Conceptual design study for the demonstration reactor of JSFR, 5; Reactor cooling system design

Kisohara, Naoyuki; Ishikawa, Hiroyasu; Futagami, Satoshi; Xu, Y.*; Shimoji, Kuniyuki*; Kawamura, Masaya*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/10

The cooling system of the JSFR adopts an integrated primary sodium pump/intermediate heat exchanger (IHX), dual structure straight tube steam generator (SG) and short elbow sodium piping layout. Since, however, this is the first experience applying these technologies to SFRs in Japan, design approaches have been evaluated and R&D has been undertaken. This paper addresses the design study of the cooling system of the demonstration reactor JSFR in terms of thermal-hydraulic and structural integrity. Recent studies have shown that these new technologies have potential to be applied to the JSFR.

A Study on time frame definition and reference evolution of the geological system for safety assessment; Case study on the Horonobe URL site

Kurikami, Hiroshi; Niizato, Tadafumi; Yasue, Kenichi

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10

Description of system evolution is an important task in safety and investigation strategies. This paper showed the method to describe system evolution based on safety functions and FEPs and its application to the Horonobe site. Based on the SDM and the Horonobe-specific FEPs, the important FEPs were put in a timeline with the main safety functions. According to the FEPs, we defined thermal and resaturation phase, steady geology phase and geologically evolution phase. In the steady geology phase, the functions of retardation and dilution in the deeper part of the Wakkanai Formation are important, therefore, advection, dispersion and sorption in the domain should be assessed based on the nuclide migration scenario. In the geologically evolution phase, the uplift and denudation are important. Thus, the uplift, denudation and the consequent THMC processes were involved in the reference evolution. Through the application and the discussion, the method was found to be applied to other sites.

Tightly coupled multiphysics simulations for prismatic reactors

Sato, Hiroyuki; Park, H.*; Knoll, D.*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10

Multiphysics core simulations for a prismatic-type VHTR are performed in this study. Our solution scheme is based on the JFNK method. As a preliminary example, a thermal-fluid calculation is performed with an idealized two-dimensional symmetric representation of the GT-MHR and compared with the RELAP5-3D simulation results. Also, a neutronics calculation is conducted using the same geometry as the thermal-fluid calculation, and using cross section data obtained from an HTGR benchmark problem. In addition, a coupled steady-state thermal-fluids neutronics calculation is performed. The calculation results showed that the developed prismatic VHTR core simulator can perform tightly-coupled multiphysics simulations efficiently.

Preliminary evaluation of JSFR achievement level to risk targets

Kurisaka, Kenichi; Okamura, Shigeki*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10

Japan Atomic Energy Agency (JAEA) has been developing the Japan Sodium-cooled Fast Reactor (JSFR) in the Fast Reactor Cycle Technology Development (FaCT) Project. Risk targets were set out as part of the safety-related design requirement: i.e., the quantitative safety design requirements on the core damage frequency (CDF) and the containment failure frequency (CFF). This paper describes a preliminary evaluation of achievement level of JSFR to the risk targets at the FaCT project phase-I: JFY2006 to JFY2010. A Level-1 PSA has been implemented preliminarily to evaluate the CDF related to internal initiators in power operation. The calculated CDF became lower than the both requirements on CDF and CFF. For seismic events, the seismic fragility of principal structures and components was evaluated in terms of core damage prevention. This evaluation was based on the seismic response analysis, which considered the seismic isolation effect and the hardening effect of the laminated rubber bearing in the isolation devices. As a result, we confirmed that the principal structures and components of JSFR have sufficient seismic margin. Based on this, we judged the risk target could be achieved against the seismic event.

Development of methodology for the characterisation of the long-term geosphere evolution, 2; Estimation of the long-term evolution of groundwater flow conditions in a Tono area case study

Kosaka, Hiroshi; Saegusa, Hiromitsu; Yasue, Kenichi; Kusano, Tomohiro; Onoe, Hironori

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/10

The methodology for estimation of the long-term evolution of groundwater flow conditions are being developed using approaches on the basis of deductive and inductive methods in the case of Tono area. Based on the studies using the approach on the basis of deductive method, it has been confirmed that the method combining physical modeling of topographic change and groundwater flow simulations is useful for estimating of changes in groundwater flow conditions in the future due to topographic and climatic perturbations. Existing information for estimation of surface hydrological conditions, which are to be used for assignment of boundary conditions for the groundwater flow simulation, has been gathered from many sources and reviewed based on modern-analogue methods. In the studies using the approach on the basis of inductive method, paleo-hydrogeological studies have been carried out on several spatial and time scales. Through the study on the largest spatial scale, a methodology needed to understand changes of groundwater flow conditions due to long-term topographic change is proposed to efficiently identify the area to be carried out site characterization involving field investigations. And then, information to estimate the paleo-topography and paleo-climate has been obtained from literature surveys and field investigations. Through these studies, it has been confirmed that these two approaches are useful for estimation of the long-term evolution of deep groundwater flow conditions.

Flowsheet study of HI separation process from HI-HO-I solution in the thermochemical hydrogen production iodine-sulfur (IS) process

Kasahara, Seiji; Guo, H.*; Tanaka, Nobuyuki; Imai, Yoshiyuki; Iwatsuki, Jin; Kubo, Shinji; Onuki, Kaoru

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/10

Flowsheet investigation of the subsection of HI separation from HI-HO-I solution in the thermochemical hydrogen production iodine-sulfur (IS) process was performed. Concentration of HI by electro-electrodialysis (EED) and distillation of HI were applied. Experimental data of the EED cell applying Nafion membrane was used to establish heat/mass balance equations for the cell. Heat/mass balance of HI distillation column was calculated using ESP, a process simulation software. HI molality at the cathode outlet of the cell, pressure in the HI distillation column, and flow rate ratio of the feed to the subsection to distillate of the column were focused as variable parameters for minimum heat demand. Parameters of the EED membrane, electric resistance and upper limit of HI molality between outlet streams, had a great effect on the heat demand; improvement of the membrane parameters is important to reduce the heat demand.

Study on flow induced vibration evaluation for a large scale JSFR piping, 4; Unsteady flow characteristics in 1/10 scale hot-leg piping experiments under undeveloped and swirl inflow conditions

Iwamoto, Yukiharu*; Kondo, Manabu*; Yasuda, Kazunori*; Sogo, Motosuke*; Tanaka, Masaaki; Yamano, Hidemasa

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10

Study on flow induced vibration evaluation for a large scale JSFR piping, 2; Vibration analysis in 1/3 scale hot-leg piping experiments under swirl inflow conditions

Baba, Takeo*; Hirota, Kazuo*; Sago, Hiromi*; Yamano, Hidemasa; Aizawa, Kosuke; Xu, Y.*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/10

A Study of applicability of JENDL-4.0 to the HTTR criticality analysis

Goto, Minoru; Shimakawa, Satoshi; Tachibana, Yukio

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 6 Pages, 2011/10

In the past, benchmark calculations of criticality approach for the HTTR, which is a Japanese HTGR, were performed by research institutes in several countries, and almost all of the calculations overestimated the excess reactivity. In Japan, the benchmark calculations were performed by JAEA, and were also resulted in overestimation. JAEA improved the calculations by revising the geometric model and replacing the nuclear data library with JENDL-3.3, which was the latest JENDL at that time. However, the overestimation remained and this problem has not been resolved until today. We performed calculations of the HTTR criticality approach with several nuclear data libraries, and found that slight difference in the neutron capture cross section of carbon at thermal energy among the libraries causes significant difference in the values. The cross section value of carbon was not concerned in reactor neutronics calculation because of its small value of the order of 1E-3 burn, and consequently the cross section value had not been revised for a long time even in the major nuclear data libraries: JENDL, ENDF/B and JEFF. We thought that the cross section should be revised based on the latest measurement data to improve the accuracy of the HTGR criticality analysis. In May 2010, the latest JENDL (JENDL-4.0) was released by JAEA, and the capture cross section of carbon was revised. JENDL-4.0 yielded 0.4-0.9%k/k smaller values than JENDL-3.3 in the criticality calculations for the HTTR critical approach, and consequently the problem of the overestimation of the excess reactivity in the HTTR benchmark calculation was resolved by replacing the nuclear data libraries with JENDL-4.0.

Development of PIRT and assessment matrix for V&V of sodium fire analysis codes

Ohno, Shuji; Ohshima, Hiroyuki; Tajima, Yuji*; Oki, Hiroshi*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10

Conceptual design study of the JSFR needs the evaluation of thermodynamic consequences in hypothetical sodium fire accident. The authors are therefore initiating systematic V&V activity for sodium fire evaluation tools. The activity is in progress with the main focuses on already developed sodium fire analysis codes SPHINCS and AQUA-SF. The present paper describes that a preliminary "PIRT" is developed at first, and an "assessment matrix" is proposed which summarizes both separate effect tests (SET) and integral effect tests (IET) for validating the computational models or whole code for important phenomena. Furthermore, the paper shows an individual validation with SET in which the spray droplet combustion model of SPHINCS and AQUA-SF predicts the burned amount of a falling sodium droplet with the error mostly less than 30%.