PWR safety/relief valve blowdown analysis experience

The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively.General approaches for generating forcing functions from thermalfluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the “acceleration or wave force” method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawback are discussed.Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.     

Criteria for determining instability due to temperature changes during transient reactor operation

In this paper we study the kinetics of uranium-graphite reactors of the Calder-Hall type taking into account the temperature coefficient of reactivity and its changes during the burning up of the fuel. We obtain the conditions of instability for the reactor. We show that in reactors of the Calder-Hall type when the fuel is burnt up 3000 Mwatt-day/m the minimum positive period of the reactor must be not less than 170 seconds.Translated from an English paper obtained from India.The author wish to express their deep gratitude to B. M. Udgaonkar for his support and valuable advice and also to A. S. Anikindi for his help in performing the calculations.     

Dynamic effects on structures and equipment due to Safety Relief Valve discharge loads

This paper presents analytical models, procedure, and results of a sensitivity study performed to investigate how the Safety Relief Valves (SRV) response of the steel containment of a Boiling Water Reactor (BWR) Mark III plant is influenced by the high frequency energy or noise associated with the idealized bubble forcing function. Various possible structural modifications to those plants already designed to have a steel containment are also presented and discussed with regard to minimizing the dynamic effects of SRV discharge loads.Special attention is given to the concept of filling concrete in the annulus between the steel containment and the shield building. The dynamic analyses of the containment structures with and without concrete in the annulus were performed and the results compared. Throughout the paper, several aspects of the SRV dynamic problem are emphasized and the relevant areas are identified for further investigations and studies.     

Numerical simulation of local non-linearities due to tube-support interaction

Flow-induced vibrations are liable for the majority of failures due to tube wear in process equipment such as steam generator and command gaps that often have weak support conditions. Local impact and sliding may occur. In order to evaluate the wear of the components, we need to compute adequately the localized contact forces and other parameters. In this paper a technique for solving friction-gap problems in tube support structures is described. The technique can be applied to various problems and allows a quite reasonable computation time.     

Determination of stress within ZIRCALOY cladding due to pellet-cladding interaction

A precise calculation of the stress distribution within the Zircaloy cladding of a water-cooled reactor fuel rod subjected to a power increase is a complex problem which, in general, requires a computer code to integrate the behaviour of both the fuel and cladding. This paper develops a simplified model which decouples the clad and fuel pellet analyses, by considering two extremes of fuel pellet mechanical behaviour, which lead to two widely different boundary conditions at the pellet-clad interface. An axisymmetric fuel rod code can be used to give the mean cladding hoop strain imposed by the thermal expansion of the pellet, and when the interfacial friction coefficient is 0.5, this information along with the frictional boundary condition can be used to determine the stress distribution within the cladding near a fuel pellet crack. Results from this simplified approach, which does not involve an integrated code, are used to study the growth of stress corrosion cracks within the cladding.     

Molten fuel-coolant interaction phenomena with application to carbide fuel safety

This paper reviews the major phases occurring during an energetic molten fuel/coolant interaction (MFCI), the categories of interaction and modes of contact between molten fuel and liquid coolant, the film boiling destabilization and collapse mechanisms, and the important fragmentation mechanisms of the melt. Two major models that describe the processes involved in an MFCI event are discussed: the spontaneous nucleation model and the pressure detonation model. Finally, the MFCI experiments involving carbide fuel and liquid sodium are reviewed and the potential for an energetic interaction between molten carbide fuel and liquid sodium is discussed. Recommendations are given for future work on MFCI phenomena relative to the carbide fuel/sodium system.     

Analytically predicted versus measured response of a free-standing steel containment vessel subjected to safety-relief valve discharge loads

Following the actuation of safety-relief valves in BWR nuclear power plants, first water then air and steam are cleared from the discharge lines through quencher devices into a suppression pool. This clearing results in water spike, air bubble, and condensation pressure loads applied to structures in the pool, and the surrounding containment vessel.The Leibstadt Nuclear Power Plant has the only free-standing steel Mark III containment vessel in the world. All other steel Mark III containment vessels have concrete backing in the suppression pool region, which dampens clearing load responses. As such, it is of interest to note how this steel vessel responds to discharge pressures, and compare these responses to analytically predicted results.The purpose of this paper is to compare the analytical results used to design the steel containment vessel with the responses measured during in-plant testing. The analytical methods considered the effects of fluid-structure interaction. The test program included initial and consecutive actuations of a single valve, and initial actuation of multiple (four) valves. The conclusion of the comparison is that, in general, there are large conservatisms in the analytical predictions versus measured responses.     

Dust particle formation due to interaction between graphite and helicon deuterium plasmas

The collection of dust particles using divertor simulation helicon plasmas has been carried out to examine dust formation due to the interaction between a graphite target and deuterium plasmas, which are planned to operate in the large helical device (LHD) at the Japanese National Institute for Fusion Science (NIFS). The collected dust particles are classified into three types: (i) small spherical particles below 400 nm in size, (ii) agglomerates whose primary particles have a size of about 10 nm, and (iii) large flakes above 1 μm in size. These features are quite similar to those obtained through hydrogen plasma operation, indicating that the dust formation mechanisms due to the interaction between a carbon wall and a plasma of deuterium, which is the isotope of hydrogen, is probably similar to those of hydrogen.     

Preliminary results of the experimental study of PFCs exposure to ELMs-like transient loads followed by high heat flux thermal fatigue

This paper deals with the empirical evaluation of the plasma facing components (PFCs) lifetime under transient events, such as type I edge localized modes (ELMs), and high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events. The first results of experiments which are ongoing in the frame of an EU/RF collaboration are presented, in particular the results of the first campaign of exposure to ELMs-like load and HHF thermal fatigue performed on PFC monoblock mock ups. For carbon-fiber composite material the erosion was determined by the analysis of the PAN fibres. The erosion of tungsten is reported as melt layer movement and crack formation. As a preliminary result of the experiments performed one can say that the mock up power handling capability seems so far not to be affected by the ELMs and HHF cycling tests.     

Development of safety analysis and constraint detection techniques for process interaction errors

Among the new failure modes introduced by computer into safety systems, the process interaction error is the most unpredictable and complicated failure mode, which may cause disastrous consequences. This paper presents safety analysis and constraint detection techniques for process interaction errors among hardware, software, and human processes. Among interaction errors, the most dreadful ones are those that involve run-time misinterpretation from a logic process. We call them the “semantic interaction errors”. Such abnormal interaction is not adequately emphasized in current research. In our static analysis, we provide a fault tree template focusing on semantic interaction errors by checking conflicting pre-conditions and post-conditions among interacting processes. Thus, far-fetched, but highly risky, interaction scenarios involve interpretation errors can be identified. For run-time monitoring, a range of constraint types is proposed for checking abnormal signs at run time. We extend current constraints to a broader relational level and a global level, considering process/device dependencies and physical conservation rules in order to detect process interaction errors. The proposed techniques can reduce abnormal interactions; they can also be used to assist in safety-case construction.     

Validation of frey for the safety analysis of LWR fuel using transient fuel rod experiments

The analysis and comparison of severe light water reactor transient experiments are presented from the FREY verification and validation effort. The purpose of this study was to validate the predictive capabilities of the code for severe transient analysis. The FREY code, developed under the sponsorship of the Electric Power Research Institute, uses a two-dimensional finite-element computational method for the thermomechanical analysis of LWR fuel rods under steady state and transient conditions. A total of 10 test fuel rods from experimental programs conducted in both the Power Burst Facility and the Transient Reactor Test Facility have been used in this study. The fuel rods were selected from the following test programs: Power Coolant Mismatch Tests, PCM-2 and PCM-4: Reactivity Initiated Accident Test, RIA 1–2; Loss-of-Coolant Accident Test, LOC-3; First Fuel Rod Failure Test, FRF-1; and Irradiation Effects Test, IE-3. The test programs used in this study cover a large range of code applications for severe transient analysis. The methods used to model the fuel, cladding, and coolant geometry are discussed in addition to experimental data comparisons. The results of the PCM-2, RIA 1–2, and FRF-1 analyses are presented to highlight the full two-dimensional modeling capabilities of FREY and to compare the thermal and mechanical measurements with FREY's prediction. The comparisons show good general agreement, with a tendency for FREY to overpredict the peak cladding surface temperature for a few cases where strong three-dimensional effects have been identified.     

Experimental verification of the fast reactor safety analysis code SIMMER-III for transient bubble behavior with condensation

Experimental verification of a reactor safety analysis code, SIMMER-III, was undertaken for transient behaviors of large-scale bubbles with condensation. The present study aimed to verify the code for numerical simulations of relatively short-time-scale multi-phase, multi-component hydraulic problems. Among these, vaporization and condensation, or simultaneous heat and mass transfer, play important roles. In this study, a series of transient bubble behavior experiments dedicated to condensation phenomena with noncondensable gases was carried out. In the experiments, a pressurized mixture of noncondensable gas and steam was discharged as a large-scale single bubble into a cylindrical pool filled with stagnant subcooled water. The concentration of noncondensable gas was taken as an experimental parameter as was the species of noncondensable gas. The characteristics of transient behavior of large-scale bubbles with condensation observed in the experiments were estimated through experimental analyses using SIMMER-III. In the experiments with steam condensation, dispersion of the gas mixture discharged into the liquid pool was accompanied by vapor condensation at the bubble surface. SIMMER-III simulations suggested that the noncondensable gas had a less inhibiting effect on the condensation of large-scale bubbles. This is a different characteristic to that of the quasi-steady condensation of small-scale bubbles observed in our previous experiments.     

Development of advanced expansion due to compression (A-EDC) test method for safety evaluation of degraded nuclear fuel cladding materials

Expansion due to compression (EDC) test has been applied to evaluate the performance of nuclear fuel claddings where pellet-cladding mechanical interaction (PCMI) is introduced by swelling of fuel pellets and is triggered by the larger hoop deformation of the pellets, especially during accidental transients. The purpose of this study is to modify the EDC test to describe PCMI, specimen volume reduction and others. Ring-shaped specimens were cut from Zry-4 cladding tubes. Cylindrical metal pellets with 8 mm in diameter and 15 mm in maximum height were used as inner pellets. Expansion of the specimens due to the inner pellet compression was performed at room temperature. The experimental data were further analyzed by finite element method. Through the survey in the variation of the specimen and core, specimen size and inner pellet geometry were optimized. Excellent reproducibility with less error was confirmed. The uniaxial tension condition in the hoop direction up to the specimen failure was confirmed. Hoop stress–hoop strain curves were successfully derived.     

Heat deposition in thick targets due to interaction of high energy protons and thermal hydraulics analysis

Heat deposition inside thick targets due to interaction of high energy protons (E_p ∼ GeV) has been estimated using an improved version of the Monte Carlo simulation code CASCADE.04.h. The results are compared with the available experimental data for thick targets of Be, Al, Fe, Cu, Pb and Bi at proton energies of 0.8 GeV, 1.0 GeV and 1.2 GeV. A more continuous heat deposition approach which has been adopted in CASCADE.04.h yields results which are in better agreement with the experimental data as compared to the ones from the earlier version of CASCADE.04. The results are also compared with the predictions of the FLUKA Monte Carlo code. Both CASCADE.04.h and FLUKA predictions are nearly similar for heavy targets and both agree with the experimental measurements. However, they do have differences in predictions for lighter targets where measurements also differ from the predictions. It is observed that the maximum heat loss in thick targets occurs at the beginning of the target due to increasing nuclear reaction contributions. This aspect is crucial in designing the window of a spallation neutron target employed in an accelerator driven sub-critical system (ADS) as this is the first material to be traversed by the proton beam and is subjected to the maximum temperature gradient. Optimization of the target-window parameters requires a careful estimation of heat deposition in the window region and this has been demonstrated through thermal hydraulic studies related to the design of a realistic lead bismuth eutectic (LBE) spallation neutron target for an ADS system.     

Variation of pressure load on flexible plate due to interaction between an expansion wave and the flexible plate

The interaction between an expansion wave and a flexible plate was investigated as a model of the fluid-structure interaction which occurs during the loss-of-coolant accident in pressurized water reactors. The test section, which was connected to an expansion tube, was divided into two regions by the flexible dividing plate, and pressure variations in both regions were measured. Two-dimensional flow equations were numerically solved coupled with an equation of the deflection of the plate. The reduction of pressure loads on the plate due to the interaction between the plate and the expansion wave only occurred within the early stage when the pressure on the back side of the plate was constant. The pressure in the back-side region of the plate was based on the volume change in that region, and the net effect of the pressure change in that region was the reduction of the load on the plate. The pressure in the outlet-side region of the plate was not affected by the pressure in the back-side region. Pressure disturbances generated by the deflection of the dividing plate were found to propagate around the plate when the test section was not perfectly divided into two regions. It was concluded that the conditions around the structure (back-side volume of the structure, for example) should be considered in the evaluation of the pressure load on the structure when the fluid-structure interaction phenomena may be important.     

Transportation safety risk for source recovery versus consequence of leaving radioactive sources in place and vulnerable due to limited transport options or denial of shipment

Abstract

Potential risks associated with transportation safety of recovered radioactive sources in normal commerce are rhetorically compared to the latent risk of not recovering disused radioactive sources due to limited transport options or outright denial of shipment. It is essential, during each phase of the recovery process, to ensure secure, timely, cost effective and reliable means to return vulnerable radioactive sources to safe and protected locations by land, sea and/or air transport. In some cases, only limited transport options exist or denials of shipment may occur that impede the recovery process. Risks associated with normal transportation of recovered sources are considered less significant than the risks related to leaving disused radioactive sources at their original location.     

A consistent approach to assess safety criteria for reactivity initiated accidents

In the context of more and more demanding reactor managements, the fuel assembly discharge burn-up increases and raises the question of the current safety criteria relevance. In order to assess new safety criteria for reactivity initiated accidents, the IRSN is developing a consistent and original approach to assess safety. This approach is based on:

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A thorough understanding of the physical mechanisms involved in each phase (PCMI and post-boiling phases) of the RIA, supported by the interpretation of the experimental database. This experimental data is constituted of global test outcomes, such as CABRI or Nuclear Safety Research Reactor (NSRR) experiments, and analytical program outcomes, such as PATRICIA tests, intending to understand some particular physical phenomena;

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The development of computing codes, modelling the physical phenomena. The physical phenomena observed during the tests mentioned above were modelled in the SCANAIR code. SCANAIR is a thermal-mechanical code calculating fuel and clad temperatures and strains during RIA. The CLARIS module is used as a post-calculation tool to evaluate the clad failure risk based on critical flaw depth. These computing codes were validated by global and analytical tests results;

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The development of a methodology. The first step of this methodology is the identification of all the parameters affecting the hydride rim depth. Besides, an envelope curve resulting from burst tests giving the hydride rim depth versus oxidation thickness is defined. After that, the critical flaw depth for a given energy pulse is calculated then compared to the hydride rim depth. This methodology results in an energy or enthalpy limit versus burn-up.

This approach is planned to be followed for each phase of the RIA. An example of application is presented to evaluate a PCMI limit for a zircaloy-4 cladding UO₂ rod at Hot Zero Power.     

Effect on non-linear soil-structure interaction due to base slab uplift on the seismic response of a high-temperature gas-cooled reactor (HTGR)

In high seismic regions it has often been the practice to use oversized base slabs for the major nuclear power plant structures in order to prevent, or at least minimize, the amount of dynamic base slab uplift which will result from the overturning moments developed during seismic ground motion. Two major reasons have been expressed as to why dynamic base slab uplift should be minimized: (1) As nuclear power plants are normally designed for seismic loadings based upon linear analysis, and since soil-structure interaction becomes nonlinear when only a portion of the base slab is in contact with the soil, linear elastic analysis may be unacceptable if base slab uplift occurs (as the resultant design loads may be incorrect), and (2) substantial uplift could cause excessive toe pressures in the supporting soil and significant impact forces when the slab recontacts the soil.The primary purpose of this paper is to evaluate the importance of the nonlinear soil-structure interaction effects resulting from substantial base slab uplift occurring during a seismic excitation. The structure considered for this investigation consisted of the containment building and prestressed concrete reactor vessel (PCRV) for a typical HTGR plant. A simplified dynamic mathematical model was utilized consisting of a conventional lumped mass structure with soil-structure interaction accounted for by translational and rotational springs whose properties are determined by elastic half space theory. Three different site soil conditions (a rock site, a moderately stiff soil, and a soft soil site) and two levels of horizontal ground motion (0.3 and 0.5 g earthquakes) were considered.Based upon the parametric cases analyzed in this investigation, it may be concluded that linear analysis (which ignores the nonlinear soil-structure interaction effects of base slab uplift) can be used to conservatively estimate the important behavior of the base slab even under conditions of substantial base slab uplift. For all cases investigated here, linear analysis resulted in higher base overturning moments, greater toe pressures, and greater heel uplift distances than nonlinear analyses. It may also be concluded that the nonlinear effect of uplift does not result in any significant lengthening of the fundamental period of the structure. Also, except in the short period region (period less than half of the fundamental period) only negligible differences exist between in-structure response spectra based on linear analysis and those based on nonlinear analysis.Finally, it may be concluded that for sites in which soil-structure interaction is not significant, as for the rock site, the peak structural response (shears and moments) at all locations above the base mat are not significantly influenced by the nonlinear effects of base slab uplift. However, for the two soil sites the peak shears and moments are, in a few instances, significantly different between linear and nonlinear analyses. As a result, linear analysis may be used to determine all structural response for rock sites even when there is substantial base slab uplift. However, for soil sites, nonlinear analyses are necessary if substantial base slab uplift occurs.