ML20028F948

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NRC Physical Research on Radiological Source Term
ML20028F948
Person / Time
Issue date: 06/30/1982
From: Kelber C
NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
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ML20027A699 List:
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FOIA-82-530 NUDOCS 8302070205
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NRC'S PHYSICAL RESEARCH ON THE RADIOLOGICAL SOURCE TERM By Dr. Charles N. Kelber l

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TITLES Figure 1:

Schematic of Plant and Models Figure 2:

Schematic of Plant and Tests Figure 3:

Schematic of Uncertainty Estimates

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Figure 4:

Schematic of Containment Loading Models i

Figure 5:

Fission Product Release and Collection System I

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Fission Product Release' Test Apparatus Figure 7:

Some Early Results from Fission Product Release Tests Figure 8:

One-Half-Kilogram Core-Melt Induction-Heated Aerosol Generator.

Figure 9:

Fuel Mock-Up in Induction Heater U

Figure 10:

Clad Fuel Attached by Silver (1400 C)

Figure 11:

Silver-Zircaloy Candling Process Figure 12:

View of the Nuclear Safety Pilot Plant Figure 13:

Plan of Fission Product Reactor Test Facility Figure 14:

View of Fission Product Reactor Test facility Figure 15:

Ba and Sr (OH) Vapor Pressure Curves e

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s ABSTRACT Inspection of data from the TMI-2 accident revealed significantly less radiciodine and other radio-isotopes released into the containment atmosphere than would have been predicted on the basis of widely used assumptions regarding fission prod)Ct release and transport, such as those used in the Reactor Safety Study (WASH-1400).

Work in progress had identified some areas of considerable conservatism, particularly aerosol settling in containment.

Further investigation was urged to determine if the radiological source term associated with accidents were properly estimated for purposes of:

site evaluation, emergency planr.ing, environmental qualification of equipment, and probabilistic risk analysis.

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An early step was an assessment of the technical state-of-the-art, a'nd i

this was done in the widely distributed report, NUREG-0772.

The research program is devoted to establishing a data base for best estimate calculations of the components and magnitudes of the radioisotopes released to the containment in a wide range of posulated accident sequences.

The verified best estimate codes will be used to benchmark and audit, faster-running, less precise probabilistic risk analysis codes.

Since the data and codes are needed for a number of uses, the research plan is organized to produce data that can be readily assimilated. Accident sequences for najor reactor types are defined through probabilistic risk analysis, and the relative frequency of the sequences is used to rank order sequences. At the same time, key dependencies of processes of production and depletion of products have been noted. For those dependencies common to most high ranking sequences (e.g. high temperature fission product chemistry) separate programs of research are in place.

Concurrently, detailed studies of accident sequences define, using best current technology, the expected source term.

It is found that in sequences dominated by decontamination in suppression pools, the source term is dominated by fission products transported through bypasses, and the decontamination factor thus need not be determined with high precision if it is more than roughly a factor of ten; if it is less, it is probably unimportant.

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Inspection of data from the THI-2 accident revealed significantly less radioiodine and other radio-isotopes released into the containment atmosphere than would have been predicted on the basis of widely used assumptions regarding fission product release and transport, such as those used in the Reactor Safety Study (WASH-1400). Work in progress had identified some areas of considerable conservatism, particularly aerosyl settling.in containment. Further investigation by the NRC was urged to determine if the radiological source tenn associated with accidents were properly estimated for purposes of:

site evaluation, emergency planning, environmental qualification of equipment, and probabilistic risk analysis.

Subsequently, the Commission, directed the staff to develop plans. to resolve the issues.

An early step was an assessment of the technical state-of-the-arp, and this was summarized in the gidely distributed report, NUREG-0772.

A parallel report, NUREG-0771, was also undertaken to summarize the regulatory i.apacts of source term assumptions. Taken together, these two reports define what the NRC needs to know and what it does know; the difference represents the goals of research. Taking into account the work being performed elsewhere, the Office of Research formulated a plan to fjll the gaps; the plan is summarized in the Long Range Research Plan.

I will describe the' highlights of the work in this paper.

The purpose of this article is to provide a review of the NRC sponsored work now well underway and selected examples of specific research.

One should keep in mind that significant work in this field is also being sponsored abroad and, in the U.S., and the Atomic Industrial Forum's IDCOR Project as well as by reactor vendors. We do, and intend to continue to, take full advantage of the work done by others.

1 Letter to NRC Chairman Ahearne from D. Campbell, A. Malinauskas, W. Stratton, August 14, 1980 2" Technical Bases for Estimating Fission Product Behavior During LWR Accidents," NUREG-0772 (June,1981) 3

" Regulatory Impact of Nuclear Reactor Accident Source Term Assumptions,"

NUREG-0771 (June,1981) 4 "Long Range Research Plan," NUREG-0784 (Revised yearly)

Figures 1 - 4 represent an effort to show in a single chart the various models and ex eriments that represent the major milestones of our work.

r The nodels, and the codes they are used in, are the major produ'ct of the work. The test program will furnish data to establish the models and verify the codes.

Best estimate codes will be used to calculate consequences of severe accidents and to benchmark and audit probabilistic risk assessments. Those calculations support such key Commission objectives as possible siting revisions and severe accident rulemaking. These figures illustrate the models and experimental programs that are directed toward analyzing fission product release and transport behavior under severe accident conditions.

Also, we will shcw our estimates of the current uncertainties associated with estimating fission product behavior within each physical transport regime (e.g. RCS, containment, etc.)

Finally, we will show the mass and energy transport models which are needed to define the accident conditions which control fission product behavior.

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The base diagram, Figure 1, presents a simplified schematic diagram of a PWR* plant which has experienced a transient initiated accident which has resulted in loss of cooling, core uncovery, core melting, and containment rupture. The release pathway from the core to the containment is through the hot leg piping, pressurizer, PORY and pressurizer quench tank (not shown for simplicity). For certain periods of the accident there may be water remaining in the pressurizer through which the steam, hydrogen, and fission product flow must pass, if the leak is through that component.

Following vessel melt-through the pathway is from the lower cavity into the main containment of compartments.

This figure also shows the models that are available, or under development, for analyzing fission product behavior under LWR accident conditions.

Two types of models have been identified. Those labeled with a "P" are f ast-running, simpler models, principally developed for plant specific probabilistic risk assessment studies. An "M" indicates those models which are detailed, or mechanistic, or first-principles models.

The mechanistic codes are intended for generic source term evaluations and for benchmarking the PRA models.

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  • Much of this discussion applies equally, in context, to a BWR.

The next figure, figure 3, shows our estimate of the uncertainty assoc'ated with predicting fission product transport behavior in the various transport regimes.

E is the current estimate cormonly used in PRAs C

today and the factors shown are rultipliers which indicate the undertainty The estimates shown are for process attenuation factors for range.

all regimes except release from the core.

In this regime the indicated parameter is the release rate.

It should be noted that the bounding values show1 here cannot be simply multiplied together (or combined statistically) to get an overall uncertainty bound since the attenuation mechanisms within each transport regime are not independent.

Examples of Current Estimates E

Regime

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- Release from fuel Calculated release rate for specific fission products from NUREG-0772 m

I RCS attenuation E = 1-(I.e. no attenuation)

C Waterpool scrubbing in RCS EC"I Containment attenuation EC"3

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leakage pathways Finally, figure 4 shows the mass and energy transport models which are required to define the accident conditions from which fission product release and transport behavior are determined.

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Models are used to calculate the initial rate of coolant loss from the

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system and the rate of core uncovery.

Follcwing core uncovery other models describe core heatup, cladding oxidation, cladding and fuel melting, and eventually fuel slumping.

Additional models describe the interaction of the fuel with residual i

water remaining in the lower RV and the attack on the vessel bottom head by the core debris.

Following " vessel failure, models are needed which describe the interaction of the core debris with concrete (and water) in the lower reactor cavity.

Througho0t this core heat-up, slumping, and vessel attack process models l

are required which describe hydrogen and steam flot: through the core and primary system and the heat transfer to structues within the RCS.

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i f*odels are needed to describe the pressure, temperature, and relative humidity of the containment atmosphere as a function of time te detemine the loads on the containment structure and to determine the environmental conditions for necessary equipment.

Models are needed to descirbe the buildup, distribution, ignition, and combustion of hydrogen in the containment.

Now I want to describe some of the test programs currently underway; I'll not describe the models in detail since refinements in current descriptions depend on test results.

Most of our work is proceeding at Oak Ridge and Sandia National Laboratories, and at Battelle Memorial Institute in Columbus.

In a few months severe fuel damage tests will also lagin in the Power Burst Facility at Idaho National Engineering Laboratory, where a miti-pin bundle will be heated in typical steam flow conditions, to conditions of major The fission products that are released will be sampled fuel damage.

at various times during the heat-up; an at smpt will be made to identify

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i products in the liquid vs. those in the gas stream.

i' At Oak Ridge four activities are taking place:

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The identification of fission products released from fuels of various irradiations under controlled, high temperatures. This work also seeks to develop correlations regarding the adsorption of the volatile products and their condensation on aerosol components.

2.

The characterization of the aerosol source term associated with core melt.

3.

The transport of aerosols in co'ntainment.

4.

Aqueous phase chemistry.

The lay-out of the fission product release test is shown in figure 5; it ir similar to past tests including tests on release at high temperatures from gas-cooled reactor fuel.

Of particular importance in developing adsorption correlations is the thermal gradient tube.

Figure 6 is a photograph of the equipment.

Some early results from these tests are shown in the next figure, More number 7; note the significant fraction of adsorbed material.

results are expected in the next few weeks.

Next, we noted on NUREG-0772 a lack of data on the character of aerosol produced by core melts. This characterizaticn is important becau:,e aerosol attenuation by agglomeration and settling is a major factor in consequence reduction.

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Core melt sanples about 1 kilogram in weight are produced by the induction heating of fuel in a large bell jar with a controlled, steam filled (Figure 8) The fuel in the induction heater is shown atmosphere.

in figure 9.

Aside from aerosol primary particle measurements, these tests show other interesting results. When control rods are formed from silver-indium-cadmium alloys, this material will melt before other The silver is particualarly aggressive in forming low temperature metals.

Figure 11 shows eutectics with the zircaloy clad, as shown in figure 10.

two steps in the melting process correlated with fuel temperature.

Finally, various test aerosols are released into the Nuclear Safety Pilot Plant (figure 12) to test the adequacy of codes such as HAARM and QUICK to r:odel agglomeration,in controlled, steam-filled, and dry atmospheres.

Aqueous phase fisison product chemistry, includhg chemical kinetics, is also being studied at Oak Ridge.

The vapor phase chemistry and kinetics at high temperatures is being The major emphasis is on ;

studied at Sandia National Laboratories.

A Iodine, Cedum, and Tellurium chemistry, but other elements are also'.

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Figure 13 is a plan of the apparatus' which uses a Laser Raman studied.

A view of the apparatus is shown in Spectroscope as the major tool.

Finally, in figure 15 we show some Barium vapor pressure figure 14.As you can tell, this tu program covers a lot of territory.

curves.

Major modeling efforts are' carried out at Battelle Memorial Institute.

The codes have been listed earlier in figure 1; containment atmosphere modeling itself is being done at Sandia, but Battelle has respons'bility for verifying the adequacy of the aerosol models as well ai supplying the interface with the TRAP-MELT models of the processes in the primary coolant system.

There are some other efforts sponsored by NRC tSat I have not described, these include:

Tests of fission product release during melt / concrete and melt water interactions (Sandia);

Evaluation of engineered saf ety feature effectiveness under severe accident conditions (BWL).

Analysis of past reactor accidents (BWL/INEL).

Taken together with work by EPRI, industry, and foreign sources (under exchange agreements which exist with most countries doing significant amounts of safety work), it is our conclusion that a sound program The Comission has charged the to resolve key issues is underway.

staff with the task of developing an improved representation of the We aim radiological source term by the second quarter of next year.

to complete this task on time.

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