Control Rod Worth Research Articles

Carbon—carbon composite as anodes for lithiumion battery systems

A detailed three-dimensional, continuous-energy MCNP4B model of the LWR-PROTEUS critical facility has been developed for the analysis of whole-reactor characteristics using ENDF/B-V, ENDF/B-VI and JEF-2.2 cross-section sets. The model has been applied to the determination of the critical loading, as well as the evaluation of reactivity worths for safety/shutdown rods, control rods, and individual driver-region fuel rods. The initially obtained results for the first configuration investigated (Core 1B) indicated that, for the same geometrical and materials specifications, the ENDF/B-V data library yields the closest critical prediction (discrepancy of 640±40 pcm), followed by ENDF/B-VI (980±40 pcm) and JEF-2.2 (1340±40 pcm). The differences in results between the three data libraries were studied by considering the contributions of individual materials to the neutron balance. 235U and 238U cross-sections from JEF-2.2, for example, explain an effect of ∼400 pcm. Refinement of the materials specifications in the MCNP4B whole-reactor model, in particular the impurities assumed for the graphite driver of the driver and reflector regions, has been shown to reduce the final discrepancy of the ENDF/B-V based keff result to ∼0.2%. The MCNP4B results for relative reactivity effects, such as control rod worths, are found to agree within experimental errors with the measured values.

High Power Transient Characteristics and Capability of NSRR

In order to study fuel behavior under abnormal transients and accidents, the control system of the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute (JAERI) was modified to achieve high power transients. With this new operational mode, called Shaped Pulse (SP), transients at the maximum power of 10 MW can be conducted for a few seconds. This new operational mode supplements the previous Natural Pulse (NP) operation at the maximum power of 23 GW for milliseconds. For high power transient operation, a simulator using a point kinetic model was developed, and characteristics of the NSRR in the new operational mode were examined through tests and calculations. With the new operational mode, new types of fuel irradiation tests simulating power oscillations of boiling water reactors (BWRs) can be conducted in the NSRR. Reactor characteristics and capability, such as control rod worth, feedback reactivity, and operational limits of the NSRR for SP operations are discussed.

Developing and modeling of the “Laguna Verde” BWR CRDA benchmark

Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant — unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The “Laguna Verde” (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTRÉE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions.

A neural network controller for load following operation of nuclear reactors

Nuclear reactors are in nature nonlinear and their parameters vary with time as a function of power level, fuel burnup, and control rod worth. Therefore, these characteristics must be considered if large power variations occur in power plant working regimes (for example in load following conditions). In this paper a neural network controller (NNC) is presented. A robust optimal self-tuning regulator (ROSTR) response is used as a reference trajectory to determine the feedback, feedforward and observer gains of the NNC. The NNC displayed good stability and performance for a wide range of operation as well as considerable reduction in computation time in regard to ROSTR and fuzzy logic controller (FAROC).

Parametric tests and measurements after shimming of a beryllium reflector in a miniature neutron source reactor (MNSR)

After 10 years operation of Pakistan research reactor-2 (PARR-2), a miniature neutron source reactor (MNSR), a beryllium reflector was added to compensate the loss of reactivity due to burn up of fuel. Beryllium shim plates have been placed at the top of the core in a tray provided for this purpose. The control rod was dismantled and withdrawn from the core and the reactor was made subcritical with cadmium shimming. To monitor the neutron population during this experiment, two additional neutron monitoring channels based on BF 3 were installed around the core. Measurement of important Parameters such as effective delayed neutron fraction, decay constant, excess reactivity, control rod worth, temperature coefficient of reactivity, thermal neutron flux, cadmium ratio was done after the addition of Be reflector. Increase in reactivity worth due to addition of Be shim was 1.0 mk.

Validation of an MCNP4B whole-reactor model for LWR-PROTEUS using ENDF/B-V, ENDF/B-VI and JEF-2.2 cross-section libraries

A detailed three-dimensional, continuous-energy MCNP4B model of the LWR-PROTEUS critical facility has been developed for the analysis of whole-reactor characteristics using ENDF/B-V, ENDF/B-VI and JEF-2.2 cross-section sets. The model has been applied to the determination of the critical loading, as well as the evaluation of reactivity worths for safety/shutdown rods, control rods, and individual driver-region fuel rods. The initially obtained results for the first configuration investigated (Core 1B) indicated that, for the same geometrical and materials specifications, the ENDF/B-V data library yields the closest critical prediction (discrepancy of 640±40 pcm), followed by ENDF/B-VI (980±40 pcm) and JEF-2.2 (1340±40 pcm). The differences in results between the three data libraries were studied by considering the contributions of individual materials to the neutron balance. 235U and 238U cross-sections from JEF-2.2, for example, explain an effect of ∼400 pcm. Refinement of the materials specifications in the MCNP4B whole-reactor model, in particular the impurities assumed for the graphite driver of the driver and reflector regions, has been shown to reduce the final discrepancy of the ENDF/B-V based k eff result to ∼0.2%. The MCNP4B results for relative reactivity effects, such as control rod worths, are found to agree within experimental errors with the measured values.

Rapid calibaration of the control rod of miniature neutron source reactor types

An innovative method of combining the well-known computer codes WIMS and CITATION in an interactive Windows environment to calculate the control rod worth and generate the calibration data/curve of single-contr-ol rod miniature source reactor (MNSR) types has been developed and tested. In the scheme, once CITATION base data have been prepared using group constants generated by WIMS, the required calculations are perform-ed automatically. The methodology eliminates errors and cumbersome procedures involved in manual regenera-tion of data after each CITATION pass associated with the calculations, thus facilitating the reactor design proc-ess, such that an important aspect of reactor design calculation that used to take days or weeks could now be more reliably done within the shortest possible time. (Journal of Applied Science and Technology: 2001 6(1-2): 7-12)

Dynamic Rod Worth Measurement

The dynamic rod worth measurement (DRWM) technique is a method of quickly validating the predicted bank worth of control rods and shutdown rods. The DRWM analytic method is based on three-dimensional, space-time kinetic simulations of the rapid rod movements. Its measurement data is processed with an advanced digital reactivity computer. DRWM has been used as the method of bank worth validation at numerous plant startups with excellent results. The process and methodology of DRWM are described, and the measurement results of using DRWM are presented.

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.

A Pressurized Water Reactor Plutonium Incinerator Based on Thorium Fuel and Seed-Blanket Assembly Geometry

A pressurized water reactor (PWR) fuel cycle is proposed, whose purpose is the elimination and degradation of weapons-grade plutonium. This Radkowsky thorium-fuel Pu incinerator (RTPI) cycle is based on a core and assemblies retrofittable to a Westinghouse-type PWR. The RTPI assembly, however, is a seed-blanket unit. The seed is supercritical, loaded with Pu-Zr alloy as fuel in a high moderator-to-fuel ratio configuration. The blanket is subcritical, loaded mainly with ThO2, generating and burning 233U in situ. Blankets are loaded once every 6 yr. The seed fuel management scheme is based on three batches, with one-third of the seed modules replaced every year. The core generates 1100 MW(electric). Equilibrium conditions are achieved with the second seed loading. For equilibrium conditions, the annual average of disposed (loaded) Pu is 1210 kg, of which 702 kg are completely eliminated, and 508 kg are discharged, but with significantly degraded isotopics (i.e., with a high percentage of even mass isotopes). Spontaneous fissions per second in a gram of this degraded Pu are ~500, resulting in significantly increased proliferation resistance.Every 6 yr the blanket discharge contains 780 kg of 233U (including 233Pa) and 36 kg of 235U. However, the blankets are initially loaded with an amount of natural uranium selected such that these U fissile isotopes constitute only 12% of the total U discharge, a percentage equivalent to 20% 235U enrichment; hence, both the discharged uranium isotopics satisfy proliferation-resistant criteria.The RTPI control variables, namely, the moderator temperature coefficient, the reactivity per ppm boron, and the control rods worth, are about equal to those of a PWR. The RTPI spent-fuel stockpile ingestion toxicity over a period of ten million years is about the same as the counterpart toxicities of a regular, or a mixed-oxide (MOX), PWR. Compared with known PWR MOX variants, the RTPI is, per 1000 MW(electric) and per annum, a significantly more efficient incinerator of weapons-grade plutonium.

Graphite-filled mixed-oxide fuel design for fully loaded PWR cores

In order to maximize the benefit of plutonium recycling in the current light water reactor (LWR) cores, recently, many researches have been performed for 100% mixed-oxide (MOX) core. However, when the LWR core is fully loaded with MOX fuel, several problems arise due to the spectrum hardening effect of the plutonium isotopes. In order to overcome the spectrum hardening effect, we propose a new fuel design for full MOX core, that consists of annular fuel material filled internally with graphite. The new design is compatible with the conventional assembly in its overall geometry, and therefore it may require only minor modifications from the conventional fuel technology. The results of assembly and core analysis with the graphite-filled MOX fuel rods indicate that the new fuel design provides desirable characteristics, namely, higher burnup with less fuel inventory, faster depletion of plutonium and less production of actinides, and better safety features in fuel temperature, control rod worth, boron worth, moderator temperature coefficient (MTC), and shutdown margin. These beneficial results come from the enhanced moderation of the neutrons due to the moderating power of graphite and its contribution to the increased moderator/fuel volume ratio.

Reactor physics calculations and their experimental validation for conversion and upgrading of a typical swimming pool type research reactor

Detailed neutronic analysis of a typical swimming pool type research reactor, Pakistan Research Reactor-1 (PARR-1), was carried out for conversion of its core from 93% highly enriched uranium to 20% low enriched uranium fuel with power upgrading from 5 to 10 MW. Standard computer codes WIMS-D/4 and CITATION were employed to calculate core excess reactivity, power defect, reactivity effect of xenon and samarium, reactivity worth of fuel element, worth of control rods, shutdown margin, reactivity feedback coefficients, neutron flux and power peaking factors. A series of low and high power tests were performed on the newly converted core to determine its performance. A comparison between the calculated and measured results is presented in this article. The agreement is generally good.

Application of modified explicit higher-order perturbation (MHP) method to PWR core analysis

A PWR core analysis was performed by using the Modified Explicit Higher-Order Perturbation (MHP) method. A new perturbation calculation code with two energy groups in a two-dimensional x- y geometry based on the MHP method TWOPEC was developed to evaluate the change both in the neutron flux distribution and the reactivity caused by a perturbation. The results were compared with those by the direct eigenvalue calculations to examine the capability of the MHP method. It was demonstrated that both the neutron flux distribution and the reactivity can be accurately evaluated by the MHP method for a string and local perturbation caused by the control rod insertion when one uses large numbers of expansion modes and perturbation orders. The MHP method was applied to the pseudo burnup calculations that regarded the change in cross sections caused by the fuel burnup as the perturbation. It was found that one can also obtain satisfactory results not only for the burnup reactivity but also for the control rod worth at each burnup step.

高温工学試験研究炉（ＨＴＴＲ）の臨界試験 （Ｉ） 試験計画，燃料装荷および核特性試験

High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium-cooled reactor which has 30MW of thermal power amd 950°C of outlet coolant-gas temperature. The fuel loading of the HTTR was started on July 1, 1998. from the core periphery. The first criticality was attained in annular type core of 19 columns on Nov. 10, 1998. The startup core physics tests consisted mainly of tests for licensing and tests for establishing the technology bases necessary for HTGRs. It was confirmed in the former tests that all fuel blocks are loaded in certain positions and the excess reactivity is less than the limit. Experimental data for the annular core are obtained in the latter tests. Also, it was confirmed that the inverse kinetics method and delayed integral counting method are useful for the measurement of scram reactivity even if it takes about 10s for rod insertions. Furthermere, control rod worth curve, axial neutron flux distribution, etc. were measured to grasp the core performance. All tests planned in the startup core physics tests had been successfully performed and were completed on Jan. 21, 1999. It was confirmed from the tests that the HTTR was capable to step up to the power ascension tests.

Monte Carlo Simulation of the Triga Mark II Benchmark Experiment

A Monte Carlo computer code MCNP4A simulation of the TRIGA Mark II benchmark experiment performed in 1992 is presented. It may be noted that this benchmark experiment is one of very few high-enrichment benchmarks available. To minimize errors due to an inexact geometry model, the TRIGA Mark II reactor core was very thoroughly modeled. All fresh fuel and control elements as well as the vicinity of the core were precisely described. MCNP4A input was prepared in such a way that any desired core configuration could be simulated easily. Continuous energy cross-section data from ENDF / B-VI and ENDF / B-V(for nat Cr, natFe, and natNi) libraries and S(α, β) scattering functions from the ENDF / B-IV library were used in our calculations. The differences between ENDF / B-VI and ENDF / B-V evaluations were examined on critical experiments. Most of the steady-state operation experiments were simulated, including two critical experiments, namely, measurements of the excess reactivity of the core, and the determination of control rod worths and fuel element reactivity worth distribution. Excellent agreement with the experimental results was observed.

96/06044 Non linear analysis of out of phase oscillations in boiling water reactors

A detailed three-dimensional, continuous-energy MCNP4B model of the LWR-PROTEUS critical facility has been developed for the analysis of whole-reactor characteristics using ENDF/B-V, ENDF/B-VI and JEF-2.2 cross-section sets. The model has been applied to the determination of the critical loading, as well as the evaluation of reactivity worths for safety/shutdown rods, control rods, and individual driver-region fuel rods. The initially obtained results for the first configuration investigated (Core 1B) indicated that, for the same geometrical and materials specifications, the ENDF/B-V data library yields the closest critical prediction (discrepancy of 640±40 pcm), followed by ENDF/B-VI (980±40 pcm) and JEF-2.2 (1340±40 pcm). The differences in results between the three data libraries were studied by considering the contributions of individual materials to the neutron balance. 235U and 238U cross-sections from JEF-2.2, for example, explain an effect of ∼400 pcm. Refinement of the materials specifications in the MCNP4B whole-reactor model, in particular the impurities assumed for the graphite driver of the driver and reflector regions, has been shown to reduce the final discrepancy of the ENDF/B-V based keff result to ∼0.2%. The MCNP4B results for relative reactivity effects, such as control rod worths, are found to agree within experimental errors with the measured values.

96/03844 Viscosity of coal-derived liquids. 1. A group contribution method for pure model compound liquids

96/03844 Viscosity of coal-derived liquids. 1. A group contribution method for pure model compound liquids

Estimation of Measured Control Rod Worths in Fast Breeder Test Reactor — Effect of Different Delayed Neutron Parameters

Control rod worths have been measured by the inverse kinetics method in the small PuC-UC core of the Fast Breeder Test Reactor at Kalpakkam. Delayed neutron fractional yields based on Tuttle`s data, ENDF/B-VI data, and the full summation approach of Brady and England have been used to get measured control rod worths. Unreasonably large reductions in control rod worths are obtained by the ENDF/B-VI data. It is suspected that the procedure, of normalizing fractional yields obtained by the summation approach to earlier evaluated total yields, is inconsistent.

96/01535 Application of the rod-insertion method for control rod worth measurements in research reactors

96/01535 Application of the rod-insertion method for control rod worth measurements in research reactors

96/00351 Estimate of total control rod worth based on lower dimensional modelling

96/00351 Estimate of total control rod worth based on lower dimensional modelling