Prototype Fast Breeder Reactor Research Articles

Design and performance evaluation of Core Flow Monitoring Mechanisms for PFBR

In this paper, the scheme adopted for off-line monitoring of fuel/blanket sub-assemblies, by measuring the coolant flow rate at sub-assembly exit, in a typical sodium cooled fast breeder reactor is presented. In Prototype Fast Breeder Reactor (PFBR), the temperature of sodium exiting from all the fuel sub-assemblies is monitored by core temperature monitoring system for detecting any flow blockage during reactor operation. However, such online monitoring of blanket sub-assembly outlet temperature was not envisaged. Instead, offline monitoring is proposed, which involves measuring the flow rate of sodium exiting from blanket sub-assemblies during re-fuelling of reactor. A flow measuring system is proposed for measuring sodium flow from blanket sub-assemblies. In addition, the sodium flow from Fuel and storage sub-assemblies is also measured during commissioning stage in order to identify any blockage in the hydraulic paths of respective sub-assemblies.The flow measuring system employs Core Flow Monitoring Mechanism (CFMM) equipped with Eddy Current Flow Sensor (ECFS) for sodium flow measurement. The system needs two CFMMs, one for fuel sub-assemblies and another for blanket and storage sub-assemblies. The necessary experimental studies were carried out for generating data needed for sodium flow estimation and to resolve uncertainties in the design of CFMM. Experimental studies include hydraulic studies and sodium calibration of ECFS. Subsequently, integrated assembly testing of CFMM was carried out to demonstrate the safe operation of the mechanisms before clearing for erection on pile. After successful completion of all the tests, both the CFMM were erected on pile. This paper discusses the detailed design features of CFMM and the developmental activities carried out to qualify the mechanism for reactor use.

Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

The 500MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

Validation of the finned sodium–air heat exchanger model in MARS-LMR

A Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), for which the Korea Atomic Energy Research Institute (KAERI) has designed a pool type sodium-cooled fast reactor, has a decay heat removal system (DHRS). The DHRS consists of sodium-to-sodium and sodium-to-air heat exchangers and their connecting pipes. There are four loops, which have two active and two passive type sodium–air heat exchangers, that are called a finned-tube sodium-to-air heat exchanger (FHX) and a helical-tube sodium-to-air heat exchanger (AHX), respectively. Recently, Zukauskaus’s air–sodium heat transfer models have been added in MARS-LMR, which is a safety analysis code for the PGSFR. In this study, to validate the newly added heat transfer models for the FHX, two experiments are selected: one is a performance test for a sodium to air heat exchanger (AHX) in the steam generator test facility (SGTF) for the prototype fast breeder reactor (PFBR), and the other is a start-up test with a dump heat exchanger (DHX) in JOYO. All validation results indicate that Zukauskaus’s correlation slightly over-estimates the heat transfer. One possible reasons is a smaller number of rows in the test bundle, which was already mentioned by Zukauskaus and Karni. The RMS values for the prediction of sodium temperature for the PFBR and the JOYO are 16.25% and 27.5%, respectively. When a correction factor is applied, their RMS values improve to 3.19% and 20.71%, respectively. In addition, the MARS-LMR’s prediction for the JOYO shows a much better accuracy with RMS of 7.45% when corrected sodium flow rates are applied.

Impurity analysis and dissolution behaviour of plutonium bearing impure MOX scrap: an assessment of recycling feasibility

A significant quantity of plutonium bearing scrap was getting generated during the fabrication of (U,Pu)O2 mixed oxide fuel for Indian Prototype Fast Breeder Reactor. This document assesses the feasibility of carrying out recycling of the impure scrap to permit the efficient plutonium recovery and re-use. The impure scrap which constitutes nearly 5 % of the total throughput was examined for impurities present and dissolution behaviour in HNO3. On the basis of experimental results obtained recycling methodologies are recommended. This study reveals that not all types of the impure scrap need aqueous recycling methods; some can be dry recycled.

Characterization of High Temperature (873 K) Low Cycle Fatigue Damage in Modified 9Cr–1Mo Steel Using Magnetic Barkhausen Emission Technique

Modified 9Cr–1Mo (P91) ferritic steel is the steam generator structural material for prototype fast breeder reactor. During the start-ups and shutdowns, the material is subjected to conditions leading to low cycle fatigue damage. Detection of initiation of the fatigue damage is required to prevent the catastrophic failure in components of the steam generator. The present study reports the non-destructive characterization of substructural and microstructural changes occurring during low cycle fatigue (LCF) of modified 9Cr–1Mo base metal and weld joint at 873 K, using MBE measurements. MBE studies have been carried out on the samples at various levels of fatigue life by interrupting the fatigue test. MBE RMS voltage has been measured at different locations in the gauge portion of the samples to map the damage profile. The systematic changes in the MBE revealed various stages of progressive LCF damage starting from initial softening to final failure in the ferritic steel.

Modelling and simulation of inclined fuel transfer machine in prototype fast breeder reactor operator training simulator

Indira Gandhi Centre for Atomic Research, Kalpakkam has successfully designed a 500 MWe prototype fast breeder reactor (PFBR). It is a proven fact that the critical factor in ensuring safe and reliable operation of any nuclear power plant is a qualified and well-trained personnel and hence, the need for operator training using a simulator. Fuel handling is one of the most important subsystems in a prototype fast breeder reactor. Ex-vessel subassembly transfer is done using an inclined fuel transfer machine (IFTM) which operates remotely and safely. Using IFTM, fresh fuel assemblies are loaded into the reactor, and spent fuel assemblies get unloaded. This paper deals with modelling and simulation of operations of IFTM in a simulator. Virtual panel models and control logic models were modelled using modelling tools present in the simulator. The process modelling was developed in-house using platform independent C++ code with 3D models.

Optimisation of thermal baffle for liquid metal injection nozzle

SADHANA is a test facility dedicated for the experimental demonstration of passive safety grade decay heat removal in the prototype fast breeder reactor. Experiments are planned in SADHANA to characterise the behaviour of this system followed by various reactor transients. For simulating the reactor transients of hot pool, cold sodium at 310°C has to be injected to the test vessel of SADHANA at 550°C temperature. Thermal baffle provided in the nozzle of test vessel was optimised to limit the peak local stress. This paper gives the thermo-mechanical studies conducted to optimise the nozzle geometry with thermal baffles.

Optimisation of thermal baffle for liquid metal injection nozzle

SADHANA is a test facility dedicated for the experimental demonstration of passive safety grade decay heat removal in the prototype fast breeder reactor. Experiments are planned in SADHANA to characterise the behaviour of this system followed by various reactor transients. For simulating the reactor transients of hot pool, cold sodium at 310°C has to be injected to the test vessel of SADHANA at 550°C temperature. Thermal baffle provided in the nozzle of test vessel was optimised to limit the peak local stress. This paper gives the thermo-mechanical studies conducted to optimise the nozzle geometry with thermal baffles.

Thermal Analysis of Irradiated Fuel Subassemblies and Fuel Pins During Storage in Concrete Pits of Head-End Facility

Irradiated fuel subassembly (SA)/fuel pins, with significant decay heat are transported from reactor and stored in hot cells (HCs) before reprocessing. During transportation they are heavily shielded and no forced cooling is provided. The HCs are made of concrete structures, the outer surfaces of which are force cooled. During these processes, the fuel pin clad temperature and concrete temperatures are to be limited within specific safety limits. These temperatures are function of the decay power and geometric details of surrounding structures. To predict these temperatures, three-dimensional conjugate conduction–convection–radiation heat transfer analysis has been carried out. For this purpose, the computational fluid dynamics (CFD) code STAR-CD has been utilized, wherein individual fuel pins, steel cans, hexagonal wrapper, lead shielding blocks, and concrete structures have been considered in detail. Based on parametric studies pertaining to fuel pin transportation, it is established that for a decay power of 150 W, natural convection is adequate with maximum clad temperature of 686 K. From the studies related to storage in HCs, it is seen that nine fast breeder test reactor (FBTR) SA can stored in hot cell-1 (HC-1), with a decay power of 31.3 W per SA, to respect the temperature limits. For 3 prototype fast breeder reactor (PFBR) cans and 2 FBTR cans stored in hot cell-3 (HC-3), a decay power of 12.5 W per FBTR can and 44 W per PFBR can, can be handled without exceeding temperature limits.

Thermal hydraulic issues and challenges for current and new generation FBRs

Pool type sodium cooled fast reactors pose several design challenges and among them, certain thermal hydraulics and structural mechanics issues are special. High frequency temperature fluctuations due to thermal striping, thermal stratifications and sodium free level fluctuations at the liquid–cover gas interfaces are to be investigated carefully to eliminate high cycle thermal fatigue of structures. Solutions to address the core thermal hydraulics call for high power computing. Innovative concepts and methods are developed to carry out plant dynamics and safety studies. Particularly, extensive numerical and experimental simulation techniques are needed for understanding and solving the gas entrainment mechanisms and its effects on core safety. Though decay heat removal through natural convection is achievable in a pool type SFR, demonstration of design solutions conceived in the reactor and performance of diverse systems under all operating conditions, especially over prolonged station blackout situations needs advanced CFD computations and should be validated by relatively large scale simulated experiments. These issues are addressed in this paper under five broad topics: special thermal hydraulic issues to be addressed in SFR, thermal hydraulic design and analysis, plant dynamics studies, safety studies and evolving thermal hydraulic studies for the future FBRs. The 500MWe Prototype Fast Breeder Reactor (PFBR) is taken as the reference design for addressing the issues. Indian fast reactor programme is highlighted in the introduction for the sake of completeness.

Primary system thermal hydraulics of future Indian fast reactors

As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

Measurement and analysis of feedback reactivity in the Monju restart core

The purpose of this study is to develop a feedback reactivity measurement technique in the Japanese prototype fast breeder reactor Monju and to validate calculation methodology to forecast the nuclear feedback phenomena. A feedback reactivity measurement technique has been developed based on a reactivity model featuring components that depend on the reactivity coefficients, denoted as reactor power (KR) and reactor vessel inlet temperature (Kin). This technique can precisely measure the two reactivity components simultaneously and was applied to the feedback reactivity experiment conducted in the Monju system start-up test in May 2010. A thorough evaluation considering all possible biases and uncertainties demonstrated that the reactivity coefficients can be evaluated with a measurement uncertainty smaller than 3%. The evaluated reactivity coefficients were simulated considering the temperature distribution in the core. The calculated and measured values of KR agreed within 1%, and the value of Kin was consistent with that reported in a previous isothermal temperature coefficient experiment. The measured and calculated fuel subassembly outlet temperatures also agreed well within 0.2°C, which supports the validity of the temperature calculation.

Sodium flow measurement in large pipelines of sodium cooled fast breeder reactors with bypass type flow meters

Liquid sodium flow through the pipelines of sodium cooled fast breeder reactor circuits are measured using electromagnetic flow meters. Bypass type flow meter with a permanent magnet flow meter as sensor in the bypass line is selected for the flow measurement in the 800NB main secondary pipe line of 500MWe Prototype Fast Breeder Reactor (PFBR), which is at the advanced stage of construction at Kalpakkam. For increasing the sensitivity of bypass flow meters in future SFRs, alternative bypass geometry was considered. The performance enhancement of the proposed geometry was evaluated by experimental and numerical methods using scaled down models. From the studies it is observed that the new configuration increases the sensitivity of bypass flow meter system by around 70%. Using experimentally validated numerical tools the volumetric flow ratio for the bypass configurations is established for the operating range of Reynolds numbers.

Fast reactor programme in India

Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

Software development methodology for computer based I&C systems of prototype fast breeder reactor

Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

Design of high temperature induction motor for application in sodium cooled fast reactor control system

Prototype Fast Breeder Reactor (PFBR) employs induction motors of various power ratings in high temperature ambient and sometimes corrosive atmosphere. Safe operation of the reactor depends on robustness and reliability of induction motors in its critical components. Safety related systems such as shutdown and online inspection systems, utilize induction motors as actuators in their mechanisms. Since external cooling is not permitted in control and inspection motors, special design of motors is required to withstand high temperatures. This paper deals with design, analysis, fabrication and testing of a compact 50 W, three phase induction motor, suitable for operating up to 250 °C ambient with winding temperature withstand capability up to 550°C. The designed motor was analyzed with two dimensional Finite Element Model (FEM) code FEMM to study the torque slip characteristics and magnetic flux density patterns in stator and rotor.

Development of magnetic flux leakage technique for examination of steam generator tubes of prototype fast breeder reactor

For non-destructive examination of small diameter (outer diameter, OD 17.2mm) and thick walled (wall thickness, 2.3mm) ferromagnetic Modified 9Cr–1Mo steel steam generator (SG) tubes of Prototype Fast Breeder Reactor (PFBR), this paper proposes magnetic flux leakage (MFL) technique. Three dimensional finite element (3D-FE) modeling has been performed to optimize the magnetizing unit and inter-coil spacing of bobbin coils used for axial magnetization of the tube. The performance of the technique has been evaluated experimentally by measuring the axial (Ba) component of the leakage fields from localized machined defects in SG tubes. The MFL technique has shown capability to detect and image tube outside defects with a signal-to-noise ratio (SNR) better than 6dB. Study reveals that Inconel support plates surrounding the SG tubes do not influence the MFL signals. As the MFL technique can detect localized defects in the presence of support plates as well as sodium and the remote field eddy current technique is sensitive to distributed wall thinning, their combined use will ensure comprehensive inspection of the SG tubes.

A preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Abstract In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of in-vessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.

Negligence, capture, and dependence: safety regulation of the design of India’s Prototype Fast Breeder Reactor

This article contributes a case study of regulation of the design of India’s Prototype Fast Breeder Reactor (PFBR). This reactor is the first of its kind in India, and perceived by the nuclear establishment as critical to its future ambitions. Because fast breeder reactors can experience explosive accidents called core disruptive accidents whose maximum severity is difficult to contain, it is difficult to assure the safety of the reactor’s design. Despite the regulatory agency’s apparent misgivings about the adequacy of the PFBR’s design, it eventually came to approve construction of the reactor. We argue that the approval process should be considered a case of regulatory failure, and examine three potential factors that contributed to this failure: institutional negligence, regulatory capture, and dependence on developers and proponents for esoteric knowledge. This case holds lessons for nuclear safety regulation and more generally in situations where specialized, highly technical, knowledge essential for ensuring safety is narrowly held.

Simplified method for measuring the response time of scram release electromagnet in a nuclear reactor

The delay time in electromagnet clutch release during a reactor trip (scram action) is an important safety parameter, having a bearing on the plant safety during various design basis events. Generally, it is measured using current decay characteristics of electromagnet coil and its energising circuit. A simplified method of measuring the same in a Sodium cooled fast reactors (SFR) is proposed in this paper. The method utilises the position data of control rod to estimate the delay time in electromagnet clutch release. A computer based real time measurement system for measuring the electromagnet clutch delay time is developed and qualified for retrofitting in prototype fast breeder reactor. Various stages involved in the development of the system are principle demonstration, experimental verification of hardware capabilities and prototype system testing. Tests on prototype system have demonstrated the satisfactory performance of the system with intended accuracy and repeatability.