BWR Containment Research Articles

Analysis of venting strategies and hydrogen concentration evolution during a station blackout in a BWR-6 containment using GOTHIC 8.3

After the Fukushima accident, the interest on hydrogen combustion hazard management has increased considerably. Many European nuclear plants reinforced their strategies for severe accident management with the installation of Passive Autocatalytic Recombiners (PARs) and Filtered Containment Venting Systems (FCVS), among others. In this study, the hypothesis of considering a preventive venting to try to reduce the hydrogen combustion risk is analyzed for a BWR-6 Mark III containment. The scenario chosen was the in-vessel phase of a Station Blackout accident. Cases with and without PARs have been tested.The venting action has been tested at different conditions to clarify its impact on the hydrogen concentration in the containment. One novelty of this study is to consider the difference between measuring the hydrogen concentration values present in the containment in the venting pipe rather than in the instrumentation position. The purpose is to try to understand the decisions that an operational crew can take with the instrumentation available and analyze the differences by modifying the location of the plant instrumentation.For the numerical analysis, numerous simulations have been carried out using a BWR containment model with GOTHIC 8.3. It has been observed that hydrogen concentration depends to a great extent on the containment venting strategy. The main conclusion is that the venting strategy, if properly performed, can reduce the risk associated with hydrogen for BWR-6 containments. During a SBO scenario, an early venting increases considerably the hydrogen combustion hazard, even in the cases with PARs.

Potential tellurium deposits in the BWR containment during a severe nuclear accident

The release of fission products to the environment is one of the concerns with nuclear power. During an accident, the most likely released are the volatile fission products i.e., tellurium. To evaluate the behavior of tellurium in the event of an accident, it was heated under different conditions (oxidizing, inert, reducing; both dry and humidified). The formed vapor was transported to surfaces (aluminum, copper, zinc) at room temperature that can be found in the BWR-containment.All formed deposits were examined for morphology and species. Moreover, the content of sodium hydroxide liquid traps following the metal surfaces and filter was also investigated. In these traps, the highest amount of tellurium was found under humid-reducing followed by humid-oxidizing conditions. In the deposit removed from the zinc surface acquired under the latter conditions, elemental analysis observed zinc, indicating a possible reaction between tellurium and zinc. The corresponding trap showed significant amounts of zinc.

Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.1

Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher water surface temperatures than in mixed conditions and thus, to higher containment pressures. The Effective Heat (EHS) and Momentum (EMS) Source models were previously introduced to predict the effect of small-scale direct contact condensation phenomena on the large-scale pool water circulation. In this paper, the EHS/EMS models are extended by adding the effect of non-condensable gases on the chugging regime. The EHS/EMS models are implemented in the GOTHIC code to model a full-scale Nordic BWR containment under different LOCA scenarios. The results show that thermal stratification can be developed in the PSP.

Severe accident progression in the BWR lower plenum and the modes of vessel failure

Most of our knowledge base on the severe accident progression in the lower plenum of LWRs is based on the data obtained from the TMI-2 accident. It should be recognized that the lower plenum of a BWR is very different from that of a PWR. Unlike the PWR, the BWR plenum is full of control rod guide tubes (CRGTs) with their axial structural variations. These CRGTs are arranged in a cellular fashion with each CRGT supporting 4 rod bundles. There are also a large number of instrument guide tubes (IGTs), each generally placed in the middle of 4CRGTs. Both the CRGTs and IGTs traverse the thick vessel bottom wall and are welded to their extensions which come to bottom of the core. The core-melt progression in the lower plenum is controlled by the structures present and they, in turn, influence the timings and the modes of vessel failure for a BWR.The uranium oxide–zirconium oxide core melt formed in the 4 fuel bundles is directed by the structure below toward the water regions in-between the 4 CRGTs. The FCI will take place in those water regions and some particulate debris will be created, although there is insufficient water for quenching the melt. A FCI may occur inside a CRGT if and when the melt enters the CRGT at its top opening or the melt in the water region between the four CRGTs breaches the wall of the CRGT.The important issue is whether the welding holding the IGT inside the vessel will fail and the bottom part of the IGT falls out creating a hole in the vessel with release of water and melt/particulate debris from the vessel to the dry well of the BWR containment. Similarly, the failure of CRGT could have water and melt/particulate debris coming out of the vessel. These modes of vessel failure appear to be credible and they could occur before any large-scale melting and melt pool convection takes place. These modes of vessel failure and the melt release to the containment will have very different consequences than those generated by the other modes of vessel failure.Such BWR plenum melt progression scenarios have been considered in this paper. Some results of analyses performed at KTH have been described. We believe that the issues raised are important enough to consider a set of experiments for verification and validation of the melt progression in a BWR plenum. Such experiments are proposed.

Heat transfer effect of an extended surface in downward-facing subcooled flow boiling

New BWR containment designs are considering cavity flooding as an accident management strategy. Unlike the PWR, the BWR has many Control Rod Guide Tube (CRGT) penetrations in the lower head. During a severe accident scenario with core melt in the lower plenum along with cavity flooding, the penetrations may affect the heat transfer on the ex-vessel surface and disrupt fluid flow during the boiling process. A small-scale experiment was performed to investigate the issues existing in downward-facing boiling phenomenon with an extended surface. The results were compared with a bare (flat) surface. The mass flux of 244kg/m2s, 215kg/m2s, and 177kg/m2s were applied in this study. CHF conditions were observed only for the 177kg/m2s case. The boiling curves for both types of surfaces and all flow rates were obtained. The boiling curves for the highest flow rate showed lower surface temperatures for the extended surface experiments when compared to the bare surface. The downstream location on the extended surface yielded the highest surface temperatures as the flow rate was reduced. The bubble accumulation and low velocity in the wake produced by flow around the extended surface was believed to have caused the elevated temperatures in the downstream location. Although an extended surface may enhance the overall heat transfer, a reduction in the local heat transfer was observed in the current experiments

B222 BWRの格納容器ベントにおける総合的な放射性物質放出抑制について(OS3 軽水炉・新型炉・原子力安全(4))

B222 BWRの格納容器ベントにおける総合的な放射性物質放出抑制について(OS3 軽水炉・新型炉・原子力安全(4))

Experimental investigation of void distribution in suppression pool over the duration of a loss of coolant accident using steam–water two-phase mixture

Studies are underway to determine if a large amount gas discharged through the downcomer pipes in the pressure suppression chamber during the blowdown of Loss of Coolant Accident (LOCA) can potentially be entrained into the Emergency Core Cooling System (ECCS) suction piping of BWR. This may result in degraded ECCS pumps performance which could affect the ability to maintain or recover the water inventory level in the Reactor Pressure Vessel (RPV) during a LOCA. Therefore, it is very important to understand the void behavior in the pressure suppression chamber during the blowdown period of a LOCA. To address this issue, a set of experiments is conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. The geometry of the test apparatus is determined based on the basic geometrical scaling analysis from a prototypical BWR containment (MARK I) with a consideration of downcomer size, downcomer water submergence depth and Suppression Pool (SP) water level. Several instruments are installed in the test facility to measure the required experimental data such as the steam mass flow rate, void fraction, pressure and temperature.In the experiments, sequential flows of air, steam–air mixture and pure steam-each with the various flow rate conditions are injected from the Drywell (DW) through a downcomer pipe in the SP. Eight tests with two different downcomer sizes, various initial gas volumetric fluxes at the downcomer, and two different initial non-condensable gas concentration conditions in the DW are conducted. Three distinct phases, namely, an initial phase, a quasi-steady, and a chugging phase are observed. The maximum void penetration depth is observed in the initial phase. A reduction in the void penetration depth is observed in the quasi-steady phase. As a result of low non-condensable gas concentration, chugging is observed at the tail end of the experiment. Chugging provides renewed void penetrations comparable to those in the initial phase. It is determined that the void distribution and area of void penetration in the SP is governed by the gas volumetric flux at the downcomer and the non-condensable gas concentration in the downcomer.

Experimental investigation of void distribution in Suppression Pool during the initial blowdown period of a Loss of Coolant Accident using air–water two-phase mixture

During the initial blowdown period of a Loss of Coolant Accident (LOCA), the non-condensable gas initially contained in the BWR containment is discharged to the pressure suppression chamber through the blowdown pipes. The performance of Emergency Core Cooling System (ECCS) can be degraded due to the released gas ingestion into the suction intakes of the ECCS pumps. The understanding of the relevant phenomena in the pressure suppression chamber is important in analyzing potential gas intrusion into the suction intakes of ECCS pumps. To obtain the basic understanding of the relevant phenomena and the generic data of void distribution in the pressure suppression chamber during the initial blowdown period of a LOCA, tests with various blowdown conditions were conducted using the existing Suppression Pool (SP) tank of the integral test facility, called Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility, a scaled downcomer pipe installed in the PUMA-E SP, and air discharge pipe system. Two different diameter sizes of air injection pipe (0.076 and 0.102m), a range of air volumetric flux (7.9–24.7m/s), initial void conditions in an air injection pipe (fully void, partially void, and fully filled with water) and different air velocity ramp rates (1.0, 1.5, and 2.0s) are used to investigate the impact of the blowdown conditions to the void distribution in the SP. Two distinct phases of experiments, namely, an initial and a quasi-steady phase were observed. The maximum void penetration depth was experienced during the initial phase. The quasi-steady phase provided less void penetration depth and was categorized by oscillations in the void penetration. The initial void conditions in an air injection pipe and air volumetric fluxes were only found to significantly impact on the void distribution in the SP.

PANDA: A Multipurpose Integral Test Facility for LWR Safety Investigations

The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the support of a large number of organizations (regulatory bodies, technical dupport organizations, national laboratories, electric utilities, industries) worldwide. The paper provides an overview of the research programs performed in the PANDA facility in relation to BWR containment systems and those planned for PWR containment systems.

Simulation of containment thermal-hydraulics in the Marviken Blowdown 16 experiment with ASTEC and CONTAIN codes

The boiling heavy water reactor Blowdown 16 experiment, which was performed in the Marviken experimental facility, was simulated with the ASTEC and CONTAIN codes. The main purpose of the work was the assessment of the codes for simulating thermal-hydraulic phenomena in a BWR containment at accident conditions. Simulated pressures, atmosphere temperatures and wetwell pool masses are compared to experimental measurements. The results show that both codes satisfactorily reproduced the overall containment thermal-hydraulic behaviour. The simulations also allow a more detailed understanding of the governing mechanisms during the performed experiment.

The impact of assessing lower head integrity on boiling water reactor probabilistic risk assessments

Abstract Over the last several decades, much effort has been directed at estimating the likelihood of a large early release of radioactivity during a nuclear accident. This effort has culminated in the Individual Plant Examinations (IPEs) for the over 100 US nuclear power plants and the NUREG 1150 study. The large early release of radioactivity requires core damage with loss of primary containment integrity during the accident. Given a successful reactor scram, early containment failure coupled with a large release of radioactivity will only occur if the reactor core vessel is breached by core debris. Most IPE/PRA studies performed to date have not considered the possibility of quenching core debris in the lower plenum. Consequently, lower head failure is presumed to closely follow the onset of core damage. Therefore, these assessments did not address the role that in-vessel debris retention plays in preserving primary containment integrity, nor do they propose a criterion for evaluating the integrity of the vessel lower head given that core damage has occurred. Yet preserving the vessel lower head integrity is a necessary condition for satisfying the plant design and licensing basis. Therefore, a more complete treatment of the risk associated with nuclear plant operation includes an evaluation of the ability to retain the core debris in-vessel. This paper presents a performance requirement for vessel integrity to be used in probabilistic risk assessments; evaluates the impact the core damage progression and lower plenum quenching models have on the likelihood of terminating the damage progression in-vessel; documents the significant reduction in BWR containment failure probability that can occur when appropriate core damage and lower head quenching models are used; reviews the implications of core debris quenching in the lower head on BWR PRA modeling; argues why crediting the capability to maintain vessel integrity is necessary from a safety point of view. These results and conclusions are derived from consideration of a BWR 4 plant with a 251 inch vessel. However, the concepts are generally applicable and results specific to other BWR designs can be developed using the methodology presented in this paper.

Finite element dynamic bifurcation buckling analysis of torispherical head of BWR containment vessel subjected to internal pressure

In this paper the bifurcation buckling pressure for the torispherical head of the Mark II type BWR containment vessel subjected to dynamically applied internal pressure is calculated, using a finite element program for a dynamic analysis. Three kinds of dynamic loadings, that is, step loading, ramp loading and pulse loading are considered in the present analysis. The minimum bifurcation buckling pressure is predicted for the respective loadings. The minimum bifurcation buckling pressure for dynamic loading is much lower than the bifurcation buckling pressure for static loading.

Mitigation of severe accident consequences by the containment design of kwu-lwr

The design principles for safety systems according to the multilevel concept of plant safety are described and especially the safety analyses carried out to quantify the inherent safety margins regarding plant behaviour after beyond design basis accidents are addressed. The containment is particularly important in this context. The design bases of the PWR spherical steel containment and the BWR spherical steel containments are explained. To mitigate the consequences of the severe beyond design bases accidents the German PWR and BWR containments are equipped with containment venting systems. The requirements and the design data of the containment venting systems to be installed in German PWRs are summarized followed by the description of flow diagrams and corresponding arrangement of systems operating under sliding or atmospheric pressure modes.

Lifetime integrity requirements for containments in the United States

Assuring the lifetime integrity of containment structures for nuclear power plants is becoming increasingly important as existing design criteria are reexamined, as new requirements for containment inspection and testing are formulated, and as today's operating plants grow older. Regulatory requirements for containments in the United States are contained in the Code of Federal Regulations and in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. New requirements for the preservice examination and inservice inspection of Class MC (metal) containments have been published in Section XI, Subsection IWE, of the ASME Boiler and Pressure Vessel Code. Requirements for Class CC (concrete) containments have been published in Subsection IWL. Federal regulations that will require mandatory compliance with the ASME Code by nuclear plant owners in the United States are forthcoming. Parameters for extending the life of current United States nuclear plants beyond the 40 year design lifetime are presently being formulated. Two pilot plant life extension (PLEX) studies, one for a BWR and one for a PWR, serve as models for extending the life of today's aging plants. This paper presents an overview of the regulatory requirements for containments in the United States and the new preservice examination, inservice inspection and related requirements for Class MC and Class CC containments in the ASME Boiler and Pressure Vessel Code. Operational and life extension considerations for containment structures, including the findings from pilot studies of typical PWR and BWR containments, are also discussed. Together, the regulatory and Code requirements and recommendations from the plant life extension studies provide a basis for improved lifetime integrity for containments in the United States.

Integrity assessment of nuclear containment vessel steels due to missile impact load

The impact behavior of the steel plate of BWR containment vessels against missiles, that are caused by the postulated catastrophic failure of components with a high kinematic energy has been investigated. The research started in 1977 and was completed in 1979 by a project team of the Technical Research Association for Integrity of Structures at Elevated Service Temperatures, sponsored by three Electric Power Companies (Tokyo, Chubu, and Chugoku). Although the probability of the occurrence of missiles inside and outside of containment vessels is extremely low, the following items are required to maintain the integrity of containment vessels. 1. (1) The probability of the occurrence of missiles. 2. (2) The weight and energy of the missiles. 3. (3) The impact behavior of containment vessel steel plate against postulated missiles. In connection with the above item (3), an actual-scale missile test was conducted. In addition, a computation analysis was performed to confirm the impact behavior against the missiles, in order to search for wide applicability to the various kinds of postulated missiles. This research has finally aimed at obtaining a new empirical formula which carries out the assessment of the integrity of containment vessels.

Hydrogen evolution from aluminium in reactor containment spray solutions

Three different aluminium alloys were exposed to conditions similar to BWR and PWR containment spray waters at 50, 100 and 150°C. BWR de-ionized water gives corrosion rates of at most 0.05 mm/year and hydrogen concentrations less than 0.1–1%. By contrast, PWR alkaline solutions give very high corrosion rates and hydrogen contents.

Evaluation of Dynamic Loads Induced by Chugging

A dynamic load evaluation method has been proposed for chugging phenomena which are assumed to occur and produce relatively large amplitude pressure spikes in the pressure suppression pool of a BWR containment, in case of a postulated loss of coolant accident. The proposed method is based on the analysis code developed by the authors and on theseven vent full scale tests performed at Japan Atomic Energy Research Institute (JAERI CRT), considering random nature of chugging phenomena. The dynamic loads are obtained by applying the design source functions of impulsive nature to the vent pipe exists in each BWR containment analysis model. The design source functions are defined to produce dynamic pressures which reasonably envelope the design spectrum based on JAERI CRT data in frequency domain. As an application example, the dynamic loads induced by chugging have been assessed based on the proposed method and on the reported JAERI CRT data from the view point of conservative load evaluation. The applicability of the analysis code has also been confirmed, since the simulated dynamic pressures have shown features and magnitudes similar to those observed in JAERI CRT.

Evaluation of dynamic loads induced by chugging.

A dynamic load evaluation method has been proposed for chugging phenomena which are assumed to occur and produce relatively large amplitude pressure spikes in the pressure suppression pool of a BWR containment, in case of a postulated loss of coolant accident. The proposed method is based on the analysis code developed by the authors and on theseven vent full scale tests performed at Japan Atomic Energy Research Institute (JAERI CRT), considering random nature of chugging phenomena. The dynamic loads are obtained by applying the design source functions of impulsive nature to the vent pipe exists in each BWR containment analysis model. The design source functions are defined to produce dynamic pressures which reasonably envelope the design spectrum based on JAERI CRT data in frequency domain. As an application example, the dynamic loads induced by chugging have been assessed based on the proposed method and on the reported JAERI CRT data from the view point of conservative load evaluation. The applicability of the analysis code has also been confirmed, since the simulated dynamic pressures have shown features and magnitudes similar to those observed in JAERI CRT.

Iodine Removal Tests for BWR Containment Spray by Large Scale Facility

Iodine removal tests for a BWR containment spray were carried out with large-scale JAERI Model Containment Test Facility under LOCA simulated conditions. The tests consisted of two groups: “gas-phase based” tests mainly to obtain the initial iodine removal rate by the spray and “liquid-phase based” tests to obtain the iodine partition coefficient at equilibrium state. It was shown that the degree of iodine removal was largely influenced by pH-value of spray water. The results were discussed with calculated results by a code MIRA-PB using a dose reduction factor for the airborne iodine.

An evaluation of containment inerting and air dilution systems as methods for post-accident hydrogen control in BWRS

Hydrogen control is important in post-accident situations because of possibilities for containment rupture due to hydrogen deflagration or detonation. Post-accident hydrogen generation in BWR containments is analyzed as a function of engineered hydrogen control system, assumed either nitrogen inerting or air dilution. Fault tree analysis was applied to assess the failure probability per demand of each system. These failure rates were then combined with the probability of accidents producing various hydrogen generation rates to calculate the overall system hydrogen control probability. Results indicate that both systems render approximately the same overall hydrogen control failure rate (air dilution: 8.3 × 10 −2−1.1 × 10 −2; nitrogen inerting: 1.3 × 10 −2−2 × 10 −3). Drywell entries and unscheduled shutdowns were also analyzed to determine the impact on the total BWR accident risk as it relates to the decay heat removal system. Results indicate that inerting may increase the overall risk due to a possible increase in the number of unscheduled shutdowns due to a lessened operator ability to correct and identify ‘unidentified’ leakage from the primary coolant system. Further, possible benefits of inerting due to reduced torus corrosion and fire risk in containment appear to be dominated by the possible operations-related disadvantages.