Shutdown Margin Research Articles

Lowering the enrichment of the Syrian miniature neutron source reactor

An investigation of lowering the fuel enrichment of MNSR was realized. A 3-D neutronic model was developed for the analysis of the reactor. It was found that lower number of fuel elements is needed when UO 2 is used with 5.45 g of 235U content in each fuel rod. Sensitivity of the reactor to the purity of the beryllium reflector proved to be an important factor in determining the reactor neutronics as well as the weight of loaded fuel in the core. Inherent safety features of low excess reactivity and shutdown margins are preserved and enhanced. Average thermal fluxes over different zones of the core are kept very much unchanged. ©

Investigation of the k eff-variation upon water ingress in a pebble-bed LEU-HTR

Accidental water ingress in pebble-bed type, high temperature reactor (HTR) cores can result in large positive reactivity changes and in consequent reductions in shutdown margins, since such systems generally tend to be undermoderated. The prediction of such effects, which are particularly important in the case of low-enriched uranium (LEU) fuel, is a challenging task due to the complex interaction of, on the one hand, an increase in moderation and, on the other, a reduction in neutron streaming. Experiments have been carried out at the PROTEUS facility of the Paul Scherrer Institute in Switzerland to quantify water ingress effects in LEU-HTR systems and to assess the performance of certain deterministic codes in predicting them. In order to simulate water ingress, polyethylene rods were inserted between the pebbles of various deterministic configurations and a pulsed-neutron-source technique was used to measure the variation of k eff with the stepwise removal of these polyethylene rods. Generation time, as well as reactivity measurements, were compared with calculations, made with various types of streaming corrections, for three different subcritical states.

Analysis of the Reactivity during a Pressurized Water Reactor Main-Steam-Line-Break Transient

The main-steam-line-break (MSLB) transient in a pressurized water reactor (PWR) is a core overcooling event that can result in a large positive reactivity insertion. In most analyses the shutdown margin is sufficiently large that the core does not return to critical. However, some researchers have reported an increase in the core power even though the core does not return to critical. A simplified kinetics model based on the prompt-jump-kinetics approximation is reported in new work, and a single delayed neutron group is used to explain the core power increase during subcriticality. Specifically, it is shown that the multiplication of the initial delayed-neutron source as predicted by the rate of change of the reactivity during the transient is the reason for the increase in power even though the core never returns to criticality after scram. The results are demonstrated using data from a RETRAN-03 model of a hot-zero-power MSLB analysis of the Three Mile Island unit 1 PWR.

Comparison of Pressurized Water Reactor Core Characteristics for 100% Plutonium-Containing Loadings

The large quantities of reactor-grade (RG) and weapons-grade (WG) Pu accumulated worldwide could be reduced by employing 100% mixed-oxide (MOX) cores in light water reactors. The buildup of new Pu from the U present in the MOX, however, remains disadvantageous from the viewpoint of inventory reduction and also enhances the need for multiple recycling. A more effective way would be to use U-free fuel so that no new Pu is produced.A comparison is made, from the physics design viewpoint, between the potential and the possible difficulties for two different types of Pu-burning pressurized water reactor cores, namely, 100% MOX and 100% uranium-free Pu fuel. The latter employs ZrO2 as inert matrix and Er2O3 as burnable poison. In each case, RG and WG Pu have been considered separately. The characteristics of the four different cores have been studied on the basis of three-dimensional calculations for an equilibrium cycle, a real-life UO2-fueled core being considered as reference for comparison purposes.For all four Pu-burning cases, it appears possible to design a four-region core with a natural cycle length of more than 300 days. For the 100% MOX cores, the Pu mass is reduced during irradiation by ~35% of the initial Pu inventory. For the U-free cores, the consumption is about twice as much, i.e., ~60% for the RG-Pu fuel and over 70% for the WG-Pu core. The reactivity balance in going from hot full power to hot zero power conditions shows that while the 100% MOX core with RG Pu would need more effective control rods, both types of U-free cores have larger shutdown margins than the reference case. Consideration of the reactivity coefficients indicates that a steam-line-break accident could be more problematic in the MOX core with RG Pu than in the other cases. The rod ejection transient should be safe because the maximum inserted worth of a control rod is ~0.5 $. More detailed investigations of transient behavior—particularly for the U-free cores—are needed, the current study having considered feasibility mainly from the viewpoint of static physics considerations.

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

Advanced Designs and Operating Strategies to Enhance the Safety, Operability, and Efficiency of VVER-1000 Reactors

Advanced fuel, burnable absorber, and control rod designs along with advanced fuel management and power distribution control strategies will be implemented in the first operating cycle of the Czech Republic`s Temelin VVER-1000 nuclear power plants. These improvements increase safety margins, enhance operability, and improve fuel efficiency. The Westinghouse VVANTAGE 6 fuel assembly design incorporates many proven advanced fuel and core design features used extensively in western pressurized water reactors. The fuel assembly incorporates mixing vane structural grids, radial enrichment zoning, ZrB{sub 2} integral fuel burnable absorbers, axial blankets, and Zircaloy guide thimbles and structural grids. Low-leakage loading patterns are also used to reduce radial neutron leakage. The rod cluster control assembly (RCCA) design incorporates two absorber materials. The absorber tip uses silver-indium-cadmium material while the remainder of the absorber material is B{sub 4}C enriched in {sup 10}B. This design increases control rod worth as well as the usable lifetime of RCCA. The Westinghouse constant axial offset control operating strategy, improved RCCA design, modified RCCA overlap, and replacement of part-length RCCA by full-length RCCA are used to improve the axial power distribution control capability for VVER-1000 reactors. These design improvements provide thermal margin benefits, increase shutdown margin by almost 1.0% {Delta}{rho},more » reduce fuel cycle costs by nearly 30%, and improve axial power distribution control.« less

Fast Shutdown Margin Calculation Using Perturbation Theory with Regionwise Flux Expansion

A two-dimensional perturbation method with regionwise flux expansion is developed and tested for the boiling water reactor fast shutdown margin calculation. The ways of generating the two-dimensional parameters for the unrodded bundles are tested to find the one that results in the most accurate eigenvalue of the single-rod-out condition. The use of the one-bundle-per-region flux expansion method gives more accurate results than the ring-regionwise flux expansion method. The first four strongest control rods chosen by this method using one-bundle-per-region flux expansion always contain the top four strongest rods predicted by SIMULATE-3 three-dimensional calculations. The strongest rod is always correctly predicted, and the differences in shutdown margin predictions are <1 mk for all the cases tested. The time saved by using the two-dimensional perturbation method rather than the direct three-dimensional full-core calculation is a factor of ∼10 and even more for larger core loadings. By using correct two-dimensional parameters, the accuracies of the perturbation method itself in the calculations of the eigenvalue and the neutron flux distribution are also tested. It is found that the errors are very small even for such a strong perturbation in the shutdown margin calculation.

Reactivity Control Method for Extended Burnup of Boiling Water Reactor Fuel Bundles

Abstract A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity. Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the amount of reprocessing of about 60% per unit power, compared with current BWR designs.

Reactivity control method for extended burnup of boiling water reactor fuel bundles.

Abstract A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity. Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the ...

A sector symmetric core concept for BWRs

A new scheme for constructing reload patterns for BWR cores has been developed. It is obtained essentially by rotating the axes of reflective symmetry currently used for reload patterns through 45°. The high values of worths of control elements during their in-sequence withdrawal for taking the core to criticality are found to be reduced by this approach. The calculations have also shown that the new scheme leads to improved safety through higher shutdown margins, while a more uniform distribution of worths over control elements can reduce thermal shocks to the fuel. It is expected to result in higher thermal margins and hence ease of operation.

A Model for the Prediction of Shutdown Margin for Boiling Water Reactors

Multidimensional linear regression analysis is employed as a modeling technique for the prediction of boiling water reactor (BWR) shutdown margin reactivity. A comparison is made between the best models developed using regression analysis and the General Electric (GE) three-dimensional BWR core simulator code. The GE code is based on one-group diffusion theory, and its accuracy is verified by comparison with experimental data. One use of this code is the calculation of shutdown margin throughout a fuel cycle, but it requires a large computing facility not located at a reactor site. The regression models give an approximation of a core's shutdown margin based on current core physics parameters. The method can be utilized at a BWR plant site to provide information demonstrating compliance with license and technical specification requirements. The results obtained by regression predictions for the two cores studied compare favorably with current industry methods. After establishing a regression model, predictions can be made at a reactor site using a pocket calculator.

Precision of Shutdown Margin Measurements Using the Two-Frequency Reactor Noise Technique in an LMFBR

The precision of reactor noise techniques for measuring subcritical reactivity can be related to detector efficiency and parameters of the measuring instrumentation. Using a two-detector cross-correlation technique, we have determined that such measurements can be made in a certain fast reactor (the FFTF) to -3 to -5 dollars with acceptable statistics and measurement times.

Measurement of Shutdown Margin

A method was developed and investigated for directly measuring the shutdown margin in a reactor operating at a subcritical steady state. Measurements were made on the Iowa State University UTR-10 reactor with conditions ranging from almost critical to fully shutdown. The method is based on the fact that the frequency dependent transfer function of the reactor is a measure of the response of the reactor to an input disturbance. The magnitude of the transfer function of a subcritical reactor depends on the frequency of the disturbance and the steady-state multiplication factor. Since the multiplication factor is a measure of the reactivity, the amplitude of the transfer function is a direct indication of the shutdown margin. A sinusoidal disturbance was introduced by means of an oscillator operating in the center of the core. The oscillator was operated at a fixed frequency of 10 cps in order to eliminate the frequency dependence. The fluctuation was measured by a BF3 detector located against the outside of the core opposite the oscillator. The signal emitted by the detector contained bombardment noise and instrument noise in addition to the desired information. The signal was passed through a band-pass filter and was then cross-correlated with a pure sine wave to obtain the transfer function. A plot of the transfer function vs. the multiplication factor was a smooth function from which the multiplication factor could be found with a confidence of ± 0.04% near critical and with a confidence of ± 0.17% in the fully shutdown condition.

Optimization of Reactor Shutdown Margin Measurements in High Gamma Fluxes

Since the quality of subcriticality measurements using neutron noise analysis in power reactors is degraded by high gamma-ray intensities produced by residual fission products, a study was made to derive an equation that can be used to predict this degrading effect quantitatively. For simplicity, this study is based on one-detector measurements but is equally appropriate to two-detector cross-correlation measurements. Equations derived to give insight into the behavior of ionization chambers operating in high gamma environments have led to the experimental development of a relatively low-cost, optimized chamber.

Power-Spectral-Density Measurements in a Subcritical Nuclear Reactor

The power spectral density of the neutron density of a reactor is frequency dependent and related to the reactor transfer function and the power spectral density of the input disturbance. For a critical reactor, a power-spectral-density measurement can be used to evaluate the ratio (β/) where β is the effective delayed-neutron fraction and l the effective neutron lifetime. For subcritical operation, an evaluation of the reactor shutdown margin can be obtained by determining the quantity where k, the effective reproduction constant, can be determined if the effective neutron lifetime and effective delayed neutron fraction are known.The output power spectral density of the University of Florida Training Reactor, operating in the subcritical region, has been measured using a plutonium/beryllium source to provide the input disturbance. The data are then fitted by a least-squares method to a theoretical model to obtain the quantity

Evaluation of a Bank of Slab-type Control Rods

A method for evaluating a bank of slab-type control rods in enriched hydrogenous reactors considering both thermal and epithermal capture was developed within the framework of three neutron energy groups. An absorption area technique that combines diffusion theory in thc fuel moderator region and a transport condition at the control rod boundary is utilized. The effect of the absorption area is to decrease the source strength for the thermal and epithermal energy groups. The epithermal probability for a non-l/ nu absorber is found by an experimental reactivity comparison to a 1/ nu absorber. Assuming that the absorptton area is uniformly distributed throughout the core, a hand calculation can be made which determines the number, size, and composition of rods necessary to achieve a specific cold shutdown margin. (auth)

The rate of emission of alpha particles from radium

This study assesses the feasibility of a thorium-based fuel-self-sustaining Reduced moderation BWR (RBWR-Th) core. This core features a hexagonal tight-lattice fuel, high exit coolant quality, axial segregation of seed and blanket regions, and compatibility with the ABWR reactor vessel inherited from the RBWR-AC proposed by Hitachi. The RBWR-Th differs from the RBWR-AC by eliminating the internal blanket, eliminating absorbers from the axial reflectors, replacing depleted urania with thoria as the primary fertile fuel and elongating the fissile region. When the Hitachi assumptions and correlations for thermal-hydraulic analysis are used, the RBWR-Th obtains an average core discharge burnup of 61 GWd/t versus 45 GWd/t for the RBWR-AC. However, when more conservative thermal-hydraulic correlations are used, the average discharge burnup drops to 25 GWd/t. The RBWR-Th maintains negative coolant void coefficients of reactivity throughout the cycle – between -155 and -131 pcm/% – but currently does not have sufficient margin for cold shutdown. Future studies will address this shutdown margin deficit and will account for coolant pressure drop constraint.