An evaluation of containment inerting and air dilution systems as methods for post-accident hydrogen control in BWRS

Hydrogen control is important in post-accident situations because of possibilities for containment rupture due to hydrogen deflagration or detonation. Post-accident hydrogen generation in BWR containments is analyzed as a function of engineered hydrogen control system, assumed either nitrogen inerting or air dilution. Fault tree analysis was applied to assess the failure probability per demand of each system. These failure rates were then combined with the probability of accidents producing various hydrogen generation rates to calculate the overall system hydrogen control probability. Results indicate that both systems render approximately the same overall hydrogen control failure rate (air dilution: 8.3 × 10⁻²−1.1 × 10⁻²; nitrogen inerting: 1.3 × 10⁻²−2 × 10⁻³). Drywell entries and unscheduled shutdowns were also analyzed to determine the impact on the total BWR accident risk as it relates to the decay heat removal system. Results indicate that inerting may increase the overall risk due to a possible increase in the number of unscheduled shutdowns due to a lessened operator ability to correct and identify ‘unidentified’ leakage from the primary coolant system. Further, possible benefits of inerting due to reduced torus corrosion and fire risk in containment appear to be dominated by the possible operations-related disadvantages.     

Consequence evaluation for pressurized thermal shock

In order to estimate the risk associated with Pressurized Thermal Shock (PTS), a sample calculation of the core melt frequency and offsite consequences has been performed for Oconee Unit 1, a Babcock and Wilcox pressurized water reactor located in the United States. Core melt frequency was derived from through-wall-crack frequency estimates based on thermal-hydraulic and fracture mechanics analyses performed by Oak Ridge National Laboratory and Pacific Northwest Laboratory. The mode and timing of containment response was estimated from previous risk studies for Oconee Unit 3 and other plants with large dry containments.The core melt frequency was calculated to be 6 × 10⁻⁶ per reactor year for operation at the PTS screening criterion. Operation of redundant and independent containment heat removal systems results in low probability of containment failure. The risk dominant scenario involves overpressure failure of containment due to failure of containment heat removal. Prompt containment failure was assigned a very low probability (10⁻⁴), and hydrogen burn failure was not considered.The central estimate of annual risk was 5 × 10⁻⁷ early fatalities, 2 × 10⁻⁴ latent cancer fatalities and 0.7 person-rem. These values are minimal compared with other severe accident scenarios.Uncertainties and sensitivies to important parameters are discussed. The response of other types of plants is briefly described.     

A design-phase PSA of a nuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10⁻⁸ year⁻¹ which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10⁻⁸ year⁻¹, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Vapour void fraction in subcooled flow boiling

Experimental data on steam void fraction and axial temperature distribution in an annular boiling channel for low mass-flux forced and natural circulation flow of water with inlet subcooling have been obtained. The ranges of variables covered are: mass flux = 1.4 × 10⁴−1.0 × 10⁵ kg/hr m²; heat flux = 4.5 × 10³−7.5 × 10⁴ kcal/hr m²; and inlet subcooling = 10–70°C. The present and literature data match well with the theoretical void predictions using a four-step method similar to that suggested by Zuber and co-workers.     

Fission product release and its environment impact for normal reactor operations and for relevant accidents

The radioactive concentration in the primary loop and the radioactive release for both normal operations and accidents for the HTR-10 are calculated and presented in the paper. The coated-particle fuel is used in the HTR-10, which has good performance of retaining fission products. Therefore the radioactive concentration in the primary loop of the HTR-10 is very low, and the amount of radioactive release to the environment is also very small for both normal operation and accident conditions. The radiation doses to the public caused by radioactive release for both normal operations and accidents are given in the paper. The results show that the maximum individual effective dose to the public due to the release of airborne radioactivity during normal operations is only 1.4×10⁻⁴ mSv a⁻¹, which is much lower than the dose limit (1 mSv a⁻¹) stipulated by Chinese National Standard GB8703-86. For depressurization accident and water ingress accident, the maximum individual whole-body doses to man are only 7.7×10⁻² and 2.0×10⁻¹ mSv, thyroid doses only 1.7×10⁻¹ and 1.1 mSv, respectively. They are much lower than the prescribed minimum of emergency intervention level (whole-body dose: 5 mSv, thyroid dose: 50 mSv) for sheltering measures stipulated by the Chinese Nuclear Safety Criterion HAD002/03. The conclusion is that the environmental impact is very small for normal operations and accidents for the HTR-10, and the requirements stipulated in the Chinese Nuclear Safety Criterions are satisfied perfectly.     

JR curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents in oxygen-containing high temperature water at 240°C

J_R curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents (0.003 and 0.011 wt.%) were determined at 240°C in oxygen-containing high temperature water as well as in air. The tests were performed by the single-specimen unloading compliance technique at load line displacement rates from 1 × 10⁻⁴ down to 1 × 10⁻⁶ mm s⁻¹ on 20% side-grooved 2T CT specimens in an autoclave testing facility at an oxygen content of 8 ppm and a pressure of 7 MPa under quasi-stagnant flow conditions.In the case of testing in high temperature water, remarkably lower J_R curves than in air at the same load line displacement rate (1 × 10⁻⁴ mm s⁻¹) were obtained. A decrease in the load line displacement rate as well as an increase in the sulphur content of the steel caused a reduction of the J_R curves. At the fastest load line displacement rate a stretch zone could be detected fractographically on the specimens tested in air and in high temperature water and consequently J_i could be determined. When testing in high temperature water, the J_i value of the higher sulphur material type decreases from 45 N mm⁻¹ in air to 3 N mm⁻¹, much more than that of the optimized material type from 51 N mm⁻¹ in air to 20 N mm⁻¹ at 1 × 10⁻⁴ mm s⁻¹.     

Wear behavior of graphite studies in an air-conditioned environment

The wear performance of graphite used in the high-temperature gas-cooled reactor (HTR-10) was researched. The wear mechanism, worn surface and wear debris were analyzed under SEM. Under test conditions, the wear rate was 2.27×10⁻⁷ g/m for surface contact, and 1×10⁻⁶ g/m level for line contact. The main wear mechanisms of graphite were groove and fatigue. The projected area of wear debris followed the logarithm normal school, giving most wear debris as a small sphere and large flake debris as only a small part.     

Fluence dependence of the surface roughness of InP after bombardment

InP(1 0 0) surfaces were sputtered under ultrahigh vacuum conditions by 5 keV ions at an angle of incidence of 41° to the sample normal. The fluence, , used in this study, varied from 1 × 10¹⁴ to 5 × 10¹⁸ cm⁻². The surface topography was investigated using field emission scanning electron microscopy (FE-SEM) and atomic force microscopy (AFM). At the lower fluences ( 5 × 10¹⁶ cm⁻²) only conelike features appeared, similar in shape as was found for noble gas ion bombardment of InP. At the higher fluences, ripples also appeared on the surface. The bombardment-induced topography was quantified using the rms roughness. This parameter showed a linear relationship with the logarithm of the fluence. A model is presented to explain this relationship. The ripple wavelength was also determined using a Fourier transform method. These measurements as a function of fluence do not agree with the predictions of the Bradley–Harper theory.     

Thermal- and radiation-stability of hydrogen-implanted silicon standards for ion-beam analysis

The effect of thermal annealing and particle radiation on the depth profile of 25 keV H⁺ implanted into crystalline and ion-beam amorphised silicon has been studied via the ¹H(¹⁵N, αγ)¹²C reaction. Out-diffusion of hydrogen from crystalline silicon was observed after annealing at 100 °C for 45 min. The corresponding temperature for ion-beam (300 keV, ⁸⁴Kr²⁺) amorphised silicon was between 300 and 500 °C. Radiation damage produced by 3 × 10¹⁵ 6.4 MeV ¹⁵N²⁺ ion cm⁻² lead to effective trapping of the hydrogen in crystalline silicon whilst 1.1 × 10¹⁵ 300 keV Kr²⁺ ions cm⁻² gave rise to significant spreading.     

Pressure vessel integrity and weld inspection procedure

The primary objective of this study is to develop a simple methodology which, when coupled with existing observations on pressure vessel behavior, provides an interrelation between nuclear pressure vessel weld integrity and the parameters of the in-service inspection program, including inspection sample size, frequency and efficiency. The basic input information on rate of generation and development of weld flaws of different sizes and types is drawn primarily from published British and German studies taken almost exclusively from welds of non-nuclear pressure vessels. The input information is varied to reflect differences in weld quality and uncertainty of input data. A modified Markov process is employed and a computer code written to obtain numerical results. If it is assumed that the quality of nuclear reactor welds are the same as the quality of non-nuclear welds (i.e. the data base), then, based on the limitations of the model, the predicted critically sized defect concentration is about 50 × 10⁻⁷ per weld at the end of weld life for welds under both high and low stress if ASME, Section XI, In-Service Inspection Requirements are applied. Based on the British data and the less stringent inspection standards (compared to Section XI) the estimated number of critically sized defects per weld at the end of weld life is 250 × 10⁻⁷ and 170 × 10⁻⁷ per weld for high and low stressed welds, respectively. If it is assumed that the nuclear reactor pressure vessel welds have superior quality to the non-nuclear welds, then the model predicts an appropriately lower probability of critical defects at the end of weld life. A variety of other sensitivity studies are included in the report. Also, a simple methodology to provide an optimal weld inspection program which is consistent with a minimum cost criteria is outlined. It should be noted that the results of this study are based on the limitations of the simple model that was used and on a variety of corresponding assumptions.     

Microstructure alterations in the base material, heat affected zone and weld metal of a 440-VVER-reactor pressure vessel caused by high fluence irradiation during long term operation; material: 15 Ch2MFA ≈0.15 C–2.5 Cr–0.7 Mo–0.3 V

Within the scope of the EC research project Tacis ’91 (‘RPV-Embrittlement’), trepans were taken from the highly irradiated circumferential RPV-weld of the Novovoronesh power plant unit-2 of the type VVER-440/230. The cumulated fast fluence level in this position reaches up to 6.5×10¹⁹/cm² (E>0.5 MeV). In a joint research work, the mechanical properties, the chemical composition, and the microstructure of the base material, the heat affected zone (HAZ), and the weld metal have been investigated in order to study the influence of irradiation, and of post irradiation heat treatment (475°C, 560°C) on the properties. The examination of the microstructure performed by analytical transmission electron microscopy (200 kV) shows the existence of dislocation loops (‘black dots’), irradiation induced precipitates, and segregation of copper in the carbides. These changes in microstructure, which are due to service affection (neutron irradiation, temperature) have occurred more pronounced in the weld metal and the HAZ than in the base material.     

Dissolution of uranium metal without hydride formation or hydrogen gas generation

This study shows that metallic uranium will cleanly dissolve in carbonate-peroxide solution without generation of hydrogen gas or uranium hydride. Metallic uranium shot, 0.5–1 mm diameter, was reacted with ammonium carbonate–hydrogen peroxide solutions ranging in concentration from 0.13 M to 1.0 M carbonate and 0.50 M to 2.0 M peroxide. The dissolution rate was calculated from the reduction in bead mass, and independently by uranium analysis of the solution. The calculated dissolution rate ranged from about 4 × 10⁻³ to 8 × 10⁻³ mm/h, dependent primarily on the peroxide concentration. Hydrogen analysis of the etched beads showed that no detectable hydrogen was introduced into the uranium metal by the etching process.     

Probability-based design criteria for shear walls in nuclear power plants

The development of probability-based criteria for the design of reinforced concrete shear walls subjected to dead load, live load and in-plane earthquake forces in nuclear plants is described. These criteria are determined for flexure and shear limit states in a load and resistance factor design (LRFD) format. The flexure limit state is defined according to traditional principles of ultimate strength analysis, while the shear limit state is established from experimental results. Resistance factors for shear and flexure, load factors for dead and live load, and a load factor for effect of Safe-Shutdown Earthquake are determined for target limit state probabilities of 1.0 × 10⁻⁶ and 1.0 × 10⁻⁵ over a a period of 40 years. Comparisons among the proposed design criteria, ACI 349 and US NRC Standard Review Plan 3.8.4 are included.     

Reduced seismic loads for temporary conditions

A new methodology, developed under an EPRI Tailored Collaboration Project, to calculate and apply reduced seismic loads (RLSs) for evaluation of temporary conditions (TCs) in nuclear power plants using design-basis (DB) allowables is described. The methodology, which was submitted to Nuclear Regulatory Commission (NRC) through the Nuclear Energy Institute (NEI), calculates load reduction factors based on an allowed limit for time-averaged increase in seismic core damage frequency within the duration of a refueling cycle. For this allowable in the range 5×10⁻⁶ to 1×10⁻⁵ per reactor year, substantial reduction relative to DB seismic load is possible. The methodology is equally applicable to plants with and without seismic probabilistic risk analysis model.     

Fuel irradiation of the first batches produced for the Chinese HTR-10

An irradiation test of four spherical fuel elements (SFE) had been performed in the Russian reactor IVV-2M. The elements were sampled randomly from the first and second product batches which were manufactured for the 10 MW high-temperature gas-cooled test reactor (HTR-10). The maximum burnup of the irradiated fuel elements reached 107,000 MWd/tU and the maximum fast neutron fluence was 1.31 × 10²⁵ m⁻². The release-to-birth rate ratio (R/B) did not increase significantly during irradiation. However, an in-pile heating-up test of element SFE 7 in Capsule 5 led to a failure of approximately 6% of the coated particles. After the test it was estimated that the fuel temperature had very likely been much higher than the intended 1600 °C.     

Manufacture and characteristics of spherical fuel elements for the HTR-10

The R&D of spherical fuel elements for the 10 MW high temperature gas-cooled reactor (HTR-10) started in 1986 in China. A process known as cold quasi-isostatic molding was used for manufacturing spherical fuel elements, and about 20,540 spherical fuel elements were produced in 2000 and 2001. Fabrication technology and graphite matrix materials were investigated and optimized. Cold properties of the spherical fuel elements met the design specifications. The mean free uranium fraction of 44 batches was 4.57 × 10⁻⁵. In-pile irradiation test results showed that irradiation did not lead to apparent change in linear dimensional, geometrical density, porosity and strength of matrix graphite samples. No cracks and blisters were observed in spherical fuel elements. This indicated that matrix graphite and spherical fuel elements of HTR-10 met the requirement of design specifications.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

Be in commercial beryllium

¹⁰Be has been observed at a level of about 10⁻¹⁴ in four commercially available beryllium compounds. The possibility that this ¹⁰Be arises from cross-talk in the ion source has been eliminated. On the other hand, beryllium oxide extracted from beryl crystals found at a depth of 30–40 feet in a lithium mine in South Dakota shows no indication of containing ¹⁰Be. An upper limit on its ¹⁰Be:⁹Be ratio is 1.7 × 10⁻¹⁵ at the 95% confidence level.     

Effect of irradiation by 140 Mev Ag ions on the optical and electrical properties of polypropylene/TiO2 composite

Changes in the optical, structural, dielectric properties and surface morphology of a polypropylene/TiO₂ composite due to swift heavy ion irradiation were studied by means of UV–visible spectroscopy, X-ray diffraction, impedance gain phase analyzer and atomic force microscopy. Samples were irradiated with 140 MeV Ag¹¹⁺ ions at fluences of 1 × 10¹¹ and 5 × 10¹² ions/cm². UV–visible absorption analysis reveals a decrease in optical direct band gap from 2.62 to 2.42 eV after a fluence of 5 × 10¹² ions/cm². X-ray diffractograms show an increase in crystallinity of the composite due to irradiation. The dielectric constants obey the Universal law given by ε α f ⁿ⁻¹, where n varies from 0.38 to 0.91. The dielectric constant and loss are observed to change significantly due to irradiation. Cole–cole diagrams have shown the frequency dependence of the complex impedance at different fluences. The average surface roughness of the composite decreases upon irradiation.