The problem of ensuring the integrity of VVER type reactor equipment integrity is now most significant in connection with justifying the safety of the NPP units and the extension of their service life-time to 60 years and more. This issue primarily first of all concerns long term operated NPP power units with VVER-440s and VVER-1000s. The justification of the VVER equipment integrity depends on the reliability of estimation of the degree of the equipment damage. One of the mandatory requirements [1], providing the reliability of such estimation, and also the evaluation of VVER equipment life-time, is the monitoring of equipment radiation loading parameters. Relative to this requirement there is a problem the challenge of justification of such the normative parameters, used for an estimating of the reactor pressure vessel (RPV) metal embrittlement, as the fluence and fluence rate of fast neutrons with energies above 0,5 MeV. Compliance with these requirements is analyzed during regular monitoring of radiation load parameters, which is performed by SEC NRS for all Russian NPP from the regulatory point of view. As a result of this activity, SEC NRS has recently elaborated one of the new approaches aimed to monitoring the radiation load of all equipment of Russian VVERs. The paper describes these approaches and shows the way of their implementation during monitoring procedures.

JMCT-S [1] 3-D high resolution Monte Carlo neutron and photon transport code is proposed for shielding design due to its significant advantages such as complex geometry modeling, accurate calculation method, high efficient parallel computing and visible mass data analysis. To demonstrate its feasibility and benefits in engineering, this paper presents the verification of JMCT-S code based on primary shielding analysis for CAP1400 nuclear power station. Firstly, by employing JMCT-S code this paper completes the modeling and simulations of CAP1400 nuclear power station with full-core pin-by-pin and reactor pressure vessel internals inside of primary shield wall [2] . Secondly, neutron fluence rates are calculated based on the equilibrium cycle information. Particularly, the full-core neutron source is calculated from 3-D power distribution, fissionable nuclide fractions, fission spectrum, particle numbers and energy released per fission. Finally, JMCT-S code is verified by comparing results with those of the MCNP [3] code. According to the calculation results, JMCT-S code has similar accuracy to the MCNP code. The relative error of neutron fluence rate in active core is within ±1.5%, and deviation of neutron fluence rate for fast neutron fluence rate on inner surface of reactor pressure vessel is less than ±10%. Consequently, JMCT-S code is reliable in shielding design of reactor plant which lays the foundation for usage of JMCT-S code.

Neutrons and photons produced from reactor core during operation pass through the pressure vessel, reach the reactor cavity, and form the reactor cavity streaming. Reactor cavity streaming dose rates calculation during normal operation is important for the evaluation and control of the equipment dose rates in the nuclear power plant. Because reactor is great in dimension and complex in geometry, neutrons and photons fluence rates declined by several orders from reactor core to outside. Cavity streaming calculation is a deep penetration calculation with heavy computation load which is difficult to converge. Three dimensional Discrete Ordinates and Monte Carlo (S N -MC) coupling method combines the advantage of the S N method with high efficiency and the MC method with fine geometrical modeling. The S N -MC coupling method decreases the tally errors and increases the efficiency of the MC method effectively by using MC surface source generated by the S N fluence rates. In this paper, the theoretical model of the 3D S N -MC coupling method is presented. In order to fulfill the coupling calculation, a 3D Discrete Ordinates code is modified to output angular fluence rates, a link code DO2MC is developed to calculate cummulative distribution functions of source particle variables on surface source, and a source subroutine is written for a 3D Monte Carlo code. The 3D S N -MC coupling method is applied on the calculation of the CAP1400 cavity streaming neutron and photon dose rates. Numerical results show that the 3D S N -MC coupling codes are correct, the relative errors of the results are less than 20% compared with those of the MC bootstrapping method, and the efficiency is greatly enhanced.

SiC fiber composite material as fuel cladding is proposed to increase the power density and maximum allowable fuel burnup in light water reactors. Empirical models about thermal properties of the SiC material are developed as a function of operating temperature and neutron fluence. A fuel rod modeling code frapcon2-SiC based on frapcon2 is compiled to predict the performance of SiC cladding when operating. Comparison of the behavior between the SiC f /SiC cladding and Zr-4 cladding in different thickness reveals that higher temperature will get due to the poor thermal conductivity of the SiC. As far as the Conclusion shows that Poor thermal performance of SiC cladding makes the design margin becomes smaller.

Mechanisms by which fission products (especially silver [Ag]) migrate across the coating layers of tristructural isotropic (TRISO) coated fuel particles designed for next generation nuclear reactors have been the subject of a variety of research activities due to the complex nature of the migration mechanisms. This paper presents results obtained from the electron microscopic examination of selected irradiated TRISO coated particles from fuel compact 1-3-1 irradiated in the first Advanced Gas Reactor experiment (AGR-1) that was performed as part of the Next Generation Nuclear Plant (NGNP) project. It is of specific interest to study particles of this compact as they were fabricated using a different carrier gas composition ratio for the SiC layer deposition compared with the baseline coated fuel particles reported on previously. Basic scanning electron microscopy (SEM) and SEM montage investigations of the particles indicate a correlation between the distribution of fission product precipitates and the proximity of the inner pyrolytic carbon (IPyC)-silicon carbide (SiC) interface to the fuel kernel. Transmission electron microscopy (TEM) samples were sectioned by focused ion beam (FIB) technique from the IPyC layer, the SiC layer and the IPyC-SiC interlayer of the coated fuel particle. Detailed TEM and scanning transmission electron microscopy (STEM) coupled with energy dispersive X-ray spectroscopy (EDS) were performed to identify fission products and characterize their distribution across the IPyC and SiC layers in the areas examined. Results indicate the presence of palladium-silicon-uranium (Pd-Si-U), Pd-Si, Pd-U, Pd, U, U-Si precipitates in the SiC layer and the presence of Pd-Si-U, Pd-Si, U-Si, U precipitates in the IPyC layer. No Ag-containing precipitates are evident in the IPyC or SiC layers. With increased distance from the IPyC-SiC interface, there are less U-containing precipitates, however, such precipitates are present across nearly the entire SiC layer.

A temperature sensitivity evaluation has been performed for an individual test capsule in the AGR-2 TRISO particle fuel experiment. The AGR-2 experiment is the second in a series of fueled test experiments for TRISO coated fuel particles run in the Advanced Test Reactor at the Idaho National Laboratory. A series of cases were compared to a base case by varying different input parameters in an ABAQUS finite element thermal model. Most input parameters were varied by ±10%, with one parameter ±20%, to show the temperature sensitivity to each parameter. The most sensitive parameters were the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. The thermal conductivity of the fuel compacts and thermal conductivity of the graphite holder were of moderate sensitivity. The least sensitive parameters were the emissivities of the stainless steel and graphite, along with gamma heat rate in the non-fueled components. Sensitivity calculations were also performed for the fast neutron fluence, which showed a general, but minimal, temperature rise with increasing fluence.

Neutron detectors are deployed at ports of entry across the world to monitor people and cargos for smuggled nuclear materials and are often incorporated in nuclear power plant designed to monitor power levels and ensure safe operations. With the supply of Helium-3 rapidly decreasing and due to increase of terrorism threats it is vital to U.S. national security that a viable alternative material be identified, and a new neutron detector design made available, especially for portal monitoring applications. The interest to study Boron-10 as an alternative to helium-3, due to the vast natural supply that the United States possesses and its nontoxic characteristics, is increasing in different research arenas. Our work consists on taking advantage of the near 20% of boron-10 present in naturally occurring boron which (just as phosphorus) is used to dope semiconductors. Boron doped semiconductor wafers were characterized using four-point probe techniques. Resistors and transistors made with various levels of boron concentration were exposed to a thermal flux at various fluencies at Los Alamos National Laboratory. The reaction, 10 B+n → L i + α, caused by neutron irradiation, introduces impurities in the silicon lattice thus producing measurable differences in electronic parameters. These changes are likely to be proportional to the fluence of the source, and hence to the neutron flux. The results show that for irradiated resistors possessing very high values of boron concentration there is a significant reduction in resistivity. This trend is not seen for medium or low values of boron. Additionally, there was no observation of significant changes in other electronic parameters such as threshold voltage or trans-conductance, for the transistors exposed and tested.

Zirconium is important because of its mechanical and neutronics properties combined with its extraordinary resistance against corrosion. It’s mainly usage in nuclear engineering is as a nuclear fuel cladding. Even though Zr layers are very resistant to neutron fluence at reactor operating temperatures, high pressure and high radiation doses, exothermic Zr oxidation in accident scenario still occurs. Hydrogen, and a newly created layer of zirconium dioxide, are undesirable products of the high temperature steam oxidation. ZrO 2 adopts a monoclinic crystal structure at room temperature and transitions to tetragonal and cubic at higher temperatures. The volume expansion caused by the cubic to tetragonal to monoclinic transformation induces large stresses and these stresses cause ZrO 2 to crack upon cooling from high temperatures. To stabilize ZrO 2 other oxides may be added. Therefore ZrO 2 with its high ionic conductivity makes it one of the most promising ceramics to protect fuel cladding and the paper focuses on properties of ZrO 2 used on Zr fuel cladding of common power reactors.

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

Integrity and lifetime of reactor pressure vessels are practically determined by their material resistance against fast/non-ductile failure and consequently by their radiation damage resulting in irradiation embrittlement and hardening. Mechanism of irradiation embrittlement of RPV materials depends on selected materials and operation conditions but their values depend in a large extent on reactor design, i.e. on neutron flux/fluence depending on RPV wall. Generally, new RPV design allows smaller neutron fluences but absolute value of irradiation embrittlement still depends on a choice of RPV material. Even though radiation damage (especially irradiation embrittlement) is important for RPV behavior, integrity and lifetime depends, in principle, on final value of applied fracture mechanics parameter — transition temperature. Thus, its initial value as well as its shift due to irradiation embrittlement is of interest but only the embrittlement can be affected during operation. The paper discusses absolute and relative effect of both these parameters for integrity and lifetime assessment for different type of reactors — of old and new as well as the newest design. Then, role of surveillance specimen programs for the evaluation of radiation embrittlement of RPV materials is discussed, first of all from point of view of existing codes and standards. Additionally, some recommendations are given for decreasing the effect of irradiation embrittlement on final integrity and lifetime evaluation.

Fast neutron fluence on pressure vessel is a main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV). In the framework of the dosimetry surveillance program its calculated value is extrapolated from measured reaction rate value on dosimeters located on the core barrel using neutron transport codes. The EDF R&D EFLUVE3D code allows the calculation of neutronic fluence (flux over 1MeV integrated on time) received at the interface between the inner wall and the pressure vessel but also at the capsules of the pressurized water reactor vessels. The reaction rate of the dosimeters located in surveillance capsules can also be calculated by this 3 dimensionaltool. The aim of EFLUVE 3D is to reproduce on a given configuration the flux and reaction rate results resulting from an exact Monte Carlo calculations with nearly the same accuracy. These EFLUVE3D calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rate calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). Based on the adjoint flux theory, the code needs as input: - for each reaction rate, the geometric importance matrice produced for a 3D pin by pin mesh on the base of Green’s functions calculated by the Monte Carlo code TRIPOLI4 associated to the JEFF3.1.1 library; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction in function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by application of the EDF core calculation scheme and a pin by pin power reconstruction methodology. The code does the Source*Importance product which allows the calculation of the flux and the reaction rates at different axial positions and different azimutal positions of the vessel as well as at the surveillance capsules. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI4 results. Theses comparisons show a deviation of less than 0,5% between the two codes equivalent to the order of magnitude of the calculation/stochastic convergence.

Neutron energy spectrum and fluence of the devices in a reactor are important parameters for the users. In this paper, the thermal neutron spectrum of the thermal column in Xi’an pulse reactor were measurement by the method of time of flight (TOF), and the peak of the energy spectrum is 0.0248±0.0006eV and the neutrons average energy is 0.048±0.001eV. With the spectrum results, the average cross section of 235 U(n,f) and 197 Au(n,γ) were calculated are respectively 92.08barn and 538.86barn. With the average cross sections, we measured the neutron fluence with the fission chamber and 197 Au(n, γ) activation method. The measurements of the two methods are respectively 1.08×10 9 n/m 2 s and 1.13×10 9 n/m 2 s. We also calculated the standard uncertainty of the measurements: the energy spectrums’ is less than 5%, and the neutron fluence’ is less than 2.2%.

Reactor Pressure Vessel (RPV) is one of the most important components in nuclear power plant (NPP). The aging mechanism of primary concern for RPV is irradiation embrittlement, which can result in a decrease of fracture toughness of RPV steel. Prediction of irradiation embrittlement for a certain Chinese domestic manufactured A508-3 steel is performed. The calculation results given by the US safety standards, the French RCC-M standards and other international safety standards are compared with each other and compared to the data from commercial operation NPP surveillance program. The effect of neutron fluence is also investigated. Furthermore, the property of the steel against irradiation embrittlement is evaluated with the regulatory requirements in the relevant standards. It can be predicted that the Chinese domestic steel satisfies the requirements in these standards.

The increasing age of the European NPPs and envisaged lifetime extensions up to 80 years require an improved understanding of RPV irradiation embrittlement effects connected with long term operation (LTO). Phenomena which might become important at high neutron fluences (such as late blooming effects and flux effects) must be considered adequately in the safety assessments. Therefore the project LONGLIFE was initiated within the 7 th Framework Programme of the European Commission. The project aims at: i) improved knowledge on LTO phenomena relevant for European reactors; ii) assessment of prediction tools, codes, standards and surveillance guidelines. In the paper, we give an overview of the project structure and the related tasks. Furthermore we present two examples for the experimental evidence of LTO relevant phenomena: the first example is related to the flux dependence of defect cluster formation in a neutron irradiated weld material. We have found that the size of the irradiation induced defects exhibits a flux effect whereas the mechanical properties are almost independent of the flux. The second example refers to the acceleration of irradiation hardening after exceeding a threshold fluence. This effect was observed by means of both small angle neutron scattering (SANS) and tensile testing for low Cu RPV steels irradiated at a temperature of 255 °C. These examples demonstrate that LTO irradiation effects have to be investigated in more detail to guarantee the applicability of the embrittlement surveillance guidelines beyond 40 years of operation.

Visual observation of inner side of a reactor pressure vessel of Japan Materials Testing Reactor (JMTR) was carried out using an underwater camera before the JMTR refurbishment work from the view point of its long term utilization, because the reactor pressure vessel of the JMTR will be used continuously after restart of the JMTR. As a result of the visual observation, the harmful wound was not confirmed. Moreover, there was no loosening of the bolts and the screws. On the other hand, adhesion materials which can be easily removed using the gauze were observed around nozzles in a top closure of the reactor pressure vessel. A major component of the adhesion materials is an iron as a result of the componential analysis. However, no significant problem affecting the integrity of the reactor pressure vessel was observed, and then the integrity of the reactor pressure vessel was confirmed. From view points of the stress corrosion cracking, fast neutron fluence and fatigue, it became clear that the reactor pressure vessel of the JMTR can be used for more than 20 years. The visual observation by the underwater camera is to be carried out periodically to confirm the integrity of the reactor pressure vessel in future.

The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam, India. The reactor went critical in October, 1985 with a core of 23 unique high Plutonium carbide fuel subassemblies and the reactor power was rated for 10.5MWt with peak linear heat rating of fuel at 320W/cm. The extension of the target burn-up of this fuel based on Post Irradiation Examination at different stages enabled progressive expansion of the core and increase in reactor power. The reactor has been operated upto a power level of 18.6MWt/3MWe with a sodium temperature of 482°C max. The reactor has completed 24 years of operation and is currently under periodic safety review by the Atomic Energy Regulatory Board of India. As a part of the periodic safety review, equipment qualification status and ageing management studies have been presented to the regulators. Equipment qualification refers to the ability of the replaceable equipment to meet the functional requirements on demand, accomplished by periodic surveillance, maintenance and replacement. Ageing management addresses the residual life assessment of components which are passive, non-replaceable / replaceable with difficulty, taking into account their life degrading mechanisms. Over a period of time, based on the operational feedback, maintenance difficulties and obsolescence, several major components have been replaced. These include the Neutronic channels, UPS, computers of the Central Data Processing System, main boiler feed pumps, three control rod drive mechanisms, two control rods, central canal plug, deaerator lift pumps, reheaters of the steam water system, station batteries, DM plant and Nitrogen plant. The starting air system of the emergency diesel generators and isolation dampers of the reactor containment building have also been replaced. Regarding the non-replaceable components, residual life assessment has been carried out based on the operational history vis-a`-vis the design limits for each component. The life limiting mechanism of heat transport systems of FBTR are creep and fatigue. Since the reactor has operated only upto a temperature of 444°C till 2007, the creep effect is insignificant. The total number of thermal cycles seen by the reactor components as of 2007 was 163, as against the design cycle of 2000 for most of the components. Hence all the heat transport system components are as good as fresh ones. However, the major life limiting factor has been found to be the Neutronic fluence on the grid plate which supports the core. The fast flux at the grid plate location was measured using Np foils and the residual life of the reactor has been assessed to be 10.5 effective full power years. This paper details the life extension exercise being carried out for FBTR.

The mechanisms of charge coupled devices (CCD) irradiated by protons are analyzed. The simulation models of ionization damage and displacement damage are developed. The charge transfer efficiency (CTE) decreased by proton irradiation is numerically simulated. The CTE degradation caused by different traps and by protons with different energies has been studied respectively. Both surface dark signals induced by proton ionization damage and bulk dark signals induced by proton displacement damage are numerically simulated. The variability of surface dark signals, bulk dark signals, and total dark signals with proton fluence is compared. The simulation results are in agreement with the experimental results of the relevant literatures.

At present, most of the developed neutron dosimeters that have a moderator with a single counter, applied in neutron radiation fields within large range energies from thermal to MeV neutrons, are not a satisfaction to energy response. The purpose of the article is designing a suitable neutron dosimeter for the radiation protection purpose. In order to overcome the disadvantage of the energy response of the neutron dosimeters combined a single sphere with a single counter, three spheres and three 3 He counters were combined for the detector design. The response function of moderators with different thicknesses combined with SP9 3 He counters were calculated with MCNP program MCNP4C [1]. The selection of three different thicknesses of the moderating polyethylene sphere was done with a Matlab program [2]. A suitable combination of three different thicknesses was confirmed for the detector design. The electronic system of the neutron dosimeter was introduced. The fluence to ambient dose-equivalent conversion coefficient were calculated, analyzed and compared with the values recommended in the ICRP 74 Publication [3]. The calculated result explain that it is very significance to this design of neutron dosimeter, it may be applied to the monitor of the ambient dose in the neutron radiation fields, improving at present the status of the energy response of neutron dosimeters.

Austenitic stainless steels in boiling water reactor (BWR) core structures can experience significant fracture toughness reductions at high fluence levels of neutron irradiation. To understand the fracture toughness reductions of irradiated austenitic stainless steels is very important for integrity estimation of core internals. Larger size specimen is needed to obtain the valid J IC of high fracture toughness material. However, it is generally difficult for irradiated material to prepare the enough big specimens for taking valid fracture toughness in ASTM E1820–06 since the material volume is restricted. Stretched zone which is generated between fatigue crack and dimple surface on the CT specimen fracture surface is said to have good linear relationship with J IC . New method for J IC estimation from small CT specimen using stretched zone width was investigated. Cold-worked SUS304 was used for investigation. 0.25–1.5TCT specimens with similar figure of standard dimensions and the CT specimens with different thickness from standard dimensions were prepared for J IC tests. The correlations of specimen width, thickness, critical stretched zone width, J Q and J IC were investigated. J Q obtained from J-R curve tended to decrease with specimen width in the case of similar figure. However, J Q tended to increase with thickness decrease in the case of CT specimen with different thickness from standard dimensions. On the other hand, critical stretched zone width showed the tendency to decrease with thickness decrease in the case of CT specimen with different thickness from standard dimensions. The equation for J IC estimation of invalid small CT specimens was designed by regression analysis using the above relationship. The prediction accuracy achieved within plus-minus 30%.

The investigation of reactor pressure vessel (RPV) materials from decommissioned NPPs offers the unique opportunity to scrutinize the irradiation behaviour under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterisation. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I), irradiated and recovery annealed (IA) and irradiated, recovery annealed and re-irradiated (IAI). The working program is focussed on the characterisation of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T 0 following the ASTM Test Standard E1921-05 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step the trepans taken from the RPV Greifswald Unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T 0 varies through the thickness of the welding seam. The lowest T 0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T 0 shows a wavelike behaviour. The highest T 0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on sub size Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT 41J estimated with results of sub size specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The evaluated Charpy-V TT41J shows a better accordance with the irradiation fluence along the wall thickness than the Master Curve reference temperature T0. The evaluated T 0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. T 0 increases from −120°C at the inner surface to −104°C at a distance of 33 mm from it and again to −115°C at the outer RPV wall. The K Jc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during 2 campaigns operation can be assumed to be low for the weld and base metal.