Operational Experiences of Fast Neutron Source Reactor “YAYOI” as Pulsed Fast Reactor

University of Tokyo research reactor “YAYOI” is intended to be operated as a dynamic fast neutron source reactor as well as a stationary one. It is equipped with reactivity adding devices with both slow and quick action, and a LINAC PNS (Pulsed Neutron Source) to be operated with the devices mentioned above. The unique idea of fly-through type pulse reactivity addition into the core lends itself to minimizing thermal shock problems pertaining to fast burst reactors thereby increasing safety of a single shot type burst reactor.

Operational experiences of YAYOI obtained during the dynamic testing of super critical state are described here with some explanation of design aspects of YAYOI as a fast pulsed reactor.

Throughout present experiments, the super prompt critical state reactivity of about up to 29 cents was realized for YAYOI core, and it was confirmed that the sizes of pulse power were well controllable with this reactivity pulser (R-P) mode pulse operation.     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

Criterion of Criticality for Reactor With External Neutron Source

There is a neutron source with 10~9 s~(-1) neutrons in core of CEFR during start up test and operation ofCEFR. For judging the criticality of reactor with external neutron source and near criticality, it isimportant that the neutron level changes in core with time must be understood after introducing positivereactivity to core with external neutron source. Using the point model and Laplace transform, it isobtained that the neutron level changes in core with time for exchanging from one subcriticality to another     

Nuclear Safeguards Techniques Non-destructive Assaying for Attributes of Fast Neutron Reactor＇s Fresh Fuel

We have adopted active neutron coincidence method and y-ray measurement method in order to measure the attributes of fast neutron reactor＇s fresh fuel. An improved uranium neutron coincidence collar （UNCC） was used for measuring the ^235U enrichment distribution of the 2 pieces fuel assembles. The measured results show curve＇s trend is the same as the fact of distribution. Fig. 1 shows the assembly＇s neutron coincidence counting rate.     

Feynman-α Analysis for a Thermal Subcritical Reactor System Driven by an Unstable 14MeV Neutron Source

In a series of Feynman-α correlation measurements for a thermal Accelerator-Driven System (ADS) with 14MeV neutrons at the Kyoto University Critical Assembly (KUCA), an unstable accelerator condition such as a drift of beam current has been frequently observed. Neutron source instability caused by such unavoidable beam-current instability resulted in a divergent variance-to-mean ratio and, consequently, the correlation analysis failed. Nevertheless, we attempted to apply a difference-filtering technique to the correlation analysis to reduce the influence of the above instability. The present attempt resulted in consistent prompt-neutron decay constants with those obtained in a previous pulsed neutron experiment. The application of the filtering is expected to enhance the robustness of Feynman-α analysis against various instabilities of accelerator operation in actual ADS.     

Verification Experiment of Threshold Detector 107Ag(n,n′)107mAg on “YAYOI” Standard Neutron Field

The validity of the ¹⁰⁷Ag(n, n′)^107mAg reaction for reactor neutron dosimetry has been examined through an irradiation experiment using the standard neutron field in the “YAYOI”, a fast neutron source reactor. The test foil used was 99.0% ¹⁰⁷Ag enriched silver, which was irradiated along with five reference foils already well evaluated. The reaction rate of 107Ag was determined by K X-ray counting with a high-purity Ge detector. The cross section for the ¹⁰⁷Ag(n, n′) ^107mAg reaction was compiled from the excitation cross section data of ¹⁰⁷Ag published in ENDF/B-IV, in combination with the decay scheme of ^107mAg listed in the “Table of Isotopes“. Using the neutron spectrum which had previously been precisely measured, it was found that the compiled activity of ^107mAg was underestimated by 12.8±5%. This error is attributed to significant underestimation involved in the cross section of the ¹⁰⁷Ag(n, n′)^107mAg reaction.     

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO₂ fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.     

Core Restraint Analysis for 75 MW Power Ascension Tests of Experimental Fast Reactor “JOYO”

In the “JOYO” initial power ascension tests to 75 MW, unexpected large negative power coefficients were observed during ascension to new power levels beyond 50 MW, which results in approximately a 40¢ permanent loss in reactivity.

This anomaly was studied from the viewpoint of radial dimensional change in the core using a three-dimensional structural analysis code along with a whole-core thermal-hydraulic analysis code. Conclusions drawn from the study can be summarized as follows:

(1) The absolute value of the power coefficient increases sharply beyond the 50 MW power level due to the outward leaning and thermal bowing of core elements, but the calculated increase in the coefficient is less than 20% of the observed value.

(2) At most, an 80 permanent loss in reactivity is caused by this outward leaning, which is 20% of the observed value.

(3) These results confirm that the radial dimensional change can not be the main cause for the power coefficient anomaly.     

Conceptual Design of a Modular Island Core Fast Breeder Reactor “RAPID-M”

A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor performance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified refueling by replacing IFAs in which all the core and blanket fuel elements are comprised. In this paper, the 600 MWe RAPID-M design consists of 7 IFAs is presented. Significant reactor mass savings and the improvement of inherent safety features are discussed. Plant dynamics analyses using the multi-point reactor kinetics equations to accommodate the modular core configuration demonstrated a favorable transient response in case of unprotected transient over power (UTOP).     

Pulse Die-Away Experiment in Small Graphite Prisms with the “Cold” Neutron Source

Abstract

Recently a remarkable progress has been made in fast ADCs such as flash ADCs. This paper shows a method of applying these fast ADCs to the radiation measurements. Direct sampling and digitizing of output signal from preamplifier were examined using the weighted sum of sampled data in order to obtain the energy of incident radiations. As a result, this method is applied to r-ray spectroscopy by a Nal(Tl) scintillation counter to give sufficient performance. Because the present method treats the signal waveform itself, the standard waveform of input signals can be measured. After the standard waveform was obtained by preliminary measurement, a trial was made to select the similar waveforms to the standard waveform using the resembrance factor. With this selection, improvement of energy resolution was observed in the measurement of ⁶⁰Co γ-ray.     

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.