Tomohiko Ikegawa

Development of Inherently Safe Technologies for Large Scale BWRs: (4) Hydrogen Explosion Prevention System Using SiC Fuel Claddings

In the aftermath of the lessons learned from the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), including a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents.One of inherently safe technologies currently under development is a system to prevent hydrogen explosion during severe accidents (SAs). This hydrogen explosion prevention system consists of a high-temperature resistant fuel cladding of silicon carbide (SiC), and a passive autocatalytic recombiner (PAR). Replacing the zircaloy (Zry) claddings currently used in LWRs with the SiC claddings decreases the hydrogen generation and thus decreases the risk of hydrogen leakage from a primary containment vessel (PCV) to a reactor building (R/B) such as an operation floor. The PAR recombines the leaked hydrogen gas so as to maintain the hydrogen concentration at less than the explosion limit of 4 % in the R/B.The advantages of using SiC claddings in the system were examined through experiments and SA analysis. Results of steam oxidation tests confirmed that SiC was estimated to show 2 to 3 orders of magnitude lower hydrogen generation rates during oxidation in a high temperature steam environment than Zry. Results of SA analysis showed that the total amount of hydrogen generation from fuels was reduced to one fifth or less. Calculation also showed that the lower heat of the oxidation reaction of SiC moderated the steep generation with the temperature increase. We expected this moderated steep generation to reduce the pressure increase in the PCV as well as prevent excess amounts of leaked hydrogen from hydrogen disposal rate capacity using PARs.The SiC cladding under consideration consists of an inner metallic layer, a SiC/SiC composite substrate, and an outer environment barrier coating (EBC). A thin inner metallic layer in combination with a SiC/SiC composite substrate functions as a barrier for fission products. EBC is introduced to have both corrosion resistance in high temperature water environments during normal operation and oxidation resistance in high temperature steam environments during SA. Further reduction of the hydrogen generation rate in high temperature steam by improving the EBC is expected to decrease the total amount of hydrogen generation even more.Copyright

The Plant Feature and Performance of Double MS (Modular Simplified and Medium Small Reactor)

A new concept of a small and medium sized light water reactor, named the double MS: modular simplified and medium small reactor (DMS) was developed. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts, and effective use of proven technology. The decrease in the primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about 2 m compared with 3.7 m in the conventional boiling water reactor (BWR). The short active fuel length reduces the drop in core pressure and overcomes the natural circulation system. By using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to confirm transient performance, the DMS plant performance under transient conditions was evaluated using the TRACG code. TRACG code, which can treat multidimensional hydrodynamic calculations in a RPV, is well suited for evaluating the DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As a result, the maximum change in the minimum critical power ratio of the DMS was 0.14, almost the same as for the current advanced boiling water reactor (ABWRs). In order to improve safety efficiency developing an emergency core cooling system (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following: (1) eliminating highpressure injection systems, (2) adopting passive safety-related systems, and (3) optimizing distribution for the systems and power source for the ECCS. In this way, the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of the loss of coolant accident analysis, we confirmed that the core can be covered by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.

Analysis of ABWR Critical Control and Heat-Up Control Operation by TRACG Code

Most startup and shutdown operations in advanced boiling water reactors (ABWRs) are automated by an automatic power regulator (APR). Hitachi and Hitachi-GE utilized the three-dimensional transient analysis code TRACG to design and verify the APR control algorithms. To verify the algorithms, an external neutron source model that makes it possible to simulate a sub-critical initial core, a water temperature reactivity model, a startup range neutron monitor (SRNM) model, and the APR system models were developed and coded onto the TRACG code. The improved TRACG code has been tested and verified with ABWR startup test data. In the test, the criticality was achieved 40 min after beginning of control rod (CR) withdrawal. The code results, for example, CR operation timing, CR withdrawal length, and signals of the neutron sensors agreed well with the test data. In the heat-up control mode, the measured increasing rate of the reactor water temperature was well simulated for a period longer than six hours.Copyright

Analysis of ABWR Critical Control Operation by TRACG Code

Hitachi utilized three-dimensional transient analysis to design and verify a critical-control mode algorithm of an automatic power regulator (APR). TRACG has a three-dimensional neutron kinetics model based on diffusion theory and a six-equation two-phase flow model. To verify the APR critical-control mode algorithm, an external-neutron-source model that makes possible to simulate a sub-critical initial core, and an APR system model were developed and added on TRACG. The code was verified by comparison of measurements and calculation results of ABWR start-up operation under the critical-control mode. The modified TRACG could simulate neutron count rates of start-up-range neutron monitors (SRNMs), reactor period, control rod operation timing, CR withdrawal length, and time of criticality declaration, well.Copyright

Transient Performance of Medium Small LWR “DMS-400” Evaluated Using TRACG Code

A new concept of a small and medium sized light water reactor, named the DMS (double MS: modular simplified & medium small reactor) has been developed. The DMS features significantly simplified plant systems realized by adoption of a natural circulation system of coolant and a free surface separation (FSS) system that is based on the gravitational separation of steam and water. With these systems, reactor internal pumps and steam separators are not needed, reducing plant cost. In this study, the DMS plant performance under transient conditions has been evaluated using TRACG code. TRACG code, which can treat multi-dimensional hydrodynamic calculations in a reactor pressure vessel (RPV), is well suited for evaluating DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As critical transient cases, generator load rejection with total turbine bypass failure (LRNBP) and loss of feedwater heating (LFWH) were chosen to evaluate. LRNBP and LFWH are the most severely recognized events as a pressure increase event and a thermal power increase event, respectively. In case of LRNBP, heat flux increased to about 110% of rated power, and the natural circulation flow barely changed, resulting in a lower ΔMCPR than that of LFWH case. The reason that heat flux only increased to 110% was because the RPV of the DMS has a large steam region volume in the chimney region compared to the thermal power. As a result, the change in the void fraction with a pressure increase in the core was small. In case of LFWH, the maximum heat flux, calculated using the neutron flux, was 121% of rated power when a scram occurred, and ΔMCPR was 0.14, almost the same as for current ABWRs. Since the analysis conditions were set conservatively, these results show that the DMS performs as well for transient events as conventional BWRs.© 2008 ASME