Power Excursion Research Articles

Improvements in Transient Testing Reactor (TREAT) with a Choice of Filter

The safe and reliable operation of nuclear reactors has always been one of the topmost priorities in the nuclear industry. Transient testing allows us to understand the time-dependent behavior of the neutron population in response to either a planned change in the reactor conditions or unplanned circumstances. These unforeseen conditions might occur due to sudden reactivity insertions, feedback, power excursions, instabilities, and accidents. To study such behavior, we need transient testing, which is like car crash testing to estimate the durability and strength of a car design. In nuclear designs, such transient testing can simulate a wide range of accidents due to sudden reactivity insertions and helps study the feasibility and integrity of the fuel used in certain reactor types. This testing involves a high neutron flux environment and real-time imaging technology with advanced instrumentation with appropriate accuracy and resolution to study the fuel slumping behavior. With the aid of transient testing and adequate imaging tools, it is possible to test the safety basis for reactor and fuel designs that serves as a gateway in licensing advanced reactors in the future. To that end, it is crucial to fully understand advanced imaging techniques both analytically and via simulations. This paper presents an innovative method of supporting real-time imaging of fuel pins and other structures during transient testing. The major fuel-motion detection device that is studied in this dissertation is the Hodoscope which requires collimators. This paper provides 1) an MCNP model and simulation of a TREAT core with a central fuel element replaced by a slotted fuel element that provides an open path between test samples and a hodoscope detector, and 2) a choice of good filter to improve image resolution.

Prompt-critical power excursion measurements with optical fibers in the CABRI experimental reactor

This paper presents our experimental work to assess the capability of estimating the transient-induced power distribution evolution in the CABRI experimental reactor, using Cherenkov radiation in optical fibers. The CABRI reactor is designed to produce power transients from 100 kW up to 21 GW in less than 100 ms, in order to irradiate a test fuel pin under conditions representative of a Reactivity Insertion Accident in pressurized water reactors. Because of the wide response range and short response time required to follow the reactor power evolution during a complete transient, the usual means of detection, such as ionization or fission chamber, are rendered inoperable, especially for in-core or core vicinity measurements. For this reason, we suggest measuring the Cherenkov light produced within the optical fibers. The Cherenkov light emission is linked to local electron production, which is proportional to local gamma flux through the Compton or pair production cross-section. In other words, the intensity of Cherenkov radiation is related to the photon flux intensity. Knowledge of the fission photons emitted by the reactor gives direct insight into the fission rate, hence a spatial power density distribution could be reconstructed by using the measurement of Cherenkov light at different points in the reactor. The radio-luminescence measurements taken using photodiodes show very good linearity with the reactor power monitoring system despite the transient radiation induced attenuation observed in the optical fibers. The radio-luminescent light shows steady values with around 5% of variations over the 15 transients for low OH fibers and high OH fibers. Low OH tends to overestimate the FWHM of the power peak by 1–4%.

Nuclear Reactor Operations Education and Training with Virtual Reality and an Immersive Desktop Application

A virtual reality learning module to train nuclear engineering students in reactor operations to understand reactor power excursions has been developed. The learning module was taught with an Oculus-2 headset and controllers (now called Meta Quest 2). The class was comprised of 71 undergraduate students, mostly in their fourth year of the nuclear engineering curriculum at Texas A&M University. The learning module simulation of power excursion, called pulsing the reactor, was modeled after the Texas A&M Engineering Experiment Station TRIGA reactor. First, the students visited the TRIGA reactor for pulsing and answered a technical quiz on the subject. Next, the students performed pulsing in the equivalent virtual reality module developed in this work. One of the primary learning objectives in the laboratory exercise was the role of passive and active safety mechanisms in a rapid reactivity insertion and power excursion. Data from the actual reactor visit showed that most students did not understand a key passive safety mechanism during the reactor visit. However, the students showed a notable improvement in their understanding of the safety mechanisms after the virtual reality reactor visit. When asked if the virtual reality learning module would have made the quiz at the reactor easier, 96% of the students reported that at least one of the quiz questions would be have been better answerable with the virtual reality module. Students also noted that the virtual reality module needed to expand its scope to include more details and teaching components. Although most students were reluctant to completely replace the pulsing reactor visit with its virtual reality module version available at the time of the study, they appreciated it as a learning reinforcement tool. Student opinion may change more favorably in the future with continued improvements and enhancements of the module.

Development of a numerical model for disassembly phase of hypothetical core disruptive accident and application to a medium-sized mixed oxide fuelled fast reactor

Safety assessment of fast reactors involves analyzing Hypothetical Core Disruptive Accidents (HCDA) that could result in reactor core disassembly. HCDA is generally analyzed in three distinct phases that include pre-disassembly, disassembly and post-disassembly phases. In the present work, a new coupled neutronics-hydrodynamics computer code called FAst Reactor DISassembly (FARDIS-1) is developed in 2-D cylindrical coordinate geometry to model the super-prompt critical power excursion during the disassembly phase of HCDA in a sodium-cooled fast reactor (SFR). The modelling approach followed is essentially similar to the one used in VENUS-II code but with a few improvements. The code is benchmarked against the predictions of VENUS-II for a sample case of HCDA in the FFTF reactor model. This is followed by an analysis of the disassembly phase in a medium-sized (500 MWe) mixed oxide fuelled fast reactor. Two cases are assumed at the beginning of the disassembly phase: a sodium voided core (conservative) and a partially voided core (realistic) and the influence of several core neutronics parameters on the transients is elucidated through parametric studies. The peak power, thermal energy release, peak pressure, maximum temperature and work-potential are estimated and their trends are discussed.

LES investigations of turbulent heat transfer structures with exponential power escalation

The BORAX-type accident is a key scenario in the safety design of pool-type experimental nuclear reactors. It considers a reactivity insertion large enough to initiate an exponential power escalation with a period as small as tens of milliseconds, necessitating effective heat removal by the coolant system. This study presents a computational analysis of single-phase transient heat transfer under thermal-hydraulic conditions relevant to such reactors, focusing on turbulent channel flow between planar fuel plates with highly subcooled water. Our investigation uses Large Eddy Simulation (LES) performed with TrioCFD, the open-source CFD code developed by the French Atomic Energy Commission (CEA). Several modeling options are tested: incompressible flow with temperature as a passive scalar, incompressible flow with varying viscosity, and quasicompressible flow with temperature-dependent viscosity and mass density. To evaluate the accuracy of our physical models, all simulations were performed under uniform conditions, using the same experimental parameters, mesh, and numerical strategies. The specific examined scenario involves a highly subcooled turbulent channel flow at moderate pressure, experiencing an exponential power increase. We ensure detailed turbulence resolution at the Batchelor scale in the wall normal direction close to the heating wall. This communication presents a comparative study of LES simulations of the single-phase flow under exponential power excursion at the wall. Subsequent analyses focused on the impact of different velocity-temperature couplings on the prediction of wall heat transfer, incorporating comparisons with experimental infrared thermometry measurements. Early in the transient phase, temperature peaks are confined within turbulent streaks, as reported in the existing literature. However, as the exponential power transient unfolds, the viscous layer and streak structures destabilize, leading to a more dispersed distribution of hot spots. Preliminary findings suggest a transition in heat transfer modes from turbulence-driven to conduction during the final stages of the single-phase transient.

Analysis of power transients in the CABRI experimental reactor with a multi-physics APOLLO3®/THEDI coupling

The CABRI research reactor located at CEA Cadarache is dedicated to the analysis of nuclear fuel behavior during Reactivity-Injection Accident (RIA) in Pressurized Water Reactors (PWRs). The reactor experimentally simulates power pulse transients in the driver zone, which induce a RIA-representative energy deposition in a fuel sample contained in a water-loop that reproduces PWR thermal-hydraulics conditions. Power pulses are initiated by the 3He depressurization of four transient rods, which introduces reactivity into the core. The injection of reactivity causes the reactor power to increase rapidly, resulting in a rise in fuel temperature. This temperature rise mitigates the power excursion by the Doppler effect. Subsequently, heat transfer occurs between the fuel, the cladding and the moderator. The evolution of the core power is measured during power transients, and these measurements can be valorized for the experimental validation of multi-physics simulation tools. Indeed, the experimental validation of these coupling mainly relies on measurement data coming from the current industrial reactor fleet. The measurement data coming from experimental facilities for the validation of multi-physics coupling at the reactor scale are very rare.A multi-physics core modelling based on the APOLLO3® and THEDI tools within the coupling platform C3PO was developed to simulate power transients in the CABRI reactor. A step-wise validation process has been defined and applied to this multi-physics coupling, from the numerical validation against Monte Carlo calculations at the neutron lattice level, to the integral experimental validation at the core level. Two types of transients were simulated: firstly, three neutron-driven supercritical transients (in which reactivity feedbacks are negligible), and secondly, a prompt supercritical transient, which is a high-energy transient with a five-decade power increase. The Calculation/Measurement (C/M) discrepancies on core power during supercritical transients can be explained with the support of the numerical validation results. However, the C/M discrepancies on the high-energy transient will require further investigation. A methodology to propagate nuclear data covariances in the multi-physics coupled simulation has been defined and implemented for CABRI transients. The first results of the covariance propagation on the high-energy transient showed that they cannot explain the C/M discrepancies. The impact of technological and experimental uncertainties on the transient simulation have to be analyzed. Finally, only the global core power is measured by ex-core instrumentation during the transient; it would be useful to carry out new local measurements at the pin-scale in the near future.

NuScale-like SMR Model Development and Applied Safety Analyses with the Code Chain Serpent-DYN3D-ATHLET

NuScale is an integral pressurized water reactor (iPWR) operated with light water driven by natural circulation through two helical coil steam generators (HCSGs). This work reports the safety analyses of a boron dilution and steam line break accidental sequences in a developed plant computational model based on the specifications from the Final Safety Analysis Report (FSAR). Multi-physics calculations are performed with the code chain Serpent-DYN3D-ATHLET. A state-of-the-art multi-dimensional vessel topology is developed with ATHLET for the accurate representation of flow and temperature fields, as well as spatial core power distribution. The static calculation results show agreement with the reference values from the FSAR. The boron dilution sequence shows a homogeneous core power excursion by the boron feedback, and the reactor is tripped upon “high pressurizer pressure” signal. During the steam line break sequence, the affected HCSG depressurizes rapidly and the reactor is tripped upon “low main steam pressure” signal. None of the transients violate safety margins. The adopted 3-D vessel modelling approach and applied multi-physics calculations are able to capture both transient physics within the reactor domain, and conclude that symmetric arrangement of the HCSG tubes enhance coolant mixing and prevent a heterogeneous core power excursion.

Benchmarking of the SPERT-III E-core experiment with the Monte Carlo codes TRIPOLI-4®, TRIPOLI-5® and OpenMC

This paper presents a code-to-code verification of the SPERT-III reactor in its E-core configuration using the Monte Carlo codes TRIPOLI-4®, TRIPOLI-5® and OpenMC. A SPERT-III E-core model was originally developed for the Monte Carlo code TRIPOLI-4® and will be the baseline of our analysis The simulation results obtained for the main reactor physics parameters (neutron effective multiplication factor, rod worth, reactivity coefficients and kinetics parameters) with TRIPOLI-4®and OpenMC using different nuclear data libraries are compared and analyzed. Additional verifications are carried out with the next-generation Monte Carlo code TRIPOLI-5®, which is currently under development by CEA and IRSN (France). This paper summarizes a joint study that was conducted within an international research collaboration between CEA Paris-Saclay (France) and the Nuclear Research Center Negev (NRCN, Israel), and is a stepping stone towards future multi-physics simulations exploring the SPERT-III reactor power excursions, including thermal-hydraulics feedback.

GWO- and SSA-Tuned PID Control for Frequency Regulation in Multi-Area PowerNetwork Integrated with Plug-in Electric Vehicle

ABSTRACT The Plug-in Electric Vehicles (PEVs) can be influential in containing power system frequency fluctuations. This study, therefore, investigates the efficacy of frequency regulation for PEV-integrated multi-area power network using Grey Wolf Optimizer (GWO) and Salp Swarm Algorithm (SSA) optimized Proportional-Integral-Derivative (PID) control. Instant investigation not only brings out the relative competence of GWO and SSA but also examines the impact of PEV in improving the system performance. The varying operating conditions are realized by subjecting the system to step and random load variations in either or both of the areas. With the proposed control scheme and involvement of PEV, system frequency and tie-line power excursions settle quicker with their peak swings also getting restricted to a lower value, while the oscillations are arrested as well to a great extent. Further, it’s the SSA that shows its superiority over GWO as per the simulation results executed in MATLAB.

SIMMER-IV application to safety assessment of severe accident in a small SFR

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

Transient Simulation and Parameter Sensitivity Analysis of Godiva Experiment Based on MOOSE Platform

The fast-neutron burst reactor is a chain reactor that can operate in a prompt critical state. In order to ensure the operational safety of the fast-neutron-pulse reactor and prevent the supercritical pulse from causing physical damage to the material, it is necessary to simulate and analyze the pulse operating conditions of the fast-neutron-pulse reactor. Godiva-I is a spherical assembly of highly enriched uranium metal made during the 1950s. A prompt-critical transient in such a nuclear system impels a quick power excursion, which will cause a temperature rise and a subsequent reactivity reduction because of the metal sphere’s expansion. The overall transient lasts for a few fractions of a millisecond. Based on the point kinetics and Monte Carlo method, the temporal and spatial characteristics of transient input power were calculated, the difference of the average reactivity temperature coefficient between uniform density and non-uniform density was compared, and the transient power distribution condition was loaded into the thermal–mechanics calculation of the MOOSE platform; thus, the pulse process of Godiva-I with different initial reactivity periods was simulated. The JFNK (Jacobian–Free–Newton–Krylov) direct method and multi-app indirect method were used to analyze the transient response of the pulse dynamic process using the heat conduction module and tenor mechanics module, respectively. After considering the influence of the inertia effect and wall-reflected neutrons, the simulation results were much closer to the experimental values. Based on the stochastic tools module, the uncertainty propagation and sensitivity analysis of the Godiva-I model were carried out, the uncertainty of external surface displacement of Godiva under input disturbance of material properties and heat source amplitude factor was obtained, and the sensitivity of different input parameters to output parameters was quantified. The research results can lay a technical foundation for the thermal–mechanics coupling analysis and uncertainty quantification of the metal fast reactor.

Safety considerations related to boron redistribution during Loss-of-Coolant-Accident initiated beyond design basis events for the NuScale power module

The authors, as staff at the U.S. Nuclear Regulatory Commission (NRC), evaluated the potential consequences of boron dilution under beyond design basis conditions in the NuScale Power Module to inform NRC staff’s safety evaluation report for the NuScale design certification review. The authors identified the small break loss of coolant accident with coincident failure of control rods to insert as the scenario with the most adverse boron dilution and reactivity insertion consequences. The authors evaluated the sequence of events for the scenario to determine which conditions or phenomena could produce the potential for a core flow surge to yield core damage because of a super-prompt reactivity excursion. The authors screened the conditions and phenomena that could potentially cause a core flow surge based on the dynamic speed of the incursion compared to key reactivity feedback mechanisms: thermal–hydraulic coupling via rod heat transfer and convective mixing. Based on the screening process, the authors were able to identify two processes that could potentially lead to a power excursion. However, the authors’ analysis relies heavily on assumptions regarding convective mixing, making this an important phenomenon worth careful consideration in any evaluation of boron redistribution.

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

ABSTRACT The typical initiating events of a core disruptive accident (CDA) in small sodium-cooled fast reactors (SFR) are evaluated with the computational code, SAS4A. CDA is one of the hypothetical events in the safety assessment of the SFR, as the SFR has the potential to experience power excursion due to the sodium voiding and the core degradation. To improve the safety of future SFRs, the development of SFRs with low void reactivity has been promoted. Small SFRs can have a negative void coefficient of reactivity. The analysis of the CDA event sequence in small SFRs is valuable for the investigation of the reactor characteristics for the future R&D of SFRs. The event progression during ULOF and UTOP, which are the typical initiating events of CDA, is investigated through the numerical analysis. The event progression of these accidents in the low void reactivity reactor is found to be slow due to the effective insertion of the negative reactivity feedback and the absence of significant positive reactivity insertion. No power excursion occurs in the initiating phase. In ULOF, the cladding melt and motion behavior becomes more important for the evaluation of the event progression due to its positive reactivity.

Analysis of unprotected loss of flow accident in a sodium cooled fast reactor using coupled thermal hydraulics neutronics code ASTRA

The Unprotected loss of flow accident (ULOFA) is a severe accident that may lead to core disruption in a sodium cooled fast reactor (SFR). It is investigated as a part of the defense in depth concept in nuclear safety. A coupled thermal hydraulics neutronics model is required to model the severe accident scenario where reactivity is affected by the changes in the temperature of the core components, coolant boiling, and material distribution during the transient. The coupled code previously developed for the analysis of total instantaneous blockage (ASTRA) is improved for the study of ULOFA by adding various models, viz., (i) a one-dimensional two-fluid sodium boiling model, (ii) a simple primary hydraulics model, (iii) a simple clad motion model, and (iv) an improved point kinetics solver. The boiling model is validated with different loss of flow experiments. The improved code is used to analyze ULOFA in a 500 MWe medium size sodium cooled fast reactor up to the onset of fuel melting. At 21.5s, coolant boiling is initiated in the central channel. The temperature evolution of the core components, reactivity feedbacks, power evolutions, and coolant voiding propagation are calculated. The code predicts early fuel melting at 24.9s due to the power excursion by the early voiding in the fuel channels compared to the study without including the boiling model. Boiling is started in five out of ten representative fuel subassembly (SA) channels at the time of fuel melting.

Preliminary assessment of the safety performance of Westinghouse LFR

The safety performance of Westinghouse Lead Fast Reactor (LFR) is evaluated using the SAS4A/SASSYS-1 system code coupled to the GOTHIC containment code. The Westinghouse LFR is a 950 MW(th) lead-cooled, pool-type reactor with microchannel primary heat exchangers (PHEs) directly immersed in the primary coolant. The reactor vessel (RV) is surrounded by a guard vessel (GV) to safeguard against the unlikely event of reactor vessel failure. The emergency decay heat removal is performed by the passive heat removal system (PHRS), which consists of a water pool system surrounding the GV filled with enough water to remove decay heat for the first seven days, and stacks to circulate air and remove decay heat indefinitely after the water has boiled off. The responses of the reactor system to Station Blackout (SBO) and Transient Overpower (TOP) are analyzed. During SBO, when the normal decay heat removal is assumed unavailable, the passive water-cooling transitioning to air-cooling outside of GV successfully removes the decay heat indefinitely. During TOP, where inadvertent withdrawal of most reactive control rods is postulated, the negative reactivity feedback of the core limits the power excursion to 140% of the normal operating power before the passive shutdown system is actuated and scrams the reactor. Fuel melting and cladding failure are avoided. The reactor design is evolving, and the preliminary results presented in the paper demonstrate the capability of the coupled SAS4A/SASSYS-1 and GOTHIC codes and characterize the safety performance of the Westinghouse LFR.

Safety-oriented SFR core design: Methodology and application to transient bifurcations and stability maps analyses

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commission), a methodology is proposed to early consider safety requirement in the undergoing reactor design process. Before the use of mechanistic tools (CATHARE, SIMMER, EUROPLEXUS, etc.) whose input deck elaboration requires an advanced knowledge of the reactor design, the methodology proposed in this article involves several physical tools simulating phenomena likely to govern the choice of design parameters. These tools are mostly based on reduced order models (ROM), meaning that they mainly involve low-dimensional modelings (mostly 0D and 1D) that are validated versus experimental results. They are gathered in a platform that covers all kind of accidental phenomenology, from the initiator (pump trip, reactivity insertion, local flow blockages, etc.) until the reach of a stable and coolable state after corium relocation. Thus, enabling a large number of simulations in a reasonable computational time, this fast-running platform makes possible the characterization of some major accident transient bifurcations (such as boiling onset, boiling stabilization, primary power excursion, molten fuel vaporization, corium axial relocation in transfer tubes, etc.), in terms of probability of occurrence and of consequences on the transient evolution. It also enables to identify the main physical parameters causing the bifurcations in order to allow a straight feedback on the core design and to give some orientations for future R&D studies. In this article, a focus is firstly made on some scenario bifurcations to illustrate the platform capabilities. The boiling onset and possible reactor state stabilization are studied. The possibility of primary power excursion depending on the core design and the fuel vaporization possibilities are assessed. This platform is also of great help in order to rapidly compare several core designs when facing accidental transients. An example is given by comparing a CADOR core design to an ASTRID-like core design through flow stability maps in natural circulation conditions obtained by the fast-running tool platform. These applications demonstrate the efficiency of the presented methodology integrating safety at the very first design stages, and at facilitating the safety-oriented design of SFRs.

Modelling of the expansion phase of Sodium fast reactor severe accident

Improving safety is one of the current objective of the fourth generation of sodium fast reactors (SFR). In order to achieve this requirement and increase safety margins, developments are carried-out to take into account severe accidents as soon as the pre-conceptual phase. In this context, fast-running and parametric tools are used in addition to mechanistic tools for parametric studies and probabilistic margin evaluations.This paper is dedicated to MOREINa (Model Of vapouR Expansion and fuel coolant Interaction in sodium fast Reactor), a new parametric tool modelling expansion phase transients. During a severe accident in SFR integrating a low void worth core (non energetic primary phase), a meltdown of the core may occur and form a molten material pool in the vessel. After a secondary power excursion in the molten material pool, the expansion phase implies the self-vaporisation of superheated materials and their expansion towards the sodium located in the upper plenum, above the core. Fuel Coolant Interaction (FCI) will be possible during this phase when the unvaporised overheated molten materials are mixed with the coolant, implying a huge transfer of energy from the superheated materials to the coolant ; vaporising this latter. The mechanical energy released by the two aforementioned phenomena is evaluated during this phase in order to assess the potential damage on the vessel. Thus, MOREINa integrates a new expansion phase model quantifying the mechanical energy induced by this severe accident phase in order to study its consequences on the reactor vessel walls.The modelling implemented in MOREINa, based on dimensional analysis and different balance equations, is able to simulate:- fuel and steel vaporisations;- the creation of a vapour bubble with or without fission gases inside;- the expansion of the molten materials due to the pressure difference between the saturated pressure of the molten materials and the cover gas located above the sodium, at the top of the vessel;- the integration of the molten material droplet into the bubble vapour owing to Rayleigh–Taylor instabilities at the vapour–liquid interface;- the fuel coolant interaction (FCI) if the expansion leads the molten materials to contact the coolant. First, the safety context of the expansion phase study is introduced. The new parametric tool MOREINa is then presented with its main models. Validations of MOREINa against a validated tool and on available test cases are carried-out and, finally, a parametric study is presented and analysed.

State of the Art on Doppler Broadening: Modern Developments on the Fuel Temperature Coefficient

The Doppler coefficient represents the primary source of passive and instantaneous negative reactivity feedback to limit peak power excursion during reactivity-initiated accidents as well as a nonnegligible negative reactivity source that changes between cold zero-power and hot zero-power conditions. Furthermore, the mechanism behind the Doppler coefficient may also contribute to an increase in the buildup of Pu under normal operating conditions. As such, its treatment is critical in the design and evaluation of the safety and control of nuclear systems. This paper provides a brief overview of the physical source of the Doppler effect through resonance broadening from first principles as well as an exploration of some recent developments in the treatment of elastic scattering in the Monte Carlo codes Tripoli4® and MCNP. This exploration results in a detailed look at the effect different elastic scattering kernels have on the radiative capture, fission, and elastic scattering rates as they directly tie into the calculation of the Doppler coefficient via the six-factor formula. Also provided is some insight into the propagation of the a priori uncertainty of 238U resonance parameters. This work is performed pursuant to the development of a new experimental program to measure the Doppler coefficient in a zero-power reactor both more accurately and to higher temperatures (1500°C to 2000°C) than has been done in the past at the MINERVE facility at Cadarache.

Numerical studies of power oscillations induced by sodium boiling in ESFR after ULOF transient

This paper investigates numerically a new phenomenon of power oscillations, which is found to take place after unprotected loss of flow (ULOF) in a large power European Sodium Fast Reactor (ESFR) with a sodium plenum above the core.The oscillations are induced by sodium boiling and re-flooding events and can trigger a power excursion, if the oscillation amplitude becomes stronger and the reactivity exceeds the prompt criticality during transients. Numerical results show that after the sodium boiling onset in and in the vicinity of the sodium plenum, a negative reactivity effect is introduced, that reduces the power, and the sodium boiling is stopped. When liquid sodium re-floods into the voided region, introducing a positive reactivity effect, the power increases, and then sodium boils again. This process repeats, i.e. power oscillations take place. A long numerical simulation shows that the oscillations may disappear as the cover gas pressure increases up to a certain value during the transient. This gives us an idea that an increased cover gas pressure could prevent the oscillation for the considered case, simply because the sodium boiling is suppressed. An additional transient simulation with a higher initial cover gas pressure confirms this idea.

Implementation of CO2 and PbLi as working fluids in RELAP5/MOD3.3 towards accident analysis of COOL blanket for CFETR

The supercritical CO2 cOoled Lithium-Lead (COOL) blanket is an advanced blanket candidate for the Chinese Fusion Engineering Testing Reactor (CFETR). The COOL blanket has dual coolant. Supercritical CO2 (S-CO2) is used to cool the structural components, while lithium-lead (PbLi) is self-cooling in the breeding zones. To verify the reliability of the COOL blanket, it is essential to conduct the accident analysis. The system code RELAP5/MOD3.3 is developed with implementation of the two working fluids of the COOL blanket, namely PbLi and CO2. The outboard blanket segment of the COOL blanket is modeled using the modified RELAP5. The plasma pulse operation and fusion power excursion transient caused by Multifaceted Asymmetric Radiation From the Edge (MARFE) are simulated to study the thermal dynamic behavior of the COOL blanket. Furthermore, sensitivity analysis of Loss of Flow Accident (LOFA) is carried out. The results show that the outlet temperature of PbLi under steady-state operation meets its design requirements, and the temperature of the structural material does not exceed its temperature limit. The thermal hydraulic parameters have pulsed characteristics under plasma pulse operation. The MARFE event can cause sharp temperature rise of the First Wall (FW) but the material remains below its temperature limit. In case of LOFA of the CO2 system, it is suggested to restore CO2 mass flow rate in time to prevent the FW from overheating when the outlet temperature of PbLi is 973 K. The results of the preliminary accident analysis can provide the basis for the design and optimization of the blanket structure and system.