BWR 90-the advanced BWR of the 1990s

The future global role of nuclear power will be determined by its ability to provide economical and safe energy. Nuclear power, like any other substantial contributor to the world's energy needs, must be generated at an acceptable cost and with negligible environmental effects. Besides, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power. The ABB Atom approach to this situation can best be characterized as a ‘cautious evolution'; for the next decade the company will largely base its offerings to the market on its ‘evolutionary' light water reactor design, the BWR 90. This design builds closely on the experience from successful construction and operation of its predecessor, the BWR 75 design. In 1995 and 1996, plants of this design achieved an average load factor greater than the 87% set by EUR; the two BWR units at Olkiluoto in Finland are among the very best performing plants in the world, with an average load factor of 94% over the last 7 years. The continued LWR design development focuses on meeting requirements from utilities as well as new regulatory requirements. A particular emphasis is put on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimization of buildings and containment design to decrease construction time and costs, and selection of materials as well as maintenance and operating procedures to even further reduce occupational radiation exposures. Probabilistic safety assessments and life-cycle cost evaluations have become major tools in the design optimization work. The BWR 90 was offered to Finland in the early 1990s, and will now as the first BWR design be reviewed by a number of European utilities with respect to its conformance to the European Utility Requirements (EUR); a specific EUR Volume 3 for the BWR 90 will be the final result. The paper describes some of the unique characteristics of the BWR 90, with emphasis on the features that are most important for achieving improved economy and enhanced safety.     

Simulation of hydrogen deflagration and detonation in a BWR reactor building

A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm² leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis.     

Description of void effects in a large BWR by means of a stochastic theory and simulation in a zero-power reactor

We first summarize the stochastic point model developed in previous papers to describe void effects in a large BWR and also summarize our results. The most important of these is the existence of a resonance frequency in the auto-power spectral density of the neutron noise (APSD), the position of which depends on the reactor characteristics, such as power or void coefficient.

In order to check the validity of this model, we made experiments simulating heat transfer and steam fluctuations in a BWR. A stochastic interpretation of the experiments is developed, and results are found to be similar to those obtained with the BWR model. In particular, the zero-power reactor with the simulation device exhibits a resonance frequency showing an identical behaviour to the one predicted for a BWR.

The APSD resulting from experiments in the zero-power reactor CROCUS, at a given power level, is fitted on the theoretical curves by means of the least square method, which provides the resonance frequency. The behaviour of this frequency as a function of the power level agrees fairly well with the theoretical prediction.

If we suppose that feedback mechanisms are the same in a large BWR, we can also admit that the stochastic model gives correctly the resonance frequency.     

A systematic method to identify nonlinear dynamics of BWR by using the reactor noise

For the identification of the dynamics of the Vermont Yankee BWR with the reactor noise, different parametric models have been tested. The widely used ARMA model is unable to identify the nonlinearity in the noise data. A systematic method by using the NARMA model, which takes advantage of both the ANN and ARMA, is developed. Comparisons are made between the identification results with ARMA and NARMA model. The advantages of identification with NARMA model over ARMA model are demonstrated. The linear-kernels of the identified NARMA models are extracted so that the natural frequency, damping ratio and time constants of the BWR are obtained. The values of those characteristics are well corresponded with the eigenvalues calculated by the differential equations of the Vermont Yankee BWR. The damping ratio with negative value is found to be a criterion for the existence of limit-cycle, which can be seen from the impulse response on the (X_t, X_t−1) plane, in stable nonlinear system.     

Analytical modeling of flashing-induced instabilities in a natural circulation cooled boiling water reactor

A dynamic model for natural circulation boiling water reactors (BWRs) under low-pressure conditions is developed. The motivation for this theoretical research is the concern about the stability of natural circulation BWRs during the low-pressure reactor start-up phase. There is experimental and theoretical evidence for the occurrence of void flashing in the unheated riser under these conditions. This flashing effect is included in the differential (homogeneous equilibrium) equations for two-phase flow. The differential equations were integrated over axial two-phase nodes, to derive a nodal time-domain model. The dynamic behavior of the interface between the one and two-phase regions is approximated with a linearized model. All model equations are presented in a dimensionless form. As an example the stability characteristics of the Dutch Dodewaard reactor at low pressure are determined.     

Uncertainty and sensitivity analyses as a validation tool for BWR bundle thermal-hydraulic predictions

In recent years, more realistic safety analyses of nuclear reactors have been based on best estimate (BE) computer codes. The need to validate and refine BE codes that are used in the predictions of relevant reactor safety parameters, led to the organization of international benchmarks based on high quality experimental data. The OECD/NRC BWR full-size fine-mesh bundle test (BFBT) benchmark offers a good opportunity to assess the accuracy of thermal-hydraulic codes in predicting, among other parameters, single and two phase bundle pressure drop, cross-sectional averaged void fraction distributions and critical powers under a wide range of system conditions. The BFBT is based on a multi-rod assembly integral test facility which is able to simulate the high pressure, high temperature fluid conditions found in BWRs through electrically heated rod bundles. Since code accuracy is unavoidably affected by models and experimental uncertainties, an uncertainty analysis is fundamental in order to have a complete validation study. In this paper, statistical uncertainty and sensitivity analyses are used to validate the thermal-hydraulic features of the POLCA-T code, based on a one dimensional model of the following macroscopic BFBT exercises: (1) single and two phase bundle pressure drop, (2) steady-state cross-sectional averaged void fraction, (3) transient cross-sectional averaged void fraction and (4) steady-state critical power tests. The Latin hypercube sampling (LHS) strategy was chosen since it densely stratifies across the range of each uncertain input probability distribution, allowing a much better coverage of the input uncertainties than simple random sampling (SRS). The results show that POLCA-T predictions on pressure drop and void fractions under a wide range of conditions are within the validation limits imposed by the uncertainty analysis, while the accuracy of critical power predictions depends much on the boundary and input conditions.     

Determination of BWR minimum steam cooling reactor water level and associated steam flow using MAAP4 code

The minimum steam cooling reactor pressure vessel (RPV) water level (MSCRWL) is defined to be the lowest RPV water level at which the covered portion of the core is capable of generating sufficient steam to preclude peak cladding temperature (PCT) in the uncovered portion of the core from exceeding 1500 °F. The associated steam flow rate is called Wg-1500. Both MSCRWL and Wg-1500 are important parameters for safe operation of nuclear power plant. In the past, the calculations of MSCRWL and Wg-1500 were calculated by indirect way via simple model and provided by the vendor. It has to be revised quite often during the operation period of nuclear power plant. The process is time-consuming. To improve the situation, a direct and easy method of generating MSCRWL using MAAP4 code by the utilities in conjunction with the proportional and integral (PI) controller is developed in this study. The developed control loop with a PI controller is capable of generating the MSCRWL in a fast and precise manner. The MSCRWL and Wg-1500 are calculated simultaneously by controlling the PCT equal to 1500 °F. Furthermore, the adjusting process is done automatically and readily with this methodology. The effect of feedwater inlet temperature is taken into account via thermal kits of the balance of plant system. The calculated MSCRWL is consistent with the calculated Wg-1500. The sensitivity study of power level and power shape are performed. For a given power shape, the MSCRWL is less sensitive to the power level. However, Wg-1500 is almost proportional to the power level. This information is helpful for the associated EOP application. This technique can be applied for other system codes.     

A simplified model for the dynamics of a BWR

A model has been developed to derive the dynamic characteristics of a BWR with natural circulation. The model is based on the basic physical processes that govern reactor dynamics. The actual values for the model parameters are estimated from experimental and theoretical data. The model enables the computation of transfer functions of reactivity and steam flow to power and pressure. The sensitivity of these transfer functions to changes in model parameters is discussed.     

Application of continuous-energy Monte Carlo code as a cross-section generator of BWR core calculations

A continuous-energy Monte Carlo code is newly applied for the assembly calculations of actual BWR core analysis. Few-groups cross-sections and related constants (kinetic parameters) were generated by the continuous-energy Monte Carlo code MVP-BURN, and were tabulated for a core simulator. The commercial BWR, HAMAOKA-3 (1100MWe:BWR-5), was analyzed by a coupled neutronic-thermalhydraulic core simulator based on modified one-group diffusion theory using these assembly constants. The calculated core parameters showed good agreement with the results of the on-line core monitoring system of HAMAOKA-3. Consequently, it was confirmed that the present method is applicable to BWR core production calculations. The present method is a particularly attractive candidate for the analysis of advanced BWR fuel assemblies with exotic geometry and high Gd content, due to the features of the continuous-energy Monte Carlo code, i.e., high accuracy and generalized geometry treatment.     

Noise analysis of a zero power critical line reactor using Langevin's technique

Noise analysis of a critical, infinite-length, zero-power line reactor is given using the Langevin technique. A one-speed model is used and delayed neutrons are ignored. Stochastic analysis is carried out starting from the Boltzman equation with the assumption that neutrons move only in both directions in the line reactor. The power spectrum of neutron fluctuations is obtained and compared with that computed from diffusion approximation. Exact expressions for auto and cross-power spectrums of non-fission neutron detectors' outputs are also obtained.     

Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1.The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are also presented.     

An assessment of the BWR fuel box-lattice-cell homogenization procedures with the Tarapur BWR incore and excore measurements

A code system has been developed to provide the incorefuel-management guidelines to the Tarapur BWR reactors. Constant checking of the design calculational methods is rendered possible by the steady flow of operating data from the Tarapur units over the last few cycles. The operating data include cold/hot criticals and detailed flux/power maps. Besides these, the burnups and isotopic composition of a few irradiated fuel pins obtained by mass-spectrometric analyses, are also available for validation of the BWR core and lattice-cell modelling.The calculated eigen values for different power levels and at different core average burnups are found to have a spread of less than 0.25% ΔK. Analyses of a number of TIP measurements show that the core power distribution can be predicted in a satisfactory manner for uncontrolled fuel bundles and non-peripheral fuel assemblies (<10%). For prediction of cold-criticals the void-history effects are found to be unimportant.The pin burnups and isotopic densities of important U and Pu isotopes relative to ²³⁸U have been compared with mass-spectrometric measurements. The pin-burnup profile comparison is found to be good for fuel pins, which are not near water gaps. Deviation histograms of various isotopes are presented in this paper. ²³⁵U is predicted within ± 3% (r.m.s.). The total Pu is overpredicted by 5–8%, while the quality of Pu is predicted within ± 1.0% (r.m.s.).     

A novel encoding water level monitor system during and after LOCAs in a nuclear heating reactor

The water level in a nuclear reactor vessel is an important parameter during and after LOCAs. Nuclear safety specifications can not be carried out when the water level is measured using a pressurizer which does show the level in the vessel. It is difficult to monitor the water level in the vessel of a Daqing 200MW Nuclear Heating Reactor (NHR-200) using the present differential pressure transducers. Based on the heat transfer differences between water (or liquid) and steam (or gas), a novel level detector, which includes encoding heating shell thermocouples, has been developed and verified for use experimentally under pressures of 0.15–3.0 MPa. A novel encoding water level monitoring system was designed, made up of an assembly that contains several detectors, a signal encoder and an intelligent processor. Analysis and experiments have shown that the new system is correct in principle, reliable and feasible in structure for monitoring the water level in the NHR-200 reactor.     

Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

Tne analytical/experimental method has been developed to to monitor the subcritical reactivity and unfold the k_∞ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a ²⁵²Cfneutron source placed sequentially in multiple positions in the core. Therefore, it is called the asymmetric multiple-position neutron source (AMPNS) method. The AMPNS method employs the nucleonic codes to analyze in two dimensions the neutron multiplication of a ²⁵²Cf neutron source. An optimization program, GPM, was utilized to unfold the k_∞ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn. State Breazeale TRIGA reactor (PSBR). A significant result of this study has been to provide a means to plan the source and detector placements and assign core cells to the damaged TMI-2 core as well as to monitor the criticality during the recovery period.     

Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSTF), which simulated a full scale 30° sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved.     

Concerning the transitional system of a reactor as a result of recharging fuel elements

The relationship is discussed between the duration of the transitional state of a reactor (from the point of view of reactivity as well as recharging), the specified irradiation stability of the fuel elements, attainable fuel irradiation and the required rate of recharging of the fuel elements for different cycles, In particular, the advantages and disadvantages are compared of equilibrium transition, transition with delayed recharging at a constant rate, transition with constant reactivity and various combinations of these transitions. The conditions are discussed which (in comparison with equilibrium irradiation) permit saving of excess reactivity because of nonoverirradiation of the fuel elements.Translated from Atomnaya Énergiya, Vol. 16, No. 4, pp. 315–324, April, 1964     

Investigation of BeO as a reflector for the low power research reactor

Calculations of the fuel burnup, core excess reactivity, and the reactivity worths of the top beryllium shim plates for two reflector types (beryllium and beryllium oxide (BeO)) in the Miniature Neutron Source Reactor (MNSR) have been presented in this paper using the GETERA and MCNP4C codes. The results showed that the reactor infinity multiplication factors were 1.7030 and 1.6824, the core unadjusted excess reactivities were 31.9 and 5.0 mk, and the reactivity worths of the top beryllium shim plates were 22 and 19 mk for the BeO and Be reflectors respectively. Finally, using the beryllium oxide instead of the existing Be reflector in the MNSR reactor increased the core excess reactivity and reactor operation time.     

Neutron energy spectrum flux profile of Ghana’s miniature neutron source reactor core

The total neutron flux spectrum of the compact core of Ghana’s miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) × 10¹² n/cm² s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) × 10¹¹ n/cm² s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) × 10¹¹ n/cm² s. The peak values of the thermal energy range occurred in the energy range (1.8939–3.7880) × 10⁻⁰⁸ MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) × 10⁰⁹ n/cm² s at the lower energy end of the slowing down region between 8.2491 × 10⁻⁰¹ MeV and 8.2680 × 10⁻⁰¹ MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) × 10⁰⁸ n/cm² s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.     

Experimental studies of core flow fluctuations and neutron noise in a BWR

Using the instrumented fuel assemblies (IFA) installed in the Japan Power Demonstration Reactor (JPDR)-II core, fluctuations of the inlet and outlet channel flow rates were observed under both conditions of at-power operation and cold core flow circulation. The correlation analysis revealed that the flow fluctuations in any IFA channel showed almost uncorrelated cross-covariance function with other IFA channel flow. To explain the mechanism of the channel flow fluctuations, some hypothetical idea is introduced.