Core Radial Expansion Research Articles

Best estimate calculation and uncertainty quantification of sodium-cooled fast reactor using MARS-LMR code

The safety analysis of nuclear power plants has been performed using conservative approach based on conservative assumptions and boundary conditions to evaluate the safety margin of plant operation. However, this conservative approach could lead to unrealistic behavior predictions and eventually distort some phenomena in reactor systems. Therefore, the nuclear field moved towards an alternative best-estimate approach with uncertainty quantification in order to improve the phenomena prediction and to decrease the excessive conservatism in safety margins. In this study, the best estimate methodology is applied to improve the accuracy and reliability in safety analysis of an SFR. The applied best estimate methodology is based on the CSAU. This methodology is composed of three unique steps for evaluation of code capability, assessment and range of parameters, and sensitivity and uncertainty analysis. The primary purpose of this study is to evaluate the appropriateness of sensitivity parameters and its ranges, which have been determined through intensive experts’ panel discussion, by use of the data obtained from the EBR-II Unprotected Loss of Flow (ULOF) experiment. The MARS-LMR thermal–hydraulic code and the parallel computing platform integrated for uncertainty and sensitivity analysis (PAPIRUS) become the basic calculation tools in the study. Confirmation of data coverage is performed through the evaluation of coolant temperature in the instrumented subassemblies XX09. The appropriateness of parameters and its ranges are evaluated for three different cases: original parameters and ranges suggested in the MIRT, ±10% increased parameter ranges, and 200% increased axial reactivity feedback coefficient only. The case with the original parameters and ranges does not result in a valid data coverage, which means inadequate modeling accuracy for the ULOF scenario. The other two cases give complete coverage of EBR-II temperature data measured at the core top, which suggest the need of further refinement of reactivity models. The relative importance of the parameters is confirmed through the sensitivity analysis with respect to the Figures of Merit (FoM). The selected dominant parameters are the sodium density reactivity, above core load pad strain coefficient, core radial expansion reactivity coefficient and fuel axial expansion reactivity coefficient. The pump coastdown curve and the core inlet form loss are also found to be significant parameters during the transient.

An investigation on Unprotected Loss of Flow Accident in Th–Pu metal fuelled 500 MWe fast reactor

This study focuses on computing static reactivity coefficients and analyzing Unprotected Loss of Flow Accident in a Th–Pu fuelled metal reactor. An attempt is also done to compare the static and dynamic performance of the fresh core with characteristics of U–Pu–6Zr fuelled 500MWe metal reactor. Isothermal temperature coefficient and power coefficients are evaluated in the steady state and found to be negative. The excess reactivity and control rod worth requirements of Th–Pu metal core are assumed similar to that of U–Pu–6Zr metal core. In the Unprotected Loss of Flow Accident (ULOFA) analysis, with flow coast down initiated by station black out, it is found that power to flow ratio is increasing initially up to 53s and then starts to reduce continuously. Power to flow ratio is found to be less than 2 at all times thus ensuring the absence of coolant boiling in the entire core. Sodium voiding starts around 886s in the upper axial blanket and provide negative reactivity. Also it will not propagate to the core center ensuring the probability for core disruptive accident a remote one. Net reactivity feedback is negative and the major contribution is from core radial expansion. Within 12min, the power drops to 32MWt, making it possible for Safety Grade Decay Heat Removal (SGDHR) system to start heat removal from core ensuring safe shutdown of reactor. Sensitivity analysis by considering an uncertainty margin of ±10% in thermo physical properties of fuel composition shows that feedback reactivity of the Th–Pu system is insensitive and the conclusion on the safe shutdown remains unaltered. From this study it is found that inherent safety of Th–Pu metal fuel core is better than that of reactor core fuelled with U–Pu–6Zr metal type under ULOFA condition.

Inherent safety aspects of metal fuelled FBR

Static and dynamic studies of metal fuelled fast breeder reactors (MFBR) are carried out to verify the passive shutdown capability and its inherent safety parameters. Static calculations are carried out to determine the vested reactivity feedback parameters from the fuel and coolant temperature rise separately. Power reactivity decrement of metal fuel reactor is found to be small as compared to oxide fuel reactor of same size.ULOF analysis of metal (U–Pu–6% Zr) 1000MWe pool type MFBR is studied with a flow halving time of 8s. The study is also made with considering uncertainties on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behaviour of 1000MWe core is benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal System (SGDHRS) capacity before the initiation of boiling. From the study, it is concluded that if the sodium void reactivity is limited (4.6$) then the inherent safety of 1000MWe design is assured, even with 20% uncertainty on the sensitive parameters and also it is found out higher primary pump flow halving time (15s instead of 8s) can avoid cliff edge effects in 1000MWe MFBR transient behaviour.

Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500MWe and 1000MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500MWe and 1000MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000MWe design is assured, even with 20% uncertainty on the sensitive parameters.

Performance of the Lead-Alloy-Cooled Reactor Concept Balanced for Actinide Burning and Electricity Production

A lead-bismuth–cooled fast reactor concept targeted for a balanced mission of actinide burning and low-cost electricity production is proposed and its performance analyzed. The design explores the potential benefits of thorium-based fuel in actinide-burning cores, in particular in terms of the reduction of the large reactivity swing and enhancement of the small Doppler coefficient typical of fertile-free actinide burners. Reduced electricity production cost is pursued through a longer cycle length than that used for fertile-free burners and thus a higher capacity factor. It is shown that the concept can achieve a high transuranics destruction rate, which is only 20% lower than that of an accelerator-driven system with fertile-free fuel. The small negative fuel temperature reactivity coefficient, small positive coolant temperature reactivity coefficient, and negative core radial expansion coefficient provide self-regulating characteristics so that the reactor is capable of inherent shutdown during major transients without scram, as in the Integral Fast Reactor. This is confirmed by thermal-hydraulic analysis of several transients without scram, including primary coolant pump trip, station blackout, and reactivity step insertion, which showed that the reactor was able to meet all identified thermal limits. However, the benefits of high actinide consumption and small reactivity swing can be attained only if the uranium from the discharged fuel is separated and not recycled. This additional uranium separation step and thorium reprocessing significantly increase the fuel cycle costs. Because the higher fuel cycle cost has a larger impact on the overall cost of electricity than the savings from the higher capacity factor afforded through use of thorium, this concept appears less promising than the fertile-free actinide burners.

ULOF and UTOP Analyses of a Large Metal Fuel FBR Core Using a Detailed Calculation System

ULOF and UTOP analyses of a large metal fuel FBR core (1,500 MWe, averaged discharge burnup: 150 GWd/t) are conducted. The effect of core radial expansion is considered as the major negative feedback during the transient. A detailed analysis system is used, in which a transient core thermal-hydraulic code is coupled with three dimensional core radial deformation and reactivity feedback calculation codes, in order to calculate the radial expansion feedback. In ULOF analysis, the pump flow halving time is assumed to be 10 s, which is reasonably long and effective in avoiding too large power to flow ratio. The reactivity insertion during UTOP is set to be 34¢, based on the control rod reactivity design. As the analysis results, it is found that the core shows benign responses to both events, owing largely to the radial expansion feedback. No significant coolant boiling or fuel failure is predicted. The response during ULOF is compared to that of an oxide fuel core of similar design, and it is confirmed that the negative Doppler effect associated with the fuel temperature rise plays the major role in the quick power decrease.

ULOF and UTOP Analyses of a Large Metal Fuel FBR Core Using a Detailed Calculation System.

ULOF and UTOP analyses of a large metal fuel FBR core (1,500 MWe, averaged discharge burnup: 150 GWd/t) are conducted. The effect of core radial expansion is considered as the major negative feedback during the transient. A detailed analysis system is used, in which a transient core thermal-hydraulic code is coupled with three dimensional core radial deformation and reactivity feedback calculation codes, in order to calculate the radial expansion feedback. In ULOF analysis, the pump flow halving time is assumed to be 10 s, which is reasonably long and effective in avoiding too large power to flow ratio. The reactivity insertion during UTOP is set to be 34¢, based on the control rod reactivity design. As the analysis results, it is found that the core shows benign responses to both events, owing largely to the radial expansion feedback. No significant coolant boiling or fuel failure is predicted. The response during ULOF is compared to that of an oxide fuel core of similar design, and it is confirmed that the negative Doppler effect associated with the fuel temperature rise plays the major role in the quick power decrease.

Critical Experiment Tests of Bowing and Expansion Reactivity Calculations for Liquid-Metal-Cooled Fast Reactors

Experiments done in several critical assemblies of the liquid-metal-cooled fast reactor type simulated core axial expansion, core radial expansion and bowing, coolant expansion, and control driveline expansion. For the most part, new experimental techniques were developed to perform these experiments. Calculations of the experiments basically used design-level methods except when it was necessary to investigate complexities peculiar to the experiments. It was found that these feedback reactivities generally are overpredicted, but the predictions are within 30% of the experimental values.