Analyses with the FSTATE code; fuel performance in destructive in-pile experiments

Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release and redistribution computations for a wide range of possible transient conditions. In this paper recent code developments are described and applications is made to in-pile experiments undertaken to study fast reactor fuel under accident conditions. Three accident simulations, including a fast and slow ramp-rate overpower as well as a loss of cooling accident sequence, are used as representative examples, and the interpretation of FSTATE computations relative to experimental observations is made.     

Development of ‘Pad Element’ for detailed core deformation analyses and its verification

It is important to understand the deformation behaviors of FBR core in the aspects of safety and economy, and to attain more precise analyses of the core mechanics for sophistication of the core structure design. However, the interaction among subassemblies, which occurs as the result of contact between pad surfaces of adjacent subassemblies, makes the core mechanics complicated. Therefore, it is important to consider the characteristics of pads in the analysis in order to realize detailed analysis. The aim of this study is to develop special ‘Pad Element’, which owes the role to treat the interaction between subassemblies at their pads precisely. In this paper, the detailed formulation of ‘Pad Element’ and the analysis results of an example problem for verification are described.     

Sizewell ‘B’ - a one tenth scale containment model test for the UK PWR programme

To fulfil a commitment given to H.M. Nuclear Installations Inspectorate, the CEGB has carried out a pressure test on a one tenth scale prestressed concrete model of the Sizewell ‘B’ containment building in order to validate the computer codes used in the design of the full size structure.This paper describes the model used, its differences and similarities to the full size structure, the construction of the model, the instrumentation used and outlines the test predictions as compared to such test results as were available immediately after completion of testing.     

Status of developing HPGe γ-ray spectrum analysis code HYPERGAM

HYPERGAM, a new MS-Windows version of HPGe γ-ray spectrum analysis code HYPERMET, is being developed. Graphic user interface is provided and the control parameters of analysis are minimized. The semi-automatic algorithm for γ-ray peak analysis is adopted from the original HYPERMET routines whose performance is well known. The details of fitting result are graphically displayed so the analyst can determine reliability. Simpler methods for peak analysis and system non-linearity calibration are newly added. The present code is widely applicable to γ-ray spectroscopy including neutron activation analysis, but it is specifically designed to become a tool for automatic spectrum–isotopic analysis in environmental radiation monitoring.     

IAMBUS-1 — A digital computer code for the design, in-pile performance prediction and post-irradiation analysis of arbitrary fuel rods: Part I: Theory and modelling

IAMBUS-1^*, a digital computer code for the thermal and mechanical design, in-pile performance prediction and postirradiation analysis of arbitrary fuel rods, will be presented in two parts. Part I describes the theory and modelling and in Part II (to be published in a subsequent issue of Nuclear Engineering and Design) material behaviour will be discussed on a quantitative basis and some numerical results illustrating typical and diverse IAMBUS usage will be analysed.The multi-zone code IAMBUS is built around a sound but flexible mechanical analysis of fuel and cladding. A state of generalized plane strain approximates the cladding; the fuel is modelled by a state of plane stress, a state of generalized plane strain, or a combination of these two well-known stress—strain configurations depending on the macroscopic structure of the fuel prevailing. It is thus possible to follow closely the deformation of the fuel and cladding as these are subjected to varying (in part mutual) loads, beginning with a relatively loose, somewhat random assemblage of minute fuel fragments at BOL and progressing to a quasi compact continuum of fuel at EOL.Cladding analysis includes routines for plasticity, creep and swelling due to void nucleation and growth; in the fuel restructuring, plasticity, creep, swelling due to solid and gaseous fission products, fission-gas release and internal pressure build-up are modelled. Routines for friction and heat transfer between fuel and cladding are also incorporated. No strict temperature-dependent boundary is drawn between typically elastic and plastic behaviour, the multi-zone nature of the code models the gradual transition between these two types of material behaviour observed in practice with increasing temperature.Great care has been exercised in choosing numerical methods, since the most sophisticated/realistic modelling is of limited value if the effort expended in reaching a numerical solution becomes exorbitant. Multi-zone modelling lends itself readily to the method of finite differences. The finite difference equations are solved via the method of secants, modified to guarantee convergence for all IAMBUS functions in a feasible amount of computer time.     

Underground siting of nuclear power plants with emphasis on the ‘cut-and-cover’ technique

This paper provides a survey of activities in the field of underground siting in Europe. Emphasis is placed on the ‘cut-and-cover’ technique because of the geological and topographical conditions prevailing in Germany. Basic considerations show the potential of an additional level of containment, of better protection against external effects, and of better conditions for decommissioning. Design concepts for HTR systems and especially for LWR systems which do not require extensive changes in the above-ground design are defined, and a preliminary assessment is given.     

Transient analysis of Ghana Research Reactor-1 using PARET/ANL thermal–hydraulic code

PARET/ANL (Version 7.3 of 2007) thermal–hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1. The reactivities inserted were 2.1 mk, 4 mk and 6.71 mk. The results obtained are similar to experiment and theoretical studies performed to demonstrate that the reactor is safe to operate. The PARET/ANL (Version 7.3 of 2007) could not simulate the reactivities above 5 mk insertions which were successfully performed in earlier theoretical and experimental studies. This may be attributed to different fluid flow and heat transfer regimes within the flow channels of the reactor that were considered by the codes.     

Experience with the dismantling of three secondary steam generators in unit A in Gundremmingen by the ‘ice-sawing’ technique

The unit A in Gundremmingen (KRB A), was the first commercial nuclear power plant in Germany. It had an electrical power of 250 MW_e and was in operation from 1966–1977. The plant was equipped with a dual cycle boiling water reactor of a former General Electric design and includes three recirculation lines each with a big recirculation pump and a steam generator comparable with those of pressurized water reactors. Therefore dismantling experience is gained for systems and components of boiling water reactors as well as pressurized water reactors.

In early 1980, it was decided to decommission the plant. Actual decommissioning work started in 1983 with the removal of the components and systems in the turbine house. Since 1990 the decommissioning activities have been expanded to all primary water systems inside the reactor building. In 1992 , KRB A obtained a licence for dismantling the remaining activated components like the reactor pressure vessel and the biological shield.

Meanwhile more than 5200 tons of contaminated components have been dismantled. Special cutting and handling tools were tested, developed and optimized for the purpose of working in radiation fields and under water. The dismantling work of the contaminated systems and components ends up in about 6000 tons of material with a rather low amount of waste, especially due to optimized decontamination techniques Eickelpasch et al. (1992). For the dismantling of the three secondary steam generators in the reactor building the ‘ice-sawing’ technique was developed and patented.     

Some comments on the concept of ‘underground siting of nuclear power plants’ — A critical review of the recently elaborated numerous studies

This paper concerns the comprehensive problems of underground nuclear power plants (UNPP) with regard to increased safety considerations. This constructional concept is not new, but has not yet been realized, for commercial facilities it is again a matter for discussion. Recently numerous studies (especially in the USA and the Federal Republic of Germany) have been elaborated and they come to considerably different conclusions — already concerning partial subdivisions. It is the aim of this contribution to critically analyse these studies, especially with regard to the principal question of different basis design criteria, constructional concepts and impacts as well as problems of licensibility and operation. Due to the size of these analyses and first of all due to the lack of in situ experiences it seems too early to give at this time period final (pro or con) recommendations concerning the undergrounding of nuclear facilities.     

Micro-PIXE and channeling PIXE analysis of Ag-doped YBa2Cu3O7−δ thin films

Nuclear Microscopy, utilizing a 2 MeV He⁺ beam for channeling Rutherford Backscattering (RBS) and PIXE analysis, was used to characterise Ag-doped YBa₂Cu₃O_7−δ thin films and measure the lateral distribution of the Ag. The samples were prepared by in situ two-beam pulsed laser deposition in order to investigate the effects of such dopings on critical current densities [1 and 2]. Films deposited at temperatures above 650°C form needle-like surface structures with a length of up to 100 μm; these tend to align with in-plane a–b axis. Results for a sample prepared at a substrate temperature of 730°C and a maximum Ag concentration of 5 at.% are discussed. The needle-like structures were found to be rich in Ag and Cu, and the YBa₂Cu₃O_7−δ film contained 0.02 at.% Ag. Broad beam PIXE-channeling results indicate that 19% of the Ag is substitutional.     

Application of ‘natural coordinate system’ in the finite element solution of multigroup neutron diffusion equation

A simple method of generating stiffness matrices for the solution of multigroup diffusion equation by ‘natural coordinate system’ has been presented. A comparative study has been made using triangular elements with linear model, triangular elements with quadratic model and rectangular elements with bilinear model to demonstrate their relative efficiencies. The quadratic interpolation model has been shown to be superior to linear and bilinear models with respect to computing time, computer storage and relative error in K_eff for a two group diffusion example. The flexibility of the finite element treatment has been demonstrated by the calculation of the reactivity of a partially inserted control rod. Good agreement has been obtained with a perturbation calculation.     

Innovative fast breeder reactor concept ‘RAPID’ for improvement of reactor performance and proliferation resistance

The 60 MWe metal fueled fast breeder reactor concept ‘RAPID’ to improve reactor performance and proliferation resistance has been demonstrated. The reactor can be operated without refueling for up to 5 years. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly (IFA) instead of conventional fuel subassemblies. RAPID concept enables quick and simplified refueling by replacing an IFA in which all the core and blanket fuel elements are comprised. An on-site storage cask achieves on-site decay heat removal of an IFA. After 3 years of on-site storage, an IFA together with an on-site storage cask can be transported directly to the reprocessing plant without any intermediate steps. Significant improvement of inherent safety features and plant availability has been discussed. Decay heat removal capability, safety consideration on criticality of the IFA and shielding design of the on-site storage cask has been confirmed. The RAPID refueling concept possesses high resistance to state-supported removal of plutonium for nuclear weapons production.     

Verification study of the FORE-2M nuclear/thermal-hydraulic analysis computer code

The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments.     

Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code

Nowadays, the coupled codes technique, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for carrying out best estimate (BE) simulation of complex transient in nuclear power plants (NPP). This technique is particularly suitable for transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Such complex interactions are encountered under normal and abnormal operating conditions of a boiling water reactors (BWR). In such reactors Oscillations may take place owing to the dynamic behavior of the liquid-steam mixture used for removing the thermal power. Therefore, it is necessary to be able to detect in a reliable way these oscillations. The purpose of this work is to characterize one aspect of these unstable behaviors using the coupled codes technique. The evaluation is performed against Peach Bottom-2 low-flow stability tests number 3 using the coupled RELAP5/PARCS code. In this transient dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein assessed and compared against the available experimental data.     

FARST — A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows:

1. (i) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod.

2. (ii) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method.

3. (iii) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions.

4. (iv) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as “jump relocation model”.

The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR).The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant.     

Influence of thermal properties of zirconia shroud on analysis of PHEBUS FPT0 bundle degradation test with ICARE2 code

In the FPT0 test of the PHEBUS/FP program, it was observed that the fraction of liquefied UO₂ reached 50%, which is much larger than the expected maximum value of 20%. Most of the post-test analyses with various computer codes underpredicted the bundle temperature during a late phase and could not reproduce such a large core degradation. In most of the previous analyses, the shroud thermal conductivity evaluated based on the Pears' ZrO₂ specific heat data and the thermal diffusivity measured by JAERI was used. However, recent thermal property data books adopt a lower specific heat than measured by Coughlin and King's at high temperature. The present analyses with ICARE2 showed that the FPT0 bundle behavior could be mostly reproduced by using the shroud thermal conductivity based on Coughlin and King's. If the present calculation is assumed to be correct enough, the shroud thermal conductivity at high temperature could be smaller than the current evaluation based on the Pears' data. Since the shroud thermal conductivity has thus a strong effect on the bundle behavior, further measurement and evaluation of the thermal properties of the shroud are highly recommended.     

NRI Řež solution of Exercise 3 of VVER-1000 Coolant Transient Benchmark – Phase 1 with coupled codes DYN3D–ATHLET

The paper presents a solution of VVER-1000 Coolant Transient Benchmark – Phase 1 (V1000CT-1) of Exercise 3 performed with the coupled reactor dynamic code DYN3D and system code ATHLET at NRI Řež. The first part of the paper contains brief characteristics of VVER-1000 NPP input deck and describes also the applied reactor core model. The second part introduces the steady-state results and important time dependencies, compared with experimental values. The calculation results show that such type of transient can be realistically described by the coupled codes DYN3D–ATHLET.     

Measurement of β-ray maximum energy with an HPGe detector by means of β-annihilation photons coincidence

A β⁺-ray detection system free from summation of annihilation photons has been constructed for the determination of Q_EC-values. It consists of an HPGe β-ray detector and two pairs of BaF₂ scintillation detectors for annihilation photons. A Q_EC-value of 4.83(4) MeV is obtained for ¹²⁶Cs separated with the JAERI on-line isotope separator.     

Phase equilibrium of PuO2−x − Pu2O3 based on first-principles calculations and configurational entropy change

Combination of an oxygen vacancy formation energy calculated using first-principles approach and the configurational entropy change treated within the framework of statistical mechanics gives an expression of the Gibbs free energy at large deviation from stoichiometry of plutonium oxide PuO₂. An oxygen vacancy formation energy 4.20 eV derived from our previously first-principles calculation was used to evaluate the Gibbs free energy change due to oxygen vacancies in the crystal. The oxygen partial pressures then can be evaluated from the change of the free energy with two fitting parameters (a vacancy-vacancy interaction energy and vibration entropy change due to induced vacancies). Derived thermodynamic expression for the free energy based on the SGTE thermodynamic data for the stoichiometric PuO₂ and the Pu₂O₃ compounds was further incorporated into the CALPHAD modeling, then phase equilibrium between the stoichiometric Pu₂O₃ and non-stoichiometric PuO_2−x were reproduced.