BWR Reactor Research Articles

Interaction of CsOH with structural steel 316L and simulated products of the steel-cladding reaction under assumed conditions relevant for the accident at the Fukushima Dai-Ichi Unit 2 and 3

A high radiation dose rate under the shield plugs of the Units 2 and 3 reactors at the Fukushima Dai-Ichi Nuclear Power Station (1F), raised an important question on volatile radioactive Cs migration, adsorption and retention on various structural materials during the accident. The high dose rate interferes with decommissioning work. The purpose of this study is to present basic knowledge on contamination mechanisms of a BWR reactor structural materials and post-accident debris by Cs compounds to help decommissioning companies consider the ways of decontamination and debris handling during 1F debris retrieval. Having this goal, a series of tests on CsOH contamination mechanism under assumed accident conditions for Units 2 and 3 was performed. The results may improve the accuracy of estimating the situation inside the reactor from the perspectives of accident analysis and materials’ properties. The test temperature and compounds were chosen based on the probable accident scenario. In particular, Cs adsorption was checked using preoxidized 316L stainless steel, a product of control blade degradation (steel-B4C eutectic) and a product of steel-channel box and cladding interaction (Zr-Fe based multiphase material). Penetration ability of Cs into the oxidized and metallic layers of materials was evaluated. CsOH most probable adsorption temperature range and new Cs containing phases’ formation temperature range were followed.

Evaluated Material Attractiveness of Plutonium Composition from Economic Simplified Boiling Water Reactor (ESBWR)

The potential for misuse of nuclear fuel has prompted many to make efforts to prevent a country from acquiring nuclear explosive devices (weapons). Weapon utility equated with nuclear material attractiveness when applied to protection and security. Material attractiveness evaluated using an equation first developed by Masaki Saito, namely ATTR. ESBWR is one of the reactors of the BWR reactor type that is still operating. The purpose of this study is to evaluate Material Attractiveness based on the composition of plutonium in the ESBWR reactor, during reactor operation and after reactor operation. The ATTR concept utilizes the plutonium isotope composition to evaluate the material attractiveness aspect. Parameters such as BCM, Rossi-alpha, and neutron prompt life become additional aspects in evaluating material attractiveness. In the research results, the ATTR value decreased from the beginning of reactor operation until the reactor stopped operating. At the beginning of the reactor operation, it was categorizing as a weapon-grade with an ATTR value of 0.19 and when the reactor stopped operating it was categorizing as an unusable grade with an ATTR value of 0.012. At the time of cooling time up to 1x106 (one million) cooling time, the ATTR value increases by 0.019 which was categorizing as an unusable grade. After a cooling time of 1x106 (one million) years up to 1x107 (ten million) years, the ATTR value was 0.001 which categorized as an exempt level. This states that the increasing burnup value will further reduce the ATTR value. Likewise with the Rossi-Alpha and Bare Critical Mass values. Conversely, it will decrease the value of neutron lifetime. The ATTR value after the reactor operates up to one hundred years of cooling time increases and then decreases significantly after 1x106 (one million) years due to the half-life of plutonium isotopes. Likewise, the Rossi-Alpha and Bare critical mass values. However, it is different with the neutron fast lifetime.

Study on fracture assessment by GTN model for BWR reactor internals based on material properties of simulated irradiated stainless steels

The fitness for service code of the Japan Society of Mechanical Engineers (JSME) provides a fracture evaluation methodology based on linear elastic fracture mechanics (LEFM) for boiling water reactor (BWR) internal components irradiated to a neutron fluence of 3.0 × 1024 n/m2 or more. On the other hand, it is known that austenitic stainless steel maintains its ductility and toughness even after neutron irradiation up to the fluence expected for BWR internal components. Accordingly, it is preferable to apply the ductile fracture evaluation methodology for BWR internal components. The present study examined the applicability of damage mechanics analysis based on the Gurson‒Tvergaard‒Needleman (GTN) model for irradiated stainless steel. The changes of mechanical properties of the material due to neutron irradiation were simulated by cold working. As a result, in terms of material strength properties the applicability of the GTN model was confirmed for irradiated stainless steel using the simulated material for neutron fluence up to about 1.5 × 1025 n/m2.

A methodology for the analysis of In-Core fuel management configurations in BWR’s

A configuration for the integral problem of in-core fuel management consists of a fuel lattice and a fuel assembly designs, as well as a fuel reload and a control rod pattern design, so that the different criteria for the proper operation of a BWR reactor are met. Such a configuration or solution is not unique. There is an universe of potential candidate solutions that solve the problem in different ways. Departing from a database of candidate configurations, we describe and show the application of a multicriteria analysis methodology based on intervals and possibility function. The nuclear engineer must define his/her preferences in terms of the relevance assigned to the configurations’ features and next, using the methodology, those configurations are ranked. Besides this, the methodology can help us to identify the role that play some fuel lattice variables into the in-core fuel problem performance.

Recycling alternatives for minor actinide and plutonium in a boiling water reactor

Recycling is an alternative for reducing minor actinide and plutonium stockpiles from depleted fuel assemblies. In the current study, it is proposed the use of a MOX fuel assembly called MOXIM-MA. It comprises 4 rods with minor actinides, 14 UO2-gadolinium rods, 8 water rods, and 66 MOX fuel rods. This MOXIM-MA design is used in a 2027 MWth BWR, reducing 0.130 kg HM of minor actinide and 5.042 kg HM of Plutonium per fuel assembly at discharge after discharge 4-operating cycles of 18 months each. The reactor core energy produced does not suffer any penalty by using this fuel assembly. There is no need for additional changes to the original control safety system. The radiotoxicity results show almost one magnitude order increase compared with standard UO2 fuel assemblies for the same BWR reactor core. Still, this increase is the same magnitude experienced when MOX fuel assemblies without minor actinide are used. A twin reactor power plant, 2027 MWth per unit, with 20 and 25 years of operational life, is considered to propose a plutonium and minor actinide recycling strategy. Standard UO2 fuel assemblies, at discharge for this plant, produce 0.135 kg HM of MA and 1.707 kg HM of plutonium. By recycling them, at least 6 depleted fuel assemblies are needed for each MOXIM-MA fuel assembly processed. The suggested strategy starts by considering the cumulative amount of minor actinide and plutonium coming from the depleted fuel assemblies discharged during the plant's operating life at the moment of this analysis. The study shows that the MOXIM-MA fuel assemblies will be used in only one of the reactor units in a once-through scheme. It describes how the auto recycling of plutonium and MA could be done to supply the MOXIM-MA fuel assemblies during the rest of the plant's operation. Results promise to deliver the plant's whole operation life a decrease of 39.50% in plutonium inventories and 12.70% in MA inventories, achieving this goal without penalizing energy production.

Trip turbine transient and stability analysis of two different fuel assembly designs aimed for the recycling of minor actinides and plutonium in BWRs

Recycling in thermal reactors is an alternative that could help to reduce minor actinide inventories produced by depleted fuel assemblies of nuclear reactors; the current paper assesses two proposed BWR fuel assemblies to recycle MA. These fuel designs aim to no introduce any perturbation in the electricity production from the Boiling Water Reactor, and they must satisfy the reactor core safety constraints. Such constraints must be considered under steady and transient conditions without the need to make any modification to the nuclear reactor safety systems. Previously, these two fuel designs were assessed under steady state conditions. In this study, the two proposed fuel assembly designs are tested for stability analysis and trip turbine transitory conditions. These fuel assembly designs are based on a 10 × 10 standard fuel assembly, and they have four MA-bearing rods that are replacing four fuel rod assemblies in three of the six axial fuel cells that compose the fuel assembly. The first design is a UO2 fuel assembly; the second one is a MOX fuel assembly where also, 8 fuel pin rods have been replaced by water rods in three of the six axial fuel cells that compose the fuel assembly. These two proposed fuel assemblies have been designed to be equivalent by producing the same amount of energy of the BWR reactor. The stability analysis is performed considering a global reactor oscillation; results show that in the MOX fuel assembly, the control rod pattern must be modified to provide an acceptable shutdown margin and to have adequate stability conditions. Both fuel designs provide acceptable conditions under the trip turbine transient condition and provide two fuel assembly designs that could be used for MA recycling in a BWR without any perturbation in energy production.

Insights from the analysis of flooding-related international operating experience at NPPs

The structured review of operating experience provides useful insights for improved operation and regulation. This paper presents insights and the related statistical analysis from the topical study of flooding-related international operating experience during the last 20 years. The study is the result of the European Clearinghouse activity and covers four databases: French, German, US Licensee Events Reports and the IAEA/NEA International Reporting System (IRS).The methodology applied includes both a qualitative and a quantitative assessment: screening, review and statistical analysis. All relevant events (263) are characterised, and a selection of these (109) are further analysed to derive insights. The paper presents 16 generic insights synthesised from 94 specific insights. The insights are grouped based on the systems causing the events and on the purpose for corrective actions.The statistical analysis shows the contribution, the trend and the correlation between the events databases and the insights. The US and IRS events contribute the most to insights, with important contributions from two other databases. A slightly increasing linear trend of the yearly number of events is missing only for the German events. The correlation analysis for the number of events contributing to the insights shows a noticeable correlation between the US and IRS databases’ and the BWR and PWR reactor type events.The analysis of the relative contribution of the events to the insights shows a statistically significantly smaller or larger contribution to several insights from different databases and reactor types. This could be the basis for a decision about the relevance and the applicability of some insights in certain countries and to specific reactor types.The fact that the most important events contributing to the insights are well distributed over the entire 20 years period shows their continuous relevance. Equally, the fact that all four databases and reactor types contribute sufficiently to the most relevant events shows that the insights are similarly relevant for all of them. The insights from this analysis seem internationally relevant recommendations and they could help the licensees and regulatory authorities to prevent flooding events from occurring, to protect safety systems from flood and to improve flooding protection and mitigation.

Development of the reactor core isolation cooling system model for the extended station black-out accident

To accurately predict the behavior of the Reactor Core Isolation Cooling System (RCIC) in the context of the FLEX strategies under the extended station black-out accident (SBO), we present a mechanistic model of the RCIC in this work. The control volume formulation of angular momentum conservation is employed to model the Terry turbine in the RCIC system. The two-phase nozzle flow model developed in the previous work is utilized to simulate the convergent-divergent nozzles in the Terry turbine. A nozzle flow table is generated to cover the operation range of the RCIC during beyond design basis accidents (BDBAs). The mechanistic RCIC turbine-pump model and the nozzle flow table are incorporated in a TRACE BWR model. Simulations of short-term and long-term SBO transients are performed to demonstrate dynamic RCIC responses. A system analysis code with the proposed mechanistic model of the RCIC system will have a better prediction of system safety response during SBO accidents. With the addition of such more realistic models, the system code will be capable of providing expanded understanding of safety margin, which can help help to form the technical basis for the successful implementation of the FLEX and SAMG measures under Extended Loss of AC Power (ELAP) conditions for BWR or PWR reactor designs.

Study on Fracture Assessment for BWR Reactor Internals Based on Material Properties of Irradiated Stainless Steels

Study on Fracture Assessment for BWR Reactor Internals Based on Material Properties of Irradiated Stainless Steels

Scaling Design of the Pressure Response Experimental Facility for Pressure Suppression Containment

Pressure suppression containment has been adopted extensively in the BWR reactor design due to the remarkable pressure suppression capacity. In order to study the pressure response characteristics of the containment in the initial stage of the loss of coolant accident, it is necessary to build the scaling facility to simulate the prototype. The analysis of the blow down source term and the steam condensation occurred in the suppression pool are conducted respectively for the prototype PWR. The suppression pipe system resistance term is also scaled with a certain ratio. The length and the diameter of the mian pipe and the branch pipe as well as the orifice diameter are calculated. At last, the prototype and the experiment facility are modeled by MELCOR to scale the source term. The results show that the scaling method in this paper is effective to predict the pressure evolution in the prototype containment during the loss of coolant accident.

Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome BWR Reactor KERENA: Part II

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modeling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I dealt with numerical and experimental methods for modeling of condensation inside the emergency condenser and on the containment cooling condenser. This second part deals with boiling and two-phase flow instabilities.

Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome KERENA BWR Reactor: Part I

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modelling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I deals with numerical and experimental methods for modelling of condensation inside the emergency condensers and on the containment cooling condenser while part II deals with boiling and two-phase flow instabilities.

Effects of post-irradiation annealing on the IASCC susceptibility of neutron-irradiated 304L stainless steel

Post-irradiation annealing (PIA) was conducted on a 304L stainless steel irradiated to 5.9 dpa in the Barsebäck-1 BWR reactor, to investigate its effect on the mitigation of irradiation-assisted stress corrosion cracking (IASCC) susceptibility. IASCC susceptibility was measured for the as-irradiated and four PIA conditions (500 °C: 1 h and 550 °C: 1, 5, and 20 h) via interrupted constant extension rate tensile and four-point bend experiments under simulated BWR-NWC conditions. The annealing treatments were observed to progressively reduce IASCC susceptibility, as measured by the final intergranular fracture fraction (tensile) and crack length per unit area (four-point bend), with full removal of IASCC susceptibility being observed following annealing at 550 °C: 1 h for tensile specimens and 500 °C: 1h for four-point bend specimens. Among the microstructure and mechanical property parameters measured as a function of PIA, the average dislocation channel spacing was observed to decrease by ∼25% and ∼40% from the as-irradiated condition after annealing at 500 °C: 1 h and 550 °C: 1 h, respectively. The mitigation of IASCC susceptibility correlated well with the decrease in the average dislocation channel spacing and is consistent with a process in which crack initiation is controlled in part by the high tensile stress at dislocation channel-grain boundary intersections.

Insights into the sources of irradiation hardening in a neutron irradiated 304L stainless steel following post-irradiation annealing

The 304L SS was irradiated in Barsebäck 1 BWR reactor to 5.9 dpa followed by post-irradiation annealing at 500 °C and 550 °C up to 20 h. Correlations of irradiation hardening with the main hardening features of dislocation loops and Ni/Si-rich clusters were made using the least squares regression. The dispersed-barrier hardening model tended to overestimate the contribution of Ni/Si-rich clusters in the as-irradiated condition. Contribution of Ni/Si-rich clusters to irradiation hardening appeared to be more appropriately accounted for by the short-range order strengthening mechanism and was dependent on the Ni and Si concentrations in the clusters. Irradiation hardening in the as-irradiated condition was predominantly from dislocation loops with minor contribution from the Ni/Si-rich clusters. This is consistent with recent molecular dynamics simulations which showed that Ni/Si-rich clusters were weak obstacles to dislocation glide. As dislocation loops were annihilated at a faster pace than the Ni/Si-rich solute clusters, the Ni/Si-rich clusters played a more important role in the remaining hardening after thermal annealing. Contribution of Ni/Si-rich clusters to remaining hardening outweighed that of dislocation loops after thermal annealing at 550 °C for 5 h and became predominant after annealing for 20 h.

Evaluation of mechanical properties distribution for irradiated stainless steels simulated by utilizing the gradient of the cold working ratio

It is known that irradiation damage can cause a reduction in fracture toughness and an increase in the strength of austenitic stainless steels used for the internal structures of nuclear power plant reactors. A change in the mechanical properties of material, such as fracture toughness and strength, affects the fracture behavior of the components. Actual reactor internal structures, such as the top guides and core shrouds in BWR plants, are irradiated from the core in a single direction. This causes the attenuation of neutron flux, which results in the distribution of neutron fluence in the components, meaning that the distribution of mechanical properties is in the through-wall direction of the components. It is therefore possible for the fracture mode of the components to change in the through-wall direction. It would be valuable for more accurate assessment of the structural integrity of the nuclear components to make the effect of the mechanical properties distribution on the fracture mode clear. However, it is difficult to prepare large-scale test specimens with the mechanical properties distribution induced by uni-directional neutron irradiation. In this study, the mechanical properties distribution was simulated by applying the gradient of the cold working ratio using Type 316L stainless steel plate with an uneven thickness. The targeted range of the cold working ratio was designed by assuming BWR reactor internal structures, such as the top guides. The mechanical properties distribution was evaluated based on the results of hardness, tensile and fracture toughness tests. It was found that the macroscopic mechanical properties of irradiated stainless steels can be simulated by using the cold working and the macroscopic distribution of the mechanical properties of irradiated stainless steels can also be simulated by using the cold working ratio with a gradient. Furthermore, fracture tests were performed using test plates with the mechanical properties distribution assuming irradiated components with a crack. In some cases, the maximum loads obtained by the tests were almost the same value in spite of the different cold working ratio in the ligament. Based on the results of the fracture tests, it was considered that the dominant mechanical properties for the fracture load were those at the tip of the slit, not those in the ligament.

Turbopump scaling analysis and similarity level estimation for Texas A&M university RCIC system experimental test facility

The Reactor Core Isolation Cooling (RCIC) System is a safety system that provides water to the reactor pressure vessel during off-normal BWR reactor conditions, such as reactor isolation from the turbines or loss of AC power. Under loss of AC power conditions, the RCIC System is expected to fail due to battery depletion within 4–8 h of operation for many units. However, the system did not fail until about 70 h into the accident at Fukushima Dai-ichi Unit 2, which was well past the time of battery depletion. To investigate the full potential of the RCIC System, the Laboratory for Nuclear Heat Transfer Systems (NHTS) at Texas A&M University is designing and constructing an experimental RCIC System facility. A careful scaling analysis is essential to ensure proper representation of the RCIC System's key components and phenomena in the experimental testing.This paper describes a method to estimate the similarity level (SL) of the RCIC turbine-pump assembly (turbopump), along with application of the methodology using the Peach Bottom Unit 2 RCIC System as the reference full-scale system. The methodology is demonstrated with the Texas A&M University facility but can be applied to other RCIC system facilities.Zuber's H2TS (Hierarchal Two-Tiered Scaling) method has been adopted to specify the required level of detail for the scaling analysis. The analysis includes development of unitless characteristic time ratios to quantify the degree of similarity of the RCIC turbomachinery between the test facility and the prototype. The RCIC turbine and pump are focused on herein because, first, the Fukushima Dai-ichi Unit 2 RCIC System may have ceased operation at 70 h due to turbine failure, and, two, the greatest RCIC System modeling challenges are with the long-term operation of these turbomachinery components. Since the Texas A&M experiments will provide validation data for analytical models of the turbomachinery, applicability of the data must be evaluated.Based on the similarity level estimates, there is a high degree of similarity between the NHTS and Peach Bottom RCIC System turbines, with a minimum similarity level of 75%. This allows the use of the NHTS RCIC turbine for testing over the full range of Design Basis Accident conditions. Another important finding from this study is that the pump configuration initially designed into the NHTS facility had a low similarity level between the experimental and full-scale facilities. This paper identifies pump parameters in the NHTS facility that can be changed to achieve greater similarity. The similarity level analysis indicate that the NHTS RCIC turbine control volume is appropriate for representing the full-size turbomachinery and can be used to study the prototype turbine and pump behaviors under design basis operating conditions.Scaling analyses of the other RCIC System components are to follow the turbomachinery similarity evaluations. The NHTS facility is currently being prepared to test over the entire Design Basis Accident containment pressure and temperature ranges. Future efforts will also focus on modeling, experimentally and with Computational Fluid Dynamics tools, the RCIC System to provide data for beyond design basis operational conditions.

Effect of spacer with mixing vane on liquid film for two-phase annular flow in a simulated BWR reactor fuel rod bundle channel

Effect of spacer with mixing vane on liquid film for two-phase annular flow in a simulated BWR reactor fuel rod bundle channel

Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

The effect of zinc injection on the increasing of Inconel 600 TT corrosion resistances

Many failures were found in reactor pressure vessel head penetration (RPV) head material. Those failures caused by boric acid corrosion, and from visual examination were found a big hole and white deposit crystal of boric acid during shutdown maintenance at David Besse reactor. Zinc Oxide addition in BWR reactor known as Zinc Injection that has purposed to reduce radiation exposure cause of Hydrogen addition. Beside reducing the radiation exposure, Zinc injection also has an effect in reducing material corrosion. The purpose of study is to determine the effect of zinc addition, boric acid, temperature also the effects of Cobalt Nitrate and Zinc Oxide addition to Inconel 600 TT as RPV head penetration material. The result in the BWR reactor experience will be implementated at PWR reactor, weather zinc oxide addition also has an effect in reducing the corrosion of Inconel 600. The method that used in this research is to observe the corrosion rates for Inconel 600 material using Potentiostat. Examination were conducted in 30, 40, 60, 70, 80 and 80 °C using 1000, 1500, 2000, 2500 and 3000 ppm boric acid concentration. The results showed that the corrosion rate for the material were very small, but the highest corrosion rate occurred in 3000 ppm boric acid concentration at 90 °C with Cobalt Nitrate addition, around 5.210 x 10−1 mpy. In the same condition at 3000 ppm boric acid concentration for temperature at 90 °C, Inconel 600 TT corrosion rate is smaller with Zinc oxide addition, around 4.631 x 10−1 mpy.

Effect of post-irradiation annealing on the irradiated microstructure of neutron-irradiated 304L stainless steel

Post-irradiation annealing was performed on a 304L SS that was irradiated to 5.9 dpa in the Barsebäck 1 BWR reactor. Evolution of dislocation loops, radiation-induced solute clusters and radiation-induced segregation at the grain boundary was investigated following thermal annealing at 500 °C and 550 °C up to 20 h. Dislocation loops, Ni-Si and Al-Cu clusters, and enrichment of Ni, Si and depletion of Cr at the grain boundary were observed in the as-irradiated condition. Dislocation loop size did not change significantly after annealing at 550 °C for 5 h but the loop number density decreased considerably and loops mostly disappeared after annealing at 550 °C for 20 h. The average size of Ni-Si and Al-Cu clusters increased while the number density decreased with annealing. The increase in cluster size was due to diffusion of solutes rather than cluster coarsening. Significant volume fractions of Ni-Si and Al-Cu clusters still remained after annealing at 550 °C for 20 h. Substantial recovery of Cr and Ni at the grain boundary was observed after annealing at 550 °C for 5 h but neither Cr nor Ni was fully recovered after 20 h. Annihilation of dislocation loops, driven by the thermal vacancy concentration gradient caused by the strain field and stacking fault associated with the loops appeared to be faster than annihilation of solute clusters and recovery of Ni and Si at the grain boundary, both of which are driven by the solute concentration gradients.