Critical Power Ratio Research Articles

Core design options for high conversion BWRs operating in Th–233U fuel cycle

Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th–233U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating.Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design.

The transient analysis of single turbine control valve closure for Lungmen ABWR

The single control valve closure in fast (SCVCF) event is the most limiting transient in terms of delta critical power ratio (ΔCPR) for the Lungmen Plant, which is a basis to determine the operating limit minimum critical power ratio value. The partial arc mode is adopted in Lungmen Plant to control the position of the turbine control valve. However, the transient analyses presented in the Lungmen Final Safety Analysis Report (FSAR) assume that the TCVs are in the full arc mode. In this study, the Lungmen RETRAM model with partial arc mode is used to analyze the SCVCF event to offer more realistic results than the FSAR. It is concluded that the most limiting maximum combined flow limiter (MCFL) setting in RETRAN analysis is different from that of FSAR. An optimum operating range for the MCFL is suggested to gain the margin against the operating drift. Additionally, a Heat Flux Factor index is defined to appropriately determine the ranking of these cases in terms of ΔCPR.

Huitzoctli: A system to design Control Rod Pattern for BWR’s using a hybrid method

Huitzoctli system was developed to design Control Rod Patterns for Boiling Water Reactors (BWR). The main idea is to obtain a Control Rod Pattern under the following considerations: (a) the critical reactor core state is satisfied, (b) the axial power distribution must be adjusted to a target axial power distribution proposal, and (c) the maximum Fraction of Critical Power Ratio (MFLCPR), the maximum Fraction of Linear Power Density (FLPD) and the maximum Fraction of Average Planar Power Density (MPGR) must be fulfilled. Those parameters were obtained using the 3D CM-PRESTO code. In order to decrease the problem complexity, Control Cell Core load strategy was implemented; in the same way, intermediate axial positions and core eighth symmetry were took into account. In this work, the cycle length was divided in 12 burnup steps. The Fuel Loading Pattern is an input data and it remains without changes during the iterative process. The Huitzoctli system was developed to use the combinatorial heuristics techniques Scatter Search and Tabu Search. The first one was used as a global search method and the second one as a local search method. The Control Rod Patterns obtained with the Huitzoctli system were compared to other Control Rod Patterns designs obtained with other optimization techniques, under the same operating conditions. The results show a good performance of the system. In all cases the thermal limits were satisfied, and the axial power distribution was adjusted to the target axial power distribution almost completely. It is very important to point out that, even though the cycle length improvement was not the main idea of this work, the effective multiplication factor ( k eff ) at the end of the cycle was improved in all cases tested. The Huitzoctli system was programmed using Fortran 77 language in an Alpha Workstation with UNIX operating system.

The design assessment of limiting transient event by RETRAN-02 and VIPRE-01 for Lungmen ABWR

System thermal hydraulic and hot channel calculations are utilized in this paper to analyze the limiting transient event for Lungmen nuclear power plant (LMNPP). The difference of initial critical power ratio (ICPR) and minimum critical power ratio (MCPR), i.e., ΔCPR, is employed to quantify an impact caused by a transient to LMNPP. Based on the ΔCPR criterion, the Lungmen Preliminary Safety Analysis Report (PSAR) concluded that Load Rejection (LR) event was the most limiting event. However, Inadvertent Closure of One Turbine Control Valve (ICOTCV) event was specified as the most limiting core-wide transient event for LMNPP in the Lungmen Final Safety Analysis Report (FSAR). Moreover, the partial-arc TCV control mode is now applied for linear steam flow control, while the Lungmen FSAR utilized full-arc mode for ICOTCV transient analysis. Based on the above differences, it is necessary to re-analyze both LR and ICOTCV transient events for safety consideration. In this paper, ΔCPRs are calculated by the combination of RETRAN-02 and VIPRE-01 codes. The system transient parameters calculated by RETRAN-02 are furthermore submitted into VIPRE-01 for hot channel analysis. The hot channel model constructed by VIPRE-01 simulates the hottest fuel bundle. The critical heat flux (CHF) is calculated by a CHF correlation. Moreover, the different valve control modes are discussed to study the influence on ICOTCV transient. The analytic results by RETRAN-02 and VIPRE-01 are compared with Lungmen FSAR for assessment. The assessment result indicates that the RETRAN-02 and VIPRE-01 analytic modes applied in this paper are capable of simulating the transient behaviors reasonably. Based on the design changes mentioned above, the calculated maximum ΔCPRs are 0.18 and 0.20 for LR and ICOTCV by RETRAN-02 and VIPRE-01, respectively. It is concluded that, first, the impact caused by ICOTCV event is larger than that by LR event with the new fuel type design and full-arc control mode. Second, the adoption of partial-arc control mode is capable of controlling steam flow rate linearly. However it also increases the ΔCPR during ICOTCV transient. This issue should be highlighted after commercial operation.

ICONE19-43536 Development of CFD Analysis Method Based on Droplet Tracking Model for BWR Fuel Assemblies

It is well known that the minimum critical power ratio (MCPR) of the boiling water reactor (BWR) fuel assembly depends on the spacer grid type. Recently, improvement of the critical power is being studied by using a spacer grid with mixing devices attaching various types of flow deflectors. In order to predict the critical power of the improved BWR fuel assembly, we have developed an analysis method based on the consideration of detailed thermal-hydraulic mechanism of annular mist flow regime in the subchannels for an arbitrary spacer type. The proposed method is based on a computational fluid dynamics (CFD) model with a droplet tracking model for analyzing the vapor-phase turbulent flow in which droplets are transported in the subchannels of the BWR fuel assembly. We adopted the general-purpose CFD software Advance/FrontFlow/red (AFFr) as the base code, which is a commercial software package created as a part of Japanese national project. AFFr employs a three-dimensional (3D) unstructured grid system for application to complex geometries. First, AFFr was applied to single-phase flows of gas in the present paper. The calculated results were compared with experiments using a round cellular spacer in one subchannel to investigate the influence of the choice of turbulence model. The analyses using the large eddy simulation (LES) and re-normalisation group (RNG) k-ε models were carried out. The results of both the LES and RNG k-ε models show that calculations of velocity distribution and velocity fluctuation distribution in the spacer downstream reproduce the experimental results qualitatively. However, the velocity distribution analyzed by the LES model is better than that by the RNG k-ε model. The velocity fluctuation near the fuel rod, which is important for droplet deposition to the rod, is also simulated well by the LES model. Then, to examine the effect of the spacer shape on the analytical result, the gas flow analyses with the RNG k-ε model were performed for three kinds of spacers of actual fuel assemblies: a lattice type spacer, a round cell type spacer, and a grid spacer with mixing vanes, respectively. The results indicate that the swirling flow is generated by the mixing vane, and the turbulence energy increases near the surface of fuel rod. The Lagrangian method was applied to droplet behavior in turbulent gas flow around the round cell type spacer and that around a mixing vane spacer as trial analyses. It was confirmed that liquid droplet deposition is promoted in downstream from the mixing vane spacer because of swirling flow.

Effect of automatic recirculation flow control on the transient response for Lungmen ABWR plant

In this study the automatic mode of the recirculation flow control system (RFCS) for the Lungmen ABWR plant has been modeled and incorporated into the basic RETRAN-02 system model. The integrated system model is then used to perform the analyses for the two transients in which the automatic RFCS is involved. The two transients selected are: (1) one reactor internal pump (RIP) trip, and (2) loss of feedwater heating. In general, the integrated system model can predict well the response of key system parameters, including neutron flux, steam dome pressure, heat flux, RIP flow, core inlet flow, feedwater flow, steam flow, and reactor water level. The transients are also analyzed for manual RFCS case, between the automatic RFCS and the manual RFCS cases, comparisons of the transient response for the key system parameter show that the difference of transient response can be clearly identified. Also, the results show that the ΔCPR (delta critical power ratio) for the transients analyzed may not be less limiting for the automatic RFCS case under certain combination of control system settings.

The Plant Feature and Performance of Double MS (Modular Simplified and Medium Small Reactor)

A new concept of a small and medium sized light water reactor, named the double MS: modular simplified and medium small reactor (DMS) was developed. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts, and effective use of proven technology. The decrease in the primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about 2 m compared with 3.7 m in the conventional boiling water reactor (BWR). The short active fuel length reduces the drop in core pressure and overcomes the natural circulation system. By using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to confirm transient performance, the DMS plant performance under transient conditions was evaluated using the TRACG code. TRACG code, which can treat multidimensional hydrodynamic calculations in a RPV, is well suited for evaluating the DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As a result, the maximum change in the minimum critical power ratio of the DMS was 0.14, almost the same as for the current advanced boiling water reactor (ABWRs). In order to improve safety efficiency, developing an emergency core cooling system (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following: (1) eliminating high-pressure injection systems, (2) adopting passive safety-related systems, and (3) optimizing distribution for the systems and power source for the ECCS. In this way, the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of the loss of coolant accident analysis, we confirmed that the core can be covered by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.

Averaging methods of the gap heat transfer coefficients and the loss form coefficients of nuclear reactor cores loaded with different fuel bundles

When performing transient analysis in heterogeneous nuclear reactors loaded with different types of fuel bundles is necessary to model the reactor core by a few representative fuel elements with average properties of a region containing a large number of fuel elements. The properties of these representative fuel bundles are obtained by averaging the thermal–hydraulic properties of the fuel elements contained in each region. In this paper we study the different ways to perform the averaging of the thermal–hydraulic properties that can have an influence on the transient results for licence purposes. Also we study the influence of the different averaging methods on the peak clad temperature (PCT) evolution for a LOCA, and on the critical power ratio (CPR) in the hot channels for a turbine trip transient without bypass credit.

Automatic boiling water reactor loading pattern design using ant colony optimization algorithm

An automatic boiling water reactor (BWR) loading pattern (LP) design methodology was developed using the rank-based ant system (RAS), which is a variant of the ant colony optimization (ACO) algorithm. To reduce design complexity, only the fuel assemblies (FAs) of one eight-core positions were determined using the RAS algorithm, and then the corresponding FAs were loaded into the other parts of the core. Heuristic information was adopted to exclude the selection of the inappropriate FAs which will reduce search space, and thus, the computation time. When the LP was determined, Haling cycle length, beginning of cycle (BOC) shutdown margin (SDM), and Haling end of cycle (EOC) maximum fraction of limit for critical power ratio (MFLCPR) were calculated using SIMULATE-3 code, which were used to evaluate the LP for updating pheromone of RAS. The developed design methodology was demonstrated using FAs of a reference cycle of the BWR6 nuclear power plant. The results show that, the designed LP can be obtained within reasonable computation time, and has a longer cycle length than that of the original design.

Steady state thermal-hydraulic analysis of hydride-fueled grid-supported BWRs

A steady state thermal-hydraulic analysis was performed to estimate the power density attainable with hydride-fueled boiling water reactor (BWR) cores with respect to that of an existing oxide BWR core chosen as reference. The power-limiting constraints taken into account were the minimum critical power ratio (MCPR), core pressure drop, fuel average and centerline temperature, cladding outer temperature, flow-induced vibrations and power/flow ratio. The study consisted of two independent analyses: a whole core analysis and a single bundle analysis. The whole core analysis was performed, with a fixed core volume, for both hydride and oxide fuel over hundreds of combinations of rod diameter–rod pitch, referred to as “geometries”, in the ranges 0.6 ≤ D ≤ 1.6 cm and 1.1 ≤ P/ D ≤ 1.6. For each geometry, the maximum achievable steady state core power was calculated. Preliminary neutronics results derived from a companion neutronic study were then overlaid on the whole-core thermal-hydraulic results to estimate the reduction in maximum achievable power caused by the application of neutronic constraints. The single bundle analysis was performed to compare in greater detail the thermal-hydraulic performance of a limited number of hydride and oxide fuel bundles having D and P values similar to those of the reference oxide bundle, and for which the compliance with neutronic constraints was demonstrated in a companion neutronic study. The study concluded that, if the core pressure drop is not allowed to increase above the reference core value, the power density increase attainable with hydride fuel is estimated to be in the range 0–15%. If the pressure drop is allowed to increase up to a value 50% higher than the reference core value, the power density increase is estimated to be in the range 25–45%. These power density increases, which are defined with respect to the reference oxide core, decrease about 10% if the comparison is made with respect to oxide designs resembling the most recent commercial high-performance oxide cores. The power gain capability of hydride fuel is primarily due to the possibility of: (1) replacing volumes occupied by water rods and water gaps in oxide fuel cores with fuel rods, thus increasing the heat transfer area per core volume, and (2) flattening the bundle pin-by-pin power distribution. The actual achievement of the above-mentioned power density increase is however conditioned to the compliance of hydride-fueled cores to safety requirements related to core behavior during transients, hydrodynamic stability and steam dryer performance, which are fields of study not addressed in this work. A potential 25–45% power density increase justifies however interest for further investigation on this alternative fuel.

Thermal Feasibility Analyses for the 1356MWe High Conversion-Type Innovative Water Reactor for Flexible Fuel Cycle

A new reactor concept of innovative water reactor for flexible fuel cycle (FLWR) is under development at Japan Atomic Energy Agency in cooperation with Japanese reactor suppliers. A design of 1,356 MWe high conversion boiling water reactor-type FLWR core, which has an instantaneous conversion ratio of 1.04, negative void coefficient, high burnup of 65 GWd/t, and 15-month operational cycle length, has been constructed. So far, studies on thermal-hydraulic characteristics have been performed for tight lattice core. Evaluation methods for the critical power and the pressure drop under both the steady and the transient states have been established, and a modified TRAC-BF1 code has been developed for the thermal-hydraulic design of the FLWR. In this paper, the thermal feasibility of the designed 1356MWe FLWR core is analyzed by using the modified TRAC-BF1 code. The analysis is first carried out for the current core design. It is confirmed that no boiling transition (BT) occurs under the steady state. However, the minimum critical power ratio (MCPR) is only about 1.08, and the BT is confirmed occurring under the postulated abnormal transient processes. Therefore, concretizations of the conditions that ensure the thermal feasibility of a natural circulation-type FLWR and a forced circulation-type FLWR are performed. As for the results, for a forced circulation-type FLWR, the operation-limited MCPR (OLMCPR) is 1.32, and the necessary minimum core coolant flow rate is 640 kg/(m2s). For a natural circulation-type FLWR, the OLMCPR is 1.19, and the necessary minimum core coolant flow rate is 560 kg/(m2s).

Automatic Search of the Power Ascension Path for a Boiling Water Reactor Using Genetic Algorithm and Neural Network

A method that includes a genetic algorithm (GA), principal component analysis (PCA), and an artificial neural network (ANN) is adopted in the search for the power ascension path of a boiling water reactor that used to rely solely on an operator’s experiences. The power ascension path is formulated as an optimization problem with thermal limits, e.g., minimum critical power ratio, maximum linear heat generation rate, and maximum average planar linear heat generation rate, and with the stability requirement serving as a constraint. The Simulate-3 code is used to calculate the reactor core status, while the optimization problem is solved through the use of the GA. Since the search domain of the GA is relatively large, the ANN, which models the power ascension path, is developed in order to quickly select the candidate solutions for further Simulate-3 calculations, permitting the algorithm to converge effectively. Meanwhile, PCA is used to reduce the ANN input vector’s dimension, which improves the ANN training efficiency and pattern recognition capability. The results show that this method efficiently obtains the proper power ascension path required for meeting all constraints at different fuel exposure levels.

An Improved Critical Power Correlation for Tight-Lattice Rod Bundles

Developing critical power correlation is indispensable for R&D of Innovative Water Reactor for Flexible Fuel Cycle, which adopts a triangular tight-lattice fuel rod configuration and axially double-humped heated profile. In this research, the critical power correlation for tight-lattice rod bundles is improved by using two sets of different tightness 37-rod bundle data derived at Japan Atomic Energy Agency (JAEA), with an aim to predict critical power in different tightness bundles with high accuracy. Deduction for the correlation forms of critical heat flux—critical quality type for spot dryout mechanism and critical quality—annular flow length type for complete dryout one is given out. To account for the changed transverse power peaking effect when the peaking location changes, power peaking on non-peripheral rods, LPF1 , is introduced in the development of the critical power correlation. The standard deviation of Experimental Critical Power Ratio to JAEA data is 3.24%. To Bettis Atomic Power Laboratory data, this value is 9.2%. The applicable range of the correlation is: rod number 20 or 37, gap width between rods from 1.0 to 2.29 mm, mass velocity from 150 to 1,500 kg/m2s, pressure from 2 to 11 MPa, inlet subcooling from 50 to 1K, and transverse power peaking from 1 to 1.1.

Coupled Bundle-Core Design Using Fuel Rod Optimization for Boiling Water Reactors

An optimization method has been developed to determine the optimal fresh fuel rod configurations, fresh streams, and fresh bundle design placements given a known exposed fuel loading pattern and operational strategy for boiling water reactors. The optimization method is based on a first-order approximation of various core parameters, such as hot excess reactivity and critical power ratio, using fuel rod perturbations to the reference fresh bundle designs. A simulated annealing optimization algorithm is shown to produce fresh bundle designs, consisting of rods selected from a user-defined set of rod types that optimize the core design with respect to its design constraints.The method utilizes a linear superposition method based upon sensitivity coefficients to approximate core parameters. A parallel computing system was implemented to decrease wall clock time for the numerous lattice physics and core simulator calculations. A periodic update of the reference bundle design, without the computational burden of updating the sensitivity coefficients, was introduced and is shown to significantly improve the accuracy of the approximation model. Application of the method demonstrates that improved core designs are achieved when a many-fresh bundle design (i.e., stream) solution is considered as part of the design space. Six-stream (and higher) core designs that increase fuel utilization while simultaneously reducing manufacturing costs through reduction of fuel rod types fabricated, previously unattainable with existing methodologies, are now possible.

ICONE15-10332 DEVELOPMENT OF DESIGN TECHNOLOGY ON THERMAL-HYDRAULIC PERFORMANCE IN TIGHT-LATTICE ROD BUNDLES : II- ROD BOWING EFFECT ON BOILING TRANSITION

A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we will describe the critical power characteristics in a 37-rod tight-lattice bundle with rod-bowing under both steady and transient states. It is observed that no matter it is run under a steady or a transient state, boiling transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle. Steady critical power increases monotonically with the increase of mass velocity, with the decrease of inlet water temperature and with the decrease of exit pressure. These trends are same as those in the base case test without rod-bowing. The steady critical power with rod-bowing is about 10% lower than that without rod-bowing. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The TRAC-BF1 code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time within the accuracy of critical power correlation. Traditional quasi - steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight lattice bundle with rod - bowing.

An Improved Critical Power Correlation for Tight-Lattice Rod Bundles

Developing critical power correlation is indispensable for R&D of Innovative Water Reactor for Flexible Fuel Cycle, which adopts a triangular tight-lattice fuel rod configuration and axially double-humped heated profile. In this research, the critical power correlation for tight-lattice rod bundles is improved by using two sets of different tightness 37-rod bundle data derived at Japan Atomic Energy Agency (JAEA), with an aim to predict critical power in different tightness bundles with high accuracy. Deduction for the correlation forms of critical heat flux—critical quality type for spot dryout mechanism and critical quality—annular flow length type for complete dryout one is given out. To account for the changed transverse power peaking effect when the peaking location changes, power peaking on non-peripheral rods, LPF1 , is introduced in the development of the critical power correlation. The standard deviation of Experimental Critical Power Ratio to JAEA data is 3.24%. To Bettis Atomic Power Laboratory data, this value is 9.2%. The applicable range of the correlation is: rod number 20 or 37, gap width between rods from 1.0 to 2.29 mm, mass velocity from 150 to 1,500 kg/m2s, pressure from 2 to 11 MPa, inlet subcooling from 50 to 1K, and transverse power peaking from 1 to 1.1.

Application of the Post-Boiling-Transition Standard to Licensing Analysis

To ensure fuel integrity, light water reactor cores are designed to avoid the onset of boiling transition (BT) inside the fuel assembly that leads to a deterioration of the heat transfer characteristics and subsequent excessive rise of the fuel-cladding temperature in the anticipated operational occurrences (AOOs). However, some boiling water reactors’ AOO events result in immediate scram or suppression of the reactor power due to an increase in the reactor coolant void fraction. Recent studies show that a short duration of dryout inside the fuel assembly only leads to a small rise in the fuel-cladding temperature and thus does not pose a threat to fuel integrity. Many tests on BT and an improved comprehension of its mechanism have led to the development of a methodology to appropriately assess the fuel-cladding temperature after BT has been reached. The Standards Committee of the Atomic Energy Society of Japan has therefore proposed a cladding temperature criterion after BT. Applying the post-BT standard enables the value of the operating limit minimum critical power ratio (OLMCPR) to be decreased by allowing for a short duration of dryout. We calculated the fuel-cladding temperature and dryout duration in the load rejection condition without a bypass event. The calculated results show that both the fuel-cladding temperature and dryout duration meet the post-BT standard in the case of a small OLMCPR, which is determined by the loss of feedwater heating. This enables a more efficient reactor core to be designed by applying the post-BT standard to licensing analysis. The possibility of applying a post-BT standard is demonstrated from the results of this work.

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO 2 fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (< 60°C). In case of LOCA, decay heat from fuel rods is removed by conduction heat transfer to the reactor water pool. Since the cooling water does not contact a fuel rod directly, core design has much flexibility without considering dry-out limitation and Minimum Critical Power Ratio (MCPR). Additionally an effective use of spent fuel is expected.

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.

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.