Nuclear Engineering and Design

The foundations of the finite element methods and their connection to alternative numerical procedures for linear structural problems are outlined. The state-of-the-art of the finite element displacement method in structural mechanics is surveyed.Extension of the method to dynamic structural problems, to initial instability problems of thin plates and shells, to linear fracture mechanics, and to linear thermoelasticity is indicated. Examples representative of structural problems in reactor technology are included.

A semi-empirical model for inverted annular flow film boiling heat transfer was previously developed which incorporates the effects of all relevant independent parameters and agrees well with available experimental data in the pressure range from 0.1 to 2 MPa. In the present work, this model has been modified to extend its applicability to elevated pressures up to 8 MPa. New expressions for the interfacial energy transfer to the subcooled liquid jet, and for the heat transfer enhancement term due to oscillatory interfacial disturbances were derived. Steady-state film boiling heat transfer data for forced upflow of water in a vertical tube were used to determine the unknown coefficients in the new relations. The data used were obtained from steady-state experiments applying different versions of the so-called “hot-patch” technique. Finally, an explicit film boiling heat transfer correlation has been obtained which approaches Bromley's solution for zero flow and saturated conditions. This correlation has been compared with steady-state data from five different sources which cover mass fluxes ranging from 100 to 1010 kg/(m2s) in the pressure range from 0.1 to 8 MPa. Results indicate an rms error of 15.5% and a mean deviation of 12.0% between measured and predicted heat transfer coefficients for 3972 data points.

Many researchers have reported that the root cause of the Chernobyl accident has not been clarified still now. Since many of them discussed the accident without a precise thermal-hydraulic investigation, thermal-hydraulic calculations coupled with neutronic calculations have been done on the basis of the recorded result at the Chernobyl Unit-4. Plant configurations and operational conditions were given to the code on the basis of reported result and published papers. Calculation could trace plant parameters from 1:19:00 to the first power excursion without any discrepancies measured at the Chernobyl Unit-4. Reactivity slightly smaller than 1β by the positive scram is concluded as a possible direct cause of the accident, which acts as a trigger to increase the reactor power. Other possibilities as a trigger of the accident such as cavitation in pumps and pump coast-down were investigated. The importance of the calculation from the stable condition is also described in this paper in order not to bring unnecessary assumptions into the calculation.

An overview of research activities under Euratom Framework Programme (FP) research in nuclear reactor safety is presented. Some nuclear needs are proposed for discussion namely: to ensure flexibility in energy supply, maintain industrial competitiveness by preparing next generation reactors, develop sustainable solutions for fuel cycle management and waste disposal and to maintain nuclear expertise for non-energetic applications. A series of technical and socio-economical facts related to regulatory and industry organizations are presented. Improvement in the impact of Euratom research actions by enhancing their public benefit and added European value is also discussed.

Modelling the behaviour of fission product (FP) in a nuclear reactor coolant system (RCS) undergoing a hypothetical severe accident is an important step in the evaluation of radioactive release outside a nuclear power plant. The SOPHAEROS module, part of the ASTEC system, aims at simulating main FP vapour and aerosol phenomena in the RCS. New capabilities of the module are here presented, such as chemistry computation and extended circuit configuration. After a description of the models, two applications are presented. The first one is a complete severe accident sequence considering all the RCS loops. It shows a non-negligible FP retention in all the loops and outlines the need for a deterministic approach to better account of all the RCS and not only the main estimated path for FP release. The second application is focused on a complex part of the RCS generally simplified in all the FP transport model: the reactor vessel upper plenum. FP retention in this volume is not well estimated. The SOPHAEROS code applied to this volume shows that FP retention is under-estimated, possibly leading to over conservative results.

Severe accident studies for very low frequency events for VVER-1000 (V320) are carried out to estimate in-vessel damage progression under steam-rich and starved conditions. The analyses with code ASTEC, jointly developed by IRSN (France) and GRS, Germany), have shown the influence of steam environment on core heat-up followed by material relocation, hydrogen production, vessel failure and aerosol generation along with release to containment. Hydro-accumulator injection for studied transients also gives rise to a steam-rich environment enhancing the material oxidation depending on the injection time and period. The generated information along with PSA-Level 2 is helpful to decide Plant Damage State (PDS) and fruitfully develop accident management strategies for the plant.

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads.

The surveillance programmes of western power reactors include, in many cases, standard reference materials in addition to actual pressure vessel steels. These are specimens cut from standard steel plates (Heavy Section Steel Technology, JRQ, etc.) that are similar in composition and heat treatment to the base material in the respective reactor pressure vessels, and are supposed to serve as a monitor by comparing the radiation embrittlement of the plant-specific material to the reference material, and to detect anomalies in the radiation environment of the surveillance capsules.

Measurements carried out in an original-size VVER-1000 mock-up (V-1000 facility, Kurchatov Institute, Moscow) were used for the validation of three-dimensional neutron-kinetic codes, designed for VVER safety calculations. The significant neutron flux tilt measured in the V-1000 core, which is caused only by radial-reflector asymmetries, was successfully modeled. A good agreement between calculated and measured steady-state powers has been achieved, for relative assembly powers and inner-assembly pin power distributions. Calculated effective multiplication factors exceed unity in all cases. The time behaviour of local powers, measured during two transients that were initiated by control rod moving in a slightly super-critical core, has been well simulated by the neutron-kinetic codes.

Currently, license renewals and plant-life extension are important issues for nuclear industry. Pressure vessel integrity is one of the main concerns related to these issues. Pressure vessel integrity is of prime concern for pressurized reactors, since they operate at higher pressures and neutron fluxes when compared to boiling water reactors. Pressure vessel integrity analyses for two commercial pressurized water reactors are performed in this study; a Westinghouse 4 loop 1000 MW PWR and a VVER 1000/320. Two most limiting loss of coolant accidents (LOCA) for pressurized thermal shock (PTS) are considered and deterministic and probabilistic failure analyses are performed. Differences in eastern and western regulatory approaches are also taken into account. Among the vessels simulated, the maximum nil ductility transition temperatures are found to be below the relevant regulatory limit. However, the results of probabilistic analyses are observed to be above the prescribed national limits. This is attributed to the use of rather conservative assumptions used in this study. Findings of this study may help the re-evaluation efforts of PTS screening criteria.

This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Kozloduy nuclear power plant. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of symmetric main coolant pump operation. For certain flow conditions, the experimental evidence of an azimuthal shift of the main loop flows with respect to the cold leg axes (swirl) was found.Such asymmetric flow distribution was analyzed with the Trio_U code. Trio_U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic turbulent flows encountered in nuclear systems. For the presented study, a LES approach was used to simulate turbulent mixing. Therefore, a very precise tetrahedral mesh with more than 10 million control volumes has been created.The Trio_U calculation has correctly reproduced the measured rotation of the flow when the CAD data of the constructed reactor pressure vessel where used. This is also true for the comparison of cold leg to assembly mixing coefficients. Using the design data, the calculated swirl was significantly underestimated. Due to this result, it might be possible to improve with CFD calculations the lower plenum flow mixing matrices which are usually used in system codes.

Activities were carried out within the frame of TACIS-96 Program—Project R2.09/96 named “LBB Applicability Review and Basic Implementation Engineering for Primary Coolant Loop & Surge Line of VVER-1000/320 Nuclear Power Plant”. Main objective of the Project was to perform a Leak Before Break assessment for the Main Coolant Pipes and Pressurizer Surge Line of the Reactor Coolant System of a selected “reference” Unit of a VVER-1000, type V-320 Nuclear Power Plant, which was assumed the Balakovo Unit 2.

Using the cladding creep energy theory, taking into account the WWER-1000 fuel assembly four-year operating period transposition algorithm, as well as considering the disposition of control rods, the location of the axial segment limiting the fuel cladding operation time, at day cycle power maneuvering, has been found. It has been shown that the WWER-1000 fuel element cladding rupture life, at normal variable loading operation conditions, can be controlled by an optimal assignment of the coolant temperature regime and the fuel assembly transposition algorithm.Highlights► The sixth axial segment limits the WWER-1000 fuel cladding lifetime. ► Keeping the inlet coolant temperature constant stabilizes the axial offset. ► The coolant temperature and the fuel transfer method set the cladding lifetime.

The University of Pisa was involved in investigations of an Accident Management (AM) procedure based on passive feed water injection. Some experiments were performed to validate this possibility (e.g. in LOBI and Bethsy facilities) and fully analyzed by thermal hydraulic system codes. Recent activities in which the University of Pisa is engaged (also as leader) are focused on VVER-1000 safety analyses. The idea is now to use the acquired knowledge to explore if a procedure based on passive feed water injection is applicable and can provide any benefits to the Russian design pressurised plant.The postulated accident is a station blackout, in such a way only passive systems are available. The proposed AM is based on secondary and primary side depressurisation in sequence. The secondary side depressurisation performed by the BRU-A valves has the scope to feed passively the SGs with the water left in the feed water lines and in the deaerators. The primary side depressurisation, via the PORV, is foreseen to keep the plant at the lowest pressure (to reduce the energy of the system) and to maximize the “grace time” of the plant. Three cases are here considered: no operator action, application of the optimized AM sequence, application of the AM procedure at the last time when it is effective.The intention of this paper is to show that in case of an unlikely event such a SBO the implementation of a strategy based on systems not designed for specific safety application can have a large impact on the “grace time” of the plant.

In order to directly analyze the structure of oxide films on metals and alloys in high temperature water conditions, an in situ Raman spectroscopic system has been developed, utilizing Ar-laser with sapphire/diamond window assemblies as a part of an Alloy 690 autoclave. The performance of the developed in situ Raman spectroscopic system was verified using high purity NiO, NiFe2O4, Cr2O3, and NiCr2O4 powders. Obtained reference Raman spectra in room temperature air environment agreed well with published results. The developed Raman system has been applied for the characterization of the surface oxide films of a nickel-base Alloy 600 in typical primary water conditions of pressurized water reactors (PWRs); water with 1000 ppm boron and 2 ppm lithium at a pressure of 18 MPa and temperatures ranging up to 350 °C. In situ Raman spectra were collected for Alloy 600 in PWR water conditions at different temperatures up to 350 °C, while the specimens were heated and then cooled. In this study, NiO, Cr2O3, and NiCr2O4 phases were observed for Alloy 600 in PWR water conditions, in reasonable agreement with earlier results of ex situ studies.

Continuum Damage Mechanics (CDM) has developed since the initial works of Kachanov and Rabotnov. The paper gives a review of its main features, of the present possibilities and of further developments.Several aspects are considered successively: •- damage definitions and measures,•- damage growth equations and anisotropy effects,•- use of CDM for local approaches of fracture.Various materials, loading conditions and damaging processes are incorporated in the same general framework. Particular attention is given to the possible connections between different definitions of damage, especially between the CDM definition and the information obtained from material science.

Ontario Hydro has developed a leak-before-break (LBB) methodology for application to large diameter piping (21, 22 and 24 inch) Schedule 100 SA106B heat transport (HT) piping as a design alternative to pipe whip restraints and in recognition of the questionable benefits of providing such devices. Ontario Hydro's LBB approach uses elastic-plastic fracture mechanics (EPFM).In order to assess the stability of HT piping in the presence of hypothetical flaws, the value of the material J-integral associated with crack extension (J— R curve) must be known. In a material test program J-resistance curves were determined from various pipe heats and four different welding procedures that were developed by Ontario Hydro for nuclear Class 1 piping. The test program was designed to investigate and quantify the effect of various factors such as test temperature, crack plane orientation and welding effects which have an influence on fracture properties. An acceptable lower bound J-resistance curve for the piping steels and welds were obtained by machining maximum thickness specimens from the pipes and weldments and by testing side-grooved compact tension specimens. This paper addresses the effect of test temperature and post-weld heat treatment on the J-resistance curves from the welds.The fracture toughness of all the welds at 250°C was lower than that at 20°C. Welds that were post-weld heat treated showed high crack initiation toughness, Jlc, rising J-resistance curves and stable and ductible crack extension. Non post-weld heat treated welds, while remaining tough and ductile, showed comparatively lower JIc, and J-resistance curves at 250°C. This drop in toughness is possibly due to a dynamic strain aging mechanism evidenced by serrated load-displacement curves. The fracture toughness of non post-weld heat treated welds increased significantly after a comparable post-weld heat treatment.The test procedure was validated by comparing three test results against independent tests conducted by Materials Engineering Associates (MEA) of Lanham, Maryland. The JIc and J-resistance curves obtained by Ontario Hydro and MEA were comparable.

The first nuclear reactor space power system, SNAP 10A, was successfully flight tested in April 1965. This paper includes a description of the reactor-thermoelectric system and the development program that preceded the flight. Special emphasis is given to the endurance test experience gained from the nuclear ground test system. The flight test is described in detail and the knowledge gained from the test is given.

A 10 MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) at Tsinghua University of China. The helium turbine and generator system of 10 MW high-temperature gas-cooled reactor (HTR-10GT) is the second phase for the HTR-10 project. It is to set up a direct helium cycle to replace the current steam cycle. The active magnetic bearing (AMB) instead of ordinary mechanical bearing was chosen to support the rotor in the HTR-10GT. This rotor is vertically mounted to hold the turbine machine, compressors and the power generator together. The rotor's length is 7 m, its weight is about 1500 kg and the rotating speed is 15,000 rpm. The structure of the rotor is so complicated that dynamic analysis of the rotor becomes difficult. One of the challenging problems is to exceed natural frequencies with enough stability and safety during reactor start up, power change and shutdown. The dynamic analysis of the rotor is the base for the design of control system. It is important for the rotor to exceed critical speeds. Some kinds of softwares and methods, such as MSC.Marc, Ansys, and the transfer matrix method (TMM), are compared to fully analyze rotor dynamics characteristic in this paper. The modal analysis has been done for the HTR-10GT rotor. MSC.Marc was finally selected to analyze the vibration mode and the natural frequency of the rotor. The effects of AMB stiffness on the critical speeds of the rotor were studied. The design characteristics of the AMB control system for the HTR-10GT were studied and the related experiment to exceed natural frequencies was introduced. The experimental results demonstrate the system functions and validate the control scheme, which will be used in the HTR-10GT project.

Three integral effects tests (IET-1, IET-3, and IET-6) were conducted to investigate the effects of high-pressure melt ejection on direct containment heating. A 1:10 linear scale model of the Zion reactor pressure vessel (RPV), cavity, instrument tunnel, and subcompartment structures were constructed in the Surtsey test facility at Sandia National Laboratories. The RPV was modeled with a melt generator that consisted of a steel pressure barrier, a cast MgO crucible, and a thin steel inner liner. The melt generator/crucible had a hemispherical bottom head containing a graphite limitor plate with a 4 cm exit hole to simulate the ablated hole in the RPV bottom head that would be formed by tube ejection in a severe nuclear power plant accident. The reactor cavity model contained 3.48 kg water with a depth of 0.9 cm that corresponded to condensate levels in the Zion plant. 43 kg iron oxide/aluminum/ chromium thermite was used to simulate molten core debris. The molten thermite in the three tests was driven into the scaled reactor cavity by slightly superheated steam at 7.1, 6.1, and 6.3 MPa for IET-1, IET-3, and IET-6 respectively. The IET-1 atmosphere was pre-inerted with nitrogen, while the IET-3 atmosphere was nitrogen with approximately 9.0 mol% O2. The IET-6 atmosphere was nitrogen with 9.79 mol% O2 and 2.59 mol% pre-existing hydrogen. In IET-1, approximately 233 g mol hydrogen were produced but almost none burned because oxygen was not available. In IET-3, approximately 227 g mol hydrogen were produced and 190 g mol burned. In IET-6, approximately 319 g mol hydrogen were produced and 345 g mol burned. The peak pressure increases in the IET-1, IET-3 and IET-6 experiments were 0.098, 0.246, and 0.279 MPa respectively. In IET-3 and IET-6 hydrogen burned as it was pushed out of the subcompartments into the upper region of the Surtsey vessel. In IET-6, although a substantial amount of pre-existing hydrogen burned, it apparently did not burn on a time scale that made a significant contribution to the peak pressure increase in the vessel.

In order to examine the in-reactor behavior of very-high-density dispersion fuels for high flux performance research reactors, U–10wt.% Mo alloy dispersions in an aluminum matrix have been irradiated at low temperature in the Advanced Test Reactor (ATR). The alloy fuel dispersant was produced by a centrifugal atomization process. The fuel shows stable in-reactor irradiation behavior to a fission density of 5×1027 m−3 at an irradiation temperature of ∼65 °C. The fuel–matrix interaction layer growth rate is similar to that observed in uranium-silicide fuels. The fuel particles have a fine and a relatively narrow fission gas bubble size distribution. There appears to be features in the microstrucure that are the result of segregation of the microstructure in to molybdenum rich and depleted regions on solidification.

Recently, a number of unified elasto—viscoplastic constitutive theories have been proposed and employed for modelling time-dependent inelastic behaviour of initially isotropic, strain-hardening, rate-sensitive engineering materials under general loading histories that combine monotonic, sustained, and cyclic loading stages. This paper presents a theoretical and experimental comparative study of the three leading theories of Chaboche, Bodner, and Walker. The basic structures of these theories are examined, their major advantages and limitations highlighted, and the principal similarities and differences among them explained. Incorporation of dynamic strain ageing modelling capability into these theories is also demonstrated.Moreover, the theories are compared with experiments for elevated temperature application. First, the material parameters of each theory are evaluated and revised using available uniaxial test data for Inconel 718 at 1100°F (593°C). Second, the theories are employed to predict the response to monotonic tension, creep and cyclic loading. These predictions are then verified by comparison with corresponding experimental data available. The three theories appear to be capable of modelling the main features of the inelastic behaviour of Inconel 718 at 1100°F, with varying degrees of acceptability. Chaboche theory, however, seems to offer the greatest promise in this regard.

This paper summarizes the methods utilized by the staff of Sandia National Laboratories in performing four probabilistic risk assessments (PRAs) in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, and consequence analyses. The accident progression and source term analyses performed for NUREG-1150 differ from previous analyses in several ways. Some of the features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analyses. The use of sampling techniques to provide quantitative estimates of the uncertainty in risk required that new approaches be developed for the accident progression and source term analyses. The uncertainties in the consequence analysis were not included in the overall uncertainty estimates in this project. The PRAs performed for NUREG-1150 also devoted a large effort to obtaining expert opinion to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community.

This paper summarizes the findings of the probabilistic risk assessment for Unit 1 of the Surry Power Station performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, and consequence analyses, and the risk results obtained when the results of these analyses are combined with the accident frequency analysis. The results show that the risk from internal initiators is well below the safety goals and is somewhat lower than estimated by Reactor Safety Study (RSS) over a decade ago. The risk from internal initiators is dominated by bypass accidents in the current analysis. The risk from seismic initiators is comparable to or greater than the risk from internal initiators, but still less than that estimated for internal initiators in the RSS. The uncertainty band for both types of initiators is considerably greater than that estimated in the RSS.

The explicit expert judgment process used in NUREG-1150, “Severe Accident Risks: An Assessment for Five US Nuclear Plants”, is discussed in this paper. The main steps of the process are described, including selection of issues and experts, elicitation training, presentation of issues to the experts, preparation of issue analyses by the experts, discussion of issue analyses and elicitation, and recomposition and aggregation of results. To demonstrate the application of the expert judgment process to NUREG-1150, two issues are summarized: one from the accident frequency analysis, and one from the accident progression analysis. Recommendations and insights are provided to improve the use of explicit expert judgment in complex technical issues.

The response of the RBMK Accident Confinement System to a large break LOCA, medium break LOCA and small break LOCA is analyzed using the CONTAIN 11AF code. The effect of Condenser Tray Cooling System failure is investigated for the large break LOCA case. The analysis employs a best estimate mass/energy source and considers both short and long-term responses of the Accident Confinement System. Parametric studies are performed to evaluate the effects of water deposition on the short-term pressure peak and of by-pass leakage on long-term pressure increases.

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

This research placed emphasis on the computer simulated stress distribution on the surface and in the bulk of the materials which are subjected to the water impact causing erosion damage. The erosion damage was predicted by evaluating the spatial and temporal stress wave distribution generated by water impact pressure on 12Cr steel and Stellite 6B as steam turbine materials and TiN as a hard coating material. There were two distinctive stress wave behaviors. Firstly, the large tensile stress at the surface was developed by the Rayleigh wave component which appeared between the water drop and the Rayleigh wave front. After the Rayleigh wave detached from the water drop, the materials were in the tensile stress state which could be related to fracture initiation. Secondly, the largest tensile stress in the bulk was near the surface due to the Rayleigh wave generated at the surface and decreased due to the enlargement of wave front as the radial distance increased. Rayleigh wave's shape was broadened due to the difference between the contact point velocity and the wave front velocity, while its value decayed exponentially in the depth direction. Also, there may be a tendency to produce a circumferential crack by σrr near the surface and a lateral crack by σzz in the sub-surface. The tensile stresses in TiN were much lower than those in 12Cr steel and Stellite 6B due to its higher wave velocity.

This paper identifies several of the possible accident sequences including postulated reactor core melt of such low probability that typical light water reactor nuclear power stations are not specifically designed to mitigate their affects. It then identifies a conceptual design and presents a preliminary cost study for a passive containment system designed to mitigate the consequences of such low probability accident sequences. The conceptual containment design thus developed appears to offer an economic alternative to underground siting or more elaborate and redundant emergency core cooling systems.

The paper compares and discusses the results of two large scale FARO quenching tests known as L-11 and L-14, which involved respectively, 151 kg of 76.7 wt.% UO2+19.2 ZrO2+4.1 Zr and 125 kg of 80 wt.% UO2+20 ZrO2 melts poured into 600-kg, 2-m deep water at saturation at 5.0 MPa. The results are further compared with those of two previous tests performed using a pure oxidic melt, respectively 18 and 44 kg of 80 wt.% UO2+20 ZrO2 melt quenched in 1-m deep water at saturation at 5.0 MPa. In all the tests, significant breakup and quenching took place during the melt fall through the water. No steam explosion occurred. In the tests performed with a pure oxide UO2–ZrO2 melt, part of the corium (from 1/6 to 1/3) did not break up and reached the bottom plate still molten, whatever the depth of the water. Test L-11 data suggest that full oxidation and complete breakup of the melt occurred during the melt fall through the water. A proportion of 64% of the total energy content of the melt was released to the water during this phase (∼1.5 s), against 44% for L-14. The maximum temperature increase of the bottom plate was 330 K (L-14). The mean particle size of the debris ranged between 3.5–4.8 mm.

Ignalina NPP (Lithuania) comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit 1 were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for NPPs with RBMK type reactors. The purpose of this paper is to introduce the content and main results of safety analysis, focusing attention on the container used for spent fuel transportation. The structural integrity, thermal, radiological and nuclear criticality safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel-transportation container and other equipment are in compliance with functional, design and regulatory requirements, assure the required safety level for both normal operation and accident conditions and can be used for fuel transportation.

Specific activities (concentrations) of fission products (FP) and activation products in spent fuel elements of the RBMK-1500 reactor were calculated using SCALE 5 computer code. Different burnup (5.1–21.0 MWd/kg) fuel assemblies were experimentally investigated. Activities of radionuclides present in the coolant water of storage cases of defective fuel elements were experimentally measured and analyzed. Experimental results provide a basis for a quantitative analysis of radionuclide release from spent fuel of the RBMK-1500 reactor. Relative release rates of radionuclides from the fuel matrix were assessed based on a comparison of experimental results with theoretical calculations. On the basis of analysis results released fission and activation products can be divided into several groups according to their release rates from fuel; this can be generalized for radionuclides with similar chemical properties.

The main problem in nuclear energy is providing of safety at all stages of lifetime of nuclear installations in conditions of normal operation, accidents and at shutdown. Ignalina NPP, located in Lithuania, is one of the latest with RBMK reactors at highest capacity. Ignalina NPP has two units, both are closed for decommissioning now (in 2004 and 2009). Both units are equipped with RBMK-1500 reactors, the thermal power output is 4200 MW, the electrical power capacity is 1500 MW for each. In RBMK-1500 reactor the fuel assemblies remain for long time inside reactor core after the final shutdown. The paper discusses possibility of heat removal from the RBMK-1500 core at shutdown condition by natural circulation of water (1) and air (2) inside the fuel channels. In first case the decay heat from fuel assemblies is removed due to natural circulation of water and the piping above reactor core should be cooled by means of ventilation in the drum separator compartments. To warrant free access of air in to fuel channels (in the second case) the reactor cooling system should be completely dry out and the pressure headers and the steam discharge valves in steam lines should be opened. If mentioned conditions will be fulfilled, the reactor core will be cooled by natural circulation of water or air and fuel rods remain intact.

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK.Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed.This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account.

Recent modifications associated primarily with elastic analysis have been made to the creep-fatigue rules of the ASME B&PV Code for use in design of elevated temperature nuclear components. The changes involve: (1) a modified equation for predicting the total strain range per cycle based on a more accurate approximation to the Neuber equation; (2) a variable Poisson's ratio in calculating stresses and strain using elastic equations to account for local thermal strain concentrations; (3) a modified approach for determining the equivalent strain range and; (4) an adjusted design fatigue curve. The modified Neuber approach accounts for the beneficial effects of a residual stress related to the relaxation strength of the material and includes a strain concentration factor greater than the elastic stress concentration factor for geometrical notches for mechanical loads. Notch concentration factors are taken equal to the elastic stress concentration factors for peak thermal and creep strains. The background and intent for these rules are discussed. Correlations of the strain range predictions to those obtained using more rigorous detailed inelastic analysis and test data are presented.

The micro structure of the non-irradiated low-alloyed steels (15Kh2MFA and 15Kh2NMFA) was studied using different spectroscopic methods as the positron annihilation lifetime and Doppler broadening techniques, the Mössbauer spectroscopy, and the Integral low-energy electron Mössbauer spectroscopy as well as the Transmission electron microscopy. Differences in the microstructural parameters of these types of RPV steels are well detectable using all methods. It was confirmed that the heat-affected zone of these steels is the most sensitive place for thermal embrittlement in the reactor. Positron-annihilation lifetime measurements on the successive annealed specimens (XTA, YTA), which simulated the heat-affected zone, showed the rapid increase in the vacancy-type defects formation in the temperature region 525–600°C. Therefore these specimens were studied using Transmission electron microscopy in more detail.

The reliability of safety analysis of nuclear reactor pressure vessels is imperatively dependent upon the reliability of the fracture toughness values used in the analysis. Normally, existing fracture toughness results are analyzed with standard statistical means making use of empirically defined distribution functions for the fracture toughness. Recent developments in the theoretical modelling of cleavage fracture initiation have evolved a sounder statistical description of the macroscopic fracture toughness. It has thus become possible to make a theoretical interpretation of the scatter in brittle fracture toughness results. Here, the theoretical relations are applied to the statistical analysis of brittle fracture toughness results for the reactor pressure vessel steel 15X2MφA and its welds. Based on the analysis, theoretical statistical reference curves for the steels are presented and compared to the presently used standard reference curves.

Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. The current irradiation test series design, designated AFC2, includes minor additions of rare earth elements to simulate expected fission product carry-over from the electrochemical molten salt reprocessing technique. The metal fuel alloys have been fabricated by an arc casting technique. The as-cast fuel alloys have been investigated for phase and thermal properties, specifically, enthalpies of transition, transition temperatures, and room temperature phase characteristics. Results and observations related to these characteristics for the “fresh” fuel alloys are provided. The alloy compositions are based on a U–20Pu–3Am–2Np–15Zr alloy, along with additions of 1 and 1.5 wt% RE (at the expense of U) where RE denotes rare earth alloy of cerium, lanthanum, praseodymium and neodymium. Phase behavior and associated transitions have been compared to available U–Pu–Zr ternary diagrams with acceptable agreement. Enthalpies of transition were deconvoluted from heating and cooling thermal traces for relatively reliable values. The rare earth additions to the base alloy have a minimal influence on the room temperature phases present, but the room temperature phases present slightly impacted the enthalpies of transition and transition temperatures.

Within the context of a Social Compatibility Study on Energy Supply Systems a complex decision making model was used to incorporate scientific expertize and public participation into the process of policy formulation and evaluation. The study was directed by the program group “Technology and Society” of the Nuclear Research Centre Jülich. It consisted of three parts: First, with the aid of value tree analysis the whole spectrum of concern and dimensions relevant to the energy issue in Germany was collected and structured in a combined value tree representing the values and criteria of nine important interest groups in the Federal Republic of Germany. Second, the revealed criteria were translated into indicators. Four different energy scenarios were evaluated with respect to each indicator making use of physical measurement, literature review and expert surveys. Third, the weights for each indicator were elicited by interviewing randomly chosen citizens. Those citizens were informed about the scenarios and their impacts prior to the weighting process in a four day seminar. As a result most citizens favoured more moderate energy scenarios assigning high priority to energy conservation. Nuclear energy was perceived as necessary energy source in the long run, but should be restricted to meet only the demand that cannot be covered by other energy means.

The respective influence of the Von-Mises equivalent stress and of the maximum principal stress on high temperature creep damage of two industrial alloys (INCO 718 and 17-12 SPH stainless steel) are pointed out in a quantitative way through tensile-torsion biaxial tests. Through inversions of the shear component, the important part taken by the principal direction corresponding to the maximum principal stress is also shown. The results are observed to be opposite according to whether the alloy suffers cyclic hardening as 17-12 SPH does or cyclic softening which is the case of Inco 718. These results are supported by metallographic observations. They demand an anisotropic form for the damage variable D, while besides a time dependence, the kinetic equation must include the part taken by the strain.

It is shown in this article that a knowledge of the mechanical behavior of the 17-12 SPH stainless steel for different isotherms (20°C ⩽ T ⩽650°C) is insufficient to describe its behaviour correctly under anisothermal thermomechanical loadings. Indeed, this alloy possesses a temperature history memorization effect at intermediate temperatures.This phenomenon is studied in detail with experiments performed under uniaxial and biaxial mechanical loadings (in-phase or out-of-phase tensile-torsion tests) for different isotherms belonging to judiciously chosen temperature history sequences. The results of cyclic torsion tests under anisothermal thermomechanical conditions are presented where strain and temperature loading evolve simultaneously in phase, in phase opposition, in quarter- and three-quarters-phase. The analysis of the totality of results shows that the material memorizes the maximal value of the cyclic stress peak reached in the zone where the derivative of the maximal cyclic stress with respect to the temperature is positive. This observation can also be applied to both uniaxial and biaxial cyclic tests. From a physical point of view, this behaviour is closely related to the interaction phenomena between dislocations and point defects in insertion solid solutions.In order to describe this history effect, a simple mathematical formulation is proposed in the second part of this article which, after being integrated into a unified viscoplastic model developed elswhere and taking into account a few modifications, leads to an acceptable phenomenological representation of the different experiments presented earlier.

A six-point comprehensive definition of reliability is given and discussed. A fifteen-step “Design by Reliability” methodology is presented and illustrated. Methods for synthesizing the failure governing stress and strength distributions, and for calculating the associated reliability, are presented and illustrated. A method for determining the lower, one-sided confidence limit of the calculated reliability is given and illustrated. The data requirements for this design methodology are discussed, and such data generated at the Reliability Research Laboratory of The University of Arizona are presented. The methodology is compared with the conventional safety factor approach and a unification of the statistical safety factor with the reliability is given.

The Nuclear Safety Information Center (NSIC) serves as a focal point for the collection, analysis, and dissemination of information related to safety problems encountered in the design, analysis, and operation of nuclear facilities. NSIC issues state-of-the-art reports, operates a Selective Dissemination of Information (SDI) program, publishes indexed bibliographies, prepares retrospective bibliographies, answers technical inquiries, and offers counsil and guidance on safety problems. Its reference files are stored in a central computer that is accessible by means of a telecommunications station. This article describes the services offered by the Center, the manner in which it operates, and discusses some of the problems encountered in its first years of operation through 31 December 1968.NSIC's productivity and the demand for its services have increased significantly since its formation in 1963. Technical questions are answered at a rate of 609 per year. The biweekly SDI abstract service is now reaching 1500 participants. Reports and indexed bibliographies are issued routinely to a distribution of over 700. Visits to the Center for consultation and/or use of its files occur at the rate of around 150 per year.

This review of water reactor fuel performance from 1968 to 1973 shows that defect levels in Zircaloy-clad UO2 fuels have, at times, been up to 1 in 3 bundles/reactor yr or ∼1 in 100 pins. Among the consequences of such events, are considerable efforts by manufacturers, designers and operators to understand the mechanisms of defection. As a result it has been possible to feed back the operating experience to give improvement in defect level to ∼1 in 200 bundles/reactor yr or between 1 in 103 and 1 in 104 fuel pins. The various types of defects and limiting features of Zircaloy clad fuel have been classified and fitted to well established reliability technology. There is justification for establishing a statistical data bank on water reactor fuel performance, to establish models and confirm acceptable defect levels, similar to those available on reactor plant. Few defects can yet be described as ‘life limiting’ in that incore lifetime is currently decided on economic grounds. Some potential life-limiting features have been identified during the review which should be the subject of further work to improve the long term performance of water reactor fuel.

An extensive library of computer codes useful for radiation transport or shielding calculations is available from the Radiation Shielding Information Center at Oak Ridge National Laboratory. In addition to the point kernel, Monte Carlo, and discrete ordinates codes used for neutron and gamma-ray transport calculations, the collection includes cross-section libraries and codes for processing cross sections, calculating fission product inventories, proton penetration of spacecraft, electron-photon transport, and analyzing neutron activation detector data to determine spectra. A list of the most current codes is given and essential information for each is included.

In 1972 the light water reactor safety activities conducted at the Karlsruhe Nuclear Research Center (KfK) were combined under the Nuclear Safety Project (PNS). Its primary objective was to assess in quantifiable terms the safety reserves which are provided in nuclear power plant design in a conservative approach. While in the initial phase R&D work conducted under the project was largely characterized by investigations of the design basis accidents, mainly the loss-of-coolant accident, emphasis in the past decade has been shifted more and more towards severe core and core meltdown accident analysis. The activities comprise both theoretical studies and experimental investigations, often performed in adequate, large-scale facilities. All activities have been an essential part of the reactor safety research program of the Federal Ministry for Research and Technology (BMFT) and have been coordinated with a number of other programs conducted in Germany and abroad. This paper gives a broad overview of PNS contributions to LWR safety research in the past 15 years and summarizes the results, comparing them with the general goals defined. In conclusion, the attempt is made to give an outlook on remaining activities in LWR safety research being carried out by KfK.

Information is presented on the recontamination of recirculation piping in commercial boiling water reactors following successive chemical decontaminations or pipe replacement. Where the recirculation pipes were replaced several types of pipe pre-treatments have been used to reduce the rate at which radionuclides were incorporated into the oxide films on the inner pipe surfaces. These pipe treatments are briefly discussed and contamination control effects of the treatments are compared.

The future global role of nuclear power will be determined by its ability to provide economical and safe energy. Nuclear power, like any other substantial contributor to the world's energy needs, must be generated at an acceptable cost and with negligible environmental effects. Besides, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power. The ABB Atom approach to this situation can best be characterized as a `cautious evolution'; for the next decade the company will largely base its offerings to the market on its `evolutionary' light water reactor design, the BWR 90. This design builds closely on the experience from successful construction and operation of its predecessor, the BWR 75 design. In 1995 and 1996, plants of this design achieved an average load factor greater than the 87% set by EUR; the two BWR units at Olkiluoto in Finland are among the very best performing plants in the world, with an average load factor of 94% over the last 7 years. The continued LWR design development focuses on meeting requirements from utilities as well as new regulatory requirements. A particular emphasis is put on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimization of buildings and containment design to decrease construction time and costs, and selection of materials as well as maintenance and operating procedures to even further reduce occupational radiation exposures. Probabilistic safety assessments and life-cycle cost evaluations have become major tools in the design optimization work. The BWR 90 was offered to Finland in the early 1990s, and will now as the first BWR design be reviewed by a number of European utilities with respect to its conformance to the European Utility Requirements (EUR); a specific EUR Volume 3 for the BWR 90 will be the final result. The paper describes some of the unique characteristics of the BWR 90, with emphasis on the features that are most important for achieving improved economy and enhanced safety.

The following topics are discussed in a review of the 1992 performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP): (1) inclusion of potash mining and other disruptions in the definition of possible futures at the WIPP, (2) definition of future drilling rates in the vicinity of the WIPP, (3) possible non-conservatisms in analysis assumptions, (4) characterization of uncertainty in solubilities, and (5) characterization of uncertainty in distribution coefficients. With respect to (1) and (2), the 1992 WIPP PA attempted to follow guidance in the US Environmental Protection Agency's (EPA's) standard for the geological disposal of radioactive waste, 40 CFR 191. Many of the reviewer's comments relate to the EPA's recent guidance in 40 CFR 194 on assessing compliance with 40 CFR 191, which was not published until early 1996 and which makes many changes from the guidance given in 40 CFR 191. The 1996 WIPP PA, which will support a compliance certification application to the EPA, will follow the guidance in 40 CFR 194. With respect to (3), (4) and (5), the WIPP PA tries to avoid both deliberate conservatism and deliberate non-conservatism. The 1992 WIPP PA attempted to use all the information available to it consistent with the requirements to incorporate uncertainty in PA results and to use ‘reasonable expectation’ as the criterion for assessing compliance with 40 CFR 191. Since 1992, many additional results related to solubilities and distribution coefficients have become available. The 1996 WIPP PA will incorporate these new results.

This paper presents damage to industrial equipment in the 1995 Hyogoken–Nanbu Earthquake and measures which should be taken against earthquakes. The investigation covers 44 industries and plants, that is, heavy industry, machinery, precision machinery, steel mills, electric power and so on. The causes of damage are discussed and the measures which should be taken is presented.