Source terms designate typical environmental releases of radioactive substances,

Transcription

1 1 2 Source term evaluation studies for PWRs J. FLEUROT (IRSN) J.-M. EVRARD (IRSN) B. CHAUMONT (IRSN) Source terms designate typical environmental releases of radioactive substances, which are also the envelope for the various accident conditions postulated for nuclear facilities. Source terms are used to define procedures for protecting human populations. Over the past few years, source term evaluation for pressurized water reactors (PWRs) has been the subject of two complementary studies at IRSN. The first of these is a level 2 probabilistic safety assessment (PSA) for 900 MWe PWRs that systematically analyzes different severe accident scenarios (with core meltdown), as shown in figure 1. The scenarios are quantified both in terms of probability of occurrence and any resulting radioactive releases. The study deals with all possible modes in which reactor containment integrity can be lost. A preliminary version of the level 2 PSA, based on the level 1 PSA published in 1990, whose aim was to assess the frequency of accidents leading to core meltdown, has already been conducted, but only for initial reactor power states. This preliminary study has served to devise appropriate methods and identify issues for which existing knowledge or calculation tools needed improvement. The second source term study is geared to reviewing PWR reference source terms, which are particularly instrumental in devising plans for protecting the populations around nuclear power plant sites. This calls for analysis of a small number of accident scenarios considered to be envelope that are also likely, in the long term, to cause loss of containment integrity (this does not, therefore, include the most improbable transients). Both studies are currently the subject of technical discussions with the utility, who is performing similar work. Figure 1 Severe accidents in a PWR. 12 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE

2 1 Level 2 probabilistic safety assessment Grouping level 1 PSA sequences into plant degradation states To develop the interface between PSA1 and PSA2, a list of interfacing variables was compiled to permit transfer to PSA2 of all data useful in characterizing accident progression beyond the point where the core becomes uncovered. Each level 1 PSA sequence is thus assigned a given set of values for interface variables. Sequences with the same set of values for these variables are then grouped into categories of plant degradation states (PDSs) which serve as input data for level 2 PSA. Developing an accident progression event tree Level 2 PSA is based on construction of an event tree known as the accident progression event tree (APET). This tree describes an ordered sequence of events relating to systems operation, human actions and physical phenomena that influence the course of the accident. Each APET event is associated with a branch of the tree and a quantification model whose role is to determine conditional probabilities at that branch along with the values of certain variables likely to evolve along a branch of the tree. The APET is probabilistically quantified for each PDS. Each of the successive branches considered during this process constitutes a level 2 accident sequence. ( GENERAL APPROACH The level 2 probabilistic safety assessment (PSA) for 900 MWe PWRs is an extension of the level 1 assessment. While level 1 PSA is concerned with frequency of accidents leading to core meltdown, the aim of level 2 PSA is to assess the frequency and levels of environmental releases resulting from such accidents. It comprises four main stages, as depicted schematically in figure 2. function of variables that significantly affect environmental releases of radioactive substances. Assessing releases The last stage in the study consists in evaluating radioactivity releases to the atmosphere for each APF. A special probabilistic quantification code (KANT) has been developed within the level 2 PSA project. This software is notably intended to represent and quantify the severe accident event tree, represent and group together accident progression families, evaluate the release levels corresponding to each family, display results and estimate uncertainties in them. Figure 2 The aim of level 2 PSA is to assess the frequency and levels of environmental releases resulting from severe accidents. Schematic representation of the main steps in level 2 PSA. The safety of nuclear reactors Grouping level 2 PSA sequences into accident progression families On completion of a probabilistic APET quantification, level 2 accident sequences are grouped into accident progression families (APFs) as a SCIENTIFIC AND TECHNICAL REPORT

3 ( All physical phenomena contributing to a severe accident are explicitly represented in the APET. Figure 3 Computer codes used for PSA2. EVALUATING SYSTEM BEHAVIOR AND HUMAN ACTIONS A program has been inaugurated for the study of equipment such as electromechanical actuators for pressurizer relief valves, and joints in the recirculation lines of low head safety injection and containment spray systems, under the environmental conditions associated with severe accidents. This program, which revolves around functional analyses, tests and expert opinion, is now underway. Its results will be used for the final version of level 2 PSA. One detailed supporting study focuses on containment leakage due to preexisting leaks or random failures in containment-critical valves or components during severe accidents. The BETAPROB code has been developed to exhaustively identify potential containment leakage paths and quantify the probability of such leakage. A special model for human reliability assessment (HRA) has likewise been developed to evaluate the probability of failed human action during severe accidents. This model accounts for the national response organization in such situations and credits existing action guidelines for severe PWR accidents. It affords identification and weighting of influential factors (quality of available data, complexity of the decision-making process, difficulties in taking action, etc.). Quantifying the probable failure of human actions calls on the experience of engineers who have previously participated in emergency response drills. EVALUATING PHYSICAL PHENOMENA The APET explicitly represents all of the physical phenomena involved in a severe accident. This involves singling out the different situations and performing a suitably representative calculation for each. Results of these calculations are compiled in tabular form for processing at the time of probabilistic APET quantification. Figure 3 shows physical phenomena credited in this evaluation and calculation tools used. The various accident sequences leading to a core melt are studied by means of the joint IRSN-EDF- FRAMATOME thermal-hydraulics code CATHARE and IRSN s SIPA simulator. Core degradation (heatup, clad oxidation and hydrogen release, U- Zr-O interactions, etc.) is modeled by the VUL- CAIN code, which is part of the ESCADRE (IRSN) package. This code is supplemented with special capability for modeling corium flow into the vessel bottom head and vessel rupture. In some accidents, primary system pipe breaks and steam generator tube ruptures may result from a combination of high pressures and temperatures. These situations have already been studied with the NRC s MELCOR code and appropriate basic mechanical models. More detailed studies involving ICARE-CATHARE (IRSN), TRIO-U (CEA) and CASTEM (CEA) codes are now underway. 14 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE

4 1 The steam explosion phenomenon produced by contact between the corium (mixture of molten core material and steel) and water present at the bottom of the vessel is modeled by IRSN s CIGA- LON code, supplemented by special models to depict an upward moving corium slug ( missile ) and possible vessel closure head rupture. These models make use of results obtained in BERDA (FZK) experiments. Related studies are also being performed with the CEA s mechanistic MC3D code. Vessel failure at high pressure may cause largescale dispersion of corium fragments through the containment, resulting in so-called direct containment heating. This phenomenon has been modeled using the RUPUICUV code (ESCADRE), after adjustment for experimental results characteristic of French reactor pit geometry (from KAERI and Sandia SURTSEY experiments). Corium ablation of the concrete basemat has been studied using the WECHSL (ESCADRE) and the MELCOR codes. These codes predict fairly different ablation rates in the long-term phase. Modeling and validation efforts have therefore been undertaken on this subject. The JERICHO code, also part of the ESCADRE package, has been used to model thermalhydraulic developments and slow hydrogen combustion in the containment; additional studies on possible fast hydrogen combustion have been conducted using the GRS code RALOC and special models based on Russian RUT experiments. These studies are being pursued further via the CPA code from the IRSN-GRS package ASTEC. The CASTEM code has been applied to fine analysis of containment mechanical behavior. This involves non linear calculations using 3D models of an entire reactor containment and the equipment access hatch zone. Only partial results have been processed for the preliminary level 2 PSA. This version will be subsequently updated to credit results from all of the calculations described here. EVALUATING RADIOACTIVE RELEASES A simplified model has been developed for assessment of radioactivity releases to the environment. This model makes use of a set of hypotheses developed for the source term review described below. The model accounts for all phenomena with impact on environmental releases. A significant number of parametric calculations has likewise been performed with the ESCADRE package to obtain a set of simple laws and correlations. The simplified evaluation approach rests primarily on: a linear law for aerosol emission during the meltdown phase in the vessel; exponential laws for natural deposition of aerosols, spraydown of aerosols and iodine and iodine adsorption on non immersed, painted surfaces; coefficients for conversion of molecular to organic iodine, aerosol fractions deposited in primary, secondary and auxiliary systems and the efficiency of VHE filters, auxiliary building ventilation system traps and the containment depressurization/filtration system (procedure U5); leakage rates determined by containment leakage cross sections, the accident phase underway and spray system operation. Main results provided by the code used to evaluate radioactivity releases include activity levels released per accident phase and family of species considered (inert gases, aerosols or volatile iodine in molecular or organic form). This software likewise determines time of onset and duration of release (figure 4). EVALUATING UNCERTAINTIES A partial evaluation of uncertainties related to physical phenomena was included in the preliminary version of level 2 PSA. This entailed assessing uncertainties affecting physical Figure 4 Probabilistic quantification code. The safety of nuclear reactors SCIENTIFIC AND TECHNICAL REPORT

5 1 - N. Rasmusen, Nuclear safety study: an assessment of accidents in US commercial nuclear power plants, parameter models (by calling on expert opinion), then using Monte Carlo methods to propagate these uncertainties through physical calculations and in the APET. A broader evaluation is slated to take place at a subsequent stage in the project. Reviewing reference source terms The study conducted under this heading aims at reviewing level S3 of the source term that, in the 1970s, served as a technical basis for offsite emergency plans (PPIs) applicable in cases where there is risk of environmental release. Source term level S3 is an old reference, since it was derived from the WASH report, adjusted to French reactors, then partially updated at the end of the 80s, after various emergency procedures and specifically procedure U5, which called for installation of a sand filter and later a prefilter in the containment became effective. This S3 source term corresponds to releases that are deferred by some means affording a degree of radionuclide retention. Because of their low probability, situations leading to containment bypass or early loss of containment leaktightness are not considered in this study. The main objectives of reviewing the S3 source term are: to ensure the best possible integration of currently available knowledge and identify any points that, in coming years, could benefit from new experimental findings and model developments; to allow for changes in plant facilities, the most recent operating experience and any improvements made to plant equipment. Work was carried out in three major phases: choice of representative accident sequences and a set of hypotheses to be used in calculating emission of radioactive substances, their behavior in the plant and release to the environment; performance of these calculations and analysis of their results, sensitivity studies; calculation of short-term radiological consequences (plume passage). The accident sequences selected and all calculation hypotheses were selected to be reasonably penalizing in terms of environmental releases and offsite impact. Three sequences with large initial breaks were considered for 900 MWe and MWe PWRs (which have single- and double-wall containments respectively). Hypotheses were selected to enable calculation of behavior and transfer to the environment of radioactive releases, using currently available computer codes. These hypotheses concern the following key data: rates of release of the major elements from fuel, in the event of total core meltdown, and deposits upstream of the break; these values were determined by analysis of VERCORS test results (IRSN program conducted by CEA- Grenoble) and the first two Phebus tests (IRSN- Cadarache); volatile iodine production at the break (based on IRSN Phebus FP tests), molecular iodine production from the sump (distinguishing between cases where there is or is not enough silver to trap the iodine) and conversion of iodine adsorbed on painted containment surfaces to organic iodine (based essentially on IRSN experiments); retention of aerosols and nonorganic iodine in the prefilter and sand filter (procedure U5, applied to depressurize the containment in the event of significant pressure rise) and in soil, after basemat melt-through caused by corium/concrete interaction; containment leaktightness and spray system operation. Figure 5 Phebus FP results illustrated by radiography and tomograms. 16 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE

6 1 The following conclusions can be drawn for core fractions released to the environment in the sequences used as a basis for this review: noble gases are nearly always all released; aerosol releases (especially of cesium) are well below S3 levels; nonorganic iodine releases (in molecular and aerosol form) are also well below S3 levels; organic iodine releases are of the same order of magnitude as for S3. Short-term radiological impact was calculated as total effective dose (sum of external doses and the effective dose committed by inhalation) and total dose to the thyroid gland (sum of external doses and dose committed to the thyroid by inhalation) for a one-year-old child, which corresponds to the most sensitive population category in the event of significant iodine release. These calculations were based on the most recent dose coefficients (from European Union publications and those of the International Commission on Radiological Protection). It is clear that short-term radiological impact is of the same order of magnitude as that corresponding to the S3 source term, but that deposits in the soil and thus the contamination of the food chain are significantly lower than those due to S3, especially for cesium and other elements released as aerosols. Using R&D and identifying needs The two studies described in this article have made ample use of knowledge acquired experimentally by IRSN, whether through its own pro- grams or those of CEA, or via international partnerships. Review of reference source terms drew on the most recent results available from several experimental programs: IRSN s own Phebus FP tests (figure 5), along with VERCORS tests conducted by CEA on behalf of IRSN (figure 6) and HI/VI (ORNL) experiments provided a great deal of data for evaluation of rates of fission product release from fuel; and the first two Phebus FP tests enabled a preliminary estimate of the gaseous iodine fraction released at a primary system break; these same tests underscored the role played by the silver resulting from control rod degradation as an iodine sink in the containment. all of the analytical tests performed, notably at IRSN, enabled assessment of organic iodine formation during the accident. As already mentioned above, in the two studies, computer codes developed for severe accident calculations at IRSN, but also at CEA or by organizations in other countries, could be applied to a large number of accident sequences. Special models were also devised for some of these cases. The two studies led to identification of certain points for which further efforts are required to improve subject knowledge and modeling capability. These included: primary system induced breaks during severe accidents: the preliminary version of level 2 PSA showed that steam generator tube ruptures could not be excluded for accidents occurring at high pressures. However, some conservative hy- The safety of nuclear reactors Figure 6 Vercors loop. SCIENTIFIC AND TECHNICAL REPORT

7 A program has been set up to gain better knowledge of material properties and conduct relevant thermal-hydraulic and mechanical studies. ( potheses need to be reconsidered; a program covering characterization of material properties, together with thermal-hydraulic and mechanical studies, has been set up; reflooding of the degraded core: the preliminary version of level 2 PSA revealed a large proportion of cases in which water could be injected into an already degraded core. A simplified approach to this scenario is being prepared; and a longer-term program has been proposed, which could include some experiments in the Phebus reactor; interaction between the corium and the concrete basemat of the containment. Studies have shown wide dispersion in results calculated by different computer codes. Efforts are therefore underway to improve models and validate them through MACE (OECD) experiments and CEA s ARTEMIS and VULCANO programs in which IRSN also participates; containment mechanical behavior and thermal response. Results of the preliminary level 2 PSA proved to be very sensitive to the pressure level at which significant loss in containment leaktightness occurs, given the current levels of uncertainty. In-depth studies based on non linear 3D models of the entire containment and the equipment hatch air lock are now underway and should alleviate these uncertainties; physico-chemical behavior of iodine in severe accident situations: a wide-ranging experimental program, including analytical and global experiments (Phebus FP), has enhanced understanding of iodine behavior, which is the most important contributor to the short-term radiological impact of severe accidents. However, better quantification of a number of processes identified during these tests will require further efforts, both in terms of experiments and modeling resources. Future projects The level 2 PSA project will now continue in two stages. The first stage will be a revision, in 2002, of the preliminary assessment, based on results of detailed containment mechanical behavior calculations. A new version of the level 2 PSA is currently also being drafted on the basis of the updated version of level 1 PSA and will likewise cover initial shutdown states and results of all supporting studies. In the longer term, level 2 PSA will include the results of the state-oriented level 1 PSA and the Fire PSA. Studies conducted to review reference source terms already offer a basis for discussion on crisis management with the French safety authority and the utility in charge of the PWR operation. 18 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE