Development of Evaluation Code for Long-Term Cooling of Containment with Passive Safety Feature

Simplified BWRs are characterized as an adoption of a passive ECCS and a passive containment cooling system (PCCS). While a passive ECCS has a short term core cooling function, a PCCS has a long-term decay heat removal function. As a PCCS, several concepts, differing in cooling location and method employed, have been considered. From the containment thermal- hydraulic response analysis viewpoint, simplified BWRs are essentially different from the current BWRs. For evaluating and comparing the performance of several PCCSs over full break spectra, the new containment safety evaluation code TOSPAC was developed as a preliminary design tool for PCCS. This paper summarizes the thermal-hydraulic modelings of the TOSPAC code and the validity evaluation of the TOSPAC code, compared with TRAC-BF1 calculation.

From the validity evaluation concerning a main steam line break (MSLB) accident analysis for an isolation condenser (I/C) as a PCCS, it was found that the TOSPAC calculation result shows reasonable agreement with that for TRAC, even though the TOSPAC consists of simpler modelings.     

Deposition of Nickel and Cobalt Ions with Iron Crud on Heated Stainless Steel Surface under Nucleate Boiling Conditions

Deposition behaviors of Ni (II) and Co (II) ions on a heated surface using simulated Fe crud which was mainly composed of amorphous Fe(III) hydroxide have been studied under nucleate boiling conditions at 553 K and 70 atm. The deposition process of Ni (II) and Co (II) ions is divided into two stages. The first stage is the deposition of hydroxide precipitate on the heated surface by microlayer evaporation and drying out in the nucleate boiling bubble. The second is settlement by conversion of hydroxide into oxides such as NiO, NiFe₂O₄, CoO and CoFe₂O₄. The effective deposition coefficients of Ni(II) and Co (II) ions, without supplying Fe crud, are smaller than that of α-Fe₂O₃ because of the solubility of those hydroxides at a low concentration condition. Their effective deposition coefficients increase with the simulated Fe crud concentration in feedwater and saturate at a value of 0.3 which is calculated theoretically, because the Ni and Co hydroxides react with the simulated Fe crud to produce insoluble NiFe₂O₄ and CoFe₂O₄ on the heated surface. The reaction of Ni deposit with α-Fe₂O₃ does not proceed, but NiO is produced. The reaction rate of Ni deposit with the simulated Fe crud on the heated surface is higher than that of Co deposit.     

Irradiation-Induced Dissolution of Precipitates in Zircaloy-2

Calibration curves of extremely low concentrations of the alloying elements Sn, Fe, Cr and Ni in Zircaloy were obtained, using standard samples, by energy dispersive X-ray spectroscopy to measure concentration distributions of alloying elements dissolved in the Zircaloy matrix. Their detectable limits were 0.21 at% for Sn, 0.06 at% for Fe. 0.04 at% for Cr and 0.03 at% for Ni. Then concentration distributions of alloying elements in unirradiated and neutron irradiated Zircaloy-2 were measured using these calibration curves. It was confirmed that neutron irradiation increased the dissolved concentrations of Fe. Cr and Ni. Furthermore, Cr diffused slower than Fe and Ni. It was suggested that the rate limiting process of irradiation-induced dissolution from Fe, Cr-type precipitates into the matrix was the diffusion of alloying atoms in the precipitates and that the dissolution process proceeded due to displacement of alloying atoms from the precipitates into the matrix and diffusion in the matrix.     

Heat Removal Performance Evaluation of Several Passive Containment Cooling Systems during Loss of Coolant Accident

As a passive containment cooling system (PCCS), which is adopted in simplified BWRs, several concepts, differing in cooling location and method, such as the suppression chamber water wall, the drywell water wall, the isolation condenser (I/C) and the drywell cooler, have been considered. This paper summarizes the characteristics of each PCCS concept, and the analysis results of the performance for several PCCSs during a main steam line break LOCA for a reference simplified BWR plant, obtained by the newly developed containment thermalhydraulic response analysis code TOSPAC.

The performance comparison suggests that I/C and drywell cooler have good heat removal capability with regard to the smallest heat transfer area among PCCS concepts evaluated in the present analysis. I/C removes decay heat efficiently, since it absorbs steam directly from the reactor pressure vessel, which is the hottest portion inside the containment. The suppression chamber water wall is ineffective, mainly due to high non-condensable gas partial pressure in the suppression chamber, and low suppression pool temperature.

Calculations of other pipe breaks were also implemented for the reference plant adopting I/C as PCCS. The results show the effectiveness of the I/C cooling over a wide range of break spectra.     

International Conference on Water Chemistry in Nuclear Power Plants

Minor actinide (MA) transmutation performances were evaluated for BWR cores with mixed oxide (MOX) fuels of various hydrogen-to-heavy metal atom number ratios (H/HM) to decrease the work for high-level radioactive wastes (HLW) management. One effective approach to increase the MA transmutation ratio is a multi-recycle core of h4OX fuel with MA. The decrease of the MA-to-fissile Plutonium (Puf) amount ratio makes it possible to recycle MA without additional plutonium (Pu) obtained from reprocessing of LWR fuel. For this purpose, a T-ratio, which is the MA-to-Puf amount ratio change during burnup, of less than unity is required, and H/HM must be less than 3 when MA enrichment is more than 2 wt%. For multi-recycle cores with the H/HM of 3 and the initial MA enrichment of 5 wt%, more than 50% of MA load initially can be transmuted by the third recycle.     

Analytical Method for Recalibration of Incore and Excore Nuclear Instrumentation System of PWRs

In pressurized water reactors (PWRs) of nuclear power plants, it is periodically required to recalibrate the correlation between the incore and excore nuclear instruments, where the recalibration guarantees the accurate monitoring of incore axial power distributions by excore instrumentation. In order to obtain the correlation between incore power distribution and excore detector response, however, it is necessary to perturb the axial power distribution of the reactor by control rods or to introduce a xenon spatial oscillation.

In this paper a simplified analytical method that eliminates such perturbations is proposed to evaluate the correlation between incore power distribution and excore detector currents which is utilized to recalibrate the nuclear instrumentation system. This method provides as good an accuracy as that obtained by conventional methods.     

Comment on Shibata-san's “Evaluation of Neutron Nuclear Data for 3He and 4He”

As advanced monitoring and controlling systems have been developed, accuracy of real-time simulators must be improved and simulation limits must be extended. Therefore the authors have developed a distributed simulation system to achieve high processing performance using low cost hardware. Moreover, the authors have developed a thermal-hydraulic computer code, using drift-flux non-equilibrium model. This code can realize a high precision two-phase flow analysis, to be considered to have the same prediction capability as two-fluid models, while achieving high speed and stability for real-time simulators. The distributed plant simulator for PWR plants was realized as a result. The distributed simulator consists of multi-processors connected to each other by an optical fiber network. Controlling software for synchronized scheduling and data transfer was also developed. In terms of the code development, the numerical calculation method and implemented correlation chosen for this code are described. The simulation results during both code development and validation testing of the four loop PWR simulator are compared with experiment data and real plant data; the agreement was satisfactory for a plant simulator. The simulation speed was also satisfactory being twice as fast as real-time.     

Evaluation of Reactivity and Xe Behavior during Daily Load Following Operation

A boiling water reactor (BWR) has an excellent load following capability provided by a core flow control, which is used for changing a reactor power level and for compensating the subsequent Xe concentration change.

The core characteristics during load following operations are investigated in detail, using our reactor core simulator. Comparisons of changes of the Doppler reactivity, the void reactivity and the Xe reactivity during transients are performed. Also the features of Xe transient during load following operations are shown.

It has been shown that the core flow change required to compensate the Xe reactivity change produces much greater change of the void reactivity than that required for power level changes, and that the resulting local power change in the lower part of the core is greater than that in the upper part, because the Xe concentration change in the lower part is hardly compensated by the core flow control. Also the effects of power level changes, cycle patterns, and initial concentration of Xe and I on the Xe transient behavior have been investigated.     

Nuclear Design Study on a Super High-burnup Fuel Assembly for Pressurized Water Reactors Using Transuranium

A conceptual design study was carried out on a super high-burnup mixed-oxide (MOX) fuel assembly (SHB FA) for pressurized water reactors (PWRs) using transuranium (TRU). This study aims to avoid the surplus plutonium (Pu) accumulation and to reduce the accumulation of long-lived radioactive minor actinides (MAS) by utilizing the currently existing PWRs under the condition that the Japanese program to develop fast breeder reactors (FBRs) is tend to delay. For this purpose, an SHB FA with discharged burnup of ?80 GWd/t was investigated by utilizing MAS positively as both burnable absorbers and fissile suppliers and loading high-content Pu. It is possible to load the SHB FAs in a current PWR together with UO₂ FAs and to use 2.5 times as much amount of Pu as that in a standard 1/3 MOX core. Moreover, it is found to be possible to reduce the total number of fresh FAs further from that of a high-burnup (55 GWd/t in maximum) UO₂ (4.9 wt%) core and also to reduce the accumulation of MAS in the nuclear fuel cycle significantly.     

Proposal of an Effective Initial Guess Scheme for Higher Harmonic Mode Calculation of Boiling Water Reactor Cores

Response functions of a BC501A liquid organic scintillator for neutrons up to 800 MeV have been measured at the heavy-ion accelerator of the National Institute of Radiological Sciences, Japan. A thick graphite target was bombarded with 400-MeV/u C ions and 800-MeV/u Si ions to produce high-energy neutrons whose kinetic energy was determined by the time-of-flight method. The measured response functions were compared with the results obtained using SCINFUL-QMD code, and the accuracy of the code was experimentally verified up to 800 MeV. This work will contribute to extending the energies measurable with our new radiation dose-monitoring system (DARWIN), which is based on the BC501A scintillator.