Radial Power Profile Research Articles

Application of EPMA Data for the Development of the Code Systems TRANSURANUS and ALEPH

In the present article, electron probe microanalysis data for Pu and Nd is being used for validating the predictions of the radial power profile in a nuclear fuel rod at an ultrahigh burn-up of 95 and 102 MWd/kgHM. As such the validation of both the new Monte Carlo burn-up code ALEPH and the simpler TUBRNP model of the fuel rod performance code TRANSURANUS has been extended. The analysis of the absolute concentrations and individual isotopes also indicates potential improvements in the predictive capabilities of the simple TUBRNP model, based on the one-group cross sections inferred from the neutron transport calculations in the ALEPH code. This is a first important step toward extending the application range of the fuel rod performance code to burn-up values projected in nuclear power rods based on current trends.

Proton Source Efficiency for Heterogeneous Distribution of Actinides in the Core of an Accelerator-Driven System

The distribution of actinides in the core of an accelerator-driven system loaded with plutonium, americium, and curium has been studied in order to optimize the proton source efficiency ψ*. The optimization of ψ* was performed by keeping some important characteristics of the system, e.g., the radial power profile and the reactivity of the core, constant. One of the basic assumptions of the study, that the magnitude of ψ* is sensitive primarily to the composition of actinides in the inner part of the core, whereas only marginally to that in the outer part, has been confirmed. It has been shown that the odd-N nuclides (those nuclides with an even number of neutrons) in general and 241Am and 244Cm in particular have favorable properties with respect to improving ψ* if they are placed in the innermost part of the core. The underlying reason for this phenomenon is that the energy spectrum of the source neutrons in the inner part of the core is harder than that of the average fission neutrons. Moreover, it has been shown that loading the inner part of the core with only curium increases ψ* by ˜7%. Plutonium, on the other hand, in particular high-quality plutonium consisting mainly of 239Pu and 241Pu, was found to be a comparatively source inefficient element and is preferably located in the outer part of the core. The differences in ψ* are due to combined effects from relative changes in the average fission and capture cross sections and in the average fission neutron yield.

06/01617 CRONOS2 and APOLL02 results for the NEA C5G7 MOX benchmark: Moreau, F. et al. Progress in Nuclear Energy, 2004, 45, (2–4), 179–200

BGCore reactor analysis system was recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge. It couples the Monte Carlo transport code MCNP with an independently developed burnup and decay module SARAF. Most of the existing MCNP based depletion codes (e.g. MOCUP, Monteburns, MCODE) tally directly the one-group fluxes and reaction rates in order to prepare one-group cross sections necessary for the fuel depletion analysis. BGCore, on the other hand, uses a multi-group (MG) approach for generation of one group cross-sections. This coupling approach significantly reduces the code execution time without compromising the accuracy of the results.Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (TH) feedback into the calculation scheme. Recently, a simplified TH feedback module, THERMO, was developed and integrated into the BGCore system. To demonstrate the capabilities of the upgraded BGCore system, a coupled neutronic TH analysis of a full PWR core was performed. The BGCore results were compared with those of the state of the art 3D deterministic nodal diffusion code DYN3D (Grundmann et al., 2000). Very good agreement in major core operational parameters including k-eff eigenvalue, axial and radial power profiles, and temperature distributions between the BGCore and DYN3D results was observed. This agreement confirms the consistency of the implementation of the TH feedback module.Although the upgraded BGCore system is capable of performing both, depletion and TH analyses, the calculations in this study were performed for the beginning of cycle state with pre-generated fuel compositions.

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Simulation of the neutronic-thermal hydraulic stability behaviour of boiling water reactors

AbstractNeutronic-thermal hydraulic oscillations are a well known phenomenon in the operation of boiling water reactors. Despite the progresses that have been made concerning their simulation, the computational determination of the stability threshold in the power-flow map still suffers from inaccuracies. In this work, the decay ratio of neutronic-thermal hydraulic oscillations and hence the stability threshold was calculated with the SIMULATE-3K code developed by Studsvik Scandpower. It is known from experience that there is a higher probability for a less stable reactor operation in the case of one or more pumps tripped while the remaining pumps run at minimum speed. For the analysis of this state, the characteristics of the pump system were modelled carefully. It is shown that the stability behaviour is determined by the axial and radial power profile due to control rod manoeuvring and fuel burn-up and by the number of pumps running. Furthermore the calculations reveal the impacts of the insertion of shallow rods and of the control-rod group insertion on the decay ratio. A comparison of the calculations with experimental results of stability measurements shows that by a thorough analysis and proper modelling of the plant characteristics the decay ratios can be calculated with the SIMULATE-3K code.

04/01290 Iterative computation of time-eigenvalues of the neutron transport equation: Lathouwers, D. Annals of Nuclear Energy, 2003, 30, (17), 1793–1806

BGCore reactor analysis system was recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge. It couples the Monte Carlo transport code MCNP with an independently developed burnup and decay module SARAF. Most of the existing MCNP based depletion codes (e.g. MOCUP, Monteburns, MCODE) tally directly the one-group fluxes and reaction rates in order to prepare one-group cross sections necessary for the fuel depletion analysis. BGCore, on the other hand, uses a multi-group (MG) approach for generation of one group cross-sections. This coupling approach significantly reduces the code execution time without compromising the accuracy of the results.Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (TH) feedback into the calculation scheme. Recently, a simplified TH feedback module, THERMO, was developed and integrated into the BGCore system. To demonstrate the capabilities of the upgraded BGCore system, a coupled neutronic TH analysis of a full PWR core was performed. The BGCore results were compared with those of the state of the art 3D deterministic nodal diffusion code DYN3D (Grundmann et al., 2000). Very good agreement in major core operational parameters including k-eff eigenvalue, axial and radial power profiles, and temperature distributions between the BGCore and DYN3D results was observed. This agreement confirms the consistency of the implementation of the TH feedback module.Although the upgraded BGCore system is capable of performing both, depletion and TH analyses, the calculations in this study were performed for the beginning of cycle state with pre-generated fuel compositions.

Subchannel void-fraction measurements in a 6 × 6 rod bundle using a simple gamma-transmission method

A relatively simple subchannel void-fraction measurement system was developed and applied to a scaled model of a BWR fuel assembly. The measurement system is based upon the gamma-transmission technique and uses a tomographic reconstruction procedure to calculate the void-fractions in the subchannels. Data pre-processing is used to remove positioning errors of the gamma source and detector. Measurements are performed under different operating conditions and radial power profiles. The measurements are compared with the drift-flux two-phase flow model. The void-fraction in the wall and corner subchannels is lower than average. By applying the drift-flux model to each subchannel separately and comparing this calculation with measurements the presence of lateral void-drift should be detectable. However, the data do not show evidence for lateral void-drift. By comparing two chordal void-fractions, one through the fuel rods and one in between the fuel rods, a flow pattern transition could clearly be seen between a void-fraction of 30% and 40%.

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power histories and specifications of the fuel rods DH and DK of IFA-519.9 irradiated in Halden reactor were input, and calculated rod internal pressures were compared with experimental data for the range of ∼25–93 MWd kg −1 UO 2, and factors affecting pellet temperature were discussed. Also some sensitivity studies were conducted with respect to the effect of swelling rate and grain growth. As a result, it is found that the prediction is sensitive to the models of thermal conductivity and swelling rate of fuel, and FEMAXI–V analytical system proved to give a reasonable prediction even in the high burnup region.

Depletion Analysis of Mixed-Oxide Fuel Pins in Light Water Reactors and the Advanced Test Reactor

An experiment containing weapons-grade mixed-oxide (WG-MOX) fuel has been designed and is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The ability to accurately predict fuel pin performance is an essential requirement for the MOX fuel test assembly design. Detailed radial fission power and temperature profile effects and fission gas release in the fuel pin are a function of the fuel pin’s temperature, fission power, and fission product and actinide concentration profiles. In addition, the burnup-dependent profile analyses in irradiated fuel pins is important for fuel performance analysis to support the potential licensing of the MOX fuel made from WG-plutonium and depleted uranium for use in U.S. reactors.The MCNP Coupling With ORIGEN2 burnup calculation code (MCWO) can analyze the detailed burnup profiles of WG-MOX and reactor-grade mixed-oxide (RG-MOX) fuel pins. The validated code MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. Applying this capability with a new minicell method allows calculation of detailed nuclide concentration and power distributions within the MOX pins as a function of burnup. This methodology was applied to MOX fuel in a commercial pressurized water reactor and in an experiment currently being irradiated in the ATR. The prediction of nuclide concentration profiles and power distributions in irradiated MOX pins via this new methodology can provide insights into MOX fuel performance.

Fast detection of 14 MeV neutrons on the TFTR neutron collimator

Current mode operation of the NE451 ZnS scintillation detectors of the TFTR neutron collimator has enabled us to record the development of radial neutron emission profiles with much faster speed and higher accuracy than in the pulse counting mode. During high power deuterium–tritium (DT) operation, the intrinsic shot noise on the detector traces was so low that we could observe sawtooth instabilities and disruptions with good precision and, in addition, were able to identify precursor magnetohydrodynamic (MHD) activity and fishbone instabilities. These results demonstrate that in future tritium burning machines like ITER or TPX, the neutron collimator should be designed not only as a monitor of radial fusion power profiles but also as a wave detector for MHD activity.

Theoretical and practical factors influencingthe optical performance of contact lenses for the presbyope

The theoretical limitations on potential visual performance set by the optical design of different types of contact lensfor presbyopes are considered. Particular attention is paid to the influence of pupil diameter and, in simultaneous varifocal designs, to the form of the radial variation in sagittal power. The adverse effects of any unwanted transition zone on the lens surface between regions of differing optical power are considered: interferometric measurements are presented which show the presence of such zones in some current lenses. Measurements of radial power profiles for several designs of varifocal lens show that not all manufacturers are successful in producing a powerprofile which is invariant with the nominal distance prescription of the lens. The failure of measurements of high-contrast Snellen acuity to indicate the losses in image contrast that are inherent in the optical design of most types of lens is discussed.

Mixed Convective Low Flow Pressure Drop in Vertical Rod Assemblies: I—Predictive Model and Design Correlation

A predictive theory has been developed for rod bundle frictional pressure drop characteristics under laminar and transitional mixed convection conditions on the basis of the intraassembly and intrasubchannel flow redistributions due to buoyancy for a wide spectrum of radial power profiles and for the geometric arrangements of practical design interest. Both the individual subchannel correlations and overall bundle design correlations have been formulated as multipliers applied to the isothermal friction factors at the same Reynolds numbers. Standard and modified subchannel friction factors have been obtained to be used with spatial-average and bulk-mean densities, respectively. A correlating procedure has been proposed to assess the effects of interacting subchannel flows, developing mixed convective flow, wire wrapping, power skew, rod number, and transition from laminar flow. In contrast to forced convection behavior, a strong rod number effect is present under mixed convection conditions in bundle geometries. The results of this study are of design importance in natural circulation conditions because the mixed convection frictional pressure losses exceed the corresponding isothermal values at the same Reynolds numbers.

Summary of CCTF test results - assessment of current safety evaluation analysis on reflood behavior during a LOCA in a PWR with cold-leg-injection-type ECCS

The conservatism of the current safety analysis was assessed by comparing the predicted results with cylindrical core test facility (CCTF) test results performed at Japan Atomic Energy Research Institute (JAERI). The WREM code was selected for the assessment. The overall conservatism of the WREM code on the peak clad temperature prediction was confirmed against CCTF EM test which simulated the typical initial and boundary conditions in the safety evaluation analysis. The WREM code predicted the reasonable core boundary conditions and the conservatism of the code came mainly from the core calculation. The conservatism of the WREM code against CCTF data could be attributed to the following three points: 1. (i) no horizontal mixing assumption between subchannels at each elevation, 2. (ii) no modeling on heat transfer enhancement caused by the radial core power profile, 3. (iii) usage of conservative heat transfer correlations in the code.

Assessment of Core Radial Power Profile Effect Model for REFLA Code by Using CCTF Data

A simple model was developed to evaluate the effect of the radial power profile on the thermo-hydraulics in the core during reflood phase of loss-of-coolant accident in PWRs. It was incorporated into REFLA code, a reflood analysis code. The model assumes that the fluid mixes completely among subchannels and additionally that the fuel temperature is not affected each other among fuels. This model is featuring no three-dimensional treatment for the evaluation on effect of the radial power profile. The model improved the predictability of clad temperature and gave a better prediction than a method which is generally used in the safety analysis. The model predicted a poorer core cooling in high radial power region than data. The error of the prediction increased with the increase in radial power ratio. These indicates the conservativeness of the model on the maximum clad temperature prediction. The reason of the conservativeness is due to the neglection of the heat transfer enhancement by two-dimensional effe...

Criteria for the Onset of Flow Recirculation and Onset of Mixed Convection in Vertical Rod Bundles

In a vertical rod bundle with forced upflow and a nonuniform radial power profile, the potential for flow reversal into the low power side of the bundle exists due to buoyancy forces. In this paper a criterion is developed to predict the onset of flow recirculation in rod bundles. The criterion is validated with experimental data for flow recirculation in water-cooled bare rod bundles and with the results of a standard multichannel rod bundle code. A criterion for the onset of mixed convection is also given. Thus, the boundaries between the forced convection, mixed convection, and recirculation regimes are identified. Results are expressed on a flow regime map in terms of the dominant dimensionless parameters (the Reynolds number and a buoyancy parameter B*) and as a closed-form equation.

Assessment of core radial power profile effect model for REFLA code by using CCTF data.

A simple model was developed to evaluate the effect of the radial power profile on the thermo-hydraulics in the core during reflood phase of loss-of-coolant accident in PWRs. It was incorporated into REFLA code, a reflood analysis code. The model assumes that the fluid mixes completely among subchannels and additionally that the fuel temperature is not affected each other among fuels. This model is featuring no three-dimensional treatment for the evaluation on effect of the radial power profile. The model improved the predictability of clad temperature and gave a better prediction than a method which is generally used in the safety analysis. The model predicted a poorer core cooling in high radial power region than data. The error of the prediction increased with the increase in radial power ratio. These indicates the conservativeness of the model on the maximum clad temperature prediction. The reason of the conservativeness is due to the neglection of the heat transfer enhancement by two-dimensional effect in high power region.

Assessment of current safety evaluation analysis on reflood behavior during PWR-LOCA by using CCTF data.

This paper describes an assessment result on conservatism of current safety analysis concerning reflood behavior during a LOCA In a PWR by using the experimental data with cylindrical core test facility (CCTF) performed at Japan Atomic Energy Research Institute (JAERI). WREM code is selected for a representative of current safety analyses. The predicted peak clad temperature with the WREM code was higher than the data, and it was confirmed that the WREM code had the overall conservatism against CCTF data. The WREM code predicted the reasonable core boundary conditions and it was found that the conservatism of the code came mainly from the calculations on the in- core thermal hydraulics and clad temperature. In addition, it was found that the conservatism of the WREM code against the CCTF data could be attributed to the neglection of horizontal fluid mixing between subchannels, the neglection of the heat transfer enhancement due to the radial core power profile, and the usage of the heat transfer correlati...

Experimental Study of Two-Dimensional Thermal-Hydraulic Behavior in Core during Reflood Phase of PWR LOCA

Two-dimensional effects on the core cooling behavior during the reflood phase of a PWR-LOCA were experimentally studied by performing four tests with various radial core power profiles under the same total power and initial core stored energy conditions using the Slab Core Test Facility (SCTF). The heat transfer was enhanced and the cladding temperature was reduced for the higher and average power bundles in the steep radial power profile test especially at the upper elevation. The effect of radial power profile on the cladding temperature was quantitatively evaluated. For all tests with different radial power profiles, the collapsed water level in the upper plenum became higher in the hot leg side and the quench in the upper half of the core was delayed in the bundles corresponding to the outer bundles of a PWR core. The delay of the quench is considered to be caused by a flow stagnation trend in those bundles because the pressure in the outer bundles became higher than the pressure in the inner bundles ...

Experimental study of two-dimensional thermal-hydraulic behavior in core during reflood phase of PWR LOCA.

Two-dimensional effects on the core cooling behavior during the reflood phase of a PWR-LOCA were experimentally studied by performing four tests with various radial core power profiles under the same total power and initial core stored energy conditions using the Slab Core Test Facility (SCTF). The heat transfer was enhanced and the cladding temperature was reduced for the higher and average power bundles in the steep radial power profile test especially at the upper elevation. The effect of radial power profile on the cladding temperature was quantitatively evaluated. For all tests with different radial power profiles, the collapsed water level in the upper plenum became higher in the hot leg side and the quench in the upper half of the core was delayed in the bundles corresponding to the outer bundles of a PWR core. The delay of the quench is considered to be caused by a flow stagnation trend in those bundles because the pressure in the outer bundles became higher than the pressure in the inner bundles due to the nonuniform water accumulation in the upper plenum.

Core Radial Power Profile Effect on System and Core Cooling Behavior during Reflood Phase of PWR-LOCA with CCTF Data

Core Radial Power Profile Effect on System and Core Cooling Behavior during Reflood Phase of PWR-LOCA with CCTF Data