Commercial Pressurized Water Reactor Research Articles

Nonfuel Antineutrino Contributions in the High Flux Isotope Reactor.

Reactor neutrino experiments have seen major improvements in precision in recent years. With the experimental uncertainties becoming lower than those from theory, carefully considering all sources of is important when making theoretical predictions. One source of that is often neglected arises from the irradiation of the nonfuel materials in reactors. The rates and energies from these sources vary widely based on the reactor type, configuration, and sampling stage during the reactor cycle and have to be carefully considered for each experiment independently. In this article, we present a formalism for selecting the possible sources arising from the neutron captures on reactor and target materials. We apply this formalism to the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, the source for the the Precision Reactor Oscillation and Spectrum Measurement (PROSPECT) experiment. Overall, we observe that the nonfuel contributions from HFIR to PROSPECT amount to 1% above the inverse beta decay threshold with a maximum contribution of 9% in the 1.8-2.0 MeV range. Nonfuel contributions can be particularly high for research reactors like HFIR because of the choice of structural and reflector material in addition to the intentional irradiation of target material for isotope production. We show that typical commercial pressurized water reactors fueled with low-enriched uranium will have significantly smaller nonfuel contribution.

Irradiation-assisted stress corrosion cracking of Type 347 and Type 316 steels irradiated in commercial pressurized water reactors

Investigations of Type 347 and Type 316 austenitic stainless steel baffle-former bolt failures at 4-loop downflow pressurized water reactors (PWRs) in the US have identified significant amounts of transgranular cracking on the fracture surfaces along with intergranular cracking and ductile rupture. The expected failure mode for baffle-former bolts is irradiation-assisted stress corrosion cracking (IASCC), which typically results in an intergranular cracking mode. The presence of transgranular cracking raised the question of whether IASCC was responsible for the failures, or if a fatigue component was contributory. Detailed fractography and metallography of bolts with and without cracks confirmed that an environmentally-assisted cracking mechanism was responsible. Reviews of the literature identified numerous cases where stress corrosion cracks (including irradiation-assisted stress corrosion cracks) in austenitic stainless steels transitioned from intergranular to transgranular, typically as a more severe stress state developed. The potential cracking mechanisms are discussed. This allowed the conclusion that IASCC was responsible for the degradation observed in the baffle-former bolts, though a component of alternating loading cannot be excluded. The downflow plant configuration results in increased stress on neighboring intact bolts after bolt failure, resulting in the development of ‘clusters’ of failed bolts, which grew as additional bolts failed. This cluster growth contributed to the severe stress state imposed on the bolts.

MCS/TH1D analysis of VERA whole-core multi-cycle depletion problems

This paper presents the verification and validation elements of the UNIST in-house Monte Carlo code, MCS, for the multi-cycle and multi-physics analyses of high-fidelity, large-scale commercial pressurized water reactors (PWRs). Analysis on the neutronic performance with thermal/hydraulic (T/H) feedback is the key to detecting the complex behavior of an operating nuclear power reactor. The MCS solutions with T/H feedback, TH1D, of the consortium for advanced simulation of light water reactors (CASL) virtual environment for reactor applications (VERA) core physics benchmark progression problems 6 and 7 showed excellent agreement in eigenvalues, temperature and power profiles with the MC21/COBRA-IE, MC21/CTF and VERA-CS solutions for the single assembly and whole core of Watts Bar Nuclear 1 (WBN1) Cycle 1 under beginning-of-cycle and hot-full-power condition. Furthermore, the core depletion analysis is one of the most compelling advances for reactor analysis. Therefore, this work is significantly focused on the nuclide depletion simulation coupled with TH1D of the first two cycles of WBN1 to address the VERA core physics benchmark problems 9 and 10. MCS is one of the few Monte Carlo codes that have the capability of depletion calculation for both WBN1 Cycles 1 and 2. The accuracy of MCS simulation of WBN1 Cycle 1 is within 40 ppm and 30 ppm in critical boron concentration (CBC) for all burnup points, compared with the measured data and VERA-CS solutions, respectively. To demonstrate the multi-cycle refueling capability of MCS, the WBN1 Cycle 2 is simulated and compared with the solutions of VERA-CS only, because of the lack of available measured data. MCS shows excellent agreement compared with VERA-CS within 30 ppm in CBC, and the average bias for the entire Cycle 2 is approximately 20 ppm. These results provide confidence in MCS’s capability in high-fidelity, multi-cycle calculations of the practical PWR core.

Validation of UNIST Monte Carlo code MCS using VERA progression problems

Abstract This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

Automatic generation control for the flexible operation of multimodular high temperature gas-cooled reactor plants

To balance the intermittent renewable energy (IRE) and to realize its deep penetration, it is necessary for nuclear power plants (NPPs) to operate with enough flexibility. This necessity leads to the concerned questions such as which types of nuclear reactors are suitable for flexible operation and how to realize the operational flexibility of NPPs. In this paper, based on literature review and related analysis, it is shown that the modular high temperature gas-cooled reactor (MHTGR) is suitable for balancing IRE because of the advanced features such as robust fuel elements, online fuel handling, low power density and full-power-range temperature negative feedback effect. Due to the low power density, the power rating of a single MHTGR is lower than those large commercial pressurized water reactors (PWRs). To balance IRE by MHTGRs in a scalable manner, it is recommended to develop NPPs with high rated power based on multimodular scheme, i.e. multiple MHTGR-based nuclear steam supply system (NSSS) modules providing superheated steam for a common turbine. Further, to realize the flexible operation of MHTGR for deep IRE penetration, the automatic generation control (AGC) of multimodular MHTGR plants should be developed. In this paper, two AGC methods including both the adaptive AGC and extended-state-observer (ESO) based AGC are proposed, where the former one has a simple proportional-integral (PI) structure, and the latter one has an additional feedforward compensating action determined by the ESO. The corresponding control system of multimodular MHTGR plants is also designed, which is composed of the units of NSSS control, multimodular coordination control, turbine speed governor (TSG) and AGC. The AGC methods as well as the plant control system design with AGC function are then applied to two modular MHTGR plant HTR-PM. Numerical simulation results show that the newly-built multimodular control system can well stabilize the grid frequency by properly performing the power-level maneuver and maintenance of NSSS modules, and also show that the ESO-based AGC has a better frequency stabilization performance than the adaptive AGC does.

Scaling and integral effect test on transition from depressurization to long-term gravity injection

The passive safety commercial pressurized water reactors (PWRs), AP1000 and CAP1400 are equipped with the automatic depressurization system (ADS). The fourth stage of ADS (ADS-4) has the largest discharging capacity among the all ADS stages. During a small break loss-of-coolant accident (SBLOCA), only when the system is fully depressurized by the ADS-4 to near the containment atmospheric pressure, the in-containment refueling water storage tank (IRWST) can deliver the large coolant inventory for the long-term core cooling by its water head. Before the IRWST injection, the core reaches its minimum inventory. So the transition from the ADS-4 depressurization to the IRWST injection is a very challenging period in a SBLOCA. Therefore, to evaluate the passive system performance by the integral effect test (IET), the transition must be simulated properly. This work conducted the scaling analysis for the ADS-4 depressurization and the IRWST injection for designing the related components of the ACME IET facility. A uniform scaling law based on the existing T-branch liquid entrainment models was proposed. For scaling the onset of the IRWST injection, the requirement for reproducing the synergetic effect in an IET was introduced in order to preserve the important event/process occurrence time with the proper conditions, and the scaling laws for the full pressure IRWST injection were obtained. For investigating the transition thermal hydraulic behaviors and the ADS-4 configuration influences on the core safety, one cold leg break test and three double-ended direct-vessel-injection (DEDVI) line break tests with the different ADS-4 failure modes were selected from the ACME SBLOCA test matrix, and the main test data related to the core safety were analyzed and compared. It was found that the ADS-4 actuation significantly accelerated the system depressurization but worsened the core inventory depletion at the same time. In addition, the break condition and the ADS-4 failure mode could differently affect the system depressurization and the core level. The ACC injection during the ADS-4 depressurization in DEDVI was critical for the core level makeup. The insufficient ADS-4 venting capacity delayed IRWST injection, but also mitigated the core inventory depletion. It suggests the ADS-4 design can be potentially optimized in the passive safety system design.

Analysis on the emission of tritium from Qinshan nuclear power base of china in recent ten years

The emissions of tritium in airborne and liquid effluent from Qinshan nuclear power base in the period from 2007 to 2016 are counted, and compared with the global average emission level. The results show that the tritium emissions from the Qinshan nuclear power base in the past ten years have met the Chinese national standard limits. Compared with the global average emission data and the US commercial pressurized water reactor nuclear power plant emission data, the tritium emissions of PWR units are at the same level. Emissions from heavy water reactors of the Qinshan nuclear power base are lower than global average emissions (about 10% of the global average) and lower than the Canadian average emission level.

PCI analysis of Zircaloy coated clad under LWR steady state and reactor startup operations using BISON fuel performance code

Research following the earthquake and subsequent tsunami that devastated the Fukushima Daiichi nuclear power plant, causing severe damage to the reactor cores, has led to the development of accident tolerant fuel designs. Accident tolerant fuels are described as a fuel form which exhibits improved material response, when compared to traditional uranium dioxide fuel clad by a zirconium alloy, during accident (e.g., loss of coolant accident or reactivity insertion accident) while maintaining or exceeding normal reactor operational expectations. One of the primary goals of accident tolerant fuel concepts is to reduce cladding corrosion by either replacing traditional Zircaloy cladding with a new material or by applying a coating to the outer surface of the cladding with enhanced resistance to corrosion at high temperatures. However, while coating the clad may provide an increase in high temperature corrosion resistance, it also changes the mechanical state of the cladding and potentially creates alternate failure mechanisms during reactor operations or accident conditions. This research has primarily focused on the pellet-cladding interactionmechanical response of FeCrAl (Iron-Chromium-Aluminum alloy) coated cladding during light water reactor steady state and reactor startup operations. Using the MOOSE-based, finite-element fuel performance code BISON and commercial pressurized water reactor data obtained from Diablo Canyon Unit 2, a preliminary analysis was performed to evaluate the performance of FeCrAl coated cladding under steady state operation and its mechanical stability under pellet-cladding interaction during a reactor startup. This work summarizes BISON 2-D radial-axially symmetric (R-Z) and radial-circumferentially symmetric (R-θ) simulation results by comparing 20, 50, and 100-µm FeCrAl coating thicknesses as well as FeCrAl cladding to traditional Zr-4 cladding under pellet-cladding interaction condition. The results of this study will provide a feasibility analysis for steady state and transient operational response of FeCrAl coated, Zircaloy cladding. Furthermore, this work will provide a fundamental basis for the assessment of the ability of coated cladding, as an ATF candidate, to maintain or exceed current steady state reactor operating expectations.

Aqueous leaching of high burnup UO2 fuel under hydrogen conditions

ABSTRACTLeaching results on fragments of high burnup (65 MWd/kgU) UO2 fuel from a commercial pressurized water reactor are presented. The experiment was performed in simplified granitic groundwater under a hydrogen pressure of up to 5 MPa, representing conditions in a water intrusion scenario for a Swedish KBS-3 design spent nuclear fuel repository. The freshly crushed fragments were pre-washed for 6 days to remove pre-oxidized matrix and part of the instant release fraction of the radionuclide inventory, and then transferred to an autoclave for leaching under hydrogen conditions. Following an initial release of U attributed to dissolution of a pre-oxidized fuel layer caused by the aerated handling mainly during the transfer from pre-washing to autoclave, the U concentration decreased with time to levels of 2-5x10-9 M, which corresponds, approximately, to the solubility of amorphous UO2. The release of radionuclides such as Cs and Sr gradually declined indicating a transition to inhibition of the fuel matrix dissolution.

Criticality concerns of a group actinide co-crystallization separations approach to used nuclear fuel recycling

The co-crystallization of hexavalent actinides with uranyl nitrate hexahydrate has been established as a viable approach to achieving group actinide separation in used nuclear fuel recycle, while avoiding the difficulties associated with the solvent-extraction process altogether. A preliminary neutronics safety analysis has been performed to alleviate criticality concerns associated with this group hexavalent actinide co-crystallization approach. The basic uranyl nitrate hexahydrate material compositions used are representative of used fuel found at commercial pressurized water reactors all around the U.S. This material is subsequently doped with various percentages of either typical reactor grade plutonium or with the total transuranic vector, replacing the uranium atoms in the crystal lattice structure. Infinite and effective neutron multiplication factors were calculated using the state-of-the-art Monte Carlo neutron transport code MCNP in a variety of plausible material configurations, assuming an acceptable safety limit for criticality to be an effective neutron multiplication factor less than 0.95. The results indicate that the crystal structures form a very subcritical intrinsic geometry even with six water molecules being incorporated into the crystal structure. The results of this research suggest that significant opportunity exists for the safe transfer of the group actinide separation approach from a laboratory to an operational scale.

Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

Abstract The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5® is a trademark or a registered trademark of AREVA NP in the USA or other countries.

PCI analysis of a commercial PWR using BISON fuel performance code

Through the early 2000’s, nuclear fuel cladding failure resulting from pellet-cladding interaction in commercial pressurized water reactor fuel was managed successfully by incorporating power ramp restrictions. Within the past decade, however, cladding failures have again begun to appear, often correlated to missing pellet surface defects, and further aggravated by modern reactor operating strategies demanding higher fuel duty operations. In response, renewed efforts in fuel behavior modeling and simulation are being pursued to improve our understanding of the conditions that can lead to missing pellet surface related cladding failure. This paper describes a methodology used to model pellet-cladding interaction related failures observed in commercial pressurized water reactors using the BISON fuel performance code developed at Idaho National Laboratory. In order to accomplish this goal, a three-step process is followed: 1) A failure-stress threshold is developed by using BISON code to model ramp test with known pellet-clad interaction related failures, 2) BISON analysis is applied to a commercial pressurized water reactor cycle startup with one rod confirmed to contain a missing pellet surface defect, and 3) an missing pellet surface sensitivity study of a cycle startup of the same commercial pressurized water reactor which contained a failed fuel rod of un-identified type. The analysis results described in this paper serve to illustrate the BISON code capabilities for modeling pressurized water reactor fuel subjected to higher duty power events.

Development of Passive In-Core Cooling System for Nuclear Safety Using Hybrid Heat Pipe

The importance of passive safety for commercial nuclear power plants has been emphasized after the nuclear accidents that occurred at Three Mile Island and Fukushima. A combination of unexpected human errors, severe natural disasters, and defects of system designs led to the accidents, thereby highlighting the vulnerability of established safety systems of commercial nuclear power plants. Various passive safety systems are under development to mitigate design-basis accidents. However, several uncertainties and problems have been pointed out. As a solution to the problems, this paper proposes various designs for a passive in-core cooling system (PINCs) based on hybrid heat pipes. The feasibility and coolability of the PINCs as a passive safety system for commercial pressurized water reactors was investigated using experimental works and numerical analyses. The PINCs showed sufficient coolability to mitigate station blackout conditions by delaying core uncovery. Additionally, several PINCs concepts for advanced nuclear power plants such as a small modular reactor and Generation IV reactors are suggested.

Electrochemical Measurement of Li2O in Molten LiCl Salt

Nuclear fuel reprocessing is an important technology designed to increase the sustainability of nuclear power generation, as well as minimize long-lived radioactive waste production. There are multiple reprocessing schemes, most notably PUREX and pyroprocessing. While PUREX and related aqueous separation processes technically have superior maturity due to large scale implementation in France, Japan, Great Britain, and Russia, pyroprocessing has attracted substantial interest in recent years largely based on the perception that it represents a much lower proliferation threat and is ideally suited for metallic fast reactor fuel recycle. The Republic of Korea in particular has invested a tremendous amount of research funding over the last decade to develop pyroprocessing for spent fuel from commercial pressurized water reactors. Pyroprocessing utilizes a high temperature electrochemical cell called an electrorefiner to separate actinides from fission products. To apply this technology to oxide fuel, as in the case of the Korean spent fuel, an oxide reduction step must precede electrorefining. The current leading technology for oxide reduction is a high temperature electrolytic process carried out in molten lithium chloride salt. Electrolytic reduction relies on a lithium cycle, which consists of oxidation of lithium metal to lithium oxide at the cathode and reduction of the lithium oxide back to lithium metal at the anode. The net result is that uranium oxide is reduced to uranium metal, and oxygen gas is evolved. Lithium oxide concentration in the molten salt electrolyte is an important process variable. If it decreases, it can be a sign of lithium oxide accumulation in the fuel basket and will result in increased anode potential that ultimately leads to anode degradation. Currently, the only proven way of measuring lithium oxide concentration in lithium chloride involves sampling followed by acid-base titration. Measuring lithium oxide concentration in real time is an important objective to maintain optimal process control. The technique being developed for this purpose involves measuring the change in the open circuit potential (OCP) between a molten pool of a lead-lithium alloy and a Ni/NiO reference electrode. Because OCP measures equilibrium potential, this method can be combined with the Nernst Equation to continuously measure lithium oxide concentration. Mixtures of lithium chloride and lithium oxide were prepared with lithium oxide concentrations ranging from 0.00 to 2.75 wt%. A lead-lithium pool containing 2.0 wt% lithium was placed below the molten salt pool. After the salt melted it was allowed an hour to equilibrate, then an OCP measurement was obtained. As expected, the OCP increased with increasing concentration of Li2O. The measured sensitivity was about 200 mV for a Li2O concentration change from 0 to 2.5 wt%. Achieving repeatable and stable reference electrode potential is key to making accurate measurements from this method. The reference electrodes were made of high-density MgO tubes with low density, porous MgO plugs. Nickel wire and NiO powder were inserted into these tubes. The reference potential varied significantly between different sets of reference electrodes, which is hypothesized to be due to variations in the thickness of the porous MgO plugs. Methods for standardizing the reference electrodes have been attempted, and the results will be discussed. Another factor that may effect this measurement is the solubility of lithium metal in the zero oxidation state in the molten LiCl. Lithium solubility measurements were made in LiCl-Li2O with varying concentrations of Li2O. The results varied from 0.7 to 1.5 mole%. This low level of solubility is expected to cause minimal loss of lithium from the lead-lithium sensor. And experiments with different volumes of salt over a fixed volume of lead-lithium confirmed that the measured OCP is relatively insensitive to any decrease in the lithium concentration in the lithium-lead pool.

A Reduced-Boron OPR1000 Core Based on the BigT Burnable Absorber

Reducing critical boron concentration in a commercial pressurized water reactor core offers many advantages in view of safety and economics. This paper presents a preliminary investigation of a reduced-boron pressurized water reactor core to achieve a clearly negative moderator temperature coefficient at hot zero power using the newly-proposed “Burnable absorber-Integrated Guide Thimble” (BigT) absorbers. The reference core is based on a commercial OPR1000 equilibrium configuration. The reduced-boron ORP1000 configuration was determined by simply replacing commercial gadolinia-based burnable absorbers with the optimized BigT-loaded design. The equilibrium cores in this study were directly searched via repetitive Monte Carlo depletion calculations until convergence. The results demonstrate that, with the same fuel management scheme as in the reference core, application of the BigT absorbers can effectively reduce the critical boron concentration at the beginning of cycle by about 65 ppm. More crucially, the analyses indicate promising potential of the reduced-boron OPR1000 core with the BigT absorbers, as its moderator temperature coefficient at the beginning of cycle is clearly more negative and all other vital neutronic parameters are within practical safety limits. All simulations were completed using the Monte Carlo Serpent code with the ENDF/B-VII.0 library.

Burnable absorber-integrated guide thimble (BigT) – II: application to 3D PWR core design

ABSTRACTThis paper is a companion to its immediate predecessor in the “Burnable absorber-integrated Guide Thimble” (BigT) series. It aims to demonstrate potential applications of the BigT concepts in a three-dimensional (3D) commercial pressurized water reactor core, which is based on the AP1000 first core design. The study specifically compares neutronic characteristics of the reference core against a BigT-loaded design. In this study, reactivity depletion patterns of nine fuel assembly lattices in the reference core were first evaluated. Corresponding sets of BigT-loaded assemblies that yield neutronically similar characteristics (i.e., initial reactivity suppression and depletion trend) with those of the reference assembly lattices were determined next. These BigT-loaded fuel assemblies were later loaded in place of the reference fuel assemblies for the subsequent high-fidelity 3D Monte Carlo core simulations. Results of the study clearly demonstrate that the BigT-loaded AP1000 first core performs as well as the reference core since all neutronic parameters are comparable, especially in terms of reactivity depletion, power peaking factors and shutdown margin. All simulations were completed using the Monte Carlo Serpent code with ENDF/B-VII.0 library.

Coupling of an innovative small PWR and advanced sodium-cooled fast reactor for incineration of TRU from once-through PWRs

In the nuclear industry, safely managing spent fuels discharged from PWRs (pressurized water reactors) is an ongoing challenge. In this paper, a synergistic coupling of innovative small long-cycle PWRs and advanced sodium-cooled fast reactors is considered to reduce the accumulated TRUs (transuranics) by transmuting them with electricity production. In the coupling strategy, the innovative small PWRs employing UO2–ThO2 and fully ceramic micro-encapsulated fuels are used to deeply burn TRUs from commercial PWRs, while advanced SFRs (sodium-cooled fast reactors) with actinide recycling are designed to further transmute the TRUs discharged from innovative small PWRs. This work focuses on the core physics analysis of new SFR burner cores using different TRU feeds discharged from small PWRs. Additionally, quasi-static reactivity balance analyses are performed to understand the safety of the SFR burner cores. The mass flows of TRUs in the nuclear park, which is composed of PWRs, small long-cycle PWRs, and SFR burners, are analyzed to evaluate TRU inventory reduction. The results of this study show that the advanced SFR burners with all the TRU feed types discharged from the small long-cycle PWRs have a high TRU consumption rate. They satisfy all of the conditions for self-controllability under unprotected accidents with a reasonable number of control rods. This coupling strategy requires ~35% less power in conjunction with the advanced SFR burners in the nuclear park and increases the support ratio of SFR burners by ~42% than does the coupling of commercial PWRs and SFR burners. Copyright © 2015 John Wiley & Sons, Ltd.

Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10−7, compared to those of commercial PWRs ∼10−6–10−1. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14×14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of PASS fuel rods during LOCA is predominantly governed by the initial cold gap pressure; a decrease of the cold gas gap pressure is a handy design direction to assure structural safety margin for SiC cladding.

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%.One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

New calculational model for self-powered neutron detector based on Monte Carlo simulation

A new calculational model of detector sensitivity, which is defined as the ratio of electric current to neutron flux, for self-powered neutron detector (SPND) is presented in this paper. Since Warren developed a sensitivity calculation model based on a multi-step analytic approach in 1972, there have been a lot of researches aimed at improving the accuracy of his model by partially employing Monte Carlo (MC) simulations in some of the calculation steps. However, there still exist assumptions such as mono-energy neutron in those models which could increase the uncertainty of the calculated detector sensitivity. In this paper, a new model which overcomes those limitations is developed by adopting the MC method more thoroughly and thereby reducing the number of calculation steps. The verification of the new method has been performed by comparing the results against the ones from the existing methods and measured data. A detailed analysis of SPND depletion by ORIGEN-S and BETA-S codes has shown that the proposed model can predict accurately the measured depletion data in a commercial pressurized water reactor (PWR) plant.