Improvement of ex-vessel molten core behavior models for the JASMINE code

Matsumoto, Toshinori; Kawabe, Ryuhei; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 9 Pages, 2016/11

During severe accident at nuclear power stations, molten core material jet could be discharged from the reactor pressure vessel into the water pool formed at the pedestal or cavity in the containment vessel. To improve the JASMINE code, The method for determining particle diameters which follow the Rosin-Rammler distribution was implemented. The jet breakup experiments, DEFOR-A conducted by KTH (Royal Institute of Technology, Sweden) were analyzed with the code. The influence of the experimental conditions, such as water subcooling, melt jet diameter and superheat were discussed. A crust layer formation model was also implemented in the code. The analyses using the model were carried out for the melt spreading experiments, PULiMS conducted by KTH. The spreading area was overestimated. Further improvement of the melt spreading model were discussed to close the gaps by introducing additional models such as heat conduction in the substrate materials, void formed inside the melt and so on.

Development of air cooling performance evaluation method for fuel debris on retrieval of Fukushima Daiichi NPS by dry method, 2; Outline of numerical method and preliminary analysis of free convection around fuel debris

Yamashita, Susumu; Uesawa, Shinichiro; Yoshida, Hiroyuki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

In fuel debris retrieval in decommissioning of the Fukushima Daiichi NPS, dry method is under consideration. Investigation of the cooling performance of fuel debris in the dry method will be very important problem to realize the method. However, there are uncertainties in the shape and surface temperature of fuel debris. In order to evaluate the cooling performance, the investigation of the cooling performance by free convection is required. We have been developing the numerical simulation method, which can evaluate the cooling performance of the fuel debris by free convection, using the JUPITER code in JAEA. In this paper, we show the evaluation result of the thermal conductivity by the free convection from fuel debris to the atmosphere in the simplified system.

Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor; Influence of flow collector of backup CR guide tube

Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11

JAEA has been conducting a design study for an advanced large-scale sodium-cooled fast reactor (SFR). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of UIS. JAEA had performed a water experiment to examine countermeasures for the significant temperature fluctuation generated at the bottom of SFRs UIS. Meanwhile, a self-actuated shutdown system (SASS) is equipped in a backup control rod (BCR) channel to ensure reactor shutdown. The BCR guide tubes have a flow guide structure "flow-collector" to provide reliable operation of SASS. Flow-collector may affect the thermal mixing behavior at the bottom of the UIS. This study has investigated the influence of the flow- collector on characteristics of the temperature fluctuation around the BCR channels.

Development of extreme rainfall PRA methodology for sodium-cooled fast reactor

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 10 Pages, 2016/11

Development of air cooling performance evaluation method for fuel debris on retrieval of Fukushima Daiichi NPS by dry method, 3; Heat transfer and flow visualization experiment of free convection adjacent to upward facing horizontal surface

Uesawa, Shinichiro; Shibata, Mitsuhiko; Yamashita, Susumu; Yoshida, Hiroyuki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

Improvements to the simmer code model for steel wall failure based on EAGLE-1 test results

Toyooka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

A Recent experimental program to evidence in-vessel retention by controlled material relocation during core disruptive accidents of sodium-cooled fast reactors

Matsuba, Kenichi; Kamiyama, Kenji; Toyooka, Junichi; Zuev, V. A.*; Ganovichev, D. A.*; Kolodeshnikov, A. A.*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11

Molten fuel discharge through control rod guide tubes (CRGTs) is a key process that dominates the termination of core disruptive accidents of sodium-cooled fast reactors, since fuel dispersion from the core contributes to the achievement of both deeper subcriticality in the degraded core and formation of coolable debris bed. Within a framework of a collaborative research program between Japan Atomic Energy Agency and National Nuclear Center of the Republic of Kazakhstan, called EAGLE program, a new experimental program has been started with out-of-pile experiments to clarify the fuel discharge through CRGTs. This paper presents the status of the new program, including experimental results obtained so far.

Event sequence assessment using plant dynamics analysis based on continuous Markov chain process with Monte Carlo sampling assessment of strong wind hazard in sodium cooled fast reactor

Takata, Takashi; Azuma, Emiko*; Nishino, Hiroyuki; Yamano, Hidemasa; Sakai, Takaaki*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 6 Pages, 2016/11

A new approach has been developed to assess event sequences under external hazard condition considering a plant status quantitatively and stochastically so as to take various scenarios into account automatically by applying a Continuous Markov Chain Monte Carlo (CMMC) method coupled with a plant dynamics analysis. In the paper, a strong wind is selected as the external hazard to assess the plant safety in a loop type sodium cooled fast reactor. As a result, it is demonstrated that the plant state is quite safe in case of the strong wind because multiple failures of the air coolers in the auxiliary cooling system (ACS) has a quite low probability. Furthermore, a weight factor is introduced so as to investigate the low failure probability events with a comparative small number of the sampling.

Study on self-wastage phenomenon at heat transfer tube in steam generator of sodium-cooled fast reactor with consideration of thermal coupling of fluid and structure

Kojima, Saori*; Uchibori, Akihiro; Takata, Takashi; Ohno, Shuji; Fukuda, Takeshi*; Yamaguchi, Akira*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 8 Pages, 2016/11

Analytical evaluation on a self-wastage phenomenon at heat transfer tubes in the steam generator of sodium cooled fast reactors has been performed by using the sodium-water reaction analysis code SERAPHIM. In this study, a fluid-structure thermal coupling model was developed and incorporated in the SERAPHIM code to evaluate heat transfer between the sodium-side reacting flow and the outer surface of the heat transfer tube. The effect of the fluid-structure thermal coupling model on the temperature field was demonstrated through the numerical analyses.

Study on spray cooling capability for spent fuel pool at coolant loss accident, 1; Research plan

Liu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Koizumi, Yasuo; Yoshida, Hiroyuki; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 4 Pages, 2016/11

The Fukushima Daiichi NPP accident asks that the accident management of the LOCA in the SFPs must be considered to avoid occurrences of severe accident in the SFPs. To prevent the failure of the spent fuel assemblies at the LOCA, transportable spray systems are expected to be put into use to discharge water into fuel assemblies to moderate the temperature increase. To apply the spray system as a countermeasure for the LOCA of the SFP, the capability of the spray cooling system must be evaluated to keep the spent fuel rods safety. JAEA has started the research project to investigate the spray cooling capability for the SFP. In this research project, we aim to construct a numerical simulation method for evaluating the capability of the spray cooling. To develop the method, the basic key phenomena that affect the cooling performance must be clarified and the validation data required for the code development. To clarify the basic key phenomena that affect the cooling performance, that is, the CCFL and the drop size effect on the CCFL, and to obtain the code validation data, we are planning to carry out 2 experiments with two test sections, the spray visualization experiment and the spray cooling experiment. The spray visualization test section aims to get CCFL data in air-water two-phase flow and to understand the two-phase flow behavior over the upper tie plate. The spray cooling test section aims to get the CCFL data in steam-water two-phase flow and to obtain the validation data. This paper focus on the outline of the research plan for the whole research project.