A dynamic model for element reliability

A dynamic model is developed for a system element reliability distribution over a generalized strength space. A differential equation is obtained describing the time-dependence of the reliability distribution function (RDF). The equation covers a wide class of power reactor system components which perform under intense stress conditions where a standard subdivision into a “burn-in” period, a “chance failures” range and a “wear-our” period is inapplicable.The hazard distribution function (HDF) over strength is introduced within the model and it is shown that a standard hazard rate is a strength-averaged failure intensity parameter with the RDF as a weighting function.It is shown that a well-known “bathtub” form of the hazard rate function corresponds to an analytical solution of the principal RDF transfer equation under some simplifying assumptions.     

The frequency-dependent elements in the code SASSI: A bridge between civil engineers and the soil–structure interaction specialists

After four decades of the intensive studies of the soil-structure interaction (SSI) effects in the field of the NPP seismic analysis there is a certain gap between the SSI specialists and civil engineers. The results obtained using the advanced SSI codes like SASSI are often rather far from the results obtained using general codes (though match the experimental and field data). The reasons for the discrepancies are not clear because none of the parties can recall the results of the “other party” and investigate the influence of various factors causing the difference step by step. As a result, civil engineers neither feel the SSI effects, nor control them. The author believes that the SSI specialists should do the first step forward (a) recalling “viscous” damping in the structures versus the “material” one and (b) convoluting all the SSI wave effects into the format of “soil springs and dashpots”, more or less clear for civil engineers. The tool for both tasks could be a special finite element with frequency-dependent stiffness developed by the author for the code SASSI. This element can represent both soil and structure in the SSI model and help to split various factors influencing seismic response. In the paper the theory and some practical issues concerning the new element are presented.     

PSB-VVER simulation of Kozloduy NPP “loss of feed water transient”

This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions.RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient.The objective of the experiment “loss of feed water”, which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as “integral system effects” and ”natural circulation“. For assessment of the RELAP5 capability to predict the “Integral system effect” phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the “Natural circulation” phenomenon the hot and cold leg temperatures behavior have been investigated.This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Main R&D issues for fast reactor structural design standard

For realization of economical and reliable fast reactor (FR) plants, the Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC) are cooperating on the “Feasibility Study on Commercialized FR Cycle Systems”. To certify the design concepts through evaluation of the structural integrity of FR plants, the research and development of the “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)” is recognized as an essential theme. The FDS focuses on particular failure modes of FRs such as ratchet deformation and creep-fatigue damage due to cyclic thermal loads. For precise evaluation of these modes, the research and development for three main issues is in progress. First, the “Refinement of Failure Criteria” needs to be addressed for particular failure modes of FRs. Secondly, the development of “Guidelines for Inelastic Design Analysis” is conducted to predict elastic plastic and creep deformation under elevated temperature conditions. Lastly, efforts are being made toward preparing “Guidelines for Thermal Load Modeling” for the design of FR components where thermal loads are dominant.     

The use of energy absorbers to protect structures against impact loading

When a flying missible impacts a fixed structure, the interface loading is dependent on the deformation characteristics of both impacting and impacted bodies. If both are too rigid to accommodate the amount of gross deformation required to neutralize the incoming kinetic energy, or if such energy absorption has a chance to proceed in uncontrolled and unreliable ways, then there is a need to interpose a specifically designed “energy absorber” between missile and structure, from which a well-defined load time history can be derived during the course of impact.

The required characteristics of such an energy absorption material are:

• the capability to accommodate large permanent deformation without structural failure; and

• the reliable and controlled “load-deformation” (or “stress-strain”) behaviour under dynamic conditions, with an aim at an optimal square shape curve.

Consideration must also be given to environmental or other disturbing effects, like temperature, humidity, and “out of plane” loading. A short survey is presented of the wide range of energy absorbers already described in technical papers or used in a number of practical safety applications within varied engineering fields (from vehicle crash barriers to high energy pipe whipping restraints). However, with such open a literature, information is usually lacking in the specific data required for design analysis.

The following “energy absorption” materials and processes have thus been further experimentally investigated, with an a aim at pipe whipping restraint application for nuclear power plants:

1. (1) plastic extension of austenitic stainless steel rods;

2. (2) plastic compression of copper bumpers; and

3. (3) punching of lightweight concrete structures.

Dynamic “stress-strain” characteristics have been established for stainless steel bars at several temperatures under representative loading conditions. For this purpose, a test rig has been specifically designed to incorporate a number of adjustable parameters and to behave as a representative “slice” of an actual pipe whipping restraint; typical strain rates are in the 10 sec⁻¹ range. The behaviour of copper bumpers has been compared under static and dynamic conditions (using a conventional drop weight test (DWT) machine); as no significant strain rate effects were emphasized, only static tests have been further developed. The DWT rig was used again to investigate crushing or punching of cellular concrete under varying geometries and loading conditions. To remedy certain deficiencies of the regular commercial grades of cellular concrete, special lightweight mixtures have been studied to optimize material toughness and provide a wider range of specific resistance.Results of this experimental program are presented and discussed. The use of energy absorbers is then illustrated for a few typical pipe whipping restraints. The design of restraints is based on real dynamic characteristics of “energy absorption” material as produced by the test program. To derive design loads of restraints, a number of methods can be used ranging from a simplified “energy balance” graph to sophisticated plastodynamic computer analysis. Typical results are presented and discussed to compare the efficiency of these alternative methods.     

Development of constitutive equations for computer analysis of stainless steel components

Engineering application makes conflicting demands of constitutive equations which are difficult to satisfy simultaneously, so forcing considerable approximation. The difficulty is compounded by the frequent need, in anything but room temperature application, to be able to describe the behaviour of the structural material over a range of temperatures. This is illustrated by considering the spectrum of behaviour of Type 316 stainless steel from room temperature to operation at 600°C. It is found that a simple plasticity model describes the behaviour well at 400°C but is less adequate at 20°C in the presence of “cold creep”. There is a discussion of the way plasticity and creep can both be described, with a systematic interaction but without the restrictions of a single “unified relation” for all inelastic deformation.     

The development of a diagnostic and monitoring system for vibrating pipe branches

This paper deals with a diagnostic and monitoring system for assessing the integrity of pipe branches, during the operation of the nuclear power plant. This system have been developed under the concept of “easy to use without any sophisticated analysis” and “portable”. The accuracy of the diagnosis is based on the model optimization subsystem, which automatically modifies the numerical vibration model so as to fit its natural frequency to the actual natural frequency. The information obtained by this system may be reflected to a maintenance program of the plant to assure more reliable operation of the plant.     

The prediction of 2D thermal detonations and resulting damage potential

The main purpose of this paper is to introduce a new concept for the processes responsible for the escalation and propagation of steam explosions. The concept recognizes that initially only a small quantity of coolant around each coarsely premixed melt mass “sees” the fragmenting debris coming off it, hence it is called the concept of “microinteractions”. We also derive the analytical basis for it, define the nature of the requisite constitutive laws and related experimental data, and demonstrate that this concept is essential for the prediction of steam explosion energetics in large-scale premixtures in 2D geometries. We also provide the first numerical illustrations of this concept, implemented in the computer code .m. Further, we provide the first numerical results of steam explosions in large water pools, i.e. ex-vessel explosions. These results reveal two important mechanisms for explosion “venting” and thus for reducing the dynamic loads on adjacent structures. We conclude that, taken together, the “microinteractions” and “venting” make realistic predictions of steam explosion loads feasible and within reach in the near future.     

Analysis of aircraft impact to concrete structures

Analysis of aircraft impact to nuclear power plant structures is discussed utilizing a simplified model of a “fictitious nuclear building” to perform analyses using LS-DYNA software, representing the loading: (i) by the Riera force history method and (ii) by modeling the crash by impacting a model of a plane similar to Boeing 747-400 to the structure (i.e., “missile–target interaction method”). Points discussed include: (1) comparison of shock loading within the building as obtained from the Riera force history analysis versus from the missile–target interaction analysis, (2) sensitivity of the results on the assumed Riera force loading area, (3) linear versus nonlinear modeling and (4) on failure criteria.     

A method for quantifying the reliability of instrumentation for engineered safety features to improve accident sequence recognition in reactor control rooms

This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to quantitatively evaluate the reliability of the instrumentation for engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel through the conversion of system fault trees to alarm trees. In the alarm tree, possible states of each instrumented alarm are identified as “true” or “false”. In addition, a “warning” status is also defined and integrated into the alarm analysis routine. The impact of this additional status condition on the Boolean laws used to evaluate the alarm trees is examined. An application is described for a BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents.     

Advanced three-dimensional two-phase flow simulation tools for application to reactor safety (ASTAR)

This paper summarizes the cumulative work undertaken in the frame of the EU shared-cost action “ASTAR Project”—the current status and future perspectives in the field of advanced numerical simulation of three-dimensional two-phase flow processes. This 3-year running project, which started in September 2000, involves seven partner institutes from around Europe. Specific emphasis is given to the further development of characteristic-based upwind differencing (also called “hyperbolic”) numerical methods and their application to transient two-phase flow. The paper summarizes the common basis adopted for the physical and mathematical modelling of two-phase flow in the form of a single-pressure “two-fluid” model and the various numerical solution techniques developed by the partners. Several benchmark exercises are presented which have been used as verification and assessment procedures for comparing the different modelling and numerical approaches. Comments on the suitability, accuracy, numerical stability, algorithmic robustness and computational efficiency serve as indicators for the possible extension of these methods to future code development activities. Two further tasks of the ASTAR project dealt with the production of high quality experimental field data in the LINX facility of PSI, for the validation of CFD models for two-phase bubbly flow, and the coupling of a two-phase CFD module with a system code. Details of these tasks have been published separately, and will not be recalled in this paper.     

Heat flux distribution from a volumetrically heated pool with high Rayleigh number

Experimental results are presented on the heat flux distribution at the boundaries of volumetrically heated pools at high enough Rayleigh numbers to be directly relevant to the problem of retention of a molten corium pool inside the lower head of a reactor pressure vessel. The experimental facility, named COPO, is a 2-dimensional “slice”, Joule-heated and geometrically similar in shape (torispherical at 1/2-scale) to the lower head of a VVER-440 reactor. The results show that: the heat flux on the side wall (vertical portion) is essentially uniform; the downward heat flux strongly depends on position along the curved wall; and average fluxes on the side in the downward direction are in agreement with existing correlations, but somewhat underestimated in the upward direction. For the shape considered, the heat flux along the lower curved wall seems to be independent of the presence and extent of the liquid pool (contained by the vertical sidewalls) portion above it.     

Development of the evaluation method for crack propagation due to thermal striping

The evaluation code “THERST” was developed to estimate the fatigue crack propagation behavior under thermal stresses due to high-frequency temperature fluctuations, called “thermal striping”. This paper presents fundamental formulations of the evaluation method and verifications of the evaluation method by FEM analyses. Experimental data were obtained in high cycle thermal fatigue tests and the effect of a multiple crack which is characteristic for a crack under thermal stress is discussed in addition to the results of the FEM analyses. A modification of the evaluation method was performed to take multiple crack effects into account.     

The DEEPSSI project, design, testing and modeling of steam injectors

The DEEPSSI project is a steam injector research programme. Among thermal-hydraulic passive systems, the steam injectors (also called “condensing ejectors” or “steam jet pumps”) are very interesting apparatus with very specific characteristics (high velocity, very low pressure). The envisaged reactor application is the Steam Generator Emergency FeedWater System (EFWS) of Pressurised Water Reactors (PWRs). The heart of this project is the development and the testing of an innovative steam injector design. Three experimental facilities are involved: CLAUDIA in France, IETI in Italy and IMP-PAN in Poland. In these facilities, different design options have been tested and some significant improvements of the initial design have been obtained.In addition to the experimental studies, the development of a steam injector computational model has been undertaken in order to model industrial systems based on steam injectors. The one-dimensional module of the system code CATHARE2 has been chosen to be the basis of this model. The first results obtained have confirmed the capabilities of CATHARE2 to describe the steam injector thermal-hydraulics.     

Development of elevated temperature structural design guide for Japanese demonstration breeder reactor

Programs to develop the “elevated temperature structural design guide for the demonstration fast breeder reactor” (DDS) in Japan have been conducted since 1987. The DDS is to be developed on the basis of the “elevated temperature structural design guide for class 1 components of prototype fast breeder reactors” (ETSDG), by considering structural and material features of the demonstration fast breeder reactor (DFBR) and incorporating results of the latest R&D. This paper describes the progress of the R& D concept of the DDS, and discusses some typical results of current studies on the DDS.     

Air leakage characteristics in cracked concrete

In this paper a mathematical formulation for the air leakage rate through cracks in concrete is given. The formula works well as a good approximation for crack widths up to 1.30 mm and overpressures up to 0.80 MPa. The formula was found by means of systematic air leakage tests using unreinforced test specimens with one “defined single crack”. For thermodynamic formulation, isothermal changes in the gas state during the leakage through the crack was estimated and experimentally proved. By additional leakage tests using reinforced test specimens, practical usability of the leakage formula for reinforced panels with a “typical crack pattern” was checked.     

Hybrid soliton nuclear reactors: A model and simulation (encapsulated long living accelerator driven system)

We intend to explore the potential of Hybrid Soliton Reactors (Réacteur Hybride à Soliton, RHYS) for producing energy. In our case an encapsulated long living fission reactor is driven by a proton accelerator, who produces neutrons on a target. In a first part we give the mathematical approach of such a sub-critical reactor, as an extension of the “Soliton Reactor” which was recently proposed by different authors, Edward Teller, L.P. Feoktistov, and others (H. Sekimoto under the name “Candle reactor”). In a second part we give results of simulations and explore the possibilities to control such a system.     

Autonomous reactor control using model based predictive control for space propulsion applications

Reliable reactor control is important to reactor safety, both in terrestrial and space systems. For a space system, where the time for communication to Earth is significant, autonomous control is imperative. Based on feedback from reactor diagnostics, a controller must be able to automatically adjust to changes in reactor temperature and power level to maintain nominal operation without user intervention. Model-based predictive control (MBPC) is investigated as a potential control methodology for reactor start-up and transient operation in the presence of either a constant or a time varying external source. Bragg-Sitton and Holloway [Bragg-Sitton, S.M., Holloway, J.P., 2004. Reactor start-up and control methodologies. In: El-Genk, M. (Ed.), Proceedings of the Space Technology and Applications International Forum (STAIF-2004), AIP Conference Proceedings 699, pp. 614–622.] assessed the applicability of MBPC to reactor start-up from a cold, zero-power condition in the presence of a time-varying external radiation source, where large fluctuations in the external radiation source can significantly impact a reactor during start-up operations. Here the MBPC algorithm is applied using the point kinetics model to describe the reactor dynamics, with a single group of delayed neutrons and a fast neutron lifetime of 10⁻⁷ s. Controller stability is assessed by carefully considering the dependencies of each component in the defined cost (objective) function and its subsequent effect on the selected “optimal” control maneuvers. Additional analysis demonstrates the effectiveness of the controller when a lower fidelity reactor kinetics model is adopted for the model system versus using a full six-group delayed neutron representation in the point kinetics equations to represent the “real” system operation.