Operation Of Nuclear Reactors Research Articles

Challenges of Atomic Energy Regulation in Indian Context

Abstract Over the years, India has mastered all the stages of the nuclear fuel cycle, which include mining, processing & fabrication of nuclear fuel; design, construction, and operation of nuclear power reactors and research reactors; reprocessing of spent fuel and management of radioactive wastes. Ionising radiation is also used widely in medical, industrial and research areas. Since its inception, Department of Atomic Energy (DAE) was enforcing radiological safety in the country through in -house or ad-hoc committees, till a dedicated regulatory body (AERB) was set up 25 years ago. Today India is operating 19 nuclear power plants with different vintages (2 BWRs and 17 PHWRS) and another 8 (1 PFBR, 5 PHWRs and 2 PWRs) are in various stages of construction. Recently there are new evolutionary reactors (AHWRs) for which design has been completed and are on the threshold for consideration for construction. To match the rapid growth in the need for power India is also about to take up construction of large evolutionary PWRs of foreign design. This variety in the Indian nuclear power programme has come up due to a systematic evaluation and optimisation of the resources and technology available within the country. Added to this is the growing use of radiation in non -power applications. As the safety supervision of this huge programme is the responsibility of AERB, it faces various challenges, like, • Strategies for regulating wide variety of nuclear and radiation facilities with wide dispersal; • Meeting present day expectations with regard to nuclear and radiation safety and nuclear security; • The safety and security of large number of radioactive sources spread over such a vast country and of the associated import/export guidance; • Ensuring safety of old plants by periodic reviews and by prescribing adequate safety upgradation and ageing mana gement programme; • Adaptation of the regulatory system and of regulations to new and foreign design nuclear technologies and applications; • Developing competence in wide variety of technologies and different reactor designs; • Developing adequate human resource ready for safety supervision of this huge nuclear and radiation programme; • Approach to regulatory research related to new technologies. Apart from the above mentioned issues regulatory challenges may arise from policies to make all energy se ctors competitive. With growing competition, it is becoming increasingly important to reconcile commercial interests with safety requirements. For AERB, the first challenge will be to ensure that economic pressures do not erode nuclear safety. To maintain highest level of safety culture, AERB will also need to adapt to an increasingly market -oriented environment and new working relationships with utilities. Although it is a difficult task, the Board has also evolved systematically over the years and now has an efficient structure, which is capable of facing the current challenges.

Load-Following Control of Nuclear Reactors Based on Takagi-Sugeno Fuzzy Model

AbstractA Takagi-Sugeno (T-S) fuzzy control system is designed for load-following operation of nonlinear nuclear reactor whose operating points vary within a wide range. Linear models are first derived from the original nonlinear model on several operating points. Next the fuzzy controller is designed via using the parallel distributed compensation (PDC) scheme with the relative neutron density at the equilibrium point as the premise variable. Last the stability analysis is given by means of linear matrix inequality (LMI) approach, thus the control system is guaranteed to be stable within a large range. The simulation results demonstrate that the control system works well over a wide region of operation.

A new hardwire circuit for metering reactivity

Sub-criticality, criticality, and reactivity are the terms which are not annunciated in Indian PHWRs (Pressurised Heavy Water Reactors) and are thus not very clear to the reactor operators. These parameters are prime movers defining nuclear reactor and should be annunciated in control room by simple means to aid operator to view reactivity excursions while in power operation or during shutdown maintenance. A very novel circuit of a hardwire scheme to solve the coupled neutron kinetics is presented and its output is compared against analytical solutions. This scheme does not require any estimation technique and is robust as the instrument is based on basic reactor governing laws. This instrument is able to show the transients in reactivity as well as steady state values and is aimed for all Indian PHWRs for multi-decade operation. This technique is tunable for any research reactor or fast reactor too.

New life for old naval reactors

The news item “Small Nuclear Reactors Raise Big Hopes” (PHYSICS TODAY, August 2010, page 25) brings to mind the dozens of small nuclear reactors that the US Navy is taking out of service as it decommissions older submarines and surface warships. They are compact, safe, proven power plants that could just as easily turn a generator as a propeller. Currently, the navy cuts the reactors out of the ships and stores them in an open pit at the Hanford Nuclear Reservation in Washington State after removing the fuel for reprocessing in Idaho.Instead, the naval reactors could be converted to serve as electrical power plants. Generators in hull-like assemblies could be attached to the power plant sections and the “vessels” towed to their new homes and dry docked. Refueling of a vessel could be accomplished by towing it back to one of the secure navy yards where refueling is already performed.The best customers for such converted reactors would be large federal installations where controlled access would protect their cold-war construction secrets. For example, Redstone Arsenal, on the Tennessee River in Alabama, is home to a large NASA center and a major US Army command. Cape Canaveral, on the Florida coast, is home to NASA and US Air Force launch centers. The large number of former navy reactor operators could be recruited to run their old charges but with the advantage that they could go home at the end of each shift.Such small plants would reduce the demands on local utilities that must also serve growing populations outside the installations. They could be made operational on much shorter schedules than new reactors could be constructed. At Cape Canaveral, they could also power an electrolysis facility to produce liquid hydrogen fuel that must now be trucked in from distant facilities that process petroleum to get hydrogen.Old naval reactors are a national resource that should be studied as a potential power source and as the wise reuse of existing equipment, rather than being discarded.© 2011 American Institute of Physics.

Modelling of iodine-induced stress corrosion cracking in CANDU fuel

Iodine-induced stress corrosion cracking (I-SCC) is a recognized factor for fuel-element failure in the operation of nuclear reactors requiring the implementation of mitigation measures. I-SCC is believed to depend on certain factors such as iodine concentration, oxide layer type and thickness on the fuel sheath, irradiation history, metallurgical parameters related to sheath like texture and microstructure, and the mechanical properties of zirconium alloys. This work details the development of a thermodynamics and mechanistic treatment accounting for the iodine chemistry and kinetics in the fuel-to-sheath gap and its influence on I-SCC phenomena. The governing transport equations for the model are solved with a finite-element technique using the COMSOL Multiphysics® commercial software platform. Based on this analysis, this study also proposes potential remedies for I-SCC.

The transient flow in a centrifugal pump during the discharge valve rapid opening process

During the rapid opening period of the discharge valve in the pump system commonly used in nuclear reactor operation, the flow-rate of the pump increases impulsively. In this paper, we report on experiment and numerical simulations which were implemented to investigate the external transient hydrodynamic performance and the internal flow mechanism of the pump during this transient process. External and internal characteristics under different flow-rates were measured with an experimental system. The simulation for steady conditions was based on detached eddy simulation (DES) and sliding mesh was verified by comparing the simulation with test results. More importantly, the transient characteristics during the valve's rapid opening process were simulated using a similar method. Results show that the Q–H curve deviates from the steady-state value. The external characteristics are further explained by analyzing the relative velocity on the middle stream surfaces S1m and S2m between blades. The pump performance during the valve's rapid opening process is influenced both by the fluid acceleration and instantaneous evolutions of the vortex structure.

Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

Abstract During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of 235 U, 238 U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.

Extreme-value-model-based risk assessment for nuclear reactors

The safety case for continuing operation of nuclear reactors requires reliable assessment of the likelihood of the coolant temperatures exiting the fuel channels exceeding certain critical levels. Temperature measurements are typically made at a fixed sample of fuel channels and used for reactor control. No sample measurements will exceed the predetermined control limit, whereas it is likely that some of the unobserved temperatures will exceed this limit. The challenge is to use the control measurements reliably to assess the risk of the critical temperature exceedance over all channels, while also accounting for the uncertainties in the risk estimation. A novel non-stationary extreme value mixture modelling technique is developed to provide rigorous extrapolation of the risk past the observed range of the sample data. The proposed technique builds upon previous deterministic and statistical approaches, while providing more accurate risk predictions with fuller account for uncertainties in the estimation. A spatial random effects model is used to capture the non-stationary spatial structure in the temperature variation across the reactor. Bayesian inference for the extreme value mixture model parameters is undertaken, which permits assessment of all sources of uncertainty including potential inclusion of expert prior information.

Reduction of Se(IV) in Boom Clay: XAS Solid Phase Speciation

The geochemical fate of selenium is of key importance for today's society due to its role as a highly toxic essential micronutrient and as a significant component of high level radioactive waste (HLRW) originating from the operation of nuclear reactors. Understanding and prediction of the long-term behavior of Se in natural environments requires identification of the in situ speciation of selenium. This article describes an XAS-based investigation into the solid phase speciation of Se upon interaction of Se(IV) with Boom Clay, a reducing, complex sediment selected as model host rock for clay-based deep geological disposal of HLRW in Belgium and Europe. Using a combination of long-term batch sorption experiments, linear combination XANES analysis and ITFA-based EXAFS analysis allowed for the first time to identify Se0 as the dominant solid phase speciation of Se in Boom Clay systems equilibrated with Se(IV).

Separation of strontium from low level radioactive waste solutions using hydrous manganese dioxide composite materials

90Sr is one of the major isotopes present in the low level radioactive liquid waste (LLW) generated during operation of nuclear reactors and spent fuel reprocessing plants. A composite ion exchange material consisting of hydrous manganese oxide and poly methyl methacrylate (PMMA) was developed for removal of strontium from aqueous radioactive waste. The prepared composite material showed very good strontium adsorption properties in aqueous solutions. Sorption of strontium on the composite material as a function of pH, equilibration time and strontium ion concentrations were studied. The process of sorption of strontium from solution was analysed using different isotherm models like Langmuir, D–R and Freundlich. Four different error functions were employed to find out the most suitable isotherm model to represent the experimental data and it was found that Freundlich model best fits the sorption of strontium on the composite material. Analysis of the data obtained from the sorption kinetics studies showed that the data fitted better to the pseudo-second order kinetic model. Lab scale column performance study of the composite material revealed that the material could be effectively used in column operations to remove strontium from LLW solutions.

Recent Progress in the Design of a Tomographic Device for Measurements of the Three-Dimensional Pin-Power Distribution in Irradiated Nuclear Fuel Assemblies

A tomographic technique for determination of the thermal power distribution in nuclear fuel assemblies is under development. The purpose is to provide an experimental validation tool for core simulation codes. Such codes are essential for the operation of nuclear power reactors, and validation is important in the process of improving and developing the codes as well as the fuel.The tomographic method is nonintrusive and offers large amounts of data within a normal revision shutdown. In earlier experimental investigations using a test platform, the method proved useful, demonstrating results of satisfying quality. However, the measuring setup also revealed nonfeasible properties related to transport, decontamination, and background radiation shielding.In this paper, the design of a new measuring device is presented. It is based on experiences from the test platform, but its size and weight make it advantageous regarding transports and decontamination. Moreover, the design inherently allows for more efficient background shielding.The latter has been investigated in a detailed study using the MCNP simulation code. The results confirm the high levels of background radiation observed in the test platform. It is also concluded that the shielding properties in the new design are sufficient.

Dissipation-Based High Gain Filter for Monitoring Nuclear Reactors

Growing electricity requirement and the serious pollution caused by burning petroleum and coal give the current rebirth of nuclear energy industry. State observation is one of the key and basic technologies of system monitoring which is very necessary to the safe and effective operation of today's nuclear reactors. Since nuclear reactors are complex and nonlinear systems, it is quite necessary to design a nonlinear state-observer with high-performance for nuclear reactors. A dissipation-based high gain filter (DHGF) is presented for nonlinear systems in this paper, and robustness analysis is also given. The DHGF is then applied to the state-observation for a nuclear heating reactor (NHR), and simulation results show the feasibility of the DHGF.

Hydrogen Pick up in Zircaloy-4: Effects in the Dimensional Stability of Structural Components under Nuclear Reactor Operating Conditions

In the present work, the expansion coefficient due to hydrogen incorporation was measured for the axial direction of a Zircaloy-4 cooling channel, similar to that installed in the Atucha I PHWR, Argentina, trying to simulate the nuclear power reactor operating conditions. As a first step, the solubility curve of hydrogen in Zircloy-4 was determined by two techniques: differential scanning calorimetry and differential dilatometry. The comparison with classical literature curves showed a good agreement with them, although the calorimetric technique proved to be more accurate for these determinations. Dilatometry was able to detect the end of hydride dissolution from concentrations around 60 wppm-H up to 650 wppm-H, where the eutectoid reaction: α + δ→α + β takes place (at 550oC). We assume that this ability is a good indicator of the aptitude of the technique to measure dimensional changes in the given hydrogen concentration range. Then, the expansion of Zircaloy-4 homogeneously hydrided samples was measured at 300oC, the typical operating temperature of a nuclear power reactor, obtaining a relative expansion of 2.21 * 10-4% per wppm-H. Considering the relative expansion observed for Zircaloy-4 at room temperature due to hydriding, starting from a hydrogen free sample, the total relative expansion rate is calculated to be 5.21 * 10-4% per wppm-H.

Hydrogen Pick up in Zircaloy-4: Effects in the Dimensional Stability of Structural Components under Nuclear Reactor Operating Conditions

In the present work, the expansion coefficient due to hydrogen incorporation was measured for the axial direction of a Zircaloy-4 cooling channel, similar to that installed in the Atucha I PHWR, Argentina, trying to simulate the nuclear power reactor operating conditions. As a first step, the solubility curve of hydrogen in Zircloy-4 was determined by two techniques: differential scanning calorimetry and differential dilatometry. The comparison with classical literature curves showed a good agreement with them, although the calorimetric technique proved to be more accurate for these determinations. Dilatometry was able to detect the end of hydride dissolution from concentrations around 60 wppm-H up to 650 wppm-H, where the eutectoid reaction: α + δ→α + β takes place (at 550oC). We assume that this ability is a good indicator of the aptitude of the technique to measure dimensional changes in the given hydrogen concentration range. Then, the expansion of Zircaloy-4 homogeneously hydrided samples was measured at 300oC, the typical operating temperature of a nuclear power reactor, obtaining a relative expansion of 2.21 * 10-4% per wppm-H. Considering the relative expansion observed for Zircaloy-4 at room temperature due to hydriding, starting from a hydrogen free sample, the total relative expansion rate is calculated to be 5.21 * 10-4% per wppm-H.

Monte Carlo characterisation of irradiation positions of the CENM TRIGA MARK II research reactor using MCNP5

The knowledge of neutron flux distribution in different parts of the research reactor is very important both for the design of experiments and for the reactor safety. Knowing the neutron flux profiles, reactor operators can optimise the irradiation conditions and select the positions most adaptable to the desired applications. Transport calculations can often give useful information about the irradiation field characteristics. However, for complicated geometries, only the Monte Carlo calculations are able to give a reliable characterisation. In this work, calculations were performed to predict neutron flux profile at different locations in the CENM TRIGA MARK II reactor. A detailed MCNP5 model of the reactor was used. The calculations show that neutron energy spectrum depends strongly on the position in the reactor. They also show that the maximum flux decreases by a factor of 3.75 from core centre to the pneumatic transfer system in the outer ring of the core. Further simulations demonstrate that neutron energy distribution is greatly affected by materials in which they diffuse.

The Workshop on “Commendable Practices for the Safe, Long-term Operation of Nuclear Reactors―OECD/NEA Stress Corrosion Cracking and Cable Ageing Project（SCAP）”

The Workshop on “Commendable Practices for the Safe, Long-term Operation of Nuclear Reactors―OECD/NEA Stress Corrosion Cracking and Cable Ageing Project（SCAP）”

Coupling CFAST fire modeling and SAPHIRE probabilistic assessment software for internal fire safety evaluation of a typical TRIGA research reactor

Due to the significant threat of internal fires for the safety operation of nuclear reactors, presumed fire scenarios with potential hazards for loss of typical research reactor safety functions are analyzed by coupling CFAST fire modeling and SAPHIRE probabilistic assessment software. The investigations show that fire hazards associated with electrical cable insulation, lubricating oils, diesel, electrical equipment and carbon filters may lead to unsafe situations called core damage states. Using system-specific event trees, the occurrence frequency of core damage states after the occurrence of each possible fire scenario in critical fire compartments is evaluated. Probability that the fire ignited in the given fire compartment will burn long enough to cause the extent of damage defined by each fire scenario is calculated by means of detection-suppression event tree. As a part of detection-suppression event trees quantification, and also for generating the necessary input data for evaluating the frequency of core damage states by SAPHIRE 7.0 software, CFAST fire modeling software is applied. The results provide a probabilistic measure of the quality of existing fire protection systems in order to maintain the reactor at a reasonable safety level.

Interactions of Cl− and OH Radical in Aqueous Solution

There is a considerable controversy surrounding the nature of the Cl-/OH complex in aqueous solution, which appears as a byproduct of the irradiation of salt solutions in nuclear reactor operation, radioactive waste storage, medicine, and environmental problems. In this work, we report results of combined quantum mechanical molecular mechanics calculations of ground-state free-energy surfaces and absorption spectrum through the CCSDT level of theory that are consistent with the experimental data and suggest that hemibonded HOCl- species may indeed exist in bulk aqueous solution.

Radiation sensitivity of microelectromechanical system devices

The sensitivity of microelectromechanical system (MEMS) devices to radiation is reviewed, with an emphasis on radiation levels representative of space missions rather than of operation in nuclear reactors. As a purely structural material, silicon has shown no mechanical degradation after radiation doses in excess of 100 Mrad. MEMS devices, even when excluding control/readout electronics, have, however, failed at doses of only 20 krad, though some devices have been shown to operate correctly for doses greater than 10 Mrad. Radiation sensitivity depends strongly on the sensing or actuation principle, device design, and materials, and is linked primarily to the impact on device operation of radiation-induced trapped charge in dielectrics. MEMS devices operating on electrostatic principles can be highly sensitive to charge accumulation in dielectric layers, especially for designs with dielectrics located between moving parts. In contrast, thermally and electromagnetically actuated MEMS are much more radiation tolerant. MEMS operating on piezoresitive principles start to slowly degrade at low doses, but do not fail catastrophically until doses of several Mrad. A survey of all published reports of radiation effects on MEMS is presented, as well as a summary of techniques that can improve their radiation tolerance.

The case for extended nuclear reactor operation

Greenhouse gas emission is a mounting problem that threatens the future production of electricity from both coal and natural gas. In 2006, 70% of domestic electricity generation relied upon fossil fuels. Projections by the Energy Information Agency1 show U.S. demand for electricity increasing 30% to 40% by 2030. Today nuclear power is the largest source of non-greenhouse-gas emitting energy generation and will be an important source of energy production in the future. This paper considers potential construction of new plants as well as the long-term prospects of existing nuclear power plants in the United States.