Reactivity Initiated Accident Research Articles

A study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions

ABSTRACT In a part of the historical reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the condition of such fracture pattern change. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of ℇz/ℇθ = 0.5–1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio ℇz/ℇθ >0.75 and a concurrent hoop strain < 1.7% at the failure instant.

Thermal-Mechanical performance analysis of a petal-shaped fuel rod under RIA condition

This paper establishes the performance analysis model of a petal-shaped fuel rod (PSFR) under Reactivity Initiated Accident (RIA) condition based on thermal elastic–plastic theory and finite element method. The results show that the temperature changes of fuel and cladding are basically synchronized, and the temperature extremes appear in the central region, with the distribution characterized by a large middle and two small sides. The stress is concentrated in the center region of the cladding. At the beginning of operation, the stress at the lobe root of cladding is the largest. With the increase of operation time, the stress concentration at the lobe tip is gradually obvious. At 150th days of operation, a sudden RIA occurred and the power factor of the PSFR rapidly increased over 5.38 times and the cladding stress exceeded the yield strength. The axial and circumferential center cladding lobe roots and nearby may be at greater risk of plastic deformation. These results will provide an important theoretical basis for the determination of the deformation and accident operation risk of PSFRs.

Numerical simulation of transient boiling and void cross flow in non-uniformly heated 5 × 5 rod bundle

A multi-dimensional numerical simulation has been conducted for a transient boiling experiment by applying rapid and non-uniform power to fluid in a 5×5 bundle channel simulated a BWR fuel assembly in atmospheric pressure. The experiment is to investigate thermal hydraulic behavior in applying transient power to subcooled reactor coolant in a stand-by mode of a nuclear power plant such as reactivity initiated accident (RIA). The numerical simulation has been conducted by TRACE (version 5.0 / patch 5) which is a nuclear system analysis code to solve thermal hydraulics of boiling two phase flow by a two fluid model. The bundle channel in the upward vertical direction has been simulated by a three dimensional component with the Cartesian coordinate system attaching heat structures to simulate heating rods for thermal conduction and heat transfer calculation. It has been revealed that the numerical result is possible to simulate the void cross flow through the sub-channel in the bundle channel qualitatively as observed in the experiment. However, there was room for improvement for a wall heat transfer model in TRACE because the numerical result is larger than the experimental result in terms of the rising rate of the rod temperature during the rapid initial increasing of applied power, and it has been inferred that the evaluation model for an onset of nucleate boiling in TRACE also has to be improved because the void initiation time in the numerical result was earlier than the experimental result. Therefore, TRACE has been optimized by modifying the two physical models described above. The numerical results with the modified TRACE have been in better agreement with the experimental results than those with the original TRACE. Using the modified TRACE, a numerical experiment has been conducted putting an inlet velocity and subcooling of fluid and applied power as parameters in addition to the experimental conditions. The results have shown that there is a square law correlation between a void cross flow and applied power to the fluid. The experiments and analyses in this study have elucidated the behavior of transient boiling two phase flow in a fuel assembly of a nuclear power plant at the transient event such as RIA.

Thermomechanical calculation within the HERA programme

HERA (High Burnup Experiments in Reactivity Initiated Accident) is one of the programmes which should replace closed Halden reactor. As the name implies HERA is focused on the tests simulated RIA (Reactivity Initiated Accident) and the aim of the programme is to investigate the performance and behaviour of high burnup fuel in RIA transients. The first phase of HERA programme consists of two steps which are the blind calculation of different RIA scenarios (different pulse width, enthalpy increase, hydrogen content and gap size) and in-reactor test with fresh fuel.The set of thermomechanical calculation was performed with the FRAPTRAN computer code as part of the first phase of HERA. The influence of several parameters on the fuel behaviour was observed, the varying parameters were the pulse width, peak radial average enthalpy increase, hydrogen content and the hydrogen rim size and gap size. In addition, the effect of type and setting of boundary condition was investigated.

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

This paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO2. The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB2 Integral Fuel Burnable Absorber (IFBA) and B4C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd2O3-doped UO2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

Multi-physics coupled analyzes of research nuclear reactors based on steady-state and kinetics models

Nuclear research reactors are the workhorses of nuclear material irradiation and isotope production, but the complex assembly geometries and physical coupling effect adds difficulties to their accurate safety analysis. In this study, a multi-physics coupled method for assessing the steady-state and kinetic behaviors of the LVR-15 and MPRR research reactors is proposed. Two-dimensional whole-core Monte Carlo models are utilized to produce homogenized assembly cross-sections in order to describe complex core geometries. Utilizing a three-dimensional variational nodal transport method with unstructured triangular-z meshes, the neutronic field is calculated. Besides, a multi-channel model and a finite difference heat conduction model are employed to determine the temperature distributions of coolant and fuel. Based on these models, the MORPHY (Multi-purpose Reactor multi-PHYsics analysis) code platform is created and verified using the NEACRP neutronics/thermal-hydraulics benchmark. The steady-state and transient behaviors of LVR-15 and MPRR are analyzed using MORPHY. The steady-state results are found to agree well with Monte Carlo reference solutions. For LVR-15 and MPRR, the keff errors are 377 pcm/383 pcm, the maximum assembly power errors are 6.49%/10%, and the maximum control rod worth errors are 31 pcm/136 pcm, respectively. In the meantime, transient processes including unprotected loss-of-flow accidents and reactivity-initiated accidents are studied. Simulation results for both reactors exhibit reasonable negative feedback phenomena. Moreover, it is discovered that despite the fact that the two reactors have distinct core designs, their accident-related behaviors are similar due to the use of the same types of fuel material and water coolant.

Transient fuel behavior analysis of UN fuel with a two-layered SiC cladding based on multiphysics method

Recently, UN fuel has emerged as a promising alternative to UO2 fuel due to its exceptional thermophysical attributes, including heightened thermal conductivity, elevated melting point, and increased fissile density. Meanwhile, SiC is recognized as a member of the next generation of accident-tolerant cladding materials owing to its impressive strength at high temperatures, resistance to creep, minimal thermal expansion, and improved irradiation characteristics. In this study, utilizing our developed fuel performance analysis code CAMPUS, we integrate a multiphysics model featuring UO2 fuel paired with a two-layered SiC cladding into the CAMPUS framework. Subsequently, we conduct simulations and analyses of the fuel performance for combinations such as UO2-two-layered SiC, UN-two-layered SiC, UN-Zircaloy, and UO2-Zircaloy. These evaluations encompass both standard operation scenarios and potential accident conditions within a pressurized water reactor (PWR). Notably, the incorporation of UN fuel demonstrates a significant reduction in maximum fuel temperature during normal operation, as well as under Loss-of-Coolant Accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. However, upon introducing the two-layered SiC cladding, it becomes apparent that its temperature performance is somewhat inferior to that of Zircaloy under normal operation and RIA scenarios. Conversely, when subjected to a LOCA scenario, the thermal and mechanical characteristics of the two-layered SiC cladding outperform those of the Zircaloy cladding.

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

A reactivity-initiated accident (RIA)-simulated test CN-1 on a high-burnup 64 GWd/t mixed-oxide fuel rod sheathed with M5™ cladding was conducted at the Nuclear Safety Research Reactor, resulting in fuel failure. A small opening with slight ballooning deformation characterized the post-test visual appearance of the test fuel rod. Simulation using fuel performance codes FEMAXI-8/RANNS predicted rod survival under early phase loading induced by pellet-cladding mechanical interaction and subsequent boiling transition, and the cladding-surface temperature measured online confirmed the occurrence of boiling transition. The experimental observation and simulation indicate that the failure was caused by a high-temperature rupture following increased rod-internal pressure. The RANNS sensitivity analysis revealed that a mechanical state parameter dedicated to predicting plastic instability might be an effective index for evaluating the risk of rupture failure during RIAs.

Neutronic and thermal-hydraulic calculations under steady state and transient conditions for a generic pebble bed fluoride salt cooled high-temperature reactor

Here, we develop multiphysics coupled neutronics/thermal-hydraulics models for a recently available generic Fluoride-salt-cooled High-temperature Reactor (FHR) at equilibrium core burnup conditions with reactivity control mechanisms. We utilized the pebble bed reactor transient tool KP-AGREE with multigroup homogenized cross-sections generated by the continuous energy Monte Carlo neutron transport code Serpent to simulate transient events for an FHR. Comparing global reactor parameters and safety characteristics between Serpent and published results to standalone neutronic calculations in KP-AGREE verifies our initial conditions for our transient models. This work explores three separate transients: a loss of flow accident, an overcooling accident, and a total control rod withdrawal accident. The boundary conditions of these scenarios come from either transient benchmark definitions of a gas-cooled pebble bed reactor or transient analysis of a similar FHR design. For the loss of flow accident, our predicted peak fuel and coolant outlet temperatures show a similar response to published values, albeit slightly higher, with differences between values attributed to our modeling assumptions. The overcooling transient simulation predicts a 52 K temperature increase in the maximum fuel kernel temperature and an approximately 32.6% increase in total core power due to the negative coolant temperature feedback of the FHR design. Our control rod removal transients predicted a 235 K increase in maximum fuel kernel temperature and a 107.5% increase in total core power. For all transient cases, the fuel temperature remained significantly lower than the temperature limit, but coolant outlet temperatures for the reactivity-initiated accident exceeded external core material temperature limits.

Resumption of water capsule reactivity-initiated accident testing at TREAT

A series of integral reactivity-initiated accident commissioning experiments were completed in a new static water capsule in the Transient Reactor Test Facility (i.e., TREAT), marking the first such tests in the United States in more than 40 years. The test campaign included a verification test followed by five tests in the Static Environment Rodlet Transient Test Apparatus capsule. The capsule’s initial conditions varied from room temperature and pressure up to 200 °C and 2.5 MPa, with energy depositions varying between ∼ 500–1,100 J/gUO2. The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. Post-transient examinations such as gamma emission spectroscopy, profilometry, and microscopy were performed to document the end state of the fuel rods. The results from the experiments show that the fuel rodlets behaved very similar to historical tests under similar energy depositions. This paper documents the design of the capsule and highlights some results from the commissioning tests and post-transient examination.

A plasma-based pressure pulse generator for simulating pellet-cladding mechanical interaction during reactivity-initiated accident.

This paper describes the characteristics of a plasma-based pressure pulse generator and its potential use as a simulator for studying pellet-cladding mechanical interaction (PCMI) during reactivity-initiated accidents (RIAs). In this device, a transient pressure pulse is generated by rapid heating and expansion of hot, dense plasma inside a nuclear fuel cladding. Thus, the parameters of a pressure pulse, such as peak pressure and pressure rise-rate, can be controlled by modifying the electrical parameters of a pulse discharge circuit. The pulse discharge circuit utilizes a capacitor bank comprising several energy storage capacitors connected in parallel and a high-power solid-state switch. A pressure loading system is attached as a load to the pulse discharge circuit. The power and energy delivered to the load are calculated by measuring the voltage and current waveforms at one end of the loading system. A piezoelectric sensor is connected at the other end of the loading system to simultaneously measure the pressure pulse inside the cladding tube. Preliminary experiments are carried out with a stainless-steel tube to characterize the performance of the device as well as with a pre-hydrided ZIRLO™ cladding tube to demonstrate the potential of the device as a simulator for studying the failure characteristics of the cladding as a result of an RIA. The high pressurization rate of the device is expected to offer unique advantages for studying the PCMI mechanism.

Numerical simulation of mechanical interaction between pellet and zircaloy-4 cladding under reactivity-initiated accident conditions

Reactivity-initiated accident (RIA) is one of the accidents considered by fuel safety criteria. The power pulse caused abrupt expansion of the pellet, leading to pellet and cladding mechanical interaction (PCMI). Cladding ductility is gradually decreased in the process of oxidation and hydrogen uptake during irradiation. As a result, the extrusion of the pellet to the cladding causes the cladding to produce greater hoop stress and hoop strain, which leads to damage. A simulation of PCMI using ABAQUS was designed to analyze cladding mechanic properties. By comparing the calculation results of the simulation scheme with the data obtained in the reactor experiment, it is proved that the simulation calculation results are in good agreement with the experimental data in the reactor, and the rationality of the simulation scheme is verified. Based on the out-pile mandrel equipment simulated PCMI, the corresponding finite element model was established to study related test parameters.By analyzing the results of the simulation calculation outside the reactor, it is found that the pellet material, cladding temperature, compression rate, tensile load, friction coefficient and the cladding hoop stress and hoop strain have a certain relationship: (1) The choice of pellet material affects the reactor. The overall law of hoop stress-strain in the external simulation results basically has no effect. (2) The temperature of the cladding has effect on the hoop stress peak. (3) The compression rate has effect on the time for the cladding to reach the hoop peak stress. (4) Tensile load will affect the starting position of the hoop stress-strain of the cladding. (5) The friction coefficient has a significant influence on the stress and strain trend of the cladding. The outside experiment can be carried out more effectively, and it can provide guidance and a basis for the outside experiment.

Analysis of Control Element Assembly Withdrawal at Full Power Accident Scenario Using a Hybrid Conservative and BEPU Approach

Reactivity Initiated Accident (RIA) scenarios require special attention using advanced simulation techniques due to their complexity and importance for nuclear power plant (NPP) safety. While the conservative approach has traditionally been used for safety analysis, it may lead to unrealistic results which calls for the use of best estimate plus uncertainty (BEPU) approach, especially with the current advances in computational power which makes the BEPU analysis feasible. In this work an Uncontrolled Control Element Assembly (CEA) Withdrawal at Full Power accident scenario is analyzed using the BEPU approach by loosely coupling the thermal hydraulics best-estimate system code (RELAP5/SCDAPSIM/MOD3.4) to the statistical analysis software (DAKOTA) using a Python interface. Results from the BEPU analysis indicate that a realistic treatment of the accident scenario yields a larger safety margin and is therefore encouraged for accident analysis as it may enable more economic and flexible operation.

Investigation of plate fuel performance under reactivity initiated accidents with developed multi-dimensional coupled method

A reactivity-initiated accident is a typical nuclear reactor accident during which the core fission rate and power increase unexpectedly, making the temperature of the fuel element soar, which causes a decreasing strength, large deformation, and even the failure of the element. To simulate the performance of the plate-type fuel during RIA, a neutronics–thermal hydraulics–mechanics coupling method was implanted in a fuel performance code BEEs-Plates. The coupling scheme was firstly verified and validated against the MTR benchmark. Then the RIA results with the consideration of the deformation effect were compared with the benchmark case, showing that the mechanics as well as the burnup effect have little impact on the neutronic–thermal hydraulics results when using the constant thermal properties. Lastly, the thermal-mechanical analysis was conducted for the fuel plate at different burnups (0% FIMA, 10% FIMA, 20% FIMA) under the RIA. The results show that the peak stress of the cladding increases with the burnup increases while it has an opposite trend for the fuel meat. Some special features such as over-shrinkage of the fuel meat have also been noticeable and analyzed.

Sensitivity Analysis of Transient Critical Heat Flux by RIA Under High-Pressure Flow Boiling Conditions in TRTL

A reactivity-initiated accident (RIA) is a design-basis accident under which critical heat flux (CHF) is likely to be exceeded. The operational margin for RIAs is currently determined using steady-state CHF lookup tables, which provide conservative estimates relative to transient CHF phenomena. The Transient Reactor Test Loop (TRTL) facility at Oregon State University is capable of performing out-of-pile rapid heating experiments representative of a RIA at conditions representative of a pressurized water reactor (PWR). To further our understanding of and ability to predict transient CHF under PWR conditions, we performed a sensitivity analysis on a RELAP5-3D model of the TRTL facility coupled to the RAVEN code framework to define a proposed experimental test matrix to be performed at the TRTL facility. We then implemented a flow boiling CHF correlation into RELAP5-3D and performed a secondary sensitivity analysis inspecting the impact of the built-in RELAP5-3D CHF and heat transfer multipliers on both the prediction of CHF and key safety parameters, such as peak cladding temperature and heat flux. The results show that the multiplier with the highest influence toward the prediction of CHF occurrence and the safety parameters is the transient CHF multiplier. Operational performance envelopes have been developed for each of the test matrix cases and will be used for validation once the experiments are performed. The TRTL facility is currently performing shakedown testing to verify system performance prior to proceeding with the experimental campaign. Restart testing results include pump curve restart testing, pressure tests, and heater rod thermocouple transients.

Multiscale identification and NTFA reduction of the elasto-viscoplastic polycrystalline behaviour of uranium dioxide (UO2) for wide range of loading conditions

This numerical study presents a micromechanical modelling of the behaviour of uranium dioxide fuel (UO2), a polycrystalline ceramic used in nuclear power plants, for loading condition typical of a reactivity initiated accident (RIA). Above a specific temperature, UO2 has an elasto-viscoplastic behaviour with strain hardening depending on both temperature and macroscopic strain rate (T,ε¯˙).In a first step, via a full-field approach, an inverse identification of the local single crystal evolution law is performed based on experimental data obtained at the macroscopic scale [4,5]. Beyond the macroscopic behaviour of the fuel pellet, it is sometimes necessary to go down in scale to study local phenomena. A first approach would be to perform a Finite Element Method squared to two (FEM2) resolution [37]. It consists to solve a thermomechanical problem on a first mesh defined at the pellet scale where at each integration point is solved another thermomechanical problem on a Representative Volume Element. The full-field approach guarantees access to the local field at the expense of relatively high computational time. It is not conceivable within industrial codes.In a second step it appeared necessary to replace the second full-field resolution by a model reduction method based on the Nonuniform Transformation Field Analysis (NTFA) approach. This method allows access to macroscopic quantities and local fields while drastically reducing the computational time. The NTFA model [21] with tangent second order (TSO) approximation [24,21] is developed and applied to the problem at hand. Two decompositions of the hardening variable are studied. It is taken as in the reference [21], as uniform per grain or taken as decomposed, like modes associated with the viscoplastic strain tensor. Then, work is done to integrate the specific loading conditions, i.e. temperatures and strain rates of the RIA, into the NTFA model while keeping the number of internal variables as low as possible. Finally, the results from the experimental, full-field and NTFA TSO models are compared. It is checked that the numerical results are not deteriorated either macroscopically or locally at the macroscopic and local scales for uniaxial and triaxial loading.

Safety analysis of VVER fuel rod modelling in reactivity initiated accident

The paper briefly reviews the behaviour of WWER type fuel rods studied under conditions of a high rate of power rating of fuel; the investigations were carried out to validate the licensing criteria with the aim of assessing design basis reactivity initiated accidents (RIA).The procedure used to calculate thermomechanical and corrosion behaviour of fuel rods in accidents is described.The results acquired from the analysis of the thermomechanical condition of WWER-1000 fuel rods in a design basis reactivity initiated accident caused by an ejection of a regulation element of the control and protection system (RE CPS) are discussed.The aim of the analysis is to validate the fulfillment of specific safety requirements imposed on fuel rod condition in design basis accidents.Consideration is given to accidents attended with a rise of fuel rod (fuel and/or cladding) temperature in comparison to the state under normal operating conditions. This may proceed from deterioration of heat transfer from fuels, e.g., in accidents due to loss of the coolant by the primary circuit or an increase in power rating of fuel in RIA.Design procedures have been worked out to simulate fuel rod behaviour in accidents, which were implemented in the RAPTA-5 fuel performance code. Using the latter, design analysis of fuel rod behaviour under design RIA conditions is implemented; its results serve to draw conclusions whether acceptance criteria are fulfilled.The results of the RAPTA-5 code design analysis of the thermomechanical state of WWER-1000 fuel rods in that accident have demonstrated the fulfillment of all the acceptance criteria.

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.

Optimization for fuel loading and reactivity initiated accident analysis for TRR-1/M1

TRR-1/M1 reactor core consists of two types of standard TRIGA fuel, 8.5% and 20% weight uranium. Periodically, the reactor core configuration is rearranged in order to optimize its performance. The objective of this work is to show that it is safe to operate TRR-1/M1 at 1 MW steady state condition for the new core configurations of which the 8.5% weight fuels in the B ring are replaced by high burn up 20% weight fuel elements. To ensure sufficient safety margins of the reactor operation, thermal–hydraulic computer codes, i.e., COOLOD-N2 and EUREKA-2/RR, were coupled with neutronic calculation based on Monte Carlo N-Particle Transport concept in order to analyse some crucial operating parameters, e.g., coolant and fuel temperature in the reactor core as well as Departure from Nucleate Boiling Ratio (DNBR) along the fuel axial direction of the hottest channel. From this study, it can be concluded that no significant effect on the reactor operation was found when replacing the 8.5% weight fuel elements in the B ring with the high burnup 20% weight fuels. Transient phenomena of Reactivity Initiated Accident (RIA) were also performed for the rapid removal of a sample from the reactor core with reactivity insertion rate of $1.00/s. It is found that the negative temperature coefficient of TRIGA fuel provided safety for the reactor even during the prescribed reactor transients. The calculation showed slightly increase in the fuel temperature during the short period of reactivity insertion which caused insignificant energy release to heat up the fuel during the accident. The maximum fuel temperature found during the accident in this study was well below TRR-1/M1 safety.

Experimental Analysis of Fast-Transient Flow Boiling in Various Flow Conditions Using Inverse Heat Conduction Calculation Method

Insufficient thermal-hydraulic knowledge for analysis of a reactivity-initiated accident demands experiments of fast-transient flow boiling heat transfer from moderate- to high-pressure conditions. In this study, those experiments are conducted for vertical upward tube flows of pressurized water. The tube wall is joule heated by stepwise electric pulse power to achieve an abrupt wall heating condition. The applied pulse power is varied from 4.68 to 13.59 GW/m3, which is beyond the power required for steady-state critical heat flux (CHF) to occur. Rapid evolution of the boiling wall temperature is extracted from outer wall temperature data by solving an inverse heat conduction problem. As a result, with increasing the applied pulse power, the time to occurrence of departure from nucleate boiling gets shorter, and the corresponding peak heat flux increases over the steady-state CHF, which is evaluated at the same flow condition. A logarithmic relation between the wall heating rate and the CHF increment ratio is also demonstrated. The effects of pressure, inlet subcooling, and mass flux on the transient peak heat flux are also investigated. As the pressure increases, the nucleate boiling duration gets shorter with decreasing peak heat flux. On the other hand, as the inlet subcooling increases, the nucleate boiling duration gets longer, and the peak heat flux increases. Contrarily, the mass flux does not show any noticeable effects on the transient heat transfer evolution.