. (Currently there are plans to develop a specific BWR core plate model for SCDAPRELAPS based on the XR results.)

2.4 BWR Control Blade and Channel (or Canister) Models

A conceptualization of the ORNL canister/control blade materials interaction, relocation, blockage model^^^,^^ is outlined in Figure 3. This figure comprises six subfigures (a-f), which visually describe the progressive steps in the structural heatup, liquefactioddissolution, relocation, blockage, and subsequent structural failure for the BWR control blade and canister models. The failure process begins with the B4C/stainless steel reaction in the control blade absorber tubes (Fig.3b). Eventually, this material liquefies ( 120O”-125O0C), Fig.3~; and then, by either failing the outer stainless steel sheath or flowing through perforations in the sheath, relocates downward leaving a crust (fiozen relocated eutectic) on the surface of the control blade (Fig.3d). This crust builds up (Fig3e) until a blockage occurs, at which time the interaction between the relocated B4C/stainless steel and the Zircaloy canister wall begins (also lateral diversion of melt interstitially can then occur). The canister finally fails (due to the material interaction), as shown in Fig.3f7 allowing flow into the fuel assembly and down the inside of the canister wall.

These ORNL canister/control blade models were originally developed for the BWRSAR3” severe accident response code. They were subsequently adapted for the DF-4 experiment-specific codes where it was found that the models had to be modified to account for the B4C/stainless steel reaction in the control blade absorber tubes. For the CORA facility, the DF-4 code was modified for the CORA-specific features9 and in subsequent CORA BWR analyses1 the relocation, blockage, interaction models (based on Hohann’s r e ~ e a r c h ~ ~ - ~ ~ ) were added and improved. The last major improvements focused on the breach of the Zircaloy channel wall (between Figs. 3e and 3f) and are based on new German studies43 of the Fe-Ni-Zr metallic system [in the dissolution process there can be vigorous localized exothermic heat generation with interaction products melting at low temperatures (-950°C) with lowered (as compared to Zircaloy) heats of fusion].

These models have been tested and validated via the DF-4 and CORA BWR experiments, and they appear

to adequately reflect the experimental response for widely divergent test conditions (see Table 1): from low to high heatup rates, virgin and preoxidized Zircaloy surfaces, dry to wet core steaming rates. The models have been applied to the XR experiment^,^^ and, again, the models adequately simulate the thermal response of the test section structures and the - end-state material distribution in the tests.

These models have been incorporated in SCDAPRELAPS at ORNL by F. P. Griffin42 and a validation exercise44 (based on CORA-17,-3 1, and -33) has been performed.

2.5 Miscellaneous Observations

2.5.1 Fuel Assembly Spacer Models

In the CORA experiment-specific code, the test section grid spacers are explicitly modeled. This is required to adequately describe the structural (he1 rods, canister, and control blade) responses at these locations. The grid spacers appear to act l,&e radiation shields; that is, these locations are observed “hot spots” for the fuel rods. Also, the grids are constructed of Zircaloy and thus will react with steam during the test transient. 2.5.2 U02 Electrical Conductivity

In the CORA experimental analyses, it is absolutely necessary to calculate the temporaVspatia1 volumetric power generation correctly. Not only does the electrical resistivity of the tungsten heating element change dramatically with temperature [fiom normal CORA initial thermal conditions (-400”-450OC) to Zircaloy melting (-1 850°C) the tungsten resistivity increases by a factor of 51, but at high temperature the U02 fuel becomes electrically conductive and electrical shunting through the fuel occurs. Essentially, a parallel path develops at high temperature, whereas at low temperature, only a series path through the heating element exists. Models to describe this shunting phenomena have been implemented in the experiment-specific CORA code. The observed bundle voltage differential, current usage, and bundle equivalent resistance in the CORA experiments are adequately simulated by these models. In Griffin’s validation study44 the SCDAP/RELAP5 electric rod models had to be modified to reflect this phenomena in order to reasonably simulate the CORA BWR experiments.

2.5.3 Clad Failure Mechanisms

Failure of the fuel rod clad pressure boundary may occur by (1) ballooning and burst, (2) melting, or (3) materials interaction. Of these mechanisms, clad deformation and rupture have received the greatest study and for prototypic commercial LWR

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depressurized accident scenarios, clad burst would be expected to occur at temperatures <975OC. As temperatures increase, materials interaction mechanisms should be considered next in order, and fmally clad melting at -1850°C. Examples of the materials interaction failure of cladding include the well-recognized PWR Inconel spacer/Zircaloy clad interaction, and the dissolution of Zircaloy clad by molten PWR absorber material (Ag/In/Cd).

In BWRs, materials interaction failure of the clad could be caused by an attack of the clad by molten control blade/canister interaction products (B&/SS/Zr). This would occur only after breach of the canister wall by relocating B4C/SS material at -1250°C. Another possible mechanism in BWRs is perforation of the cladding by Inconel fmgers (springs) in the Zircaloy spacers, which would occur at 1350"- 1400°C. Given the accident progression and the radial temperature gradients between bundle components [100"-200"C between the heated rod (at 1200°C) and the control blade], perforation of the cladding could occur before breach of the canister and perhaps before control blade liquefaction.

These failure mechanisms are all considered in the experiment-specific analyses of the CORA experiments.

In CORA-16, a cladding pressure differential of 0.25- 0.40 MPa was measured (six rods were monitored), and cladding failure was observed to occur between 4215 and 43 10 s; the experiment-specific code predicted cladding rupture in these rods between 4250 and 4280 s.

In CORA3 1, a differential pressure across the cladding of 0.15-0.17 MPa was maintained until cladding failure. Five rods were observed to fail between 4715 and 5000 s. In the simulation, the predicted failure mechanism was perforation of the cladding by the spacer at the midpoint (525-575 mm) of the bundle between 4830 and 4883 s. At the time of failure, the maximum cladding strain was -3 percent, which is insufficient to induce failure by burst.

Within the past ten to fifteen years, the nuclear fuel fabricators [for BWRs, these are the General Electric Company (GE), Siemens Power Corporation (SPC), and ABB Combustion Engineering (ABB-CE)] have brought to market new and innovative fuel design features. Generally NPP operating trends have been towards longer operating cycle lengths (1 8-24 months) and higher discharge bumups (approaching 50000 MWd/MTU for BWRs). These trends have brought pressure on the fuel fabricators to develop fuel designs that offer higher discharge burnups, longer lived components, and provide improved plant operating margins. The following is a partial list of the improvements and their reason for being introduced: .

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. 3. ADVANCED BWR CORE COMPONENT

DESIGNS

The DF-4 and CORA BWR test assemblies were modeled on the BWR core component designs circa 1985; that is, the 8x8 fuel assembly with two water rods (fuel rod and water rods having diameters of 12.27 and 15.0 mm, respectively) and a cruciform control blade constructed of B4C-filled tubelets (see Figure 1).

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smaller (diametrically) fuel rods (i.e., 9x9 and 10x10 fuel rod arrays): allows higher bumup with a lower linear heat generation rate (LHGR), thus, lower pellet and cladding operating temperatures and lower cladding corrosion; barrier cladding (essentially a layer of zirconium or Fe/Zr alloy on the inside of the Zircaloy: together with lower LHGRs, this offers a high resistance to pellet-to-clad interaction (PCI); cold worked, stress-relief annealed cladding: higher stress threshold for failure and improves resistance to PCI; p quenched cladding: improves resistance to nodular corrosion; improved inside surface finish of cladding: improves resistance to PCI; improved pellet design (i.e., uniformity, microstructure, etc.): better fuel performance, lower fuel swelling, minimize interaction with cladding

changing fiom Zircaloy-4 to Zircaloy-2 channels: geater corrosion resistance and reduces hydrogen pickup thus less embrittlement; optimized channel geometry (thinner walls and thicker comers): increases rigidity, better neutron economy; new spacer and tie plate designs to capture water- borne debris: reduce fietting damage to fuel and cladding; reduced pressure drop components (tie plates, spacers, etc.): off-sets increased pressure drop through smaller rodded bundles, improves stability, decreases pumping requirements; partial length fuel rods: decreases pressure drop, improves stability, increases cold shutdown reactivity margin; larger water rods (or more water rods, or water crosses): increases hot excess and cold shutdown

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reactivity differences, improves neutron efficiency, improves moderation; using high purity stainless steel tubing in the control blade: increases rod life, decreases B4C/stainless steel swellingkracking problems; using hafnium at the control blade wing edges and at the top of the control blade: reduces swelling at high burnups (as compared to B4C), longer rod life; solid control blade construction (i.e., no outside blade sheath).

The list of improvements made by the fuel fabricators is extensive (the above is only a partial list); and, in general, our current SFD experimental data base does not specifically address any of the above design changes. However, our systems level SFD codes can easily handle some of the changes (dimensions for example), and the effects of some improvements (such as low pressure drop components) are insignificant in core damage progression. Intuitively, if the component material has not changed but only the surface finish or texture or heat treatment has changed, then at elevated SFD temperatures, it is not expected that the melting, relocation, material interactions characteristics would differ from the existing data base.

When major changes in assembly dimensions or new materials are introduced that are not consistent with the existing data base, then uncertainties in the core melt progression and phenomena increase. Two of the above improvement^^^ will be cited as examples of this argument.

First, hafhium has excellent neutron-absorbing properties (it does not swell, like B4C; it converts to other neutron absorbing isotopes with irradiation; and it depletes more slowly over time), and it is an excellent improvement to the BWR control blade. The hafnium is placed at high flux regions in the control blade; it will serve to protect the inner B4C tubelets and will increase the operating life of the control blade. However, what new material interactions can be expected from the introduction of hafkium in the BWR core, and how will that affect core melt progression andor accident management procedures?

Reviewing the B,C/stainless steel control blade design (see Sections 2.2 and 2.4), rapid and complete liquefaction occurs by 1250°C with subsequent rapid relocation to the lower portions of the core (and core plate). This has caused the NRC concern with possible recriticality in BWRs if the core is reflooded afker a partial meltdown (control blades have liquefied and relocated, but the clad and fuel are still intact). Given the constituents of the control blade (Le., B,

C, Fe, Ni, Cry and minor impurities) and referring to standard reference~,4~,~~ several binary combinations (B/Fe and B/Ni) show low melting eutectics (from 1000" to 1 150°C), and this is the r e a s 0 n ~ ~ 3 ~ ~ that the control blade liquefies -200°C lower than the melting range of stainless steel.

Adding Hf to the control blade constituents, and again consulting the standard references, H a n ~ e n ~ ~ shows no low melting eutectics (other than the B/Fe and B/Ni pairs); but supplement to Hansen's text, indicates that Hf may form low melting eutectics with Fe and Ni, although these systems are less definitive than the boron systems. Thus, if Elliott is correct, then the new BWR control blade (with hafhium) may behave the same as the control blade as currently modeled42; however, there is the possibility that the hafnium may not interact with the stainless steel sheath of the control blade. For this postulate, the inner portion of the blade (where the B4C-filled tubelets are positioned) will probably liquefy at 1200"-1250°C and relocate (interacting with the control blade and Zircaloy channel wall at lower elevations); but the blade wing tips (containing the hafnium) might remain intact in the core until the stainless steel or the hafhium melts. For this case, the recriticality issue is again raised, since neutron-absorbing material (hafnium) might remain in the core after the B4C portion of the control blade has exited the core; also, for this case, even the advanced control blade models42 are not applicable.

To f m l y quantify the effect of including hafnium in the control blade on SFD degradation behavior, at the least, separate-effects studies employing hahidstainless steeYZircaloy combinations (as originally done by Hofmann) should be performed. Only then will we know if our systems level codes are adequate, as is, or if they will need to be modified.

which is a later

The second example (cited for M e r study) is the increased water volumes within the fuel assembly itself; that is, in the 8x8 assembly on which our BWR SFD data base is developed, the water rods are only slightly larger than the surrounding fuel rods; but, in the new fuel assembly designs, larger water rods (replacing a 3x3 cluster of fuel rods) or water boxes, or water crosses are being employed. These designs have operating advantages: that is, increased hot-excess and cold-shutdown reactivity differences, improved neutron efficiency, improved moderation. However, in severe accident conditions, these are large, open, empty spaces in the fuel assembly.

As our current BWR SFD data base has shown, metallic melts (especially the control blade material and, subsequently, channel wall material) relocate

8

rapidly to lower colder portions of the core [and in our most prototypic experiments, even out of the core area (i.e., onto or through the core plate)] with some of the relocation being aided by the relatively open interstitial region (outside the fuel assembly). In general, the material relocation is binodal; that is, metallic melts accumulating low in the test sections with oxidic melts (U-Zr-0) higher in the test section. However, with large open spaces within the fuel assembly, there exists the possibility of rapid relocation of oxidic melts (the walls of the water rods offer little resistance to attack by melts, and the thermal capacitance of the walls will not accommodate much refrozen material before failure) to lower portions of the core. These structures offer little impedance to relocating melts and actually provide a large highway for relocation. The impact is that instead of melts with just phase-change energy, melts with phase-change energy and internal power generation (high enthalpy melts) will attack the BWR lower core structures rapidly and earlier. Our systems level codes are not capable of handling this phenomena sequence change.

The improvements made by fuel fabricators on the BWR fuel assemblies and core components should be reviewed by SFD experts, and the effect of these “improvements” on our capabilities (Le., codes) to predict core damage phenomena and their timing sequence must be assessed. If there are ‘Lhole~” in our BWR SFD data base, then separate-effects tests and even new integral tests should be designed and funded to fill these “holes.”

REFERENCES

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2.

3.

4.

S. R. Kinnersly, J. N. Lillington, A. Porracchia, K. Soda, K. Trambauer, P. Hofinann, Y. Waarenpera, R. Bari, C. Hunt and J. Martinez, In-Vessel Core Degradation in L WR Severe Accidents: A State of the Art Report to CSNI, Januay 1991, NEA/CSNI/R(9 1)12, November 1991.

T. J. Haste, B. Adroguer, R. 0. Gauntt, J. Martinez, L. J. Ott, J. Sugimoto and K. Trambauer, In- Vessel Core Degradation Code Validation Matrix, NEA/CSNI/R(95)21, 1996.

,

S. A. Hodge and L. J. Ott, Boiling Water Reactor Severe Accident Response (B WRSAR) Code Description and Assessment, letter report to Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, February 1, 1989.

L. J. Ott, “Advanced Severe Accident Response Models for BWR Application,”

Nuclear Engineering and Design, 115,289- 303, 1989.

5 . L. J. Ott, “Advanced Severe Accident Response Models for BWR Application,” presented at the Fifteenth Water Reactor Safety Information Meeting, National Bureau of Standards, Gaithersburg, Maryland, October 29, 1987.

R. M. Summers, et al., MELCOR 1.8.3: Computer Code Manual, Volume I : Primer and Users ’ Guides, Sandia National Laboratories, NUREG/CR-6 1 19, September 1994.

6 .

7. E. C. Johnson, Editor, SCDAP/RELA P5/MOD3. I Code Manual, Volumes I to IV, NUREG/CR-6150, October 1993.

8. L. J. Ott, Post-Test Analysis of the DF-4 BWR Experiment Using the B WRSAWDF-4 Code, letter report (ORNL/M-1020) to Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, August 10, 1989.

9. L. J. Ott, Description of the CORA BWR Experiment-Specific Code, letter report (ORNL/NRC/LTR-90/23) to Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, September 30, 1990.

R. 0. Gauntt, R. D. Gasser, and L. J. Ott, The DF-4 B WR Control BladdChannel Box Fuel Damage Experiment, NUREGICR-467 1 , November 1989.

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11. L. J. Ott, Posttest Analyses of the CORA-I6 and CORA-I 7 BWR Experiments, letter report (ORNL/NRC/LTR-92/17) to Dr. A. Behbahani, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, July 10, 1992.

L. J. Ott, Posttest Analysis of the CORA-31 Slow Heatup B WR Experiment, letter report (ORNL/NRC/LTR-92/29) to Dr. A. Behbahani, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, December 3 1, 1992.

12.

13. L. J. Ott, Posttest Analysis of the CORA-33 Dry Core B WR Experiment, letter report (ORNL/NRC/LTR-93/21) to Dr. A. Behbahani, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, August 31, 1993.

9

14. L. J. Ott, Posttest Analysis of the CORA-28 Preoxidized B WR Experiment, letter report (ORNL/NRC/LTR-93/26) to Dr. A. Behbahani, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, September 30, 1993.

15. V. F. Urbanic and T. R. Heidrick, “High- Temperature Oxidation of Zircaloy-2 and Zircaloy-4 in Steam,” J. NUC. Math., 75, 251- 261, 1978.

16. J. V. Cathcart, et. al., Zirconium Metal- Water Oxidation Kinetics IV - Reaction Rate Studies, ORNLMUREG-17,1977.

17. T. J. Haste, et al., Zircaloy Oxidation Kinetics in the Temperature Range 700- 13 00 “Cy IAEA-TC-657,4.7 , September 1988.

18. T. J. Haste, et. al., UK Analysis of the CORA Melt Progression Experiments, paper presented at the 1992 International CORA Workshop, FZK, Karlsruhe, Germany, October 1992.

F. J. Erbacher, et al., “Burst Criterion of Zircaloy Fuel Cladding in a Loss-of- Coolant Accident,” Zirconium in the Nuclear Industry; Fifth Conference, ASTM STP 754,271-283, 1982.

19.

20. M. Firnhaber, et. al., International Standard Problem 31, COR4-13 Experiment on Severe Fuel Damage, Preliminary Comparison Report, presented at the ISP-3 1 Workshop, FZK, Karlsruhe, Germany, August 1992.

2 1. D. Olander, “HJH,O-Zr-UO2 Chemical Interactions in Severe Fuel Damage Accident Analyses,” ANS Proceedings, 1993 National Heat Transfer Conference, Atlanta, Georgia, August 1993.

P. Hofinann, V. Noack, and L. Schmidt, Physico-chemical Behavior of Zircaloy Fuel Rod,Cladding During Quenching, presented at the Cooperative Severe Accident Research Program meeting at Bethesda, Maryland, May6, 1996.

S. Hagen and R. W. Ostensen, Scoping Test on B&-Absorber Rod Behavior Under Severe Fuel Damage Conditions, Severe Fuel Damage Program Review Meeting, Oak Ridge, Tennessee, April 7-10, 1986.

S. Hagen and P. Hohann, Behavior of B4C- Absorber Roc& Under SFD Conditions, Severe

22.

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Fuel Damage Program Review Meeting, Rockville, Maryland, October 21-24, 1986.

25. P. Hofmann, et. al., Reasons for the Low- Temperature Failure of B WR Absorber Elements, Severe Fuel Damage and Source Term Research Program Review Meeting, Idaho Falls, Idaho, April 10-14, 1989.

P. Hofmann, et. al., Low-Temperature Liquefaction of L WR Core Components, Severe Fuel Research Program Partners Review Meeting, Brookhaven National Laboratory, Upton, New York, April 30-May 4, 1990.

26.

27. P. Hofinann, et. al., Results of Separate- Efsects Tests: Influence of Cladding Oxidation on Chemical Interactions with other Bundle Components, 1990 International CORA Workshop, FZK, Karlsruhe, Germany, October 1-4, 1990.

28. P. Hofmann, et. al., Reactor Behavior of BdC Absorber Material with Stainless Steel and Zircaloy in Severe L WR Accidents, KfK-4598, July 1989.

29. W. Hering, Global Analysis of CORA Tests to Establish a Broad Data Base for Calculation, 1989 International CORA Workshop, FZK, Karlsruhe, Germany, September 25-27, 1989.

30. W. Hering and K. Muller, VeriJication of the KESS-111 Code System using CORA Experiments, Part I: Interpretation of P WR Specijk CORA Experiments, 1990 International COR4 Workshop, FZK, Karlsruhe, Germany, October 14,1990.

31. R. 0. Gauntt, et. al., A Program Plan for the Ex-Reactor Metallic Melt Relocation Experiments, S N L letter report to Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, June 22, 1990.

32. R. C. Schmidt, Porous Media Type Modeling of Late Phase Melt Progression, Cooperative Severe Accident Research Program Review Meeting, Bethesda, Maryland, May 6-10, 1991.

33. R. 0. Gauntt, First ResultsfLom the Ex- Reactor Metallic Melt Relocation and Blockage fiperiments, 1993 CORA International Workshop, FZK, Karlsruhe, Germany, September 27-29, 1993.

10

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R. 0. Gauntt, et. al., Data Report on the XRl- 1 and XRI-2 B WR Metallic Melt Relocation Experiments, SNL letter report to Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, January 18, 1994.

L. J. Ott and F. P. Griffin, Development of the B WR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments, letter report (ORNL/NRC/L,TR-94/38) to Dr. Yi-Shung Chen, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, October 31, 1994.

L. J. Ott, Description of the XIUBWR Experiment-Specific Code and Preliminary Posttest Analyses of the XRl Tests, letter report (ORNL/NRC/LTR-95/5) to Dr. Yi-Shung Chen, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, February 28, 1995.

R. 0. Gauntt and L. Humphries, Final Report for the XR.2-1 fiperiment, draft report to Dr. Yi-Shung Chen, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, September 1996.

S. A. Hodge, Role of the Core Plate in BWR Severe Accident Progression, letter report (ORNL/NRC/LTR-90/16) to Dr. Yi-Shung Chen, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, September 1990.

L. J. Ott and W. I. van Rij, A Survey of Current Models of B WR Core Plate Failure used in the Severe Accident Codes APRIL, B WRSAR, MELCOR, MELPROG, and SCDAP/RELAP, letter report (ORNL/NRC/LTR-90/14) to Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, June 29, 1990.

B. S. Cowell, Recommendations for Implementation of a B WR Core Plate Region Model Within SCDAP/RELAPS, letter report (ORNL/NRC/L,TR-92/7) to Dr Yi-Shung Chen, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, March 13, 1992.

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11

F. P. Griffin, BWR Control BladeIChannel Box Model for SCDAP/RELAPS: Damage Progression Theory ana’ User Guide, letter report (ORNL/NRC/LTR- 96/20) to Dr. Yi- Shung Chen, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, July 12, 1996.

H. Wang, R. Luck, and B. Predel, ‘‘Thermodynamic Investigation on Liquid Iron- Nickel-Zirconium Alloys,” Journal of Phase Equilibria, 14, No.1, 48-53, 1993.

F. P. Griffin, Validation of SCDAP/RELAP5 Against CORA B WR Test Results, letter report (ORNL/NRC/LTR-97/1) to Dr. Yi-Shung Chen, Accident Evaluation Branch, Division of Systems Research, RES, USNRC, January, 1997.

M. Hansen, Constitution of Binary Alloys, McGraw-Hill Book Company, 1958.

R. P. Elliott, Constitution of Binary Alloys, First Supplement, McGraw-Hill Book Company, 1965.

C ; ~ ~ ~ % ~ R O D S ~ OFUEL RODS

CONTROL BLADE

* FUEL CLADDING

FUEL ROD INTERIM SPACER

LOWER TIE PLATE

FUEL ASSEMBLY

INSULATION: UNHEATED ' / ! / ZrtL.FIBCR

WINDOWS 21Q

Figure 1 BWR core components (8x8 he1 assemblies) and the DF-4 and CORA BWR experimental cross-sections.

F' 3000'C

2850'C 4 4-1 Melting of U02

t Melting of ZrO2 - Formation of ceramic (U,Zr,O) melt

Formation ofa-Zr(O)/UO2 and UNO2 monotectics

McMng of B4C I

1

t burnable poison rod: Ai203 + B4C)

fonnation of metallic

Formation of firs? FelZr and NilZr eutectics

Figure 2 Chemical interactions and formation of liquid phases in a LWR fuel rod bundle with increasing temperature26.

/-

d. SSlBrC EUTECTIC

RELOCATION WITH FREEZING I

UUQ

Figure 3

b. C.

ABSORBER TUBE UOUEFACTION

I

e. BLOCKAGE WITH

SSA34CRr INTERACTION

f CANISTER BREACH WITH

SSIB4CIzr EUTECTIC RELOCATION

Conceptualization of structural liquefaction, material dissolution, and eutectic relocation for the ORNL BWR control bladekanister models.

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, make any warranty, express or implied, or assumes any legal liabili- ty or responsibility for the accuracy, completeness, or usefulness of any information, appa- ratus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessar- ily state or reflect those of the United States Government or any agency thereof.