Containment venting as an accident management strategy for BWRS with Mark I containments

A quantitative evaluation of primary containment venting was performed to assess its risk reduction potential. A boiling water reactor with a Mark I containment was evaluated by developing simplified containment event trees for its risk dominant sequences. Risk results were benchmarked with those from the NUREG-1150 risk rebaselining effort, and sensitivity studies then were performed. It was found that for station blackout sequences, containment venting by itself does not significantly reduce overall risk. For sequences involving loss of long-term decay heat removal or failure to scram, however, venting is potentially an important mechanism in preventing or delaying core melting. Subsequent studies show that when venting is combined with other potential containment improvements, there is a large potential for risk reduction.     

US NRC research on the integrity of piping in light water reactors

The paper describes the activities underway in NRC on the subject of LWR piping integrity as of the summer and fall of 1983. The paper is necessarily vague on certain topics of policy because they are either under review or are under development. Particularly in the area of BWR pipe cracking, events are very rapid so that positions and actions described in this paper may well be obsolete by the time it is published. Nevertheless, this paper is useful to show the intentions of NRC in the area of research for LWR piping, and it is also useful to document the status of the regulations on piping for which the research is being performed.     

PHEBUS FP: a severe accident research programme for current and advanced light water reactors

The ongoing PHEBUS FP programme is the centrepiece of an international co-operation investigating, through a series of integral in-pile experiments, key-phenomena involved in the progression of a postulated severe accident in a light water reactor (LWR). The dedicated PHEBUS facility offers the capability to study the degradation of real core material, from the early phase of cladding oxidation and hydrogen production up to the late phase of melt progression and molten pool formation. The subsequent release of fission products (FPs) and structural materials is also experimentally studied, including their physicochemical interactions, their transport in the cooling system, and their deposition in the containment. The revolatilisation of iodine due to radiochemical effects in the water of the sump and the amount of low-volatility FPs and transuranium elements reaching the containment are receiving a special interest, as large uncertainties related to their modelling subsist. FPT-0 and FPT-1, the first experiments of the programme, performed in December 1993, and July 1996, respectively, have demonstrated that the PHEBUS FP facility is capable of successfully attaining these objectives. They reached very advanced states of degradation, comparable to what was observed in TMI 2, and generated a wealth of results on FP behaviour. The resulting database has been—and will be—applied to develop and validate the computer codes used to assess the safety of the currently operating plants and to check the efficiency of accident management procedures. They will also support the design of future plants having the capacity to confine core melt-down accidents within their containments.     

Analysis of strategies for containment venting in case of severe accidents

During a core melt accident, a pressurization of the containment has to be expected, which could lead to a failure of the containment due to overpressurization. This failure mode is expected to be the most likely one for large dry containments under accident conditions. Also during a core melt accident, a large quantity of hydrogen may be generated, giving the potential of a loss of containment integrity due to violent hydrogen combustion. Timely venting of the containment atmosphere can prevent overpressurization and may perhaps make the hydrogen situation in the containment less severe. This paper discusses the thermodynamic consequences of different vent strategies for a large German PWR during core melt accidents.     

Hydrogen management and overpressure protection of the containment for future boiling water reactors

New design and evaluation method for hydrogen management of containment atmosphere have been developed for application in the future boiling water reactor (BWR). These are intended as a part of consideration of severe accidents in the course of design so as to assure a high level of confidence that a large release of radioactivity to the environment that may result in unacceptable social consequences can reasonably be avoided. Emphasis on hydrogen management and protection against overpressure failure is based on the insights from probabilistic safety assessments (PSAs) that late phase overpressure (and associated leakage) and molten corium concrete reaction (MCCI) need attention to ensure that containment remains intact, in case energetic challenges to the containment such as DCH (direct containment heating) or FCI (fuel coolant interactions) are practically eliminated by design or resolved from risk standpoint of view. The authors studied the use of palladium-coated tantalum for hydrogen removal from containment atmosphere in order to avoid pressurization of the containment with small free volume by non-condensable gas and steam. Its effectiveness for ABWR (advanced boiling water reactor) containment was evaluated using laboratory test data. Although further experimental studies are necessary to confirm its effectiveness in real accident conditions, the design is a promising option and one that could be backfitted upon necessity to existing plants for which pressure retaining capability cannot be altered. Also new evaluation method for flammability control under severe accident conditions was developed. This method employes a realistic assessment of the amount of oxygen and hydrogen gases generated by radiolytic decomposition of water under severe accident conditions and their subsequent transport from water to containment atmosphere.     

Studies on innovative hydrogen recombiners as safety devices in the containments of light water reactors

In order to prevent the containment and other safety relevant components from incurring serious damage caused by a detonation of the hydrogen/air-mixture generated during a severe accident in light water reactors (LWR) passive autocatalytic recombiners (PAR) are used for hydrogen removal in an increasing number of European plants. These devices make use of the fact that hydrogen and oxygen react exothermally on catalytic surfaces generating steam and heat.

Experimental investigations at several research facilities indicate that existing PAR systems bear the risk of igniting the gaseous mixture due to an overheating of the catalyst elements caused by strong reaction heat generation. Innovative devices could overcome existing limitations making use of the knowledge deduced from experiments performed at the REKO facilities at Forschungszentrum Juelich (FZJ).

The paper analyses the mechanisms of the thermal behaviour of catalytic plate-type recombiners and presents experimental results on existing and innovative devices for hydrogen removal introducing the modular recombiner concept.     

Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

The U.S. Nuclear Regulatory Commission recently identified a possible safety concern for pressurized water reactors. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: (1) describes the important aspects of the problem, (2) provides an initial analysis of the consequences, and (3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of Westinghouse design, the concern is greatest for those plants. There is less concern for most plants of Combustion Engineering design, and likely no concern for plants of Babcock and Wilcox design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a Westinghouse pressurized water reactor. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed.     

Risk assessment in setting inservice inspection priorities at light water reactors

As part of the Nondestructive Evaluation Reliability Program, sponsored by the U.S. Nuclear Regulatory Commission, Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish inservice inspection plans for nuclear power plant components. The method first uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The acceptable level of risk from structural failure for important systems and components is then apportioned as a small fraction of the total PRA estimated risk for core damage. This process determines the target (acceptable) risk and failure probability values for individual components. The Surry Unit 1 Nuclear Power Station was selected for pilot applications of the method. The specific systems addressed are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants.     

An overview of reactivity initiated accident behavior of rock-like fueled pressurized water reactors

Reactivity initiated accident (RIA) analyses of plutonium rock-like oxide (ROX) fueled PWRs have been carried out with the point kinetics calculations. As a result, the analyses have shown a very severe transient behavior of the ROX fueled PWR, which is unacceptable without any improvement. It was also found that the RIA behavior of ROX fueled PWRs can be improved by increasing the negative fuel temperature coefficient (_f). For this improvement, the additives in the ROX fuel such as UO₂ and ThO₂ were considered, as well as a ROX assembly partial loading UO₂ core. With UO₂ additive, it was successful to have satisfying _f and RIA behavior of ROX fuel core, while the partial loading core must be further improved. Besides the ROX-PWR RIA analytical study, the actual behavior of the ROX fuel pin under RIA condition has been experimentally investigated at the Nuclear Safety Research Reactor (NSRR) of JAERI. Though the ROX fuel pin failure mechanism with fuel melting seems quite different from that of UO₂ pin with cladding melting, the ROX pin failure threshold was found to be roughly the same as that of UO₂ in terms of accumulated energy per unit fuel volume.     

Development of methodology to optimize management of failed fuels in light water reactors

Fuel cladding is one of the key components in a fission reactor that confines radioactive materials inside a fuel tube. During reactor operation, however, cladding is sometimes breached, and radioactive materials leak from the fuel pellet into the coolant water through the breach. The primary coolant water is therefore monitored so that any leak is quickly detected; coolant water is periodically sampled, and the concentration of radioactive iodine 131 (I-131), for example, is measured. Depending on the measured leakage concentration, the faulty fuel assembly with leaking rod is removed from the reactor and replaced immediately or at the next refueling. In the present study, an effort has been made to develop a methodology to optimize the management for replacement of faulty fuel assemblies due to cladding failures using measured leakage concentration. A model numerical equation is proposed to describe the time evolution of an increase in I-131 concentration due to cladding failures and is then solved using the Monte Carlo method as a function of sampling rate. Our results indicate that, to achieve rationalized management of failed fuels, higher resolution to detect a small amount of I-131 is not necessarily required, but more frequent sampling is favorable.     

Escape of fission products from defective fuel rods of light water reactors

The escape behaviour of various fission product isotopes from defective fuel rods in PWRs and BWRs is analyzed.Diffusion in the UO₂ is the rate controlling step for the release of noble gases from defective fuel rods. The escape of fission iodine from defective fuel rods is controlled by a mechanism which includes migration and additional delay steps, probably in the nature of a chemical reaction.The inferred effective diffusion constants for fission gases are noticeably higher for defective fuel rods than for intact fuel rods. The difference is about two orders of magnitude. The enhancement of diffusion in defective fuel rods is believed to be due to the increase in the -ratio of the UO₂ in the defective fuel rods.     

Identification and evaluation of facilitation techniques for decommissioning light water power reactors

This paper describes a study sponsored by the US Nuclear Regulatory Commission to identify practical techniques to facilitate the decommissioning of nuclear power generating facilities. The objectives of these “facilitation techniques” are to reduce public/occupational exposure and/or reduce volumes of radioactive waste generated during the decommissioning process.The paper presents the possible facilitation techniques identified during the study and discusses the corresponding facilitation of the decommissioning process. Techniques are categorized by their applicability of being implemented during the three stages of power reactor life: design/construction, operation, or decommissioning. Detailed cost-benefit analyses were performed for each technique to determine the anticipated exposure and/or radioactive waste reduction; the estimated cost for implementing each technique was then calculated. Finally, these techniques were ranked by their effectiveness to facilitate the decommissioning process.This study is a portion of the NRC's evaluation of decommissioning policy and supports the modification of regulations pertaining to the decommissioning process. The findings can be used by the utilities in the planning and establishment of the activities to ensure all objectives of decommissioning will be achieved.     

Radiation-induced ‘long-cell’ (macro-cell) corrosion in light water reactors

In previous papers, the author has established various ‘long-cell’ (akin to ‘macro-cell’) corrosion configurations that exist in nuclear power plants. Among these, the radiation-induced corrosion cell is an important mechanism since it plays a major role in the corrosion problems found in primary water of the nuclear power plants. There are numerous experimental evidences indicating a potential difference induced by radiation, however, the exact mechanism of such phenomena has not been clarified. The author investigated the basic mechanism by combining radiation chemistry, electrochemistry and corrosion science to confirm the existence of radiation-induced ‘long-cell’ action.By performing a competition kinetic study, , reacting mainly with stable molecules are found to be responsible for inducing a large portion of the potential difference both in the PWR and BWR water chemistry environments. The hydrated electrons react at a cathodic half-cell thereby inducing reductive reactions in the mixed cell configuration. This method reproduces the reported and experimentally observed redox potential variation to a certain extent (observed in the INCA Test Loop in Sweden and the NRI-Rez BWR-2 Loop in Czech Republic). The author believes the results support the assumed corrosion mechanism although details are still debatable.     

Radiotoxicity hazard of inert matrix fuels after burning minor actinides in light water reactors

Ingestion radiotoxicity hazard index of inert matrix spent fuels are investigated after burning minor actinide (MA) isotopes in LWRs and compared with the hazard index of MOX and MA burning MOX (MOX+MA) spent fuels. As U-free fuels, ROX: (PuO₂+ZrO₂) and TOX: (PuO₂+ThO₂), are considered, in which MA's are added as oxides. The radiotoxicity hazard index of ROX+MA spent fuel is less than that of TOX+MA and MOX+MA spent fuels due to the lower density of actinides in spent fuel. Some of cooling years the toxic yield of ROX+MA spent fuel is even less than that of MOX spent fuel, if the initial loaded MA in ROX is about 0.5 at %.     

A fundamental approach to specify thermal and pressure loadings on containment buildings of sodium cooled fast reactors during a core disruptive accident

Reactor Containment Building (RCB) is the ultimate barrier to the environment against activity release in any nuclear power plant. It has to be designed to withstand both positive and negative pressures that are credible. Core Disruptive Accident (CDA) is an important event that specifies the design basis for RCB in sodium cooled fast reactors. In this paper, a fundamental approach towards quantification of thermal and pressure loadings on RCB during a CDA, has been described. Mathematical models have been derived from fundamental conservation principles towards determination of sodium release during a CDA, subsequent sodium fire inside RCB, building up of positive and negative pressures inside RCB, potential of in-vessel sodium fire due to failed seals and temperature evolution in RCB walls during extended period of containment isolation. Various heating sources for RCB air and RCB wall and their potential have been identified. Scaling laws for conducting CDA experiments in small-scale water models by chemical explosives and the rule for extrapolation of water leak to quantify sodium leak in reactor are proposed. Validation of the proposed models and experimental simulation rules has been demonstrated by applying them to Indian prototype fast breeder reactor. Finally, it is demonstrated that in-vessel sodium fire potential is very weak and no special containment cooling system is essential.     

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs.     

Spherical steel containments of pressurized water reactors under accident conditions: Investigation program and first results

The investigations will deal with the mechanical behavior of a free standing spherical containment shell built for the latest type of a German pressurized water reactor. The diameter of the containment shell is 56 m. The wall thickness is 38 mm. The material used is the ferritic steel 15MnNi63.The investigation program includes theoretical as well as experimental activities and concerns four different accidents which are beyond the scope of the common design and licensing practice: containment behavior under quasi-static pressure increase up to containment failure; containment behavior under high transient pressures; containment vibrations due to earthquake loadings (consideration of shell imperfections); containment buckling due to earthquake loadings. First results concerning the containment behavior under quasi-static pressure increase are presented. It turns out that the mechanical failure of the containment shell is controlled by plastic instability. A computer program to describe this problem has been developed and membrane tests to check the computational methods have been carried out.