Structural integrity of lmfbrs including leak before break

The German Basis Safety Concept is an approach which allows the possibility of catastrophic failures to be excluded. It was developed in Germany to render the probabilistic approach unnecessary for safety cases relating to nuclear power plants. The process of evaluation started in 1972, and in 1979 the Basis Safety Concept was officially published and thus became a legal requirement for LWR plants. With appropriate modifications in regard of the particular features of LMFBR, this concept has also been applied to SNR 300. The “Structural Integrity Demonstration Concept” of SNR 300 is based on five principles:

• - Principle of quality by design and fabrication

• - Principle of multiple examination

• - Principle of worst case consideration

• - Principle of operating surveillance and documentation

• - Principle of verification and continuous development.

The same principles are taken over for SNR 2.The specific requirements on the components relevant to safety have to be defined at an early stage so that the components can be designed appropriately to the feasibility of the measures required by the concept.     

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.     

Two-phase flow models in unbounded two-phase critical flows

Referring to a Loss-of-Coolant Accident situation in LWRs, an analysis of the two-phase region just downstream from the broken pipe, in which a two-phase critical flow takes place, has been performed. A characterization of the flow pattern inside the unbounded two-phase jet is given considering:

• - jet's external shape, obtained by means of photographic pictures;

• - pressure profiles inside the jet, obtained by means of a movable “Pitot” gauge;

• - jet phase's distribution information, obtained by means of X-ray pictures.

Jet's X-ray pictures show the existence of a central high-density “core” gradually evaporating all around, which gives place to a characteristic “dart flow” the length of which depends on stagnation thermodynamic conditions.     

Short-term degradation mechanisms of piping

The operation of PWR-type EDF plants has shown that components have been subjected to loadings higher than the design basis loads. For example, localized degradations on nuclear system pipes were found after relatively short times (10-10⁴ hours). The main damage mechanisms involved are:

• - erosion—cavitation arising from the type of hydraulic flow,

• - vibrational fatigue arising from the flow or operation of mechanical

• - corrosion fatigue occurring in some confined spaces (dead ends).

This paper addresses these damage modes and the mitigating steps taken to cope with them, together with the initiatives taken for future reactors.     

PSA as a tool for decision making

The question on “How safe is safe enough?” is being responded presently by deterministic criteria. Probabilistic criteria in support to more rational and less emotional decisions in regulatory and licensing issues, rationalization of resource allocation and research prioritization, among others, have a potential which is only marginally being explored.This paper discussed PSA limitations and proposes three areas for the use of PSA in decision making, namely:

1. (a) preventing accidents,

2. (b) mitigating accidents, and

3. (c) defining regulatory requirements.

Current activities of the International Atomic Energy Agency in these areas are mentioned.PSA studies depict clearly the uncertainties and this is viewed as a positive aspect, which is unique to the use of probabilistic methods.     

Standardized French PWR programs and the application of mechanical analysis

The implementation of the French PWR construction programs is marked by the following factors:

• - the construction of similar plants by series, together with the standardization of plant layout and component design,

• - the various aspects of the plant operation policy adopted by Electricité de France (EdF), the main client,

• - the special provisions of French licensing regulations,

• - the continuous development of technical experience required to support an active export policy.

The effect of these factors on the organization of the programs and on technical achievements will be examined, with special attention being given to the mechanical aspects, which constitute the main subject of the SMiRT conference.     

The application of reliability-based techniques to Technical Specifications

As required by 10 CRF 50.36, plant Technical Specifications (TS) for power reactors are to include: (1) safety limits and limiting safety system settings, (2) limiting conditions of operation (LCO), (3) surveillance requirements (SR), (4) design features, and (5) administrative controls. Purportedly to make the Technical Specifications process (which operating license applicants are required to utilize) more effective and efficient, the NRC developed and required, on a forward-fit basis, the use of Standard Technical Specifications (STS) since 1975.In part, and in attempting to comply with the elements contained within TS, engineering judgment, codes and standards requirements, and manufacturer's recommendations are used as the basis for establishing testing and surveillance policies specified in the SR element of TS. However, situations have arisen which indicate that test and surveillance intervals that are either too short or too long could, through different mechanisms, be adverse to safety.On August 3, 1983, an interoffice and interdisciplinary Executive Directors Office (EDO) Task Group of the US Nuclear Regulatory Commission was formed to address some of these concerns. Their findings, documented in NUREG-1024, resulted in a directive wherein Brookhaven National Laboratory was charted to conduct the needed research to firmly anchor future TSs, TS changes, and reviews of TSs to a probabilistic risk basis. To scope out the project several issues were identified to evaluate the safety implications of technical specifications. In increasing order of importance, they are:

• - What are the possible approaches for evaluating technical specifications?

• - What are the possible measures of technical specification performance?

• - What technical specifications aspects (attributes) need to be evaluated?

• - What are the possible objectives of technical specifications?

This paper describes the development of the program and some of its early findings.     

The HTR module pressure vessel unit, design criteria and safety philosophy

A modular-helium-cooled high temperature reactor system for the cogeneration of electricity and process heat has been developed by Siemens—Interatom.Design, manufacture and operation of the pressure vessel unit will conform to German nuclear codes and standards for LWR's, some deviations or peculiarities for their application to HTR's are unavoidable. These are for instance:

• - The main steam nozzle, through which the steam line at 530°C penetrates the steam generator pressure vessel with a nominal design temperature of 350°C.

• - The pressure test concept in which the preservice pressure test will be performed in complete accordance with the codes and standards at 1.3 times the design pressure of 70 bar using water. Afterwards, the presence of graphite structures, ceramic insulation and, of course, the pebble bed core has to be considered. Pneumatic pressure tests are performed at 1.1 times design pressure accompanied by more detailed ultrasonic examinations.

• - The position of operational material irradiation surveillance specimens has to be chosen carefully. Design postulates concerning the incrase of ΔRT_NDT will pe confirmed in a separate program.

In general, the requirements of the assured safety concept, aimed to rule out catastrophic failure of the pressure vessel unit during lifetime are fulfilled.     

The use of the uniform risk spectra in the seismic PSA for a PWR power station

This paper describes a simple method for incorporating the effects of the uniform risk spectra (URS) in the seismic probabilistic safety assessment (PSA) for a pressurized water reactor (PWR) power station. The “traditional” fragility parameters for a range of critical equipment items in a PWR power station on two typical UK sites are modified to incorporate the URS using this simple method and the effect on the high confidence low probability of failure (HCLPF) acceleration levels and seismic-induced failure probabilities of the equipment items is examined. The results illustrate the potential benefit of using the URS in the seismic PSA for a PWR power station.     

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.     

Probabilistic seismic safety study of an existing nuclear power plant

This study was conducted as part of an overall safety study of the Oyster Creek nuclear power plant. The earthquake hazard was considered as an initiating event that could result in radioactive release from the site as a result of core melt. The probability of earthquake initiated releases were compared with the probability of releases due to other initiating events.Three steps are necessary to evaluate the probability of earthquake initiated core melt.

1. (1) estimate the ground motion (peak ground acceleration) and uncertainty in this estimate as functions of annual probability of occurrence;

2. (2) estimate the conditional probability of failure and its uncertainty for structures, equipment, piping, controls, etc., as functions of ground acceleration; and

3. (3) combine these estimates to obtain probabilities of earthquake induced failure and uncertainties in such estimates to be used in event trees, system models, and fault trees for evaluating the probability of earthquake induced core melt.

This paper concentrates on the first two steps with emphasis on step 2. The major difference between the work presented and previous papers is the development and use of uncertainty estimates for both the ground motion probability estimates and the conditional probability of failure estimates.The ground motion capacity of a structure, component, etc., is treated for simplicity and clarity as a product random variable A given by , where is the best estimate of the median ground acceleration capacity, _R and _U are lognormal random variables with unit median and logarithmic standard deviation β_R and β_U, respectively. β_R expresses the dispersion in the ground acceleration capacity due to underlying randomness from such sources as (1) the variability of an earthquake time-history and thus of structural response when the earthquake is only defined in terms of the peak ground acceleration; and (2) the variability of structural material properties associated with strength, inelastic energy absorption and damping. Essentially, β_R represents those sources of dispersion which cannot be reduced by more detailed evaluation or more data. Uncertainty concerning the ground motion capacity is expressed by β_U which results from such things as (1) lack of complete knowledge of structural material properties; and (2) errors in calculating response due to approximate modelling. This paper presents a methodology (with examples) for estimating , β_R, and β_U for structures and components. These estimates are then used to estimate conditional probabilities of failure with confidence bounds on these estimates.The conclusion is that a rational approach exists for estimating earthquake induced probabilities of failure. Confidence bounds on such estimates can be developed to express uncertainty in the parameters used. Such an approach is preferable over one in which dispersion due to underlying randomness, and due to uncertainty in the data are combined into a single probability of failure estimate with no estimate of the uncertainty in this probability.     

A complex study on the reliability assessment of the containment of a PWR Parts I–III

The intention of this paper is to contribute to the development of methods to be used for the quantification of the risk of nuclear power plants. For this purpose a reliability analysis of a structural component, i.e. a reactor containment structure is carried out. Detailed information in various fields had to be developed and compiled. The project consists of three parts: Part I concentrates mainly on the analysis of the load condition of the steel hull following a Loss of Coolant Accident (LOCA). Part II deals with the material aspects of the design properties of containment steels and furthermore the behaviour of concrete under impact load conditions are discussed. Part III of the paper is concerned on the one hand with external load conditions, and on the other hand with assembling the information of the previous parts to a reliability analysis.The methodology is exemplified by applying the general and theoretical results to the containment of the PWR-plant “Biblis B”.     

An advanced seismic analysis of an NPP powerful turbogenerator on an isolation pedestal

This paper presents a detailed seismic analysis of a powerful high-speed Russian turbine within a Nuclear Power Plant. Although dozens of these turbines have worked reliably since the 1970s worldwide, until the last decade, only simplified structural analyses were available due to the turbines’ complicated overall structure and internal design. The current analysis considers the detailed geometry of the turbine itself and the vibration and seismic isolation system within the turbine's pedestal and the full range of operational, accident and seismic loads like high pressure, outside loads induced by pipelines and so on.To solve the problem of the turbine seismic qualification, the following steps have been taken. The first step was to create detailed finite element models of the turbine's high and low pressure parts and rotor system with bearings. Using such models, corresponding simplified models were developed to be included into the coupled model of the system: “Building–Vibroisolation Pedestal–Turbine” (BVT). The second step was the analysis of that coupled system. Soil–structure interaction was considered using actual soil conditions. Three components of time history acceleration were used to define seismic excitation. As the result of BVT system analysis, a full picture of time history displacements and loads was determined. At the same time, a problem of rotor gaps was solved. In the final step, determined loads were applied to the detailed models of the turbine's parts for seismic qualification of the whole structure.     

Simplified seismic analysis methods in Belgium

Seismic design and analysis of nuclear plant systems, structures and components have requested huge effort and tremendous costs in the past two decades. The extended use of sophisticated, linear response type methods (modal analysis, spectral response) and the associated conservatism are responsible for the significant stiffening of the piping systems and the multiplication of supports and snubbers. The remedy used against the seismic risk seems worse than the pain itself, and safety might be impaired rather than improved. Indeed, system stiffening increases the average load level in normal operation (stresses, fatigue, nozzle loads, etc.); supports do not behave ideally as assumed (friction, rust, etc.) and snubbers are remarkably unreliable. On the other hand, experience with actual earthquakes shows that industrial facilities designed using very simplistic seismic techniques, or even no seismic requirement at all, suffer essentially no damage, even in the case of a large earthquake. This paradox challenges the traditional seismic design techniques, and appeals for revised seismic qualification methods of piping systems. When the assumption of the occurrence of an earthquake event is made in a plant in operation, which has not been designed against seismic criteria, the use of the standard seismic qualification techniques is still more questionable; simplified (quasi-static) techniques offer in this case a valuable and economically justified alternative. The paper describes the application of the quasi-static “modified load coefficient method” to the seismic assessment of the piping in a nuclear plant in operation, designed during the pre-seismic era.     

Design values of Inconel 617 for high temperature reactors operating at temperatures above 800°C

For extrapolation of the time-dependent stress values the following two methods are introduced:

• -extrapolation with constitutive equations

• -extrapolation with time-temperature parameters.

The first method is a system of coupled equations with creep and damaging parts (a physical-mathematical method), the second is an extrapolation method which combines three well-known time-temperature parameters (a statistical method). On the basis of the latter method time-dependent stress intensity values for a high temperature reactor material are derived and as an example the time safety against rupture is shown for a heat exchanger tube at a temperature of 950°C.     

Current perspective on fast reactor containment

The concept of “containment” is to provide a series of physical barriers between the radioactive products of the fission process and the public. All nuclear reactors have several such barriers and LMFBRs have more than most. These barriers are, successively:

1. fuel, which retains fission products;

2. fuel cladding, which encloses the fuel;

3. sodium coolant, which absorbs fission products released through fuel caldding;

4. primary coolant boundary, which has energy absorption and leakage control capabilities;

5. containment building, hereafter referred to as containment, which provides the final engineered barrier for control of radioactive releases;

6. exclusion distance, which provides space for natural attenuation of radioactive releases before reaching the public.

These barriers, along with the design approaches and features which protect their integrity under normal and accident conditions, assure that the public is adequately protected from the potential hazards of radioactivity residing in the core. It is only in the case of hypothesized core disruptive accidents (HCDAs) that these successive barriers can be sufficiently threatened as to pose a significant threat to the public. These HCDAs involve an extremely low probability sequence of successive failures resulting in core cooling imbalances which lead to fuel overheating. Under such conditions, the fuel and cladding barriers can be lost and energy sources can be generated which threaten the primary coolant boundary and containment. This paper addresses current perspectives on containment of HCDAs with emphasis on the approach and programs in the US.     

PISC full scale reactor pressure vessel validation of non-destructive examination

The PISC III Programme involves validation of techniques and procedures and, within this programme, evaluation has now started on the ability to discriminate service induced defects from indications produced by fabrication defects in A 508 Class 2 material when sensitive techniques are used.Action No. 2 of PISC III: Full Scale Vessel Testing is designed for the performance demonstration of three groups of inspection procedures:

• - Mechanized ASME type procedures with variable recording level and complementary techniques

• - Industrial full ISI procedures (mechanized);

• - Several detailed evaluation procedures (generally mechanized) based on advanced techniques to be used on defective areas detected by usual inspection.

These procedures, typical for ISI in most of the cases, are applied in four situations which could be typical of old and new LWR pressure vessels:

• - vessel material and welds containing important service and fabrication defects but mixed with base material defects and small welding defects;

• - nozzle to shell welds with typical service defects, often well isolated and distant from other defective areas in rather clear material and/or welds;

• - nozzle inner radius defects;

• - artificially heat and unbranched fatigue defects in the test blocks assembled to simulate a PWR pressure vessel wall portion.

The paper summarizes the PISC II programme results which stress the characteristics of capable NDT techniques, in opposition to material characteristics like acceptable base material defects. It describes the full scale pressure vessel components available to conduct the PISC III exercise with improved ultrasonic techniques.     

Testing to determine relay seismic ruggedness

The seismic qualification of equipment in operating nuclear plants has been identified as a potential safety concern in U.S. Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment in Operating Nuclear Power Plants”. In response to this concern, the Seismic Qualification Utility Group (SQUG), with support from the Electric Power Research Institute (EPRI), has undertaken a program to demonstrate the seismic adequacy of essential equipment by the use of actual experience with such equipment in plants which have undergone significant earthquakes and by the use of available test data for similar equipment. An important part of this program is the development of the methodology and test data for verifying the functionality of electrical relays used in essential circuits needed for plant shutdown during a seismic event. This paper describes the EPRI supported relay testing program to supplement existing relay test data. Many old relays which are used in safe shutdown systems of SQUG plants and for which seismic test data do not exist have been shake-table tested. The testing performed on these relays and the test results for two groups of relays are summarized in this paper.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 3. Evaluation of seismic safety margin

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), had conducted a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate an actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, was used for this test. The test model and the results of pressure and leak tests are described in Part 1. Test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load–deformation relationship are described in Part 2. Part 3 reports the seismic design safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 will report simulation analysis results by a stick model with lumped masses.