Severe Accident Conditions Research Articles

Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted.

Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

Within the framework of the International Source Term Programme (ISTP), the VERDON programme aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a light water reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO2 fuel and FP releases at very high temperature (≈2873 K) in a reducing atmosphere. Post-test qualitative and quantitative characterisations of the VERDON-1 sample led to the proposal of a scenario explaining the phenomena occurring during the experimental sequence. Hence, the fuel and the cladding may have interacted which led to the melting of UO2-ZrO2 alloy. Although no relocation was observed during the test, it may have been imminent.

Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

Critical heat flux for downward-facing pool boiling on CANDU calandria vessel

Under a severe accident condition in CANDU reactors, damage to the reactor is expected such that a significant amount of core material can melt and relocate downward to the bottom of the vessel. It is essential that the vessel remains intact and mechanically strong to accommodate the core melt. However, a thermal creep failure is postulated to occur if sustained critical heat flux is instigated in the surrounding shield tank water. A mechanistic model is developed to predict the critical heat flux variations along the downward facing outer surface of calandria vessel. The hydrodynamic model considers a liquid macrolayer beneath an elongated vapor slug on the heated surface. Local dryout is postulated to occur whenever the fresh liquid supply to the macrolayer is not sufficient to compensate for the liquid depletion within the macrolayer. A boundary layer analysis is performed, treating the two phase motion as an external buoyancy driven flow, to determine the liquid supply rate and the local critical heat flux. Model shows good agreement with the available experimental data. The model has been modified to take into account the effect of subcooling and has been validated against the empirical correction factors.

Recent developments in the GENESIS code based on the Legendre polynomial expansion of angular flux method

Abstract This paper describes recent development activities of the GENESIS code, which is a transport code for heterogeneous three-dimensional geometry, focusing on applications to reactor core analysis. For the treatment of anisotropic scattering, the concept of the simplified Pn method is introduced in order to reduce storage of flux moments. The accuracy of the present method is verified through a benchmark problem. Next, the iteration stability of the GENESIS code for the highly voided condition, which would appear in a severe accident (e.g., design extension) conditions, is discussed. The efficiencies of the coarse mesh finite difference and generalized coarse mesh rebalance acceleration methods are verified with various stabilization techniques. Use of the effective diffusion coefficient and the artificial grid diffusion coefficients are found to be effective to stabilize the acceleration calculation in highly voided conditions.

Assessment of Ultimate Load Capacity of concrete containment structures against structural collapse

Nuclear safety regulations in India require assessment of containment structure for Ultimate Load Capacity (ULC) for internal pressure. ULC and associated behavior of containment is evaluated with respect to various performance indicators based on defined criteria. Non-linear FE analysis of the containment structures was carried out considering non-linear material properties of concrete and steel using ABAQUS and ANSYS software. Evaluation of ULC was undertaken to understand containment performance under severe accident conditions, as a round robin analysis based on 1:4 containment model test results conducted at SANDIA national Lab, USA. AERB also participated in the international round-robin exercise organized by BARC to assess the ULC of BARC’s containment test model (BARCOM). The test model is a 1:4 scale replica of 540MWe PHWR pre-stressed concrete inner containment structure of Tarapur Atomic Power Station Units 3&4. Based on the knowledge gained from these studies VVER 1000 reactor building constructed at Kudankulam (KKNPP) was assessed for ULC of its primary containment as a regulatory requirement. VVER containment study was first of its kind as it is a lined containment with cellular basement structure. Non-linear FE analysis in ABAQUS using layered shell model was performed to assess the ULC of the containment structure. Three times of design pressure is estimated to be ULC of the structure.

High-temperature oxidation of advanced FeCrNi alloy in steam environments

Alloys of iron-chromium-nickel are being explored as alternative cladding materials to improve safety margins under severe accident conditions. Our research focuses on non-destructively investigating the oxidation behavior of the FeCrNi alloy “Alloy 33” using synchrotron-based methods. The evolution and structure of oxide layer formed in steam environments were characterized using X-ray diffraction, hard X-ray photoelectron spectroscopy, X-ray fluorescence methods and scanning electron microscopy. Our results demonstrate that a compact and continuous oxide scale was formed consisting of two layers, chromium oxide and spinel phase (FeCr2O4) oxides, wherein the concentration of the FeCr2O4 phase decreased from the surface to the bulk-oxide interface.

Probabilistic Assessment of Severe Accident Consequence in West Bangka

Probabilistic dose assessment for severe accident condition is performed for West Bangka area. Source-term from WASH-1400 reactor analysis is used as a conservative release scenario for 1000 MWe PWR. Seven groups of isotopes are used in the simulation based on core inventory and release fraction. Population distribution for Muntok district and the area within a 100 km radius is obtained from 2014 data. Meteorological data is provided through cyclic sampling from a database containing two-year site-specific hourly records in 2014-2015 periods. PC-COSYMA segmented plume dispersion code is used to investigate the assumed the consequence of the accident scenario. The result indicates that early or deterministic effect is important for areas close the release point while long-term or stochastic effect is related to population distribution and covers area of up to 100 km from the release point. The mean annual expected values for early mortality and late mortality for the population within 100 km radius from Muntok site are 2.38×10−4yr−1 and 1.33×10−3 yr−1 respectively.

MELCOR severe accident analysis for a natural circulation small modular reactor

A natural circulation small modular reactor (NCSMR) is designed and the severe accident of the NCSMR was analyzed with MELCOR code in this work. The NCSMR is a 330 MWt light water reactor no reactor coolant pumps and no active safety injection systems similar to the design of NuScale reactor. A MELCOR code model was developed for the reactor. The MELCOR model shows consistent results in steady state operation with the design values. The NCSMR severe accident analyzed in this work was induced by the small break loss of coolant accident in the reactor pressure vessel and a crack in the containment pool wall. The decay heat removal system is assumed to fail for the main steam line break. The results indicated that the RPV pressure decreases quickly while the containment pressure increases followed by the opening of reactor vent valves. The core exposes and heat up with the coolant leakage and finally collapsed to the lower plenum, leading to the lower head breach. High temperature core corium quickly broke the relatively thin containment vessel wall, and was discharged to the containment pool floor, leading to the molten corium concrete interaction. The fission products release during the accident process was also analyzed. The severe accident sequence shows that the NCSMR has much longer time duration for the reactor vessel lower head failure to occur compared to conventional commercial nuclear power plant under severe accident conditions. The analysis results help identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents in NCSMR.

Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

Abstract The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

The accumulator effects on in-vessel severe accident progression of a three loop PWR nuclear power plant in a SBLOCA without safety injection system

This paper aims to evaluate the effects of coolant injection into a three-loop Westinghouse plant reactor vessel by the accumulators during small break loss of coolant accident, under severe accident condition. Using RELAP5/SCDAP mod3.4 computer code, a best estimate simulation from initiating SBLOCA event of 2, 3, and 4-in. without active part of the emergency core cooling system, in conditions with and without accumulators, is conducted until the rector pressure vessel failure. The results show that, the actuation of the accumulators lead to postponing the reactor vessel failure. Therefore, although the accumulators are designed for a large break LOCA, their actuation is very effective on the delay of a vessel failure.In this paper, the melt core concrete interaction is analyzed using CONTAIN2.0 code for the 4-in. break size without ACCs case as the most serious cases for the time of lower plenum failure. The result for 52.8h is reported. The in-vessel molten material retention strategy for this case is studied using RELAP5/SCDAP code. The results show that lower plenum integrity is preserved during the severe accident that causes reactor core melt.

The influence of the crust layer on RPV structural failure under severe accident condition

The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

Investigation of Zircaloy-2 oxidation model for SFP accident analysis

The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study.

Current Severe Accident Research and Development Topics for Wolsong PHWR Safety in Korea

In Korea, pressurized heavy water-cooled reactors (PHWR) account for 17% of operating units and have taken an important role in providing national energy supply. The recent biggest issue in domestic PHWR community was the continued operation of the Wolsong-1 CANada Deuterium Uranium (CANDU) plant, which has recently been approved to operate for 10 more years after a 30 year design life. In relation to this issue, various actions from both post-Fukushima lessons and Wolsong-1 (WS1) stress test results are being taken. In KAERI R&D, the following topics are studied to support the basis for these actions. First, PHWR severe accident issues such as (1) primary heat transport system (PHTS) overpressure protection capability, (2) containment overpressure protection capability, and (3) bypass source term are evaluated. Second, a computer tool (called MAAP–ISAAC) has been modified and updated to support analyzing Wolsong severe accident issues. Third, a decision supporting tool, called Severe Accident Management Expert (SAMEX)–CANDU, has been developed to aid emergency response experts under severe accident conditions.

Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900–1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

Influence of insoluble aerosols on the heat-exchanging characteristics of the SPHR CV in the case of severe accidents

Analysis of the processes in severe accidents at nuclear power plants has allowed distinguishing four characteristic stages of the accident, which differ in condensation rates, concentration and composition of aerosols (table 1) [1]. The released aerosol may condense on the heat transfer surfaces and reduce the effectiveness of the work of PCHRS under severe accident conditions at the deposition of insoluble aerosols on its heat transfer surfaces.

Study on effect of steam condensation on aerosol diffusiophoresis in severe accident condition

Study on effect of steam condensation on aerosol diffusiophoresis in severe accident condition

Analysis of hydrogen product for reactor core loaded annular fuel in severe accident condition

Analysis of hydrogen product for reactor core loaded annular fuel in severe accident condition

Dose rate applied to spent fuel damage degree evaluation system under severe accident condition

Dose rate applied to spent fuel damage degree evaluation system under severe accident condition