Reduced-Moderation Water Reactor Research Articles

A feasibility study on long-life reduced-moderation water reactor with highly protected Pu breeding by doping with minor actinides

The present paper is focused on the feasibility of reduced-moderation water reactor (RMWR) which could simultaneously achieve the extension of core life-time, and highly protected plutonium (Pu) breeding performance with short compound system doubling time (CSDT) by neptunium-237 (237Np) or americium-241 (241Am) doping of the inner blanket of the RMWR. As a preliminary analysis of the RMWR, a simplified fuel pin configuration was analyzed. In case of 60% doping of 237Np and 241Am in the inner blanket, the maximum available effective full power days (EFPDs) were much longer and were about 15,100 and 14,200 EFPDs, respectively. For the proliferation resistance of Pu produced in the blanket, the total Pu in the inner and lower/upper blanket was below the practically unusable criterion for an explosive device proposed by Pellaud, and was technically unfeasible for high-technology hypothetical nuclear explosive devices (HNEDs) proposed by Kessler and Kimura. The protected Pu breeding performance was analyzed and in case of 60% doping of 237Np and 241Am, CSDT was short by approximately 40 and 50years. Therefore, the feasibility of RMWR was shown which could simultaneously achieve the extension of core life-time and have highly protected Pu breeding performance with short CSDT by doping the inner blanket with minor actinides (MAs). The objective of this study was thus to evaluate the performance of the concept with respect to the core life-time, proliferation resistance of Pu and CSDT. The technological feasibility of such core concept will have to be evaluated by further dedicated analyses.

Design Study to Increase Plutonium Conversion Ratio of HC-FLWR Core

The innovative water reactor for flexible fuel cycle (FLWR) is an advanced reactor concept based on the well-developed light water reactor (LWR) technology. It is to be introduced in two stages to achieve effective and flexible utilization of the uranium and plutonium resources. In the first stage, the high-conversion-type reactor concept (HC-FLWR) is to be introduced, with a core that achieves a fissile Pu conversion ratio of 0.84. Then, in the second stage, the reduced-moderation water reactor (RMWR) concept can be introduced, with a breeder-type core that achieves a fissile Pu conversion ratio of 1.05. From the viewpoint of effective introduction of the high-conversion-type reactor, such as the introduction capacity of the reactor, HC-FLWR is required to further raise the fissile Pu conversion ratio to [approximately]0.95.This study aims to develop a new core design concept for the high-conversion-type core, HC-FLWR+, to achieve the higher fissile Pu conversion ratio of [approximately]0.95 under the framework of UO2 and U-Pu mixed-oxide (MOX) fuel technologies for LWRs. For raising the fissile Pu conversion ratio and controlling the void reactivity characteristics of the core, the concept of FLWR/MIX fuel assembly, which uses MOX and enriched UO2 fuel rods, is utilized.The relationships between the main design parameters and the core performance index parameters are clarified in this study. When the fuel rod diameter and the clearance range from 1.23 to 1.28 cm and 0.25 to 0.20 cm, respectively, under the same pitch of 1.48 cm, the fissile Pu conversion ratio and the core average discharge burnup range from 0.89 to 0.94 and 53 to 49 GWd/tonne, respectively (the fissile Pu conversion ratio and the burnup are subject to a trade-off). Furthermore, when 235U enrichment in the UO2 fuel rods is increased from 4.9 to 6 wt%, the fissile Pu conversion ratio improves to 0.97.From these relationships, two representative core designs with fissile Pu conversion ratios of 0.91 and 0.94 and one optional design with a ratio of 0.97 were obtained. Hence, the flexibility of HC-FLWR+ concept to achieve a higher fissile Pu conversion ratio of [approximately]0.95 has been revealed. Together with the standard HC-FLWR design, the concept covers a wide range of needs on fissile Pu conversion ratio from 0.84 up to 0.97, with design variations that are expected to be within the scope of current boiling water reactor and MOX fuel technologies.

低減速軽水炉の燃料集合体局所出力分布の限界出力への影響評価

A reduced-moderation water reactor (RMWR) is an innovative light water reactor comprising tight-lattice fuel assemblies with a gap clearance of around 1.0 mm for a reduction in water volume ratio to achieve a high conversion ratio. By taking into account recent experimental findings from 37-rod tight-lattice bundle thermal-hydraulic tests indicating that the critical power tends to be higher for a peripheral peak local power distribution than for a flat power distribution, the viability of fuel assembly designs with fewer types of plutonium enrichment of MOX fuel, which may result in a large local peaking factor in peripheral rods, was assessed in this report. Critical powers of 217-rod bundles with peripheral peaks for upper and lower MOX regions of the double-flat core of the RMWR were calculated by a subchannel analysis code NASCA. Peripheral peaking with the corresponding local peaking factor for a uniform plutonium enrichment design yields almost the same critical power as that for a flat power distribution. Thus, a reduction in fuel fabrication burden may be possible by decreasing the number of types of plutonium fuel enrichment while maintaining the same thermal-hydraulic margin as the fuel assembly design with five enrichment types of MOX fuel.

Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241Pu content in the initial fuel, and the decay heat mainly depends on 238Pu and 244Cm. The contribution of 244Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from the waste disposal point of view, the characteristics of the heat generation FP elements, the platinum group metals, Mo and the long-lived FPs (LLFPs) were also investigated.

Conceptual Design of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) and its Recycle Characteristics

A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been proposed based on the well-experienced Light Water Reactor (LWR) technologies. The concept aims at effective and flexible utilization of the uranium and plutonium resources through the plutonium multiple recycling by the two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from the current LWR and coming MOX-LWR technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0, being useful for the long-term sustainable energy supply through the plutonium multiple recycling. The FLWR is a BWR-type reactor, and its core design is characterized by the short flat core, which consists of hexagonal-shaped fuel assemblies in the triangular-lattice fuel rod configuration with the highly enriched MOX fuel and Y-shaped control rods. The cores in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of the natural uranium resource or establishment of an economical reprocessing technology of the MOX spent fuel. Detailed investigations have been performed on the core design, in conjunction with the other related studies, and the results obtained so far have shown the proposed concept is feasible and promising.

ICONE15-10555 CONCEPTUAL DESIGN OF INNOVATIVE WATER REACTOR FOR FLEXIBLE FUEL CYCLE (FLWR)

In order to ensure the sustainable energy supply in the future based on the matured light water reactor (LWR) technologies, a concept of innovative water reactor for flexible fuel cycle (FLWR) has been investigated in JAEA. The concept utilizes the tight-lattice core loaded with the plutonium mixed oxide (MOX) fuel, and consists of two steps in the chronological sequence. The first step realizes a high conversion type one (HC-FLWR), which is basically intended to keep the smooth technical continuity from the current LWR / MOX-LWR technologies. The second represents the reduced-moderation water reactor (RMWR) concept, which realizes a high conversion ratio over 1.0 and is preferable for the long-term sustainable energy supply through plutonium multiple recycling. In the present paper, investigation results on the FLWR conceptual design are presented mainly from the neutronics point of view. The design of the HC-FLWR core has been recently improved, and detailed core properties have been evaluated by neutronics and thermal-hydraulics coupled calculations. The core can achieve the average burn-up around 55GWd/t as well as the negative void reactivity coefficients. The size of its fuel assembly is the same as in RMWR, and hence, it can be replaced with the highly tight one for RMWR.

Conceptual Design of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) and its Recycle Characteristics

A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been proposed based on the well-experienced Light Water Reactor (LWR) technologies. The concept aims at effective and flexible utilization of the uranium and plutonium resources through the plutonium multiple recycling by the two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from the current LWR and coming MOX-LWR technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0, being useful for the long-term sustainable energy supply through the plutonium multiple recycling. The FLWR is a BWR-type reactor, and its core design is characterized by the short flat core, which consists of hexagonal-shaped fuel assemblies in the triangular-lattice fuel rod configuration with the highly enriched MOX fuel and Y-shaped control rods. The cores in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of the natural uranium resource or establishment of an economical reprocessing technology of the MOX spent fuel. Detailed investigations have been performed on the core design, in conjunction with the other related studies, and the results obtained so far have shown the proposed concept is feasible and promising.

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It has a very tight triangular fuel rod lattice and a high coolant void fraction. The RMWR core axially has two short and flat uranium plutonium mixed oxide (MOX) regions with an internal blanket region in between, in order to avoid a positive void reactivity coefficient. The MOX regions are sandwiched between upper and lower blanket regions, in order to increase a conversion ratio. In this small reactor core, leakage of neutrons is expected to be larger than in a large core. Therefore, a core design concept different from that for a large core is necessary. Core burnup calculations and nuclear and thermal-hydraulic coupled calculations were performed in the present study with SRAC and MOSRA codes. MVP code was also used to obtain control rod worth. Because of its large neutron leakage, keeping the void reactivity coefficient negative is easier for S-RMWR than RMWR. Thus, the heights of MOX region can be taller and the plutonium enrichment can be lower than in RMWR. On the other hand, to achieve the conversion ratio of 1.0, radial blanket and stainless steel reflector assemblies are necessary, whereas they are not needed for RMWR.

Concept of innovative water reactor for flexible fuel cycle (FLWR)

In order to ensure sustainable energy supply in the future based on the matured light water reactor (LWR) and coming mixed oxide (MOX)-LWR technologies, a concept of innovative water reactor for flexible fuel cycle (FLWR) has been investigated in Japan Atomic Energy Research Agency (JAEA). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming MOX-LWR technologies without significant technical gaps. The second part represents the reduced-moderation water reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-developed LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the future fuel cycle circumstances during the reactor operation period around 60 years. At present, reprocessed plutonium from the LWR spent fuel is to be utilized in MOX-LWR. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or MOX-LWR. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the MOX-LWR utilization. When the fuel cycle for plutonium multiple recycling with MOX fuel reprocessing is realized, the fuel assembly will be replaced with another type of the tight-lattice one for RMWR with different rod diameter, rod gap width and so forth even in the same reactor system, being flexibly corresponding to the fuel cycle circumstances. Investigation on the core for both the parts of the FLWR concepts has been performed, including the core conceptual design, the core characteristics under Pu multiple recycling, the thermal hydraulic investigation in the tight-lattice core, and so forth. Up to the present, promising results have been obtained.

Pressure Drop Experiments using Tight-Lattice 37-Rod Bundles

In order to design a Reduced-Moderation Water Reactor (RMWR) core from a thermal-hydraulic point of view, an evaluation method on the pressure drop in a tight-lattice rod bundle is required. In this study, axial pressure drops in tight-lattice 37-rod bundles were measured under conditions of 2-9 MPa in exit pressure and 200-1,000 kg/(m2·s) in mass velocity. The measured pressure drops were compared with calculated ones by the evaluation method with the Martinelli-Nelson's correlation. The comparison shows that a single-phase friction factor can be applied not only to a circular tube but also to a tight-lattice bundle except for an extremely small gap width. Then two-phase friction loss is a dominant component and accounts for about 60% of the pressure drop under an RMWR nominal operating condition. The evaluation method can evaluate effects of the flow area configuration (rod number, rod diameter, gap width) and axial power distribution under a wide range of flow conditions, and it can yield a good prediction of the pressure drop in a tight-lattice bundle.

Critical Power Experiment with a Tight-Lattice 37-Rod Bundle

Since most of critical power or CHF data have been collected in tube, annulus, or BWR geometries under BWR flow conditions, critical power data for highly tight and triangular lattice bundles under low mass velocity are indispensable for thermal-hydraulic design of Reduced-Moderation Water Reactor. Large-scale thermal-hydraulic experiments which use a basic 37-rod bundle test section (rod diameter: 13.0 mm, gap width between rods: 1.3 mm) were therefore carried out in this study within range of 2-9 MPa in pressure and 150-1,000 kg/(m2·s) in mass velocity. Fundamental characteristics of boiling transition were investigated through effects of flow parameter on critical power and those of rod number. It was confirmed that the fundamental characteristics in 37-rod bundle are similar to those in 7- rod bundle and in case of the BWR geometry. The results of the transverse non-uniform power distribution test and subchannel analysis suggest that the critical power becomes higher when the transverse local quality distribution closes to uniform.

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It has a very tight triangular fuel rod lattice and a high coolant void fraction. The RMWR core axially has two short and flat uranium plutonium mixed oxide (MOX) regions with an internal blanket region in between, in order to avoid a positive void reactivity coefficient. The MOX regions are sandwiched between upper and lower blanket regions, in order to increase a conversion ratio. In this small reactor core, leakage of neutrons is expected to be larger than in a large core. Therefore, a core design concept different from that for a large core is necessary. Core burnup calculations and nuclear and thermal-hydraulic coupled calculations were performed in the present study with SRAC and MOSRA codes. MVP code was also used to obtain control rod worth. Because of its large neutron leakage, keeping the void reactivity coefficient negative is easier for S-RMWR than RMWR. Thus, the heights of MOX region can be taller and the plutonium enrichment can be lower than in RMWR. On the other hand, to achieve the conversion ratio of 1.0, radial blanket and stainless steel reflector assemblies are necessary, whereas they are not needed for RMWR.

Pressure Drop Experiments using Tight-Lattice 37-Rod Bundles

In order to design a Reduced-Moderation Water Reactor (RMWR) core from a thermal-hydraulic point of view, an evaluation method on the pressure drop in a tight-lattice rod bundle is required. In this study, axial pressure drops in tight-lattice 37-rod bundles were measured under conditions of 2-9 MPa in exit pressure and 200-1,000 kg/(m2·s) in mass velocity. The measured pressure drops were compared with calculated ones by the evaluation method with the Martinelli-Nelson's correlation. The comparison shows that a single-phase friction factor can be applied not only to a circular tube but also to a tight-lattice bundle except for an extremely small gap width. Then two-phase friction loss is a dominant component and accounts for about 60% of the pressure drop under an RMWR nominal operating condition. The evaluation method can evaluate effects of the flow area configuration (rod number, rod diameter, gap width) and axial power distribution under a wide range of flow conditions, and it can yield a good prediction of the pressure drop in a tight-lattice bundle.

1187. 大規模シミュレーションによる稠密炉心内気液二相流特性の解明,(III)

The reduced-moderation water reactor (RMWR) core adopts a hexagonal tight-lattice arrangement with about 1mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermalhydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing an advanced interface tracking method to improve the conservation of volume of fluid.The present paper describes a part of the verification works of the two-phase flow simulation code TPFIT. The numerical results applied to liquid film falling down on an inclined flat plate are shown and compared with existing experimental results. In the results of numerical simulation, the development of the wave structures could be reproduced by the numerical simulation. The predicted average local film thicknesses almost agreed with Nusselt's mean film thickness. Moreover, the probability density function of local film thickness agreed well with the experimental result. In addition, the trends related to the film velocities were consistent with the theoretical results. From these results, it was confirmed that the TPFIT code is applicable to predict the liquid film flow behavior.

Critical Power Correlation for Tight-Lattice Rod Bundles

Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced-Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration and axially double-humped-heated profile. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux-critical quality type. For low mass velocity region, it is written in critical quality-annular flow length type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: Rod number lower than 37, rod gap from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2,000 kg/m2·s and pressure from 2 to 11 MPa.

OS8-04 稠密37本バンドル熱特性試験とギャップ幅効果の評価(OS8 軽水炉・新型炉・核燃料サイクル)

A thermal/hydraulic feasibility project for Reduced-Moderation Water Reactor (RMWR) was started since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes are being carried out. In this paper, results of the large-scale thermal-hydraulic tests which simulate the RMWR bundle are summarized. Especially, gap width effect on critical power which was investigated using two 37-rod bundle test sections with 1.3 and 1.0mm in gap width is reported. As a result, it is confirmed that (1) fundamental characteristics of flow parameter impact on critical power are similar between the results using two test sections with 1.3mm and 1.0mm in gap width, (2) effect of cross-sectional power distribution could be expressed by the power factor of subchannel where BT (Boiling Transition) would occur, (3) quasi-steady-state predictive method for BT is applicable for postulated transients and (4) proposed evaluation method of pressure drop could yield good predictions in the tight-lattice bundle.

3526 稠密格子ロッドバンドルにおける限界出力相関式の改良と過渡限界出力試験の解析(S49-3 流動,原子炉用機器(2),S49 原子炉システムおよびその要素技術)

To develop the design correlation for the critical power in rod bundles is indispensable for R&D of Reduced-Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration. In this research, critical power correlation developed for tight-lattice rod bundles is improved by using newly obtained 37-rod bundle data with 1.0mm gap between rods. Compared to the existing correlation, which based on the data obtained in 37-rod bundle data with 1.3mm gap, the improved one expands its adaptability to tighter rod bundles whose rod gap is 1mm, by revising the dependencies on mass velocity and equivalent heated diameter. The improved correlation is confirmed being able to give good critical power predictions to both its based datasets and Bettis Atomic Power Laboratory (BAPL) data. The standard deviation of ECPR (Experimental Critical Power Ratio) to JAERI 37-rod data and BAPL data are 7.3% and 8.4%, respectively. Moreover, the improved correlation is implemented into TRAC code, with which the analyses of transient processes postulated for the RMWR are carried out. For the postulated transients for the RMWR, it has been confirmed that BTs can be predicted within the accuracy of the implemented critical power correlation.

Critical Power Correlation for Tight-Lattice Rod Bundles

Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced-Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration and axially double-humped-heated profile. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux-critical quality type. For low mass velocity region, it is written in critical quality-annular flow length type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: Rod number lower than 37, rod gap from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2,000 kg/m2·s and pressure from 2 to 11 MPa.

Critical Power in 7-Rod Tight Lattice Bundle

The Reduced-Moderation Water Reactor (RMWR) has recently becomes of great concern. The RMWR is expected to promote the effective utilization of uranium recourse. The RMWR is based on water-cooled reactor technology, with achieved under lower core water volume and water flow rate. In comparison with the current light water reactors whose water-to-fuel volume ratio is about 2-3, in the RMWR, this value is reduced to less than 0.5. Thereby, there is a need to research its cooling characteristics. Experimental research on critical power in tight lattice bundle that simulates the RMWR has been carried out in Japan Atomic Energy Research Institute (JAERI). The bundle consists one center rod and six peripheral rods. The 7 rods are arranged on a 14.3mm equilateral triangular pitch. Each rod is 13mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under G=100-1400kg/m2s, Pex=2-8.5MPa. Effects of mass velocity, inlet temperature, pressure, radial peaking factor and axial peaking factor on critical power and critical quality are discussed. Compared with axial uniform heating condition, the axial non-uniform heating condition causes an obvious decrease in critical quality. Arai correlation, which is the only correlation that has been optimized for tight lattice condition, is verified with the present experimental data. The correlation is found to be able to give reasonable prediction only around RMWR nominal operating condition.