Prototype Fast Reactor Research Articles

Numerical analysis of duct wall failure in a prototype FAIDUS design-based generation IV SFR fuel subassembly under flow blockage accident

An ingenious prototype Fuel Assembly Inner DUct Structure (FAIDUS) design is conceptualized under the Generation IV safety philosophy to prevent large-scale corium pool formation following total coolant flow blockage accident. The early failure of the duct wall provides a built-in guideway for molten material relocation out of the core region, ensuring inherent safety robustness. Therefore, understanding the duct wall failure mechanism is critical for evaluating accident mitigation potential of the FAIDUS design. In this regard, a transient enthalpy-based multiphase computational model is developed to investigate the heat transfer and fluid flow behavior involved during the duct wall failure in 3-D domain. The numerical model is employed initially to simulate wall failure in the EAGLE ID1 in-pile test. The model results are validated against the in-pile test data. Subsequently, FAIDUS duct wall failure in a prototype fast reactor fuel subassembly (SA) is investigated employing the present model. The study reveals that the large heat flux from the molten pool to the duct is the main reason for the duct wall failure. The duct failure starts with an initial rupture across the corners of the hexagonal-shaped duct and progresses to total failure of the duct wall by the expansion of the rupture over the whole duct wall area with time. The outer hexcan wall sustains only minor thermal damage during the duct wall failure process. The present study explores the associated thermal-hydraulics, molten pool dynamics, crust formation behavior, and event sequences involved in duct wall failure in great detail. The simulation findings suggest that the FAIDUS SA design facilitates premature failure of the slim duct wall prior to the hexcan wall failure. This could potentially enable early discharge of the molten corium away from the core region, limiting the accident progress.

Evaluation of Reaction Rate Distribution for Shielding Region in the Prototype Fast Reactor Monju

The reliability and usefulness of the reaction rate distribution data measured in the prototype fast breeder reactor Monju were examined through a comparison with a calculation using JENDL-4.0, mainly focusing on shielding regions around the reactor core. The reaction rates of 238U (n,f) and 58Ni (n,p) sensitive to high-energy neutrons were all judged reliable. The calculation-to-experiment values were slightly worse in the shielding regions; however, those for the 58Ni (n,p) reaction rates were improved by employing JEFF-3.3 instead of JENDL-4.0. A different tendency was observed between the two reactions, probably due to the 238U (n,f) cross section in the energy range of around 700 eV. The reaction rates of 235U (n,f), 239Pu (n,f), 238U (n,γ), and 197Au (n,γ) sensitive to the lower-energy neutrons were mostly judged reliable. The data in the lower axial shielding region are less reliable but may be acceptable for the shielding calculation.

Evaluation of Control Rod Interaction Measurement in the Prototype Fast Reactor Monju

For evaluating control rod (CR) interaction, CR worth values with and without the insertion of interaction CRs were measured in six CR interaction patterns by the continuous period method and the compensation method in the prototype fast breeder reactor Monju. The measured CR worth was evaluated in detail by conducting corrections, such as the thermal contraction of CR drive shafts, and compared with calculations to examine the reliability. The calculation-to-experiment values (C/Es) of examined CR worths with and without the interaction were 0.96 to 0.98 ± 0.03. The values of the measured CR interactions were obtained from those examined CR worth measurements as a change in target CR worth by the insertion of interaction CRs: 5.0 ± 0.6% to 10.1 ± 0.1%. The C/Es of the interaction in those cases were 0.78 to 0.93 within 5% to 27% uncertainty. Through this study, the consistency and reliability of the measurements and the calculations of the CR interaction were confirmed, and the measurements were judged useful.

Evaluation of Sodium Radioactivity in the Primary System of the Prototype Fast Reactor Monju

In this study, the sodium radioactivity of 24Na and 22Na in the primary system measured in the prototype fast breeder reactor Monju was evaluated, and the reliability of the measurements and calculations was examined. The calculated-to-experiment values and their uncertainties for 24Na and 22Na radioactivities were 0.97 to 1.07 and 8.1% to 11.0% and 1.03 to 1.16 and 23.3% to 24.1%, respectively, using the JENDL-4.0 nuclear data library. The 22Na radioactivity calculated with ENDF/B-VIII.0 was larger by 40% than those calculated with JENDL-4.0 and JEFF-3.3 due to the 23Na(n,2n) cross-section discrepancy. The importance of the 22Na neutron capture effect is also confirmed herein for the accurate evaluation of the 22Na radioactivity. The experimental data were judged to be useful for validating the calculation method for improving the reliability of future designs of sodium-cooled fast reactors.

Quantifying water in spent fuel assemblies with neutrons: A simulation-based approach

A simulation-based study concerning three non-destructive approaches by which water in spent nuclear fuel assemblies might be quantified with neutrons is described. Three fuel types have been considered: thirty-six spent Advanced Gas-cooled Reactor (AGR) fuel pins contained in a stainless-steel can; a prototype fast reactor (PFR) spent fuel assembly and a light water reactor (LWR) spent fuel assembly – with the PFR and LWR assemblies containing mixed oxide fuels. The three approaches are investigated with MCNP5 based on neutron interrogation: fast neutron detection of the perturbed fast neutron flux and two approaches based on the 2.223 MeV γ-ray photon emission resulting from the 1H(n,γ)2H capture reaction, stimulated with moderated and unmoderated neutrons. For each of these approaches the following perspectives have been considered: neutron transmission, the influence of radiation emitted by the associated spent fuel inventory, and the dependencies on water quantity and location. The moderated γ-ray assay technique is observed to return the most significant distinction between scenarios with and without water in the fuel assemblies, notwithstanding the need for efficiency response functions to be overlaid on the neutron models as part of the interpretation of the MCNP model results. Water present in all of the spent fuels considered is observed to perturb the incident neutron flux to a degree significant within the simulation uncertainties. The AGR and PFR cases are not undermined by the intrinsic neutron field from the fuel; whereas for the LWR-MOX case, the simulations suggest that this will limit its potential. Successful detection of water quantities down to 10 g is anticipated for AGR, whereas reduced mass sensitivity is forecast for PFR and LWR-MOX, where it is assumed that the water will be more dispersed. Water located in fewer locations is observed to be harder to discern, suggesting that small, localised quantities of water could pose a challenge for these techniques.

Assessment of Impact Strength of Welds Produced by Cold Wire and Hot Wire Gas Tungsten Arc Welding (GTAW) Processes

Modified 9Cr 1Mo material is used as principal material of construction of Steam Generators (SG) for 500MWe Prototype Fast Breeder Nuclear Reactor (PFBR), at Kalpakkam, India. One of the major problems associated with welding of modified 9Cr 1Mo steel is poor impact strength of weld metal produced from the processes that involves the use of flux. Due to this reason, 100% Gas Tungsten Arc Welding (GTAW) process is chosen for fabrication of PFBR Steam Generators.The major limitation of conventional cold wire GTAW process is low deposition rate which makes it highly uneconomical especially for welding component with large section thickness. Therefore, hot wire GTAW process is selected for welding of PFBR Steam Generators wherever possible in addition to the conventional cold wire GTAW process. The impact properties of modified 9Cr 1Mo welds produced by hot wire GTAW process are compared with impact properties of cold wire GTAW welds. The impact strength of welds produced by hot wire GTAW process is found to be inferior compared to that of welds produced by cold wire GTAW process. Nevertheless, hot wire GTAW system has the benefit of high weld deposition rate with excellent profile of weld with reduced dilution and porosities

A robust multi phase model to investigate molten material relocation during total flow blockage in SFR fuel subassembly

A transient, multi-phase enthalpy based computational model is developed to analyse the streaming and freezing characteristics of the molten fuel in a blocked fuel subassembly (SA) during Total Flow Blockage (TFB) at the foot of the SA. During TFB accident progression, a molten material pool is formed following the meltdown of fuel pins. The molten fuel has the potential to relocate axially through the available subchannels in the downward direction. The mathematical model comprises of two modules; a Fluid Dynamics module and a Heat Transfer module. The model adopts both implicit and explicit type of finite difference techniques employing a variable grid system for boundary tracking. The mathematical model has been validated against benchmark experimental data and SIMMER III code prediction reported in open literature. The developed model is employed to simulate the relocation of molten fuel in a prototype fast reactor fuel SA under accident condition. The simulation results indicate that the molten fuel is unable to penetrate the axial blankets in fuel SA without freezing and forming a blockage mainly owing to the rapid freezing of melt. Further, a parametric analysis is carried out to ascertain the influence of critical parameters on the relocation and freezing characteristics. It is observed that the design of the fuel SA has a potential effect on the streaming behaviour and enhancement in fuel relocation is substantially achieved by dwindling the fuel pin size, especially axial length and diameter of the pin, thus preventing the accident progression to the neighbouring SAs.

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

ABSTRACT As a part of safety assessment or design of steam generators of sodium-cooled fast reactors, it is necessary to evaluate the water leak rate under sodium–water reaction accident. The computer code called LEAP-II calculating a design basis water leak rate during long-time event progress including self-wastage, target-wastage, wastage-type failure propagation, water leak detection, and water/steam blowdown was developed for the prototype fast reactor in the past studies. In this study, a numerical analysis method to predict occurrence of overheating tube rupture was constructed and integrated into this code to expand its application range. The newly constructed method consists of the elemental analysis models for temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer at the tube wall, temperature and stress of the tube, and failure of the tube. Applicability of the method was investigated through the numerical analysis of the experiment on water vapor discharging into liquid sodium pool under the actual condition of the steam generator. The numerical analysis demonstrated that the method could provide the appropriately conservative result on the overheating-rupture-type failure propagation.

Coupled system dynamics and computational fluid dynamics simulation of plant transients in sodium cooled fast reactors

Abstract Analysis of transient thermal hydraulic behavior of the plant during various design basis events is important for (i) demonstrating safety, (ii) design of plant protection system, (iii) deciding safety parameters, (iv) design specification on response time of systems and instruments, (v) estimation of transient thermal loading on various structures and (vi) meticulous planning of plant operating strategy. Traditionally these studies are carried out using one-dimensional system dynamics models. Due to the presence of large sodium pools in the primary circuit of prototype fast reactors, there is a possibility of significant stratification effects which may govern the transient under certain plant conditions. Traditionally, sodium pools are treated as single well mixed volumes or network of one dimensional control volumes in system dynamics models. One-dimensional models fail to simulate stratification effects leading to unrealistic transient simulation. An integrated code system with coupling of one dimensional and multi-dimensional models is an optimum approach to overcome this difficulty. This paper discusses development of multi-dimensional transient CFD code for turbulent flow simulations in stratified sodium pool of a fast reactor and integration of the same with one dimensional system dynamics models for the rest of the plant. The predictions of the improved code are compared against those of system dynamics model of entire plant.

Numerical Simulation and Analysis for the Decay Heat Exchanger of China Prototype Fast Reactor

Decay heat exchanger (DHX) is an essential piece of equipment of the decay heat removal system (DHRS) in a pool-type sodium-cooled fast reactor. The main function of DHX is to transfer heat from primary circuit fluid (liquid sodium) to secondary circuit fluid (liquid sodium) of DHRS, which can reduce the high temperature caused by decay heat in the reactor core. In this paper, CFD method was used to simulate the statistic three-dimensional flow field of DHX which is used in China Prototype Fast Reactor under several different operating conditions. The numerical results under rated operating conditions were compared with the design values and the comparison was fit in with expectation, which demonstrated the validity of the numerical model. The study of the flow field of DHX under designed operating conditions was conducted to obtain the heat transfer features and the flow characteristics. The results show that the design of DHX in China Prototype Fast Reactor is sufficient to the purpose to timely bring the decay heat out of the reactor core and DHX meets the safety design requirements from the perspective of heat transfer. The work in this paper provided references for further design improvement of the DHX in a pool-type sodium-cooled fast reactor.

Immobilisation of Prototype Fast Reactor raffinate in a barium borosilicate glass matrix

The vitrification of Dounreay Prototype Fast Reactor (PFR) Raffinate in a barium borosilicate glass matrix was investigated, with the aim of understanding process feasibility and the potential benefits over the current baseline of cement encapsulation. Laboratory scale glass melts demonstrated the production of homogeneous glasses incorporating at least 20 wt% simulant PFR waste (on an oxides basis), with no detectable crystalline accessory phases. The hardness and indentation fracture toughness of the simulant PFR waste glasses were determined to be comparable to those of current UK high level waste glass formulations. The normalised dissolution rate of boron from the simulant PFR glasses was determined to be 3 × 10−2 g m-2 d-1, in 18.2 MΩ water at 90 °C and surface area/volume ratio of 1500 m−1, only a factor of two greater than the French SON-68 simulant high level waste glass, under comparable conditions. Consequently, the simulant PFR waste glasses show considerable promise for meeting envisaged waste acceptance criteria for geological disposal. Overall, the superior stability of vitrified PFR wasteforms could enhance the safety case for long term near surface storage of radioactive wastes, mandated by current Scottish Government policy.

Numerical study on structural integrity of inner barrel caused by thermal stratification in upper plenum of Monju

In Japanese prototype fast reactor, Monju, an inner barrel with several flow holes is placed at an upper plenum adjacent to a core outlet. When the reactor scram occurs, a cold coolant flows into the bottom of the upper plenum through the core outlet and thermal stratification will appear at the upper plenum. And thus, the inner barrel may be damaged by a thermal stress due to thermal stratification. In this study, a structural integrity assessment method is developed based on fluid-structure interaction analysis and cumulative damage rule. First, a three-dimensional thermal-hydraulics analysis is conducted to simulate a turbine trip test from 40% power operation. Full power output conditions are also simulated by modifying conditions of 40% power output conditions. Next, the thermal stress analysis is modified by adding a practical condition, such as a bending stress. Then, the thermal stress is calculated at each location of the inner barrel. Finally, cumulative damage is evaluated by using the present method. It is concluded that a main factor of cumulative damage is a stress near flow holes that causes stress concentration. It is also found that thermal transient within several hundred seconds after the reactor scram is an important factor.

Application of a new passive acoustic leak detection approach to recordings from the Dounreay prototype fast reactor

A new approach for passive acoustic leak detection in sodium fast reactors without using a priori knowledge on the leak noise is introduced. The new approach is tested on recordings of argon and water injections from the Dounreay prototype fast reactor under digital mixing with two types of additional noise. It is estimated that the new approach is able to detect injection of argon into sodium in a stable background noise at signal to noise ratios between -9 and -17dB with a low false alarm rate and with few free parameters in the signal processing. For detection of water into sodium injection the corresponding signal to noise ratios range from -9 to -18dB.

Application of statistical method for FBR plant transient computation

It is obvious that design tolerances, errors included in operation, and statistical errors in empirical correlations effect on the transient behavior. The purpose of the present study is to apply above mentioned statistical errors to a plant system computation in order to evaluate the statistical distribution contained in the transient evolution. A selected computation case is the turbine trip test conducted at 40% electric power of the prototype fast reactor “Monju”. All of the heat transport systems of “Monju” are modeled with the NETFLOW++ system code which has been validated using the plant transient tests of the experimental fast reactor Joyo, and “Monju”. The effects of parameters on upper plenum temperature are confirmed by sensitivity analyses, and dominant parameters are chosen. The statistical errors are applied to each computation deck by using a pseudorandom number and the Monte-Carlo method. The dSFMT (Double precision SIMD-oriented Fast Mersenne Twister) that is developed version of Mersenne Twister (MT), is adopted as the pseudorandom number generator. In the present study, uniform random numbers are generated by dSFMT, and these random numbers are transformed to the normal distribution by the Box–Muller method. Ten thousands of different computations are performed at once. In every computation case, the steady calculation is performed for 12,000s, and transient calculation is performed for 4000s. In the purpose of the present statistical computation, it is important that the base regions of distribution functions should be calculated precisely. A large number of computations are performed in order not to reduce the precision of the base regions of the distributions. Through a set of computation, the effect of errors on plant transient is obtained as the statistical distribution. In addition, the reduction method of trial numbers is considered. And results using the method express base regions of the distribution.

Photon transport effect on intra-subassembly thermal power distribution in fast reactor

In order to accurately predict intra-subassembly thermal power distribution in a fast reactor, neutron and photon transport calculations are carried out with a multi-purpose reactor physics calculation code system CBZ. All the fission fragment nuclide are treated explicitly during fuel depletion, and irradiation time-dependent energy spectra of delayed fission γ-rays emitted from all the fission fragment nuclides are precisely simulated. Time-dependent delayed β-ray emission and transmutations of fission fragment nuclide by neutron-nuclide reactions are also taken into account. A fuel subassembly model of Japanese prototype fast reactor Monju is used for numerical calculations, and their two-dimensional geometric feature is precisely modeled by a ray-tracing-based collision probability method implemented in CBZ. When the photon transport is considered, total thermal powers in fissile material regions are reduced by about 1.5% except at the beginning of fuel depletion.

R&D in Support of ASTRID and JSFR: Cross-Analysis and Identification of Possible Areas of Cooperation

The Commissariat à L’Énergie Atomique et aux Énergies Alternatives (CEA) and the Japan Atomic Energy Agency (JAEA) intend to develop prototype or demonstration sodium-cooled fast reactors (SFRs) within two decades. The common final goal of their respective programs is SFR commercialization. The target of a commercial SFR for CEA and JAEA is basically consistent with Generation IV goals. Based on their industrial backgrounds and feedback from past and existing reactor experiences, CEA and JAEA have selected pool and loop configurations for the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) and the Japanese Sodium-Cooled Fast Reactor (JSFR), respectively. CEA and JAEA have cross-analyzed both pool and loop concepts (ASTRID and demonstration JSFR, respectively). The analysis results show that both concepts are technologically feasible and meet design goals. From the viewpoint of collaboration, the analysis identifies a wide range of collaborative areas in several fields: design principles (e.g., design target and design standard); development of components and systems; development of component parts, devices, or subsystems; design methods; simulation tools; etc.

Acoustic Leak Detection Technology for Water/Steam Small Leaks and Microleaks into Sodium to Protect an SFR Steam Generator

This technical note presents the results of an experimental study of the role of water in sodium leak noise spectrum formation and at various water/steam leak rates of <1.0 g/s. The conditions and ranges for the existence of bubbling and jetting modes in water/steam outflow into circulating sodium through an injector device were determined to simulate a defect in the wall of the heat-transmitting tube of a sodium-water steam generator (SG). Based on experimental leak noise data, the simple dependency of the acoustic signal level on the leak rate of a microleak and small leaks at different frequency bands was presented for the principal analysis to develop an acoustic leak detection methodology for a KALIMER-600, 600-MW(thermal) reactor (K-600) SG, with the operational experiences for noise analysis and measurements of the Bystry neutron (fast neutron) reactor BN-600. Finally, the methodology was tested with the Korea Atomic Energy Research Institute (KAERI) acoustic leak detection system using sodium-water reaction signals of the Institute of Physics and Power Engineering and background noise of the Prototype Fast Reactor (PFR) superheater for methodology development of KAERI, and it was able to detect a leak rate of under 1 g/s and a signal-to-background noise ratio of −22 dB, using this system and methodology.

Technology Gap Analysis on Sodium-Heated Steam Generators Supporting Advanced Burner Reactor Development

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The Advanced Burner Reactor (ABR) is a fast reactor concept that supports the GNEP fuel cycle system. Since the Integral Fast Reactor (IFR) and Advanced Liquid Metal Reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP ABR program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology.A sodium-heated steam generator is one of the key components in the fast reactor system since it provides interface between sodium and water. In this gap analysis, information of fabrication and operation experiences in reactor plant steam generators and prototype steam generators was carefully reviewed, for example the Enrico Fermi Atomic Power Plant, the Prototype Fast Reactor (PFR), and Phénix steam generators; the Babcock & Wilcox helical coil tube, 70 MW; the Westinghouse double-wall tube, 70 MW; the Clinch River Breeder Reactor (CRBR) full-scale evaporator; the Superphénix prototype helical coil tube, 45 MW; the SNR-300 prototype straight tube, 50 MW; the SNR-300 prototype helical coil tube, 50 MW; and the Monju prototype helical coil tube, 50 MW. The results of this evaluation indicate that straight and helical coil tube steam generators are the best immediate candidate designs for producing reliable steam generators for future sodium fast reactor applications. Though the design comparison suggested that the straight tube type has the advantages of compactness and ease of inspection, prototype tests revealed more technical problems than the helical modules. From the viewpoint of tube material, 2¼Cr steel has been well established, and Incoloy® 800, 9Cr, and 12Cr steels are available as higher-performance materials.

PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt/20 MWe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 MWe; and the China demonstration fast reactor (CDFR) with a power of 2 500–3 750 MWt/1 000–1 500 MWe. With regards to the fuel cycle, a 100 t/a PWR spent fuel reprocessing pilot plant and a 500 kg/ a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

The Dounreay PFR Liquid-Metal Disposal Project

The UKAEA Prototype Fast Reactor at Dounreay had a liquid sodium-cooled core. Following its shutdown in 1994, the liquid metal is being removed from the reactor and other vessels by means of specialized equipment and reacted with an aqueous solution of sodium hydroxide in a special vessel. The reaction products are neutralized with hydrochloric acid to produce a saline solution.The reactor sodium delivery and processing equipment is all of novel design. As sodium has been withdrawn from the vessel, it has been necessary to switch off the primary sodium pumps (used to heat the sodium), and the reactor is now kept at temperature by a purpose-designed electric heater and a NaK loop heater.A primary sodium extract pump has currently removed [approximately]450 tonnes of primary sodium from the reactor. As the level falls special equipment will be used to punch a hole in the primary circuit pipe work and to drill the strongback to allow trapped sodium to drain for extraction.