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Nuclear Technology Applications and Innovations

2010;():1-5. doi:10.1115/ICONE18-29042.

Nowadays, boron acid is used widely in nuclear power station as the chemical controlling remedy by making use of 10 B as the effective neutron absorbent. But the natural element boron includes 10 B just 20% (mass fraction), all the rest is 11 B which is very weak at absorbing neutron. This paper brings forward a way to improve the effect of chemical controlling remedy on the nuclear power station through taking 10 B enriched boron acid as the substitution of the natural one. If this technological measurement is taken, the operation control of nuclear reaction would get at high level. Not only the amount of boron acid waste would drop suddenly, but the amount of all kinds of chemical agents added in the cooling circulation would lower down sharply. So, many complex chemistry problems in the cooling system could be solved, such as erosion of devices, worse quality of wastewater and consume of chemical agents, etc.

Commentary by Dr. Valentin Fuster
2010;():7-15. doi:10.1115/ICONE18-29047.

At present, most of the developed neutron dosimeters that have a moderator with a single counter, applied in neutron radiation fields within large range energies from thermal to MeV neutrons, are not a satisfaction to energy response. The purpose of the article is designing a suitable neutron dosimeter for the radiation protection purpose. In order to overcome the disadvantage of the energy response of the neutron dosimeters combined a single sphere with a single counter, three spheres and three 3 He counters were combined for the detector design. The response function of moderators with different thicknesses combined with SP9 3 He counters were calculated with MCNP program MCNP4C [1]. The selection of three different thicknesses of the moderating polyethylene sphere was done with a Matlab program [2]. A suitable combination of three different thicknesses was confirmed for the detector design. The electronic system of the neutron dosimeter was introduced. The fluence to ambient dose-equivalent conversion coefficient were calculated, analyzed and compared with the values recommended in the ICRP 74 Publication [3]. The calculated result explain that it is very significance to this design of neutron dosimeter, it may be applied to the monitor of the ambient dose in the neutron radiation fields, improving at present the status of the energy response of neutron dosimeters.

Topics: Design , Dosimeters
Commentary by Dr. Valentin Fuster
2010;():17-21. doi:10.1115/ICONE18-29071.

Pu material can generate surface corrosion and self-radiation effect during storage, leading to the creation and recoil of uranium and helium ions, which produce defects through displacement cascades, these self-irradiation defects tend to change plutonium properties. To study these aging behavior, calculations at the spin unrestricted generalized gradient approximation (GGA) level of density functional theory (DFT) have been performed using the DMol3 programs. Relativistic effects, such as mass-velocity, Darwin term, are considered in this code. Some conclusions are draw as follows: 1) Band structure of the (100) surface of γ-Pu is very narrow around the Fermi level, showing that the eigenstate of this level is mainly composed of local atomic orbital, the local property of electrons in this band is very strong, while the band around the Fermi level is mainly constituted by 5f narrow band; 2) DOS of the (100) surface of γ-Pu are mainly composed of the density of states in −48–41eV, −23–16eV, −3–2eV; 3) Contribution of s shell to the total DOS is mainly distributed in the first interval, and p shell is mainly in the second interval, while d and f shells are mainly in the third interval.

Commentary by Dr. Valentin Fuster
2010;():23-27. doi:10.1115/ICONE18-29072.

Neutron tube is a kind accelerator neutron source, which has been applied extensively. It is mainly composed of an ion source system, an accelerator system, and a target system. The target system is one of the most important parameters of the neutron tube, and it impacts directly the yield, the lifetime, and the stability. The pure titanium film was evaporated on the ceramic target by using the evaporation technique. The impact of the thickness of the titanium film on the yield of the neutron tube was studied, and the yield of the neutron tube was on optimistic state if the thickness of titanium film was 2.2μ m. The properties of training and stability of neutron tubes with different thick titanium film target were also present in this paper.

Commentary by Dr. Valentin Fuster
2010;():29-36. doi:10.1115/ICONE18-29080.

It has been experienced that service life of reinforced concrete structures is often limited due to lack of durability of cement-based materials. One major reason for this durability problem is the penetration of water and compounds dissolved in water into concrete. Therefore, there is an urgent need to study water penetration into concrete in order to better understand deterioration mechanisms. Neutron radiography provides an advanced non-destructive technique with high spatial resolution. In this contribution, neutron radiography was successfully utilized to study the process of water absorption of two types of concrete with different water-cement ratios namely 0.4 and 0.6. It is shown that it is possible to visualize migration of water into concrete and to quantify the time-dependent moisture distribution with accurately and with high spatial resolution by means of neutron radiography. In concrete with high water-cement ratio, water penetrates much quicker than in concrete with lower water cement ratio. Water penetration depth obtained from neutron radiography is in good agreement with corresponding values obtained from capillary suction tests. Experimental results obtained by means of neutron radiography on water penetration into concrete will be presented and discussed in this contribution. Results will provide us with a solid basis for a better understanding of deteriorating processes in concrete and other cement-based materials. These results may be considered to be a first step to improve durability of concrete.

Commentary by Dr. Valentin Fuster
2010;():37-42. doi:10.1115/ICONE18-29094.

The full energy peaks always overlap each other when carrying out radioactive spectrum measurement. In this paper, combining GMM statistical clustering model with EM iterative algorithm, the overlapping peaks of nuclear spectrum with no background has been decomposed. No matter how many overlapping peaks we can use this decomposition method, and ensure that the statistical errors close to zero. It has great practical value for radionuclide with a quantitative or qualitative analysis.

Commentary by Dr. Valentin Fuster
2010;():43-48. doi:10.1115/ICONE18-29100.

In the present study, a type of power generators of cantilever beam structure has been studied. A dynamic model describing the power generator, which takes into consideration the vibration of the mechanical structure, the power harvesting circuit, and the coupling between them, is proposed. A set of partial differential equations describing the dynamic model are obtained. Stability analysis is then carried out to analyze elementary properties of the generator. By using modal decomposition and state space representation, the set of partial differential equations could be easily solved no matter what the external excitation is, harmonic or non-harmonic, transient or continuous, at or off resonant frequency. The characteristics of the generators are analyzed and some recommendations on design of efficient power generators are provided.

Topics: Modeling , Generators
Commentary by Dr. Valentin Fuster
2010;():49-54. doi:10.1115/ICONE18-29126.

Gamma-ray spectrum analysis was essential for detecting the elemental abundance and distribution in lunar science. However, for the low-energy region of gamma-ray spectrum, weak peaks were implicated in the fast-decreasing background, and it was difficult to extract characteristic information from original spectra. In order to get a better analytic result, based on wavelet and FFT filtering methods in frequency domain, we had processed the gamma-ray spectrometer (GRS) data of Chang’E-1 (CE-1), and well extracted some useful information of spectral characteristic peaks. Then we preliminarily mapped the distribution of net peak counts for potassium on lunar surface, which indirectly reflected the distribution of elemental abundance. At last, we compared our analytic result with that of Apollo and Lunar Prospector (LP), and found some consistencies and differences.

Commentary by Dr. Valentin Fuster
2010;():55-59. doi:10.1115/ICONE18-29150.

There are significant relations between temperature and the electrical performances of betavoltaic cell. Two silicon diodes as the energy conversion device of betavoltaic cell were irradiated by Ni-63, checked the relations between temperature and the electrical performances such as Voc , Isc , Pmax etc. Isc increased very little as the temperature increasing but Voc decreased significantly. The changing values of Voc were −3.1 mV/k and −3.0 mV/k respectively within the temperature range of 233.15 K∼333.15 K. As a result of this, the Pmax , η etc. decreased markedly too.

Topics: Temperature
Commentary by Dr. Valentin Fuster
2010;():61-66. doi:10.1115/ICONE18-29166.

The problems of incorporating radioactive material into PIN-diode and those of encapsulating the nuclear battery were waiting for being solved in the phases of developing of beta voltaic effect nuclear cell. Aiming at the need of loading sheet radioisotope source Ni-63 at present, this paper designed its loading method on PIN-junction and encapsulation manner of nuclear batteries with energy transforming structures based on the photoelectric detector. Some parameters relating with encapsulation of prototype cell were advanced.

Commentary by Dr. Valentin Fuster
2010;():67-76. doi:10.1115/ICONE18-29176.

The most abundantly available fossil fuel on Earth is coal. For countries like China, the USA, South Africa, or Germany, coal plays a dominant role as energy resource. The introduction of nuclear energy into coal refinement processes would be a significant contribution to the saving of resources, lowering specific carbon emissions and reducing dependencies on oil and natural gas imports. In Germany, comprehensive R&D activities were conducted within the project “Prototype Plant Nuclear Process Heat” (PNP) to investigate the utilization of nuclear energy from a pebble-bed HTGR in both steam-coal gasification and hydro-gasification. A major component to be newly developed was the gas generator. Its operation on semitechnical scale confirmed the feasibility of allothermal, continuous coal gasification under nuclear conditions. A key problem remained the selection of appropriate high temperature materials for gas generator and other high temperature heat exchanging components. The project was accompanied by comprehensive safety studies targeting tritium contamination and consequences of potential explosions of flammable gas mixtures. Future activities could take benefit from a reevaluation of the studies conducted in the past by comparing HTGR process heat applications against current technologies. Fossil fuel market conditions and environmental effects shall be considered. Superior safety features and high reliability are prerequisites for the introduction of nuclear process heat and nuclear combined heat and power.

Commentary by Dr. Valentin Fuster
2010;():77-81. doi:10.1115/ICONE18-29184.

Bonner Spheres neutron spectrometer has been widely applied as neutron dosimeter, however the derivation of neutron energy spectrum from its measurement data is still a significantly difficult task. This unfolding problem is proved to be ill-posed, under-determined and have no exact solution. Two major require of the unfolding methods are accuracy and stability. Most unfolding methods try to search the solution that best fit the measurement data and the response function. As a universal optimization tool Genetic Algorithm shows its potential to solve this kind of problem. Through gene operation of every generation, GA could find the global optimal among the searching space. A new fitness function which contains a distance part and a penalty part was constructed in this research. The distance part is the square distance between the individual and the measurement data. The penalty part which is a function associated with the continuity of individual is used to avoid intensively change of unfolded data. Five classical neutron spectra were chosen as benchmark input spectra. The product of the benchmark spectra and the response function played as input measurement data of the unfolding program. The unfolded results showed good agreement with the real ones. The measurement data could be well reproduced by the unfolded results though the results had some difference with the real spectra.

Commentary by Dr. Valentin Fuster
2010;():83-88. doi:10.1115/ICONE18-29191.

How to get the position, direction and width of buried fault effectively is still very difficult when finding buried faults by measuring 222 Rn radioactivity. In this paper we established a technique to carry out buried fault investigation. It was based on Fick’s fist law, Darcy’s law and theory of clusters to analyze radon transportation and simulate 222 Rn transportation in ideal conditions. The feasibility indicates that measuring or investigating the concentration of radon to find abnormal region can help people find buried faults. 218 Po and 214 Po, daughter products of 222 Rn, are generally considered to be proportional to initial concentration of 222 Rn. 218 Po and 214 Po have short half-life of 3.05 min and 164us respectively which is very suitable for actual measure work. So in order to accumulate alpha particles effectively, soil gas sampling period is set about twice half-life of 218 Po. The established model is applied to analyze two buried fault areas in Southwest China and the results are really much better.

Commentary by Dr. Valentin Fuster
2010;():89-92. doi:10.1115/ICONE18-29223.

Boron loaded plastic scintillator could detect both fast neutrons (thanks to hydrogen) and slow neutrons (thanks to 10B). The large cross sections of both reactions lead to high detection efficiency of incident neutrons. However, gamma rays must be rejected first as the scintillator is also sensitive to them. In the present research zero crossing method was used to test neutron-gamma discrimination performance of BC454 boron loaded plastic scintillator. Three contrast experiments were carried out and different thermalization degrees lead to different time spectra in the MCA. Further analysis proved that three Gaussian curves could be used to fit the spectra; they corresponded to gamma rays, fast neutrons and slow neutrons respectively. The slow neutron curve could be clearly separated from the gamma curve. Discrimination performance for fast neutrons became poor, but their peaks could also be separated.

Topics: Neutrons
Commentary by Dr. Valentin Fuster
2010;():93-98. doi:10.1115/ICONE18-29267.

The Japan Atomic Energy Agency has been conducting R&D on thermochemical water-splitting Iodine-Sulfur (IS) process for hydrogen production to meet massive demand in the future hydrogen economy. A concept of sulfuric acid decomposer was developed featuring a heat exchanger block made of SiC. Recent activity has focused on the reliability assessment of SiC block. Although knowing the strength of SiC block is important for the reliability assessment, it is difficult to evaluate a large-scale ceramics structure without destructive test. In this study, a novel approach for strength estimation of SiC structure was proposed. Since accurate strength estimation of individual ceramics structure is difficult, a prediction method of minimum strength in the structure of the same design was proposed based on effective volume theory and optimized Weibull modulus. Optimum value of the Weibull modulus was determined for estimating the lowest strength. The strength estimation line was developed by using the determined modulus. The validity of the line was verified by destructive test of SiC block model, which is small-scale model of the SiC block. The fracture strength of small-scale model satisfied the predicted strength.

Commentary by Dr. Valentin Fuster
2010;():99-104. doi:10.1115/ICONE18-29272.

The recent advances in utilization of nuclear heating reactor technology are to couple with seawater desalination process, in which process steam and fresh water is combined to produce, steam turbine is used, thermal and membrane processes are joined. In this technical concept, the “zero discharge” to environment will be realized due to using nuclear energy as the heat source for process steam and seawater desalination. Combined process steam and fresh water produce will make the product’s cost lower and make the whole system economic results better. The special design features of nuclear heating reactor will ensure that any radiological contamination can be excluded from fresh water. As joining thermal process and membrane process seawater desalination, the step utilization of thermal energy and reuse of waste heat can be realized.

Topics: Heating
Commentary by Dr. Valentin Fuster
2010;():105-109. doi:10.1115/ICONE18-29302.

The mechanisms of charge coupled devices (CCD) irradiated by protons are analyzed. The simulation models of ionization damage and displacement damage are developed. The charge transfer efficiency (CTE) decreased by proton irradiation is numerically simulated. The CTE degradation caused by different traps and by protons with different energies has been studied respectively. Both surface dark signals induced by proton ionization damage and bulk dark signals induced by proton displacement damage are numerically simulated. The variability of surface dark signals, bulk dark signals, and total dark signals with proton fluence is compared. The simulation results are in agreement with the experimental results of the relevant literatures.

Commentary by Dr. Valentin Fuster
2010;():111-116. doi:10.1115/ICONE18-29314.

The compact pulsed neutron source facility can play an important role in the research, education, user training, and development of the advanced neutron scattering instruments. The materials for the target, moderator, reflector (TMR) and their configurations must be optimized to get the optimal yield of neutrons with energy in the range of 1 meV to eV order which depends on the proton energy and its nuclear reaction. Several kinds of materials of the TMR, their configurations, and their dimensions are investigated by the Monte Carlo simulation and optimized for developing the compact pulsed neutron source. The results would contribute to the construction of the Compact Pulsed Hadron Source (CPHS) of Tsinghua University.

Topics: Neutron sources
Commentary by Dr. Valentin Fuster
2010;():117-122. doi:10.1115/ICONE18-29325.

The production of clean water in the US as well as other countries is a critical need along with non-greenhouse gas electrical power generation. Low-temperature waste heat from nuclear power plants can be used to produce the large quantities of clean water for reactor cooling (∼25,000 acre-ft/yr), potable water for culinary and agricultural use and many other applications. Cogeneration of nuclear electrical power and clean water is reviewed and discussed in this paper. These cogeneration systems can utilize grey and/or brackish water that can markedly extend potential sites for future nuclear plants in areas where only poor water sources are available. A steam adsorption system for on-line production of clean water and refrigeration using nuclear power plant waste heat is also proposed and discussed. This improved design for more energy-efficient use of the steam adsorption cooling has the potential to substantially reduce the intense electrical power consumption for food processing and storage, ice- and snow-making and air-conditioning.

Commentary by Dr. Valentin Fuster
2010;():123-128. doi:10.1115/ICONE18-29334.

A computer code named Fitting for Diffusion Parameters (FDP) based on Mathematica 6.0 has been developed for modeling through- and out-diffusion experiments. FDP was used to determine the diffusion coefficients (De ) and the rock capacity factors (α) for tritiated water (HTO) and 22 Na+ and the distribution coefficient (Kd ) of 22 Na+ in Opalinus Clay (OPA). The values for De and α were obtained by fitting the results of experimental data of both transient and steady-state phases to the analytical solution of accumulated activity. The quality of the parameters De and α was tested by using them as input parameters in the equation of flux. Moreover, the diffusion parameters of HTO and 22 Na+ were determined also by out-diffusion experiments. Under ambient condition at pH 7.6, the De value of (1.5 ± 0.1) × 10−11 m2 /s for HTO is lower than that of (1.9 ± 1.1) × 10−11 m2 /s for 22 Na+ , which could be explained by the electrostatic attraction between the negative surface charge of OPA and the sodium cations. For the non-sorbing species HTO, α was 0.15 ± 0.01. For the weakly sorbing species 22 Na+ , α was 0.50 ± 0.02 and Kd equaled (1.5 ± 0.3) × 10−4 m3 /kg. The obtained diffusion parameters for HTO and 22 Na+ in OPA are in good agreement with previous results by Van Loon et al. [1, 2]. FDP developed in this study has been used successfully to determine the parameters De and α for the diffusion of 237 Np(V) in OPA [3].

Commentary by Dr. Valentin Fuster
2010;():129-133. doi:10.1115/ICONE18-29359.

Japan Atomic Energy Agency has been conducting research and development on a thermochemical water-splitting cycle featuring iodine- and sulfur-compounds (called an IS process) as one of promising heat utilization systems of High Temperature Gas-Cooled Reactors. We have prepared polymer electrolyte membranes by the radiation-induced graft polymerization and cross-linking methods and then have investigated their applicability to electro-electrodialysis (EED) for concentrating HI in an HI-I2 -H2 O mixture. For practical applications, EED membranes are required to be stable in the severe environment of high-temperature strongly acidic solutions. We thus examined thermal, chemical and electrochemical stabilities of the radiation-grafted membranes under the conditions of the actual EED operation over 100 hours, while measuring the time evolution of a cell voltage and a change in the ion exchange capacity between the EED experiment. The results showed that chemical cross-linking in the graft chains could largely improve the membrane stability.

Commentary by Dr. Valentin Fuster
2010;():135-142. doi:10.1115/ICONE18-29456.

The mathematic models of reactor, once-through steam generator and turbine are built based on the mass, energy and momentum conservation theorem. Because of serious coupling and different dynamic characteristic, the coordinated control that solves big system problem is presented to apply into the nuclear power plant after researching deeply the variety feature and coupling relation of primary parameters of the nuclear power plant. The coordinated control system is filled with manage control, coordinated control and bottom controller. The simulation is processed by changing turbine load. Compared with non-coordinated control system, the coordinated control system improves briefly the dynamic feature of nuclear power plant. The fuzzy decoupled control strategy between once-through steam generator and turbine is proposed. The fuzzy decoupled frame including a compensator and design method of the decoupled compensator are given, the fuzzy rules are applied in the decoupled compensator. Finally, a fuzzy decoupled control system is designed in detail with a two inputs and two outputs’ system, which is applied in the coordinated control system of the nuclear power plant. The simulation results show that the coordinated control system based on the decoupled strategies is better than the coordinated control system, which weakens the couple connection, reduces the fluctuation of exit steam pressure by adjusting the feedwater flux.

Commentary by Dr. Valentin Fuster
2010;():143-148. doi:10.1115/ICONE18-29459.

The 1.2MV, 70ns FWHM induction cell is developed for a 3MV Induction Voltage Adder (IVA) accelerator with three series connected cavities through a high voltage, vacuum insulating transmission line (VITL) driving rod-pinch diode (RPD) for radiography. The experimentally measured maximum relative permeability of IVA used amorphous material under pulse excitation is consistent with saturation wave model with flux density changing rate dB/dt greater than 10T/μs up to 32 T/μs. The remanence ratio Br /Bs is experimentally measured for IVA pre-annealed amorphous cores. Upon experimental results, cores in each cell are designed and determined according to current transfer efficiency and volt-second integral necessary for each induction cavity, and peak field preventing VITL bore negative surface from electron emission. The field analysis is carried out for designing VITL vacuum stack and oil cavity that contains cores and azimuthal transmission line with operation field stress being about 50% critical breakdown field. The prototype cell tests validate electric field safety and magnetic core performance. With magnetization inductance and resistance of eddy current calculated for inductive cell, the IVA accelerator circuit model is set up, and simulation predictions approximately according with experimental results are presented.

Commentary by Dr. Valentin Fuster
2010;():149-156. doi:10.1115/ICONE18-29477.

A new moderator cell structure based on the single-phase thermo-siphon loop is proposed for the Cold Neutron Source (CNS) of China Advanced Research Reactor (CARR). Different from previous single-phase moderator cells, the design scheme of a crescent-shape hydrogen layer together with a helium cooling jacket is adopted in this paper. Theoretical analyses on cold neutron gain factor, heat transfer performance and structural strength of the moderator cell are carried out aiming at optimizing the preliminary design scheme. Results show that the modified moderator cell has high cold neutron gain factor, good heat transfer capacity and low stress, which could meet the function requirements of the CNS of CARR.

Topics: Design
Commentary by Dr. Valentin Fuster
2010;():157-162. doi:10.1115/ICONE18-29479.

Neutron energy spectrum and fluence of the devices in a reactor are important parameters for the users. In this paper, the thermal neutron spectrum of the thermal column in Xi’an pulse reactor were measurement by the method of time of flight (TOF), and the peak of the energy spectrum is 0.0248±0.0006eV and the neutrons average energy is 0.048±0.001eV. With the spectrum results, the average cross section of 235 U(n,f) and 197 Au(n,γ) were calculated are respectively 92.08barn and 538.86barn. With the average cross sections, we measured the neutron fluence with the fission chamber and 197 Au(n, γ) activation method. The measurements of the two methods are respectively 1.08×109 n/m2 s and 1.13×109 n/m2 s. We also calculated the standard uncertainty of the measurements: the energy spectrums’ is less than 5%, and the neutron fluence’ is less than 2.2%.

Commentary by Dr. Valentin Fuster
2010;():163-169. doi:10.1115/ICONE18-29486.

Template measurement is an important method in deep nuclear disarmament. The gamma-ray spectrum of Plutonium pit shows unique property due to age, abundance, amounts and thickness of the Plutonium pit; that is, same designed pits yield similar gamma-ray spectra while different design give distinct spectra. Useful information is extracted from gamma-ray spectrum generated by the reliable Plutonium pit radiation as ‘template’. Comparison of the data from inspected objects with the template can give conclusion whether they are of the same type. This paper studies how to choose template data from gamma-ray spectrum and discusses the limits of the gamma-ray measurement. Because of the strong self-absorption of Plutonium, some characteristics of Plutonium pit can’t be identified only by gamma spectrum. MCNP simulation was employed to prove that in some cases, template depending on gamma-ray spectrum from the reliable Plutonium pit alone can’t effectively distinguish the spurious objects. And a further approach indicates that enhancing neutron counting rate of spontaneous fission of Plutonium can improve the problem. Neutron counting rate can be indirectly acquired by spontaneous fissile neutrons bombarding a 10B target. 478 keV γ rays are concomitant with the nuclear reaction 10B(n,α)7Li* from 7Li* nuclei’s deexcitation. Neutron information is gathered by detecting 478 keV γ photons. Using HPGe γ detector can both detect γ-ray spectrum and acquire neutron counting rate. This method efficiently increases confidence of template measurement and also ensures the dismantling process without revealing sensitive nuclear warhead design information.

Topics: Gamma rays
Commentary by Dr. Valentin Fuster
2010;():171-173. doi:10.1115/ICONE18-29492.

The Compact Pulsed Hadron Source (CPHS) of Tsinghua University will produce neutrons by the Be(n,p) reaction through bombarding a proton beam with 13MeV/50Hz/1.25mA from a LINAC system on a beryllium target. One of the purposes of this neutron source facility is to provide the neutron scattering capability for characterization of materials, especially soft matters and biological systems. Cold neutrons (wavelength > 4 Å) are essential to characterize the structure of these materials over the length scale of ∼100 nm with good resolution. We discuss the design and optimization of a cold neutron source (CNS) which employs a solid methane moderator for cold neutron generation. The moderator configuration, the associated cryogenic system, and operation conditions will be discussed.

Topics: Neutrons , Design
Commentary by Dr. Valentin Fuster
2010;():175-178. doi:10.1115/ICONE18-29588.

The research of radon detection based on air mesh and pulse chamber at normal is presented in details. The Instrument consists of air mesh, pulse chamber, signal amplification and discriminator and MCU. Radon of ambient air can directly make a free diffusion to the high-pressure chamber, because The Instrument uses a new type of atmospheric air-gridded pulse ionization chamber, ventilated with ambient air. A voltage pulse in the Resistance connected to Collecting electrode is produced when the charged particles moving through the gas in high-pressure chamber, can make the air ionize to create the charged ion-pairs (negatively charged ion and positively charged ion), which drift to the cathode and anode respectively with the effect of the directional electric field. The voltage pulse height represents the ionizability of an incident particle. And the relative counting can be made for the incident particles, reflecting the changes of the radon concentrations in air, and the total measurement too. But the pulse signal, from Front-end detector, is very weak, so they need to be amplified and shaped by amplification and discriminator circuit and RC-CR band-pass filter, compared with a given threshold voltage through the Pulse discriminator to produce the digital signals which micro-controller system can handle and for which SCM can make the counting, storage and display. Then, this Instrument is a design of appropriate devices for short-term measurements, and some criteria used in the design of this instrument are field measurements applicability, portability, convenience and reliability.

Commentary by Dr. Valentin Fuster
2010;():181-185. doi:10.1115/ICONE18-29606.

The principle of radiographic X-ray spot size measurement with Bread piece was presented. The output responses of the Bread piece for uniformly-distributed X-ray sources are calculated numerically. The relationship between FWHM of the response curve and the spot size is obtained. Preliminary experiments were performed on the inductive voltage adder (IVA), and the Bread piece is used to measure the axial spot size of the rod-pinch diode (RPD) radiographic source. Under the hypothesis of uniformly-distributed line source, the axial spot size value is obtained, consistent with the measuring result of the pinhole camera, verifying the applicability of the Bread piece for radiographic X-ray spot size measurement.

Commentary by Dr. Valentin Fuster
2010;():187-190. doi:10.1115/ICONE18-29745.

Neutron radiography uses the unique interaction probabilities of neutrons to create images of materials. This imaging technique is non-destructive. MCNP Monte Carlo Code has been used to design an optimized neutron radiography system that utilizes 241 Am-Be neutron source. Many different arrangements have been simulated to obtain a neutron flux with higher amplitude and more uniform distribution in the collimator outlet, next to image plane. In the final arrangement the specifications of neutron filter, Gamma-ray shield and beam collimator has been determined. Simulations has been Carried out for a 5Ci 241 Am-Be neutron source. In this case 43.8 n/cm2 s thermal neutron flux has been achieved at a distance of 35cm from neutron source.

Commentary by Dr. Valentin Fuster
2010;():191-203. doi:10.1115/ICONE18-29801.

Complex nuclear plants have high costs. Documenting and maintaining information contributes a large share of those costs. Improving methods to design, construct and operate nuclear plants lowers costs. Improving design processes could help manage nuclear project costs, birth-to-death. Improving methods that manage design basis information before, during and after construction would not only lower initial nuclear costs, but also provide more complete, accurate information for operations over plant life. Nuclear safety would benefit. The Department of Energy’s (DOE) Next Generation Nuclear Plant (NGNP) offers a unique opportunity to advance nuclear design methods to improve cost performance. Better design methods could automatically track changes to critical design content reducing laborious effort to update designs, speeding development. They would enhance information sharing. The commercial NGNP project is a significant opportunity to improve commercial nuclear plant safety design. The U.S. DOE should promote innovative nuclear design processes with the NGNP. Providing a single, consistent design framework on a modern data platform would make the NGNP the nuclear industry’s “Go to the Moon” project.

Commentary by Dr. Valentin Fuster
2010;():205-209. doi:10.1115/ICONE18-29936.

Tomographic Gamma Scanning (TGS) method is one of the most advanced non-destructive assay (NDA) methods. But for measuring heterogeneously distributed media with medium- and high-density, there are still three main problems: experiment’s method of the calibration of detection efficiency of TGS is more difficult and complicated because of large voxels, “point-to-point” model and average model can’t calculate high-density samples accurately in transmission image reconstruction and computational cost is very large for correction factor in emission image. Calibration of detection efficiency using Monte Carlo method shorten calibration cycle greatly, a new Monte Carlo statistical iteration in TGS transmission image reconstruction method which is based on MC calculation and Numerical Analysis is presented, give a chance for measuring high-density samples; the division method and pre-calculation method in reconstructing TGS emission image is used which saves a great lot of computation time and provide a fast reconstruction algorithm for emission image. Above methods apply to TGS experiment device, the relative errors between experiment and MC calibration were less 5%; the relative errors between reconstructed values and the reference values were less than 4% in transmission image; the corrected experimental results were compared to the standard values and the relative deviation was found to be 7%. It took no more than one hour to complete the reconstruction of TGS emission image for a sample model with 3×3×3 voxels using a 2.0G computer.

Commentary by Dr. Valentin Fuster
2010;():211-217. doi:10.1115/ICONE18-29976.

129 I is a long-lived (15.7M year) radioisotope of iodine. It can be used as a tracer for monitoring nuclear proliferation and the 129 I/127 I ratio can be used to evaluate the radiation contamination level. Nowadays a great number of nuclear power plants will be built in China, but the data of 129 I concentration in environmental samples around nuclear power plants are limited. Accelerator Mass Spectrometer (AMS) whose detection limit is about 10−14 , is one of the best instruments for analyzing environmental samples. The methods of making 129 I target samples for AMS measurement from different type samples were studied, and the processing system for water and soil samples were established. Six surface seawater samples were collected at different distance away from a nuclear power plant in China. These samples were measured by Xi’an AMS. The ratios of 129 I/127 I in the seawater samples are between 0.829 × 10−10 and 9.451 × 10−10 , and the average value is about 3.518 × 10−10 . The ratios of 129 I/127 I in these samples are compared with other measurement results under different circumstances in other parts of the world. The results show that this nuclear power plant has not released superfluous 129 I into environment after several years’ operation. Since the AMS and sample processing system are established, we will do much work on nuclear technology application with 129 I tracer.

Commentary by Dr. Valentin Fuster
2010;():219-222. doi:10.1115/ICONE18-29977.

A CsI(T1) scintillation detector with a two-dimension position-sensitive photomultiplier readout at RIBLL is developed. The position resolution of the detector was obtained with RIB 17 N, and the performance of detector was simulated by using GEANT4. The simulation results accorded with the experimental ones. The reflection factor of the aluminum foil outside the scintillator plays an important role on the position resolution.

Topics: Sensors
Commentary by Dr. Valentin Fuster
2010;():223-228. doi:10.1115/ICONE18-30018.

This paper reports a novel radioisotope microbattery structure that integrates a betavoltaic converter with a work function converter. The battery collects energy from radioisotope and environment vibration. A model is developed to simulate the mechanism of the proposed battery, and select parameters to improve its efficiency. Using the proposed model, the battery is designed with structures optimized for the environment vibration frequency in the range of 100–400Hz and 63 Ni of 11mCi. The theoretical output power is on the order of 200nW. The output power collected from the radioisotope is close to that from environment vibration. Since the vibration beam frequency of the work function converter is much larger than the environment frequency, the output power of the battery keeps stable when the environment frequency changes significantly.

Commentary by Dr. Valentin Fuster
2010;():229-233. doi:10.1115/ICONE18-30027.

In this article, the defect of measurement method using pulse-counting for total measurement is discussed. Time-to-count measurement principle is introduced and Monte Carlo method was successfully used to simulation. The author designs a personal dosimeter with that principle, which eliminates the influence of dead time and other factors and improves the performance of the entire instrument.

Topics: Dosimeters
Commentary by Dr. Valentin Fuster
2010;():235-241. doi:10.1115/ICONE18-30117.

On January 2008, the US NRC issued the Generic Letter 2008-01 [1], “Managing gas accumulation in emergency core cooling, decay heat removal and containment spray systems”. Among other responses, this letter requires an evaluation of locations sensitive to accumulate gases in several safety systems. In order to get accurate data related to the real slope of horizontal pipes and other geometrical parameters needed for this evaluation, laser scanning and 3D modeling techniques have been applied in Spanish Nuclear Power Plants. From October 2008 to December 2009, five Spanish units have been scanned and modeled. As a result of these activities, the plants have obtained detailed 3D models as well as 2D as-built drawings of the selected components. These models were integrated in 3D web servers which give a panoramic view of the scanned areas and permitted measurements in the local coordinate system of the plant. Moreover, the 2D elevation drawings included accurate and useful information for the plants in order to make decisions related to the GL-2008-01 requirements. The geometric information generated in the frame of the GL-2008-01 activities is being currently used for alternative applications. For instance, laser scanning technology is being used to enhance design modification procedures. A pilot project on the MSRs replacement is being currently carried out with successful results. This technology has the advantage that new components from CAD software can be updated in the as-built models obtained through laser scanning. In addition to this, it’s very easy to check fitting and interferences, and also to make accurate measurements and handling simulations. The potential applications in personnel training and radiological protection are also very important. The panoramic viewers on 3D web servers are versatile and could fit the specific requirements of each organization. Regarding staff training, virtual tours and component seekers are being currently developed. These tools provide a significant save of time and dose and also give independence for each person to get to the working place without external help or time-consuming paper consulting. Integration with existing plant databases is also possible through the panoramic viewers and is currently being developed for In-Service Inspections and Maintenance applications. The main advantage of these products is their accessibility with free visors which don’t need specific training. Therefore, the implementation of these tools doesn’t need additional investments. In conclusion, Laser 3D Technology Applications set the first step on the democratization of these powerful 3D environments among common users as integrated tools in their daily work.

Commentary by Dr. Valentin Fuster
2010;():243-247. doi:10.1115/ICONE18-30138.

We present here an ellipsoidal timing detector in Radioactive Ion Beam Line in Lanzhou (RIBLL). The photons induced by radioactive beam ions passing through a thin plastic-scintillator foil BC422, emit from the foil center corresponding to one focal point of an aluminum ellipsoidal mirror and are reflected to another focus point at which the cathode of a photomultiplier tube locates. A time resolution of about 115ps is achieved for 12 N and the counting rate up to 108 pps is allowed. The simulation was carried out using GEANT4 Monte Carlo toolkit. The photons total collection efficiency following projectile from different position, photon collection efficiency and time resolution of photon to photocathode of 3 different cases were calculated. Also the main factors influencing the detector’s time resolution and some proposals are given.

Topics: Sensors
Commentary by Dr. Valentin Fuster
2010;():249-258. doi:10.1115/ICONE18-30201.

Thermochemical hydrogen cogeneration using heat of molten salt nuclear reactors (MSRs) is discussed in this paper. Sulfur-iodine and copper-chlorine cycles are taken as typical examples for analysis and discussion. It is found that the heat exchanger design is predominately determined by the maximum and range of temperatures of themochemical hydrogen production cycles with MSRs. Copper-chlorine (Cu-Cl) thermochemical cycles can link with most MSRs, but sulfur-iodine (S-I) cycles can only link with very high temperature MSRs. The location of extracted heat from MSRs to S-I and Cu-Cl cycles is investigated, and its influence on the layout of nuclear reactor coolant loop is discussed. Some conceptual designs of heat exchangers are proposed to transfer heat from MSRs to Cu-Cl and S-I cycles. The available heat quantity at different hours of a day and corresponding hydrogen production scales are determined. It is found that the available heat at most hours of power demand in a day is equivalent to the hydrogen cogeneration capacity of an industrial scale steam methane reforming plant, if an MSR power station is operating at an invariable maximum power, independent of an electrical load throughout a day or year.

Commentary by Dr. Valentin Fuster
2010;():259-264. doi:10.1115/ICONE18-30275.

Neutron diffraction and Scanning Electron Microscope (SEM) were employed to study complex two-phase coexistence structure and surface morphology of CaO-Al2O3-SiO2(ZnO-BaO-Na2O) glass ceramics prepared under different cooling conditions. With the rapid cooling temperature decreasing from above 850°C to 750°C and 300°C, the length of the needle-like precipitated β-wollastonite crystal decreased from 30 μm to 15 μm and 5 μm, respectively. Meanwhile, the transgranular fracture appeared and the grain boundary became indistinct in the sample rapidly cooled to 750°C, and the microcracks appeared in the samples rapidly cooled to 300°C and below. These phenomena contribute to the decrease of bending strength for the rapid cooling. Neutron diffraction revealed that the unit cell of precipitated β-wollastonite crystal elongated along its three axes and its volume increased at different cooling conditions. With the decrease of the cooling temperature, the elongation of axes and increase of volume were enhanced, implying that the tensile stress of the β-wollastonite crystal increased. At the same time, intensity of the crystal diffraction peaks increased and atomic temperature factors decreased, which revealed that defects inside a smaller size of crystal granular were less than that in larger one. Amorphous peaks at low diffraction angle did not changed with cooling temperature, showing that the middle-range-order inside residual glass phases were almost the same for all cooling conditions, while intensity of amorphous peaks at high diffraction angle increased notably for samples rapidly cooled to below 850°C, showing that rapid cooling may result in severe short-range-order in residual glass phase, which induced tensile stress of crystalline phase from around amorphous phase and therefore lead to occurrence of transgranular fracture and microcracks. This study suggests that rapid cooling to below 850°C should be avoided in order to obtain preferable mechanical properties for CaO-Al2O3-SiO2(ZnO-BaO-Na2O) glass ceramics.

Commentary by Dr. Valentin Fuster
2010;():265-271. doi:10.1115/ICONE18-30288.

Neutron tube is a widely used small-scale accelerator-type neutron source, it is mainly made up of the storage, ion source, acceleration system, target and other parts. One of the important parameters of neutron tube is ion source current and its stability, which has a direct impact on the amount of neutron tube production and stability. Ion source current depends on storage heating current and it is hysteresis controlled object. This paper presents a kind of nonlinear PID controller and the control method is applied to ion source current control by using B&R PCC controller, the stability of current is not more than 2%, which meet the requirements of high-quality neutron tube.

Commentary by Dr. Valentin Fuster
2010;():273-277. doi:10.1115/ICONE18-30294.

In the Specimen Reconstitution Technology, the empirical fitting equation of Boltzman function can be adopted for fitting the relationship between Charpy impact absorbed energy & percent ductile fraction and temperature, and the judgment index of goodness-of-fit is adjusted coefficient of determination (Adj.R). The t/t′ hypothesis testing can be adopted to check the identity of ductile-brittle transition temperature before & after reconstitution, and in advance, F testing should be carried out on the homogeneity of two groups of samples’ variances.

Commentary by Dr. Valentin Fuster
2010;():279-283. doi:10.1115/ICONE18-30298.

The 89 Zr radioisotope is used in the field of tumor diagnostics, tumor therapy and the investigation of the biokinetic. The present work is investigated a suitable reaction to produce 89 Zr..The Zirconium-89 excitation function via 89 Y(p,n)89 Zr, 89 Y(d,2n)89 Zr, nat Zr(p,pxn)89 Zr, nat Sr(α,xn)89 Zr and 90 Zr(n,2n)89 Zr reactions were calculated by ALICE-91 and TALYS-1.0 codes and the reaction of 89 Y(p,n)89 Zr has been selected. The calculated excitation function of 89 Y(p,n)89 Zr reaction was compared with the reported measurement and evaluations. Requisite thickness of targets was obtained by SRIM code for all above reactions except the 90 Zr(n,2n)89 Zr reaction. The 89 Zr production yield was evaluated with attention to excitation function and stopping power for all above reactions except 90 Zr(n,2n)89 Zr reaction.

Commentary by Dr. Valentin Fuster
2010;():285-289. doi:10.1115/ICONE18-30328.

Excitation functions were calculated by the ALICE/91 and TALYS-1.0 codes for nat Rb(a,xn)88 Y, nat Zr(p,pxn)88 Y, nat Sr(a,xn)88 Y, 89 Y(p,n)88 Y and 88 Sr(p,n)88 Y reactions. The calculated cross sections were compared with the experimental data. The suitable energy ranges for the production of 88 Y for each reaction is reported. From the excitation functions, integral yields of the products were calculated. Finally the suitable reaction was selected for the production of 88 Y.

Commentary by Dr. Valentin Fuster
2010;():291-300. doi:10.1115/ICONE18-30349.

Weld repair sites in nuclear systems are very often located in high radiation areas in difficult to reach sites. Additionally it is usually a pre-requisite to shut down the reactor while the repair is performed, ideally during a scheduled maintenance outage. To minimize the risk of an extended outage it is of paramount importance to ensure the repair time is optimized and the technique is reliable. Technical solutions to complex remote weld repairs are rarely available “off the shelf”. While some commercially available products can be adapted to suit an application, the constraints of space, distance and radiation usually require novel and unique designs involving a variety of engineering skills. Of equal importance to the weld repair is the requirement to perform pre and post repair nondestructive examinations (NDE). The application of ultrasonic, eddy current and liquid penetrate examinations requires a similar level of engineering. This paper outlines some of the welding and NDE techniques developed to address unique repairs to nuclear systems.

Topics: Maintenance
Commentary by Dr. Valentin Fuster
2010;():301-309. doi:10.1115/ICONE18-30365.

A numerical simulation of logging while drilling density (LWD) is investigated by using Monte-Carlo (MC) method. The response of azimuth imaging is performed with in a highly deviated well with 60 degree slope. And then, the influence of counting rates of detectors was discussed due to characteristics of the inhomogeneous stratum. The numerical result shows that the azimuthal density image can intuitively display the characteristics of metrical stratum. Moreover, it can really display the thickness and location of the reservoir when drilling. The density image visually realized the visualization of the results and achieved application of geo-steering while drilling in theory. A further numerical study was carried out by using a Spine-and-Rib plot and the Far Detector Count Rate Compensation to analyze the influence of the standoff distances. And the numerical simulation results were analyzed with different Be-windows design parameters. The numerical result shows that the compensated density can satisfy the precision requirement with the two methods when the standoff distance is less than a predetermined value. Otherwise, just the far detector count rate can be compensated the stratum density with the cone-shape Be-windows.

Commentary by Dr. Valentin Fuster

Safety and Security

2010;():311-324. doi:10.1115/ICONE18-29002.

Flow-induced acoustic resonance in a piping system containing closed tandem side-branches was investigated experimentally in this study. Velocity perturbation was induced at the mouth of the cavity using two pumps and a block. An uncommon acoustic mode change, from a higher mode to a lower mode, was observed when the flow rate in the main pipe increased. This phenomenon was examined by high-time-resolved Particle Image Velocimetry (PIV). The instantaneous velocity field in a cross section was visualized two-dimensionally using PIV technique, simultaneously with the pressure measurement at multi-points around the cavity by microphones. The fluid flows at different points in the cavity interact, with some phase differences between them, and the relation between the fluid flows was clarified. The phase difference of the acoustic pressure fluctuation at different points around the cavity was also obtained. Consequently, phase delays of oscillation at different points were obtained two-dimensionally. 2-D phase map was helpful to discuss the feedback mechanism of the self-induced vibration. This is the first research which can obtain this kind contour map using the time sequential instantaneous velocity fields for closed tandem side-branches system which has long side-branches (L/D ≫1) and high inflow velocity at high resonant frequency.

Commentary by Dr. Valentin Fuster
2010;():325-328. doi:10.1115/ICONE18-29018.

Radiation shielding material investigations is an important subject which relates to human safety, for those engaging in medical radiation, radiology, and facilities related with nuclear radiation, manned spacecraft and so on. Therefore necessity of finding effective shielding materials which have low weight, and cost, and have stability against radiation, with high temperature resistant, and no biological damage,[[ellipsis]] is obvious. On the other hand, because of the multi-variability of the problem, it is experimentally difficult to determine the optimum values of the concerned variables. For this reason, simulation for discovering optimized shielding material is important. Parameters such as attenuation, thickness, and cost of gamma radiation shielding materials were optimized by using genetic algorithm combined with MCNP code. Multi objective genetic algorithm (GA) is one of the scientific solutions for this problem.

Commentary by Dr. Valentin Fuster
2010;():329-335. doi:10.1115/ICONE18-29032.

Emergency Action Level (EAL) is an effective basis and criteria for nuclear power plant emergency classification which is a pre-determined, site specific, observable threshold for a plant initiating condition that places the plant in a given emergency classification level. Systematic approaches have built a set of generic EAL guidelines, together with the basis for each, such that they could be used and adapted by each utility on a consistent basis. EALs information is presented by Recognition Categories (RC), in which “A” Recognition Category refers to abnormal Rad levels/Radiological effluent. The methodology of EALs development is reviewed first and the technical bases for “A” Recognition Category according to classic EAL methodology, i.e. NEI 99-01 series and IAEA-TECDOC-955 are introduced. The statue of Chinese nuclear power plants’ “A” Recognition Category EALs development is summarized after that, which intends to raise some challenging questions for classic “A” Recognition Category EALs development. One opinion is that some “A” Recognition Category EALs in NEI 99-01 series are not conservative. To explain the reason, the discussion mainly focuses on the interpretation and comparison of NEI 99-01 and IAEA-TECDOC-955 technical bases. Finally, it is recommended to select one suitable guideline as reference when developing plant specific EALs.

Commentary by Dr. Valentin Fuster
2010;():337-343. doi:10.1115/ICONE18-29038.

Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM) with single module power of 250MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper, e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not, and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature and Xenon concentration are studied and compared in detail between these different cases. The calculating results show that, the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.

Commentary by Dr. Valentin Fuster
2010;():345-351. doi:10.1115/ICONE18-29077.

Uncertainty analysis is imperative for criticality risk assessments when using computational methods to predict the multiplication factor (keff ) for fissionable material systems. For the validation of criticality safety computer codes, code accuracy and precision are determined by the computational bias and uncertainty in the bias to account for experimental, computational and model uncertainties. For the application of criticality safety computer codes in the criticality safety design of fissionable material systems, a minimum margin of subcriticality (MMS) must be included to provide additional assurance of subcriticality for any unquantified or unknown uncertainties [1]. Because of a substantial impact of the MMS on nuclear fuel cycle operations, recently increasing interests in reducing the MMS make the uncertainty analysis in criticality safety computations more risk-significant. This paper provides an overview of two most popular keff uncertainty analysis methods for Monte Carlo neutron transport based criticality safety computations: (1) sampling-based methods, and (2) analytical methods. Examples are given to demonstrate their usage for the code validation against benchmark experiments and for the criticality safety design evaluation.

Commentary by Dr. Valentin Fuster
2010;():353-356. doi:10.1115/ICONE18-29134.

The loss of off-site power supply for nuclear power plants may lead to a core melt accident, with large potential risk. So the evaluation of loss of off-site power is important for the PSA (Probabilistic Safety Assessment). Uncertainty of the original parameters of electric grid components may be caused by the shortage of the statistical information, the approximation to model parameters or the statistic error. In order to deal with the uncertainty problem in electric grids, interval analysis method is introduced in this paper. The original reliability parameters that are changing in a certain range are treated with interval number. Thereby the uncertainty of the parameters can be calculated in the whole process of the probabilistic assessment. The proposed method has been tested on the reliability analysis evaluation of T power plant loss of off-site power. The results demonstrated the effectiveness and practical value of the interval analysis method.

Commentary by Dr. Valentin Fuster
2010;():359-367. doi:10.1115/ICONE18-29141.

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation and in the suppression of a steam explosion. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-inches large break loss of coolant accident without safe injection. The corium spreading regime was estimated by an asymptotic calculation. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water injection system was an effective corium cooling method in the ex-vessel core catcher to preclude a possible steam explosion and to suppress the quick release of steam.

Topics: Cooling , Vessels
Commentary by Dr. Valentin Fuster
2010;():369-373. doi:10.1115/ICONE18-29159.

Probabilistic safety assessment (PSA) uses a systematic approach to estimate the reliability and risk of a nuclear power plant (NPP). Over the past few years, severe accident management guidance (SAMG), which delineates the mitigation actions of core melt accidents of an NPP, has been developed to support operators and staff in the technical support center in dealing with those misfortunes. It can be expected that the implementation of SAMG will reduce the amount of radionuclides released to the environment during the accident. The plant studied is a three-loop pressurized water reactor (PWR) with large dry containment. The RCS depressurization and reactor cavity flooding can be used as an accident management strategy. Then, the decrease of LERF (Large and Early Release Frequency) is quantified using PSA approach. It can be found that strategy of RCS depressurization and reactor cavity flooding can mitigate the result of severe accident effectively.

Topics: Cavities , Floods
Commentary by Dr. Valentin Fuster
2010;():375-380. doi:10.1115/ICONE18-29165.

As components in an actual nuclear system have multi states and are dependent on each other, a method for system reliability analyse based on Markov theory and Bayes network is presented in this paper. The method resolves the problems of multi-state and dependence simultaneously, by utilizing Markov theory and Bayes network synthetically. Applying the method to analysing reliability of power-supply system in nuclear power plant, the system failure probability, RAW and posterior probability of each component are obtained, which are useful for system reliability evaluation, improvement of weakness and fault diagnosis. Synchronously, the method is proved feasible and effective for solving system reliability that contains dependent and multi-state components.

Commentary by Dr. Valentin Fuster
2010;():381-389. doi:10.1115/ICONE18-29186.

After the September 11th terrorist attacks, the Convention on the Physical Protection of Nuclear Material was revised, and many countries have enhanced their regulatory regimes about the management of sensitive information, especially in the physical protection system. Japan also amended the Nuclear Reactor Regulation Law in 2005 in step with this global movement. The major areas of this revision which are associated with sensitive information are as follows: formulation of the Design Basis Threat (DBT), introduction of inspection system of physical protection and obligation of confidentiality of the secret of physical protection. Through this amendment, the responsibilities of the national government and the utilities have been clarified. However, there is no prescription which ordains the role and responsibility of the local governments. In fact, the local governments receive various information from the utilities through the “Safety Agreements” which are concluded between the local governments and the utilities, and the Public Safety Commissions of prefectures are involved in the transportation of nuclear materials. Moreover, the Act on Special Measures concerning Nuclear Emergency Preparedness provides the engagement and the responsibility of the local governments in case of nuclear disaster. In addition, the Civil Protection Law also provides the formulation of local governments’ plans for a response to national emergencies including nuclear disaster which is caused by terrorist attacks. As described above, the local governments are in a position where they can or have to touch the sensitive information in a variety of ways. Originally, the local government employees have obligation of confidentiality by the Local Public Service Act. Thus, about the sensitive information, they have duty to keep secret. However, we are hard to say that there are complete systems to check this obligation, so we can point out that its effectiveness is doubtful. Especially, the sensitive information which is related to nuclear materials is vital for security of the nation as a whole. Under such awareness, we’re studying the change of the local governments’ way of the management of sensitive information accompanied by the strengthening of Japanese nuclear regulation, and the actual condition of it. Now, we interview some local governments’ departments in charge where nuclear facilities are located. In this paper, we discuss the actual condition and the problems around the local governments’ management of the sensitive information.

Topics: Governments
Commentary by Dr. Valentin Fuster
2010;():391-395. doi:10.1115/ICONE18-29247.

The availability of alternating current (ac) electrical power is essential for the safe operation and accident recovery of commercial nuclear power plants (NPPs). And reliable off-site power is one key to minimizing the probability of severe accidents. This paper mainly issues the risk of SBO resulting from Loss Of Offsite Power (LOOP) on the typical M310 type stations (3-Loops, 1000MWe, pressurized NPP) of China during full power operation and normal shutdown/RHR states with constructing SBO event trees and fault trees, and Human Reliability Analysis (HRA) and data analysis are also taken into account in this analysis. Further more, Based on these work, the Core Damage Frequency (CDF) values and associated Minimal Cut Sets (MCSs) are gained. The results from the analysis give us some useful information that can help to find the weaknesses of NPP design or operation and to improve the safety level of NPPs operation.

Commentary by Dr. Valentin Fuster
2010;():397-404. doi:10.1115/ICONE18-29256.

It is the important feature of passive system and the basic difference from the active system that nuclear plant can be driven to safe state or shutdown by inherent safety characters of the reactor and physical principles, independent of human interfere or the operation of outside equipments, when the reactor is in abnormal condition. So passive system is widely used in new generation nuclear power plant (NPP) such as high-temperature gas-cooled reactors and AP1000 NPPs. While physical process failure become one of the important contributors to the system operation failure since system operation is depending on natural force but not on outside power and both the driven force and resistance are influenced by many uncertain factors. Then finding the key factors for the system operation, analyzing the development of the passive system combining with the accident scenarios are the main steps of the analysis of the passive system reliability, and the important content of the probability safety assessment (PSA) of nuclear plant with the passive design. In this paper, a model for analyzing the passive system reliability is described, in which variance decomposition and analytic hierarchy process (AHP) methods are used to select the key factors for the system operation, and Monte Carlo simulation and dynamic event tree methods are used to evaluate the system reliability according to the accident scenarios. Finally, Passive Residual Heat Removal System in the High Temperature Gas-Cooled Reactor (HTGR) is analyzed as an example.

Commentary by Dr. Valentin Fuster
2010;():405-411. doi:10.1115/ICONE18-29273.

The Chemical and Volume Control System (CVCS) in nuclear pressurized power plant are in charge of coolant pump seal injection, primary loop volume control and plant safety or security. The characteristics of the system in normal operation and incident situation will leave a remarkable impact on the power plant safe operation and security. The numerical verification of CVCS behavior in normal operation shows that the model established in this paper can predict the thermal hydraulic characteristics of the system accurately. CVCS and SIS are numerical investigated with the Computational Software, Flowmaster V7.5, and the models are validated with design data. If SGTR takes place, the flow limits of letdown, charging and coolant pump seal injection will significantly influence the reactor safety. In this accident situation, the safety injection mode of the Reactor Safety Injection System will be switched into the charging mode of the Chemical and Volume Control System, the charging and seal injection maximum flowrates variations with Reactor Coolant System pressure are predicted, and different charging modes are simulated and compared in detail. The comparison results show that one low head safety injection pump of SIS will increase the charging and coolant pump seal injection flow significantly, but two low head safety injection pumps of SIS will not increase these two flows more greatly than the one low head safety injection pump mode.

Topics: Control systems
Commentary by Dr. Valentin Fuster
2010;():413-419. doi:10.1115/ICONE18-29287.

Compacted Na-bentonite blocks, with the original water content of 9% and compacting density of 1800kg/m3 , from the Gaomiaozi (GMZ) deposit in Inner Mongolia Autonomous Region have been experienced about two months experiment in order to reveal the transmission patterns of temperature and the heat induced moisture movement in GMZ Na-bentonite blocks under high-level radioactive waste repository-like conditions. Based on the design, temperature and humidity composed micro-sensors have been used and allocated in different positions in the bentonite blocks as the check points of the experiment. There are two thermal conduction stages in GMZ bentonite blocks in a closed cylinder heat conduction system. In the first stage, the temperature at the check points in the bentonite blocks near the heater in the centre of the cylinder increased to 55°C in 20 hours, and the temperature at the check points distributed near the edge of the cylinder reached about 48°C in the same period. The temperature was maintained at about 60°C for about 35 days in the first stage. The heat induced moisture movement in bentonite blocks in the cylinder is different from changes of the temperature. The relative humidity at the check points near the heater increased quickly at the beginning of the heating, and then slowly decreased with the temperature maintained at 60°C. The average radial temperature gradient (GT ) and radial relative humidity gradient (GHR ) in bentonite blocks are 0.85°C/cm and 1.32%/cm respectively in the first stage. At the end of the first thermal conduction stage, the temperature of heater was slowly increased into 85°C and maintained this temperature for about 25 days as the second thermal conduction stage. The transmission patterns of temperature and the heat induced moisture movement in bentonite blocks are similar to the first stage. However, the radial temperature and humidity gradients are higher than that in the first stage. The average radial temperature gradient (GT ) and radial relative humidity gradient (GHR ) in bentonite blocks are 1.57°C/cm and 1.89%/cm respectively in the second stage.

Commentary by Dr. Valentin Fuster
2010;():421-428. doi:10.1115/ICONE18-29294.

To protect the sodium cooled FBR plant against the hazardous effects of sodium leak into the ambient, one of the passive protection devices used is the Leak Collection Trays (LCT) below the secondary sodium carrying pipelines in the Steam Generator Building (SGB). The design of LCT is based on immediate channeling of burning liquid sodium on the funnel shaped ‘sloping cover tray’ to the bottom ‘sodium hold-up vessel’ in which self-extinction of the fire occurs due to oxygen starvation. In the secondary heat transfer circuits of FBRs, leakage of liquid sodium from the pipelines is postulated as one of the design basis accidents with probability of occurrence at 10−2 per reactor year. LCT collect the leaked sodium in a hold up vessel, suppress the sodium fire due to oxygen starvation and guide the sodium to an inerted ‘sodium transfer tank’ located at the bottom most elevation of the SGB. The procedure of draining the leaked sodium into the transfer tank has been envisaged as a defense in depth measure against the handling of un-burnt sodium and to guard against larger leak rates than that can be handled by the LCT effectively. Towards this, a network of carbon steel pipelines are laid out connecting all the LCT and the transfer tank through headers in strategic locations, each having a fusible plug. The fusible plug separates the air environment in LCT and argon environment in sodium transfer tank. Woods metal is the preliminary choice for the fusible plug. It is an alloy of 50% Bi, 25% Pb, 12.5% Sn and 12.5% Cd with a melting point of 72°C. The transfer tank is filled with argon at ∼ 0.03 bars-g pressure. Both the header and the tank are at room temperature during normal conditions. Leaked sodium by virtue of its high temperature has to heat up the fusible plug to melt the same and drain into the transfer tank. Transient thermal hydraulic investigations have been carried out to predict the fusing characteristics of woods metal plug. The numerical results have been validated against analytical solutions for idealized conditions. Detailed parametric studies have been carried out with plug thickness as a parameter. It is established that effective melting of the plug and trouble free draining of the leaked sodium is possible for a 3 mm thick fusible plug.

Commentary by Dr. Valentin Fuster
2010;():429-431. doi:10.1115/ICONE18-29332.

Probabilistic safety assessment (PSA) on a specific reactor are often implemented without considering the applicability of generic reliability data, and doubt about such assessments is aroused because of the lack of plant-specific reliability data. The applicability of generic reliability data is analyzed in present paper, in order to remove the doubt in a way. Several sets of reliability data composing from different sources are researched. The following analysis evaluate a fault tree for a typical example of a reactor, using several sets of reliability data and show the differences in the results. Additionally, a comparison is made with a procedure of analysis using reliability data ranges. The results show that the probabilistic safety analysis on a specific reactor using reliability data which come from different sources is feasible. The differences are slight for most components, only a few key components should be separated in the first place and concentrated more attention on them. The superiority of plant-specific data should be advocated. In the mean time, the lack of data should not be a barrier for PSA on a specific reactor. And the analysis facing a lack of data is advised to be encouraged as an approach to improve the safety of specific reactors.

Topics: Reliability
Commentary by Dr. Valentin Fuster
2010;():433-438. doi:10.1115/ICONE18-29346.

Lower hybrid current drive (LHCD) is an efficient method for noninductive current drive in fusion devices. The LHCD system has been constructed on the Experimental Advanced Superconduct Tokamak (EAST). It is a complex system due to lots of devices involved. Each device has possibility of faults, which causes great difficulties in fault diagnosis. Consequently, a fault diagnosis expert system is essential for a safe and steady operation of the LHCD system. This paper proposes an expert system called LFDES (lower hybrid current drive fault diagnosis expert system) to aid operators in diagnosing and analyzing abnormal situations of the LHCD system. After a brief description of the structure of LHCD system, the LFDES architecture, the knowledge base, the inference engine and the database are presented in detail. Based on an empirical knowledge, the diagnostic tree of LHCD system is built. A fuzzy group multiple attribute decision making method is used to determine the priorities of nodes in the diagnostic tree. KDevelop tool, QT Designer tool and Linux operation system have been used in developing the proposed system. In the study, satisfactory results were obtained. The analyses of the results indicated that LFDES can provide reliable, efficient and economical service.

Commentary by Dr. Valentin Fuster
2010;():439-442. doi:10.1115/ICONE18-29352.

After the construction of nuclear power plants (NPPs) the main aim is to achieve high level of safety. For this purpose different methods and techniques such as defense in depth, safety culture, human reliability analysis (HRA), human factor engineering (HFE), fault tree analysis (FTA), event tree analysis (ETA), deterministic safety analysis (DSA), and probabilistic safety assessment (PSA) etc. have been used for many years and also in present days. Although these methods are suitable for safety of NPPs but with the passage of time, changes occur in components reliability and operating procedures, which continuously modify configuration of NPP. In order to handle these situations, living probabilistic safety assessment (LPSA) and risk monitoring (RM), as an application of PSA, play an important role in updating and maintaining level of safety. The objective of this paper is to summarize the history of LPSA and RM. The study also highlights a newly developed risk monitor called Risk Manager that has been presented in this paper.

Topics: Safety
Commentary by Dr. Valentin Fuster
2010;():443-450. doi:10.1115/ICONE18-29395.

In the fast reactors, rapid and accurate detection of fuel failures as well as subsequent identification of failed fuel location are essential to achieve their safety operation and high plant availability. The gas tagging method, currently employed in the prototype fast breeder reactor Monju, is one of the efficient ways for the failed fuel detection and location (FFDL) technique, the principle of which is the isotope analysis of the argon (Ar) cover gas that includes, in case of fuel failure, a partial amount of leaked krypton (Kr) and xenon (Xe) originally loaded into each fuel pin. We propose a new type of FFDL technique using laser resonance ionization mass spectrometry (RIMS) for the isotope analysis of the cover gas in view of selective ionization of a specific element to obtain high S/N ratio. Nevertheless, the actual experimental data shows the existence of Ar and Ar2 non-resonant ionization by the photoelectron generated in the vacuum chamber to hinder precise measurement of Kr and Xe. We could successfully decrease the effect of these ions by one to two orders of magnitude by applying both a set of a neutralization apparatus and a Brewster window, and an electrode with a slit-type hole in the ion acceleration region, resulting in reliability improvement of RIMS in the FFDL system.

Commentary by Dr. Valentin Fuster
2010;():451-460. doi:10.1115/ICONE18-29473.

The plant dynamics and the interactions of plant systems and operator with the random evolution of parameters led to the development of dynamic reliability methodologies. The present paper approaches a dynamic PSA methodology by making use of the thermal–hydraulic model of CANDU 6 reactor, implemented in RELAP5 Mod.3.3, and of IDDA code. IDDA (Integrated Dynamic Decision Analysis) is a software code that with the help of enhanced dynamic event tree methodology provides results in terms of system unavailability. The coupling of the two above mentioned codes allows a full representation of the plant operational states, as well as of all the possible occurrence patterns that complete the spectrum of possible probability-consequence conditions. The plant transient considered for the CANDU 6 thermal–hydraulic model is the total Loss of Feed Water (LOFW) supply to the secondary side of steam generators. That is followed by depletion of water inventory and subsequent cool down via Emergency Water System (EWS). The present dynamic PSA approach reveals those situations where the correct intervention of protective equipment could bring to unexpected events. This allows taking the most appropriate decisions for the given plant configuration.

Commentary by Dr. Valentin Fuster
2010;():461-466. doi:10.1115/ICONE18-29483.

Baseline Risk Index for Initiating Events (BRIIE) is an integrated industry-level initiating event performance indicator that is risk informed. BRIIE is a performance indicator that provides a mechanism for determining the risk significance of changes in performance, at both the individual initiating event level and at the integrated cornerstone of safety level. This paper describes the process to use BRIIE to evaluate the initiating events in Daya Bay NPP. We established two indexes for BRIIE, one is based on NRC baseline, the other is based on the plant specific PSA initiating events frequency. Both indexes of BRIIE are calculated in two tiers.

Commentary by Dr. Valentin Fuster
2010;():467-474. doi:10.1115/ICONE18-29488.

Digitalized nuclear instruments and control systems have become the main stream design for the main control room (MCR) of advanced nuclear power plants (NPPs) nowadays. Digital human-system interface (HSI) could improve human performance and, on the other hand, could reduce operators’ situation awareness as well. It might cause humans making wrong decision during an emergency unintentionally. Besides, digital HSI relies on computers to integrate system information automatically instead of human operation. It has changed the operator’s role from mainly relating operational activity to mainly relating monitoring. However, if operators omit or misjudge the information on the video display units or wide display panel, the error of omission and error of commission may occur. Therefore, how to avoid and prevent human errors has become a very imperative and important issue in the nuclear safety field. This study applies Performance Evaluation Matrix to explore the potential human errors problems of the MCR. The results show that the potential problems which would probably affect to the human performance of the MCR in advanced NPPs are multiple accidents, pressure level, number of operators, and other factors such as working environmental.

Commentary by Dr. Valentin Fuster
2010;():475-482. doi:10.1115/ICONE18-29601.

The development of a corium pool in the lower head and its behavior is still a critical issue and is of great importance to assess the severe accident progression consequences to ensure the nuclear plant safety. Therefore, experimental efforts are a vital element of the assessment process, providing hard data and insights of the complicated multi-component, highly turbulent corium pool dynamics. It is essential to consider the whole evolution of the accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behavior after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at the Karlsruhe Institute of Technology (KIT) is to study these phenomena experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behavior. The LIVE-L4 experiment was performed using a non-eutectic melt (KNO3 -NaNO3 ) as a simulant fluid. Besides the transient behavior, for which the LIVE-L4 test provides qualified data on temperature evolution in the molten pool and crust growth rates, the experiment addresses other important phenomena, such as the local distribution of heat flux, and the influence of solidification on the thermal-hydraulics of the pool, i.e. the possible existence of a mushy region and its impact on the heat transfer. In the post-test analysis crust thickness profile along the vessel wall, the crust composition and the morphology were determined. The results of this experiment also allow a comparison with findings obtained earlier in other experimental programs. The LIVE-L4 experimental results are being used for the assessment of correlations and development and validation of mechanistic models for the description of molten pool behavior. These calculations are complemented by analyses with the CFD code CONV (thermal hydraulics of heterogeneous, viscous and heat-generating melts) which was developed at IBRAE. The CONV code was applied to simulate the LIVE-L4 test: a) assuming homogeneous heat generation in the liquid and b) accounting for wire heaters used to simulate the heat generation in the melt. Though the results of calculations demonstrate satisfactory agreement with the experimental measurements, deficiencies in the code prediction have been identified regarding e.g. the prediction of the crust thickness. The paper summarizes the objectives of the LIVE program, the main results obtained in the LIVE-L4 experiment and the results of the post-test calculations performed with the CONV code.

Commentary by Dr. Valentin Fuster
2010;():483-489. doi:10.1115/ICONE18-29621.

Visualization test was conducted by designing and setting up a special experimental apparatus to investigate characteristics of thermal-hydraulic interaction between lead-alloy droplet and subcooled water in pool water tank. The violent boiling phenomena were observed by a high-speed camera and local transient pressure was measured by high frequency piezo pressure transducer. The results showed that violent boiling mainly occurred when interface temperature between droplet and water was higher than homogenous nucleation temperature of water and subcooling temperature of water was higher than 40K. A violent boiling model for lead-alloy droplet/water interaction was proposed. In the model, partial contact of lead-alloy droplet with water caused by vapor film instability was taken into account to simulate fragmentation due to rapid evaporation.

Commentary by Dr. Valentin Fuster
2010;():491-496. doi:10.1115/ICONE18-29642.

Fuel-Coolant Interaction can be divided into two stages including premixing and explosion. In the stage of premixing, the liquid corium flowing from the core fragmented into drops, and the drops may fragmented into fine fragmentations leads to steam explosion in the condition of triggering. In the severe accident, the contact of melt-fuel and coolant may result into steam explosion, which may threaten integrity of containment. The prediction of steam explosion must be based on results of experimental research and on simulations with computer code. In this work, the experimental result of low temperature fuel-coolant interaction is compared with the computing result of FCI-CMFD code MC3D, analyzing the phenomenon of FCI, the diameter of droplet and the pressure wave of explosion. Then, the MC3D is used in the analysis and prediction of fuel coolant interaction in CPR1000 NPP model.

Commentary by Dr. Valentin Fuster
2010;():497-504. doi:10.1115/ICONE18-29643.

In order to investigate the magnitude and distribution of pressure and impulse in the reactor cavity during the process of fuel-coolant interaction includes both the premix and steam explosion, a three dimensional multiphase fluid simulation code MC3D has been used based on the structural and environmental characteristic of the sever accident state of 1000 MW level PWR Linao Nuclear Power Station Phase II with the consideration of both pressure variation and impulse variation. The results indicate that the process of premixing has a low level pressure variation with a long duration and the affect of this process is negligible. On the contrary, explosion stage creates a instant and sharp pressure pulse which threatens the integrity of the reactor cavity structure. In addition, the mass of the melting drop contacts the coolant all over the calculating grids has a significant effect regarding the magnitude of the impulse created by steam explosion.

Commentary by Dr. Valentin Fuster
2010;():505-510. doi:10.1115/ICONE18-29665.

Firstly this article describes how to analyze accident sequence precursor in U.S. nuclear regulatory commission. Following this method, the licensee operational events, internal operational events and inspection reports of Daya Bay and Lingao Nuclear Power Plants were reviewed to identify the precursors with support of probabilistic safety assessment. All the identified precursors were calculated, documented and ranked. Then the trends on precursors can be obtained. Finally this article analyzes the trends available and gains many beneficial insights.

Commentary by Dr. Valentin Fuster
2010;():511-517. doi:10.1115/ICONE18-29681.

Safety reports have shown that tons of solid particles would be generated as dusts in the operation of ITER facility. The dust particles include carbon, beryllium and tungsten with diameters ranging from a few to a few hundreds microns. The particles deposit downwards and mostly accumulated on the surfaces of the diverter on the bottom side of the vacuum vessel (VV). In accident scenarios, e.g., loss of vacuum accident (LOVA), the potentially combustible dust particles can be suspended by the air ingress and entrained into the whole volume of the VV, and impose a risk of dust explosions in case of unintentionally ignition to the whole ITER facility. Therefore the mechanism of particle resuspension was investigated theoretically in the work. A force balance approach and numerical fittings have been utilized to develop a semiempirical particle resuspension model based on a group of particle resuspension experimental data. The model has been applied into a three-dimensional computational fluid dynamics code, GASFLOW. The model validation has been done by comparison of the numerical predictions about particle resuspension rates in given incoming flows against the corresponding experimental data. The comparisons have proved the validity of the developed model about particle resuspension.

Commentary by Dr. Valentin Fuster
2010;():519-528. doi:10.1115/ICONE18-29716.

Nowadays, there is a general consensus that establishing nuclear regulations concerning human factors engineering (HFE) is an important issue. NUREG-0711, original version published in 1994, was developed under such assumptions. And it soon became a common reference for nuclear power plant reviewers and designers. Lungmen NPP is the first ABWR plant in Taiwan and is under construction now. Taipower Company signed the contract for Lungmen Project with General Electric Company in 1995. By Lungmen Project Bid Specification, GE should take the responsibility to design the main control room according to the last version of HFE regulation that is NUREG-0711 version zero. Up to the present, NRC has modified NUREG-0711 twice on the basis of evaluating experiences and users’ feedback from different fields. But the Lungmen NPP has not finished yet. No doubt, the modifications not only make the regulation state-of-the-art but practicable. How to cope with this asynchronous problem between contracts and modification is a critical concern. In this article, we present our resolutions on this issue. Step one; comparing the differences between NUREG-0711 version zero and two. Step two; figuring out what meanings and intent are behind these changes. Step three; following the version zero regulation and taking advanced principle into consideration at the same time. Implementation according to old version regulation and taking the advanced intent and principle from step 2 is a practice resolution from the experience of Lungmen NPP. Those experiences will be helpful for human factors engineering activities on update the advanced main control room of nuclear power plant in the near future.

Commentary by Dr. Valentin Fuster
2010;():529-534. doi:10.1115/ICONE18-29722.

Nuclear safety culture has been one of the most important aspects for safe and reliable operations of nuclear power plants worldwide. Japanese nuclear operators started promoting Safety Culture using the PDCA (Plan-Do-Check-Action) cycle in December 2007. Chubu Electric Power Company has been developing the system of promoting safety culture and challenging to develop evaluation tools for safety culture.

Topics: Safety
Commentary by Dr. Valentin Fuster
2010;():535-541. doi:10.1115/ICONE18-29723.

The magnetic properties will be lost when the temperature of Curie point alloy is above the Curie point temperature. The physical properties of Curie point alloy are used for Passive Actuated Shutdown System (PASS) in the Fast Reactor, for example American, Japan, the shutdown systems are called Self Actuated Shutdown System (SASS). The shutdown device is composed of the up assembly and the down assembly. The up one is a magnetic material and the down one is Curie point alloy. The reliability of PASS in normal condition, and whereabouts of sensitivity are determined by the properties of Curie point alloy. The Reactor-type Experiment Furnace and related circuits were designed to carry out static and dynamic experiments under the high temperature. Cylindrical alloy functions were investigated through the high temperature experiments outside of the reactor. The results showed that improvement was necessary because of long response time. In order to improve the sensitivity of PASS, different shapes were designed for Curie point alloy, under the premise that PASS was reliable. The response times of different sharps were calculated by finite element software, ANSYS, in abnormal condition. Comparison of response times of cylinder, quincunx and fins model under the same conditions, quincunx structure had the shortest response time when the temperature of Curie point ally was above the Curie point temperature. The results obtained from the high-temperature experiments, also showed that the quincunx has the shortest response time. These results showed the quincunx structure was superior to the other models.

Commentary by Dr. Valentin Fuster
2010;():543-547. doi:10.1115/ICONE18-29753.

The computation of initiating event frequencies has received much attention in recent years. This attention reflects the importance of support system failures in the computation of core damage frequencies from internal events for nuclear power plants. Multi-train support system failures, such as a total loss of service water or a total loss of component cooling water, potentially represent significant impacts for many nuclear plant designs. The contribution to basic event importance from such initiator models has also been recognized, especially as part of the Mitigating Systems Performance Index (MSPI) program; Reference 1.

Commentary by Dr. Valentin Fuster
2010;():549-558. doi:10.1115/ICONE18-29784.

Buildings, equipment and pipe networks, herein below called “structures”, within nuclear and classic objectives are affected by the dynamic actions of earthquake, shocks and vibrations type, herein below called“ excitations”. The repeated action of excitations on structures most often lead to important built-up of kinetic and potential energy in oscillating systems made-up of such structures. Such a built-up is leading to tens of times increase of the amplitude of the dynamic response in accelerations and displacements of structures as to the excitation amplitude. In its turn, such an increase is leading to the exceeding of the efforts and distortions of the structures material, accompanied by the occurrence of damages or destroy. Because in most cases the excitation cannot be reduced or eliminated, the only technical solution available for the engineers is to reduce the dynamic response of the structures on the existing excitations. The innovative SERB-SITON solution developed and applied within the Subsidiary of Technology and Engineering for Nuclear Projects (SITON) is based on the use of SERB-SITON isolation behavior devices with low friction and controlled stiffness on horizontal plane to cut-off the dynamic action transfer from the excitation to the structure and telescopic devices with controlled elasticity and damping both to reduce the dynamic response of the structures, as well as the dissipation of the energy transferred to the structure or the limitation of structure distortion upon the dynamic actions. The paper presents the performances of SERB-SITON mechanical devices used to protect buildings, equipment and pipe networks against dynamic actions and an adequate method to evaluate the dynamic response of structures (by using SERB-SITON devices), in real time inclusively. For large-size buildings such as nuclear power plants, old buildings, churches, bridges, etc., SERB-SITON isolation devices with friction by sliding or rolling with a very small and adjustable, stiffness installed under the structure on any horizontal direction, are presented. For reinforced concrete or metal framework buildings such as high buildings, industrial halls, towers, etc., SERB-SITON telescopic devices with controlled stiffness and damping large, force inclusively are presented. For equipment and pipe networks, SERB-SITON supports that are capable to overtake large permanent loads with relative displacements on two directions for thermal displacements, and also capable to elastically overtake and damp dynamic actions, are presented.

Commentary by Dr. Valentin Fuster
2010;():559-568. doi:10.1115/ICONE18-29789.

The paper presents three types of a passive safety containment for a near future BWR. They are tentatively named Mark S+ , Mark D and Mark X containments in the paper. They all have a leak tight secondary containment vessel (SCV) in order to meet the reactor site criteria without relying on an active standby gas treatment system at a DBA LOCA. One of their common features is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The containment pressure can be limited within the design pressure. Even if a large amount of hydrogen is generated at a severe accident, it can be released into the SCV. Hydrogen detonation or deflagration is completely prevented without using igniters. Another feature is the capability to submerge the PCV and the RPV above the core level without relying on accident management. The core debris is completely submerged not only ex-vessel but also in-vessel. The third feature is robustness against external events such as a large commercial airplane crash. All the containments have built-in passive safety systems (BIPSS) including a passive containment cooling system (PCCS) and a passive cooling core catcher that has radial cooling channels. The Mark S+ and Mark D containments are applicable to a large power BWR up to 1830 MWe. The SCV is made of steel-concrete composite. The PCV can be vented into the inerted part of the SCV at a severe accident. The Mark X containment has the steel secondary containment vessel (SSCV) and can be cooled by natural convection of outside air. It can accommodate a medium power BWR up to about 1000 MWe and has a permanent grace period without replenishing the PCCS pool. In all cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides in-depth protection against severe accidents and also enables N+2 design. All the three containments coupled with the IDHS can potentially provide an evacuation free plant at a severe accident caused by severe natural disasters such as a giant earthquake, a tsunami, a mega hurricane, and so on.

Commentary by Dr. Valentin Fuster
2010;():569-575. doi:10.1115/ICONE18-29818.

In-vessel retention (IVR) of core melt through external reactor vessel cooling (ERVC) is a key severe accident management strategy to ensure that the vessel head remains intact and eliminate consequent major threats to containment integrity. To maintain the margin against the failure of the reactor vessel and make sure of the feasibility of IVR for its role to confine the molten corium with 1400MW power, CAP1400, Chinese version of large passive PWR, systematic investigations including experimental and analytical researches of IVR are very important to the development of CAP1400. This paper briefly reviews the progress and tasks of a four-year project which was planned to analyze, evaluate, improve and validate the effectiveness of IVR employed in the CAP1400.

Topics: Vessels
Commentary by Dr. Valentin Fuster
2010;():577-581. doi:10.1115/ICONE18-29819.

Uncertainties are addressed in the special context of assessing and managing risks from rare, severe-consequence hazards. Risk Oriented Accident Analysis Methodology (ROAAM) is used to analyze uncertainties during severe accidents analysis in nuclear power plants. In-vessel Retention (IVR) is one of the mitigations for severe accidents which will cause core damage. By external reactor vessel cooling (ERVC), the integrity of the reactor vessel is preserved. The success criterion for IVR is the local heat flux on the wall of lower head is less than the critical heat flux (CHF). This paper analyzes the uncertain parameters which decide the mitigation to be successful or fail. Two bounding structures and 4 molten pool steady states are defined. And the success probability of IVR is evaluated with a molten pool heat transfer model. Then the effectiveness of IVR-ERVC under the two bounding structures is evaluated.

Commentary by Dr. Valentin Fuster
2010;():583-588. doi:10.1115/ICONE18-29833.

The analytical and experiment research of In-Vessel Corium Retention (IVR) in the Chinese Pressurized-water Reactor 1000 MWe (CPR1000) are introduced. The IVR research consists of preliminary phase and detailed phase. The analysis of thermal failure, structural failure and penetration failure of Reactor Pressure Vessel (RPV) and the experimental research of External Reactor Vessel Cooling (ERVC) are performed at preliminary phase. Analysis results show that the RPV failure is the dominated by thermal failure mode and the probability of the thermal failure is very low. Test results show that the IVR success probability for CPR1000 is about 99% if the Critical Heat Flux (CHF) of CPR1000 is the same as that of AP600. Further works, including the ERVC enhancement design, the CHF test of the RPV outer wall and the recalculation of the IVR success probability for CPR1000, will be performed at detailed phase in the near future.

Topics: Vessels
Commentary by Dr. Valentin Fuster
2010;():589-596. doi:10.1115/ICONE18-29857.

The project of nuclear station LNPP-2 with a reactor power plant VVER type by electrical power 1200 MVt involves a number of new design solutions to increase of parameters of safety. The passive containment heat removal system and heat removal system via steam generators is including of number of such solutions. Passive heat removal system via steam generators (PHRS/SG) is assigned for remove of residual heat of reactors core to final heat absorber (atmosphere) through a secondary circuit at DEC accident. The system PHRS/SG duplicates cooling-down system via SG to final heat absorber in case of impossibility of realization of its design functions. Containment heat removal system (PHRS/C) is assigned for remove of residual heat from containment in accidents with heat-transfer emissions from primary circuit. PHRS/C duplicates functions of a spray system to reduce of pressure under containment in case of spray system failure. In the substantiation of passive security systems the complex in SPbAEP of computational and experimental analysis was executed, the main results of which are shown in the present report.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2010;():597-603. doi:10.1115/ICONE18-29860.

The equipment of LNPP-2006 for severe accident’s managements includes the device of core melt localization–core catcher. By analogy with core catcher of the Tainvan NPP in China, the LNPP-2006 core catcher has been developed based on the crucible concept that combines corium retention inside the vessel (passive water cooling of metal surfaces that form the boundary of the corium localization zone) and the control of physical and chemical properties of the corium (by the use of sacrificial material). By present time design works in which the basic design parameters of the core catcher are defined have been performed. It was allowed to execute the verification calculations of core catcher processes taking into account behaviour of reactor’s cavity materials. For the analysis of core catcher severe accident’s behaviour code HEFEST-CC was used, for its adjustment three-dimensional hydraulic codes were applied. As a result of this analysis the quantity of the additional materials, which supplying from reactor’s cavity during concrete destructions under the influence of radiation was adjusted. The cooling ability of a passive water system of a melt surface has been confirmed. Introduced paper is devoted the analysis the formation dynamics and parameters of the molten core bath inside the core catcher vessel, to destruction the fracture dynamics of a thermal protection and response time the water passive feeding system on a melt surface, to definition the crisis store before heat exchange on the water-cooled vessel.

Commentary by Dr. Valentin Fuster
2010;():605-612. doi:10.1115/ICONE18-29879.

We examine the prediction of real accident and event probability in the absence of prior data and/or with partial knowledge, when the human contribution is properly included. We now know that the major cause of all real accidents (not postulated ones) is actually the unforeseen human contribution, as an integral and inseparable part of the technological system. The real events we actually will experience or observe in our lives may be spectacular plane, train, space shuttle or stock market crashes. In every case, they are unexpected occurrences, they seemingly appear randomly, and how often they happen, or the rate of such events, covers the whole spectrum from frequent to rare. Because so-called rare events do not happen often, they are also widely misunderstood and do not follow the expectations or the same “rules” governing many or frequent events, and are always due to some apparently unforeseen combination of circumstance, conditions, and combination. Usually in safety analysis, a distinction is made between “probabilistic” safety analysis (PSA), based on examining so-called risk dominant accident sequences, and “deterministic” safety analysis (DSA). Intended to be complementary, PSA provides insights into risk scenarios and allowing numerical estimation of outcomes for transients, such as loss of offsite power (LOOP) or station blackout (SBO), and the resulting core damage frequency (CDF) or large early release frequency (LERF) with some estimated uncertainty in the calculated probabilities of occurrence. In contrast, the DSA provides a standard set of stylized events, such as large breaks (LOCA) and transients (ATWS), as a means of setting safety margins and design criteria, as also proposed in Theofanous’s ROAMM, where extremes of knowledge are postulated as a test of the robustness of the design and safety systems. These methods can produce statements of margins and uncertainties, and converge in the area known as “risk informed regulation” (RIR), where the insights gained are proposed to derive limiting Farmer-type “tolerable risk” boundaries or frequency-consequence (F-C) curves. Conversely, real accidents are often unknown sequences, with no priors or precursors, and/or include possibly unforeseen initiators (for example, undetected pressure vessel corrosion) and the key role of the human. In this paper, we address the question of the quantitative prediction of such real and rare events, their occurrence probability and hence the risk.

Topics: Safety , Accidents
Commentary by Dr. Valentin Fuster
2010;():613-621. doi:10.1115/ICONE18-29882.

Shutdown seismic PSA was performed for the plant cold shutdown condition. Normally, automatic ECCS signal is blocked after the Permission Signal is indicated during the plant cooldown process of PWR. In response to this, PSA model including manual actuation of ECCS signal on the postulated LOCA condition during cold shutdown state was developed based on the actual plant Emergency Operation Manual, which consists of several steps, such as, 1) LOCA event diagnosis, 2) isolating RHR system, 3) manual actuation of ECCS signal, 4) switching RHR to Low Pressure Injection mode. This PSA model with evaluated human error rates, was quantified demonstrating that risk level for the cold shutdown condition is in the acceptable range, however, suggesting some room for model improvement considering more realistic manual ECCS actuation procedure depending on the size of the LOCA.

Commentary by Dr. Valentin Fuster
2010;():623-628. doi:10.1115/ICONE18-29888.

The Examination Guide for Seismic Design of NPP was revised in 2006 in Japan. In response to the revised guide, utilities are required to establish the seismic design acceleration of the NPP site and evaluate the residual risk of individual NPP for earthquakes exceeding the seismic design acceleration. JNES is developing seismic PSA models of typical NPPs to support the regulatory review of utility’s residual risk evaluation report. Trial analysis of seismic PSA is performed for a typical BWR4 plant and a typical BWR5 plant. Dominant accident sequences and dominant initiating events are obtained, and sensitivity analysis is performed to evaluate the influence of analytical conditions on the core damage frequency (CDF) such as LOCA model and system mitigation effect. Analysis result shows that the CDF profile changes largely depending on the seismic intensity of the site.

Commentary by Dr. Valentin Fuster
2010;():629-634. doi:10.1115/ICONE18-29890.

Deterministic health effects can be prevented and the risk of stochastic health effects can be reduced by taking protective actions before or shortly after a release. These actions must be based on plant conditions and then refined subsequently based on environmental measurements. Operational intervention levels (OILs) are some calculated values (e.g., ambient dose rate or radionuclide concentration) measured by instruments or determined by laboratory analysis that correspond to a GIL or GAL. Through the use of the OILs, the environmental data are assessed primarily, which are quantities directly measured by the field instrument. Default OILs have been calculated in advance on the basis of the characteristics of severe reactor accidents. These default OILs are used to assess environmental data and take protective actions until sufficient environmental samples are taken and analyzed to provide a basis for their revision. This approach allows data to be quickly evaluated, and decisions on protective actions to be promptly made. A decision support system of the off-site emergency protective actions based on the OILs for nuclear emergencies was discussed in this paper. The system accesses the environmental data through the default OILs and the revisions of OILs. It is applied to Daya Bay Nuclear Power Plant. In the early release, according to the characteristic of the plant, the system provides the approach to calculate the default OILs based on the accident source terms described in Reactor Safety Study of USA. Also some real factors are considered, including the meteorological parameters. When sufficient environmental samples are taken and analyzed to provide a basis for their revision, the default OILs can be revised or recalculated by them. In the entire emergency planning zone, the environmental data will be assessed through the use of those OILs to provide the advice of protective measures.

Commentary by Dr. Valentin Fuster
2010;():635-640. doi:10.1115/ICONE18-29900.

The RELAP5/SCDAP Mod3.2(am5) code is employed to simulate the OSU-AP1000-05 test conducted in the A dvanced P lant Ex perimental (APEX) test facility at Oregon State University (OSU). The APEX-1000 test facility is an one-fourth height, one-half time scale, and reduced pressure integral systems facility to simulate the Westinghouse Advanced Passive 1000 MW (AP1000) pressurized water reactor. OSU-AP1000-05 is a two-inch break at the bottom of cold leg #4 with 3 out of 4 ADS-4 valves of OSU-APEX-1000 facility. RELAP5 predictions are compared to the experimental data generated by the test. The comparison shows good agreement between the predicted and measured sequence of events of some key parameters during the transient. From the comparison results, it could be preliminary concluded that the RELAP5/SCDAP Mod3.2(am5) code are suitable to simulate the small LOCA of APEX.

Commentary by Dr. Valentin Fuster
2010;():641-646. doi:10.1115/ICONE18-29906.

Nuclear off-site emergency response for nuclear power plants in Guangdong province is a complicated process, which is finished by many emergency response organizations. In a nuclear emergency, the source, format, content, take-over means and processing means of the information passed among these organizations are different. The aim of nuclear emergency response is to protect public and environment by taking some proper urgent protective action. The essential information from the original information is necessary for decision-making in the processing of nuclear response emergency, thus the key of emergency response is incepting and analyzing the information. An effective method of processing information is using a computer-based information system to aid the traditional handwork means. Nuclear Emergency management Information System in Guangdong Province is such systems, which covers preparedness, response and recover phase. At first, this paper analyzed the system functions according to the responsibilities of the emergency organizations in Guangdong province. Then the system frame is decided, which is composed of operation, information management, file processing, consequence assessment, operational intervention level (OIL) computing subsystem, geographic information system and a system database. A useful method for decision is applying the OILs in consequence assessment, and default OILs need be revised in a real nuclear emergency. Developing OIL computing subsystem is helpful to make decision with measurement result directly.

Commentary by Dr. Valentin Fuster
2010;():647-652. doi:10.1115/ICONE18-29926.

This paper analyses the shielding design of medical accelerators. To find the best plan among them, which need the lowest cost to obtain the same effect, some theoretical calculations according to the recommended methods of NCRP 151 report have been done. The paper also compares the effects on dose equivalent at the maze door when beam direction is parallel or perpendicular to the maze wall separately. Then it concludes that the photon dose at the maze door will be lower by a factor of one magnitude if the door locates at the side of the maze but not at the terminal of it. In addition, the neutron capture γ-ray and the photo neutron dose will also be greatly reduced in this situation. It should be pointed out that the beam directly hitting the maze wall is not a recommended design, although it can meet the final dose standard. Since more and more medical accelerators are used in radiotherapy, this work may be helpful to the medical accelerator shielding design both for maze and treatment room.

Commentary by Dr. Valentin Fuster
2010;():653-659. doi:10.1115/ICONE18-29946.

In Japan, a lot of efforts have been made on severe accident study, and development and application of the probabilistic safety assessment (PSA) technique. The PSA was applied to the examination of the accident management (AM) plan in the beginning of 1990s and was performed for all the nuclear power plants (NPPs) to evaluate the effects of the AM. Furthermore, the PSA has been performed as part of periodic safety review (PSR) to review the safety of individual plant. In recent years, discussions have started to apply risk information to the safety regulation or safety related activities to improve rationality, accountability and transparency. As the technical foundations, the Nuclear Safety Commission (NSC) showed the safety goals and policy toward risk informed decision making (RIDM), and the Nuclear and Industrial Safety Agency (NISA) developed guidelines for risk informed regulation (RIR). Consensus standards have been developed in the Atomic Energy Society of Japan (AESJ), the Japan Society of Mechanical Engineers, and the Japan Electric Association. Especially, the AESJ established the Standards Committee in 1999 and has made several PSA standards. Now, there are seven PSA standards. The most important one is the seismic PSA standard. It was developed ahead of the world as a concrete manual. Lots of illustrations and useful example are included for ease of use and to make decision adequately. There were needs to develop a standard, which provides the basic requirements and specific procedures commonly applicable to respective fields of utilization of RIDM regarding changes in safety related activities. Responding to such needs, an implementation standard has been developed on use of risk information in changing the safety related activities. It stands over individual standards that will be developed in future, and shows the common and basic rules. It requires being consistent with the defense-in-depth philosophy, to maintain sufficient safety margins, and to clarify the influence to safety by comparing with some criteria. And it also requires as a final step that a comprehensive decision be made by considering various items, e.g. the defense-in-depth, safety margins, risk indices, and implementation and monitoring program. We will continue to make an effort toward RIDM and develop the standard to assess dominant risk hazards, e.g. fire risk and internal flooding risk. Moreover, it is necessary to develop the standard for individual applications in future.

Topics: Decision making
Commentary by Dr. Valentin Fuster
2010;():661-665. doi:10.1115/ICONE18-29963.

The emergency shelter with disaster prevention facilities provides a safe place by construction for emergency situations such as earthquake, nuclear accident etc., which can supply emergency evacuation and temporary residence for refugees. The disaster prevention facilities of emergency shelter are complete sets of measurements which ensure the daily essentials for refugees. The paper classifies the emergency shelters and puts forward their scales and service scopes; it also proposes the setting requirements of disaster prevention facilities for every kind of emergency shelter. The study is not only has advantages to the general plan and distribution of various emergency shelters, but also makes a significant sense of improving the level of disaster resistance and rescue.

Commentary by Dr. Valentin Fuster
2010;():667-671. doi:10.1115/ICONE18-30007.

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800°C, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1242.4°C. It was a little higher than 1230°C–the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation raised to 1620°C, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM because they have the same outlet temperature of the core.

Commentary by Dr. Valentin Fuster
2010;():673-680. doi:10.1115/ICONE18-30038.

In a postulated core melt accident, if a molten core is released outside a reactor vessel despite taking mitigation actions, the core debris would relocate in the reactor cavity region and attack the concrete wall and basemat of the reactor cavity. This will potentially result in inevitable concrete decompositions and possible radiological releases. To prevent direct contact of the melt and basemat concrete of the cavity, a core catcher concept is suggested, which can passively arrest and stabilize the molten core material inside the reactor cavity. The core catcher system includes a retention device for the molten core material, a cooling water storage tank, and a compressed gas tank. Upon ablation of the sacrificial layer on top of the retention device while molten core material is discharged, a mixture of water and gas is injected from below. It is expected that a simultaneous injection of water and gas could prevent a possible steam explosion/spike. It could also suppress the rapid release of steam which might result in fast over-pressurization of the containment. A test facility for the core catcher using a thermite reaction technique for the generation of the melt was designed and constructed at KAERI. The first series of tests were performed by using a mixture of Al, Fe2 O3 , and CaO as a stimulant. As a first try, only water was injected from the bottom of the melt through five water injection nozzles when the melt front reached the water injection nozzles. In this paper, the core catcher concept and the related provisions are suggested. A description of the test facility for the core catcher, the thermite composition, and the methods of experiment is included. The first experimental results with only water injected from the bottom of the melt are discussed.

Commentary by Dr. Valentin Fuster
2010;():681-689. doi:10.1115/ICONE18-30040.

In order to develop licensing LOCA analysis program platform for nuclear power plant (NPP), a relatively realistic technique have been adopted in this paper, which is to satisfy the requirements of conservative evaluation models (EM) in related regulation (10CFR50, Appendix K) by modifying related models or correlations of the best estimate (BE) program RELAP5/MOD3, in an effort to form a licensing LOCA analysis tool. By using this “advanced program platform plus conservative EM” technique, LOCA analysis on the Chinese 300MW NPP, of which LOCA used to be conservatively evaluated, is attempted so that further potential of margins should be achieved. By comparing with Appendix K, 10 models or correlations in RELAP5/MOD3 code are identified, which need to be modified, invoked, evaluated or verified. Modifications or control have been separately introduced on such models or correlations as fission product decay model, correlations for critical heat flux (CHF), post-CHF heat transfer correlations, logic of preventing from returning nucleate boiling and transition boiling heat transfer prior to reflood, discharge model, metal-water reaction model, emergency core cooling (ECC) bypass model and related models for pressurized water reactor (PWR) refill and reflood. With the accomplishment of modification or invoking the above-mentioned models, conservatism verifications by utilizing proper standard curve or separate test data have been carried out. The standard curve or separate effect tests data includes: (1) 1971 fission product decay standard curve; (2) related ORNL THTF test data; (3) Marviken test-22 data; (4) Cathcart oxidation test data; and (5) Westinghouse FLECHT - SEASET test data, etc. Further, for the integral effect test LOFT L2–5, calculations with both original RELAP5/MOD3 program and the program of which some related models modified in compliance with the requirement of 10CFR50 Appendix K have been made. The results indicate that, on the one hand, ability of the original BE program simulating the accident progress is identified; and on the other hand, the modified models are confirmed to be conservative. Moreover, both individual contributions and combined effects of the model modification to conservatism of the integral test analysis are observed and analyzed. To assess effects of the “advanced program platform plus conservative EM” approach on LOCA analysis, finally, calculations with the modified code for both hypothetical small break and large break LOCA of the Chinese 300MWe NPP are conducted. Compared with the calculated results of original RELAP5/MOD3 program and that delivered in Final Safety Analysis Report (FSAR) of the NPP, mutual and integrated effects with different model modifications are investigated. And finally, safety margin achieved by the present calculation is preliminarily discussed.

Commentary by Dr. Valentin Fuster
2010;():691-695. doi:10.1115/ICONE18-30070.

Two approaches have been proposed to solve the large-scale fault trees or event trees for Probabilistic Safety Assessment in a nuclear power plant. The first one consists in MCS/ZBDD, which uses ZBDDs (Zero-suppressed Binary Decision Diagrams) to implement classical MCS (Minimal Cut Sets) algorithm. The second consists in designing heuristics and strategies to reduce the complexity of the BDDs (Binary Decision Diagrams) construction. This paper was motivated to combine the MCS/ZBDD and designing heuristics for ZBDDs together. A heuristic, which took the failure rate of basic event into account and utilized that truncation could be implemented on ZBDDs during the calculating process, was proposed. This heuristic accelerated the analysis progress by bringing forward the truncation and reducing the complexity of the intermediate ZBDDs. RiskA, a Zero-suppressed Binary Decision Diagram package extended to safety and reliability analysis, has adopted this heuristic. RiskA’s truncation strategies, which had some relations with the ordering scheme, were also introduced. The correctness and efficiency of this new heuristic were verified by some practical models’ analyses.

Commentary by Dr. Valentin Fuster
2010;():697-701. doi:10.1115/ICONE18-30072.

Risk monitor, which has been widely used in the progress of RID (Risk Informed Decision) in a NPP (Nuclear Power Plant), is a plant specific real-time analysis tool to determine the instantaneous risk based on actual plant configuration. Based on wide investigation of challenges and technical issues during the development of a risk monitor, a prototype named Risk Angel has been designed by FDS Team in collaboration with several institutes and universities. An overview of the architecture and main functions of Risk Angel were introduced in this paper, as well as the quantitative approach, the calculating engine and the model development. Risk Angel has been applied in a nuclear power plant in P.R. China.

Commentary by Dr. Valentin Fuster
2010;():703-707. doi:10.1115/ICONE18-30147.

The research on radiological impact of tritium is highly concerned in high temperature gas-cooled reactors. In order to better assess the environmental performance of HTR-PM (HTR demonstration project with 2 × 250MW plants), analysis of tritium behavior in HTR-PM is conducted in this paper. The main production sources of tritium are the ternary fission in the fuel and neutron capture reactions of some nuclides. Based on the tight interactions between tritium sources and sinks, differential equations are built to describe tritium behavior in primary and secondary loop. Specific analysis is conducted to tritium permeation through heat exchanger walls to secondary loop, considering the oxidation of alloys used for heat exchanger. Applied with the parameters of HTR-PM, tritium concentration in primary and secondary loop is calculated, and the amount of tritium released to the environment is evaluated. The evaluation shows that the amount of tritium released to the environment is less than the limit value prescribed by Chinese regulation on radiation protection. The calculation results can also be applied to the safety analysis and used to guide the design of relevant systems and equipments for the HTR-PM.

Commentary by Dr. Valentin Fuster
2010;():709-718. doi:10.1115/ICONE18-30181.

In the framework of the research activities of the EURATOM FP6 project named ELSY (European Lead-cooled System), aimed at demonstrating the possibility of designing a competitive and safe fast critical reactor based on the Generation IV Lead Fast Reactor (LFR) concept, the study of the lead-water interaction following an incidental SGTR (Steam Generator Tube Rupture) event is an important issue to address. To simulate such event, an experimental test has been carried out on the LIFUS 5 facility at the ENEA Brasimone Research Centre, in order to assess the physical effects and the possible consequences connected to this kind of interaction. The experiment has been conducted by injecting water at the pressure of 185 bar and with a temperature of 300 °C into a volume of 80 l of Lead Bismuth Eutectic (LBE) kept at atmospheric pressure and at a temperature of 400 °C. The experimental facility has been suitably modified in order to reproduce as close as possible the operating conditions of the ELSY Steam Generator Unit (SGU), in which a free volume of cover gas (argon) is foreseen at the top of the system, with the objective to dampen the pressure waves inside the SGU itself. The experimental test has been supported through a numerical modelling campaign performed at the University of Pisa by means of the SIMMER code within both 2-D (SIMMER III) and 3-D (SIMMER IV) models. Pre-test simulations have been carried out to aid the design of the new facility configuration and to select the test conditions which could better reproduce the behaviour expected for ELSY. In addition, a post-test analysis has also been accomplished, allowing to compare the numerical and experimental results, so as to validate and assess the performance of the code when employed for this kind of applications.

Topics: Water
Commentary by Dr. Valentin Fuster
2010;():719-721. doi:10.1115/ICONE18-30252.

Rapid bioassay methods for the determination of actinides in urine samples at ultra-trace levels are needed for both emergency and routine radiation exposure monitoring. Several rapid actinide urinalysis methods have been recently developed at the AECL Chalk River Laboratories. These methods employ hydrous titanium oxide co-precipitation followed with actinide separation using chromatographic columns; the actinide isotopes are analyzed by alpha spectrometry and inductively coupled plasma mass spectrometry. The chemical recoveries, procedural blanks, and achieved detection limits for these bioassay methods are also presented.

Commentary by Dr. Valentin Fuster
2010;():723-728. doi:10.1115/ICONE18-30259.

The presented work is part of a joint research project performed in cooperation between the Forschungszentrum Dresden and University of Applied Sciences Zittau/Goerlitz. The paper deals with experimental investigations concerning the influence of an impinging jet on sedimented insulation material in the building sump of the reactor containment. One of the main tasks in reactor safety research is the safe heat dissipation from the reactor core and the containment of light-water reactors. In the case of loss of coolant accident (LOCA) the possibility of the entry of insulation material into the containment and the building sump of the containment and into the associated systems to the residual heat exhaust is a serious problem. In the long-term phase of a LOCA the coolant ejected by the pipe leakage falls several meters onto the sump water surface. On this way, the jet is mixed with air. Furthermore, the impinging jet will entrain air bubbles into the building sump of the reactor containment. The entrained air bubbles will rise and have an additional influence on the flow field in the sump and of the sedimented insulation material. Hence, the impinging jet has an influence of the insulation material transport in the sump. To investigate the influence of an impinging jet a special test facility was designed. The test facility “Tank” was build up with acrylic glass. Thereby it is possible to use laser PIV to measure the flow field and high-speed video to analyze the jet-structure in the test facility. With help of this instrumentation the following experiments were performed: • Experiments without air entrainment and different distances between pipe outlet and water surface in the test facility. • Experiments with air entrainment, different distances between pipe outlet and water surface in the test facility and different flow velocities of the impinging jet. • Experiments with a cold impinging jet and hot water in the test facility for a defined distance between pipe outlet and water surface and a defined jet velocity of the jet. The main goal of the experiments is to study the physical phenomena of the impinging jet and provide experimental data for verification of CFD models. The corresponding CFD investigations are reported by Krepper et al. [KRE10].

Commentary by Dr. Valentin Fuster
2010;():729-734. doi:10.1115/ICONE18-30293.

Equipment reliability is a key factor to achieve excellent production capability and maintain high security for nuclear power stations. As the modern nuclear power industry develops in China, much effort is devoted to promoting the research on application of advanced equipment reliability management concepts and technologies. Besides the introduction and convergence of external advanced methods, some domestic innovations on equipment reliability technologies have also been accomplished recent years based on the long-term research and operating experience. These new ideas and technologies are fostered within the rapid development of China nuclear industry in the near decade, and aim at solving the emergent and common issues. As a typical representative of advanced equipment reliability management technology, INPO (Institute of Nuclear Power Operations) AP-913 equipment reliability process is accepted universally in global nuclear power industry and has been recommended by EPRI (Electric Power Research Institute), INPO and other authoritative organizations worldwide. It was originally introduced by many nuclear power stations in China as a reference model to establish their own equipment reliability management systems. As continuous research, attempts and ameliorations conducted, domestic innovations have been performed to develop more comprehensive and adaptive equipment reliability management technology, including integrating many existing reliability technologies, such as RCM, CCM, TCM, PFU, COMIS, and so on. This paper introduces some new research achievements on implementation of INPO AP-913 equipment reliability process at nuclear power stations in China. For each section contained in AP-913 process, the primary plans and suggestions proposed to meet the basic intent of AP-913 is firstly introduced to establish the practicable access for further development. And then, some improvement-step strategies and technologies are also presented to perform more sophisticated and effective management to improve equipment reliability in routine work. Finally, based on the actual conditions and demands of China nuclear power industry nowadays, some comprehension and advice beyond AP-913 process itself, which can also be considered as the localized modification, are presented. Moreover, the instance of AP-913 based equipment reliability management and development in Da Ya Bay Nuclear Power Station is also introduced as illustration. For Chinese nuclear power stations, these innovative attempts not only impelled the sustainable improvement of equipment management, but also exploited a feasible and compatible way to progress in Chinese nuclear industry pattern independently instead of seeking external support overseas.

Commentary by Dr. Valentin Fuster
2010;():735-740. doi:10.1115/ICONE18-30305.

This study focuses on developing a new method to remove uranium from aqueous solution. Chitosan and ferrous ions were used together to remove uranium ions from aqueous solution. Through two-step pH adjustment, the uptake behavior of chitosan and ferrous ions toward uranium in aqueous solution using batch systems were studied in different experimental conditions. The experimental results indicated that the removal of uranium by synergetic effect of chitosan and ferrous ions was more effective than the way of adsorbing uranium ions by chitosan alone. Under the given experimental conditions, the concentration of the residual uranium in the effluent after chitosan and ferrous ions treatment could meet the discharge standard (< 0.05mg·l−1 ) when initial concentration of uranium ions was 10 mg·l−1 or 100 mg·l−1 . The synergetic effect of chitosan and ferrous ions including adsorption, coacervation and coprecipitation, are responsible for the high removal rate of uranium.

Topics: Ions , Uranium
Commentary by Dr. Valentin Fuster
2010;():741-745. doi:10.1115/ICONE18-30334.

Every day thousands of shipments of radioactive materials are transported on international and national routes. These consignments, which are carried by road, rail, sea, air and inland waterway, can range from smoke detectors and cobalt sources for medical uses to reprocessed fuel for use in electricity generation. The transport of radioactive materials worldwide is governed by stringent regulatory regime, which includes standards, codes and regulations that have been continuously revised and updated over the past four decades. The safety measures have been developed to protect the general public, transport workers, emergency response teams and the environment against the risks posed by the cargoes. These risks include the radioactivity itself and other chemical risks that the cargoes may pose, such as toxicity or corrosivity. In addition to the safety regulations, the regulatory regime addresses other, related issues such as physical protection and liability. It was recognized that these standards should provide a uniform, global regime to ensure that all parties apply the same provisions. Since 1961, the UN (United Nations) has published and periodically reviewed and updated the regulations for the safe transport of radioactive material. These regulations are used today by more than 60 countries as the basic for their national regulations. In addition, the main international modal organizations responsible for the safe transport of dangerous goods by road, rail, sea, air and inland waterways have incorporated the relevant parts of the UN regulations into their own instruments. This paper will discuss and outline the principal regulations that apply to the transport of radioactive materials such as the UN regulations for the safe transport of radioactive materials, The UN regime governing the international transport of dangerous goods, the principal modal regulations governing the transport of dangerous goods and achievement of a more harmonized regime. and the international organizations responsible for their development and implementation.

Commentary by Dr. Valentin Fuster
2010;():747-752. doi:10.1115/ICONE18-30340.

Reactor Pit Flooding System (RPF) is adopted under the severe accidents situation in CPR1000+ units. It can move the heat generated from the reactor core via external reactor vessel cooling (ERVC) to keep the integrity of RPV and achieve the in-vessel corium retention (IVR). But if IVR function of RPF is failed, there is Ex-Vessel Steam Explosion (EX-SE) risk. The Ex-Vessel Steam Explosion is analyzed by MC3D software which is for fuel and cooling interaction (FCI). The physical model of CPR1000+ for Steam Explosion is built firstly and then the phenomenon of Ex-Vessel Steam Explosion under typical severe accident is analyzed. The conclusion of this study is that the impulse load of pressure on the cavity wall induced by steam explosion is about 310KPas ∼ 440KPas. Referencing the structure capacity of AP600 containment, if the structural capacity of CPR1000+ containment is equal to AP600, the impulse load of pressure is lower than it. So it could be preliminarily estimated that steam explosion will not threaten the integrality of CPR1000+ containment.

Topics: Explosions , Steam , Vessels
Commentary by Dr. Valentin Fuster
2010;():753-758. doi:10.1115/ICONE18-30353.

Developing the advanced nuclear power plant design to meet the demanding safety, efficiency and environmental goals of electric utilities requires great efforts. In this paper, a design of the safety systems for the large-power PWR units is introduced, which is deemed a optimal combination of the passive safety systems with the active safety systems. The typical design basis accidents are analyzed for this safety system design, such as the Small Break LOCA, SGTR, SLB and Loss of Flow Accidents (LOFA). The results show that the safety systems of the passives combined the actives can mitigate effectively these typical accidents in large-power PWRs. PSA results also show that the passive safety systems contributes to the reduction of the CDF. It is preliminarily concluded that the passive combined active safety system is designed in balance.

Commentary by Dr. Valentin Fuster
2010;():759-768. doi:10.1115/ICONE18-30367.

A sodium-cooled fast reactor (SFR), SFR-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and non-proliferation. SFR-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. Preliminary level 1 PSA models and the results for the metal fuel SFR-600 conceptual design are introduced here.

Commentary by Dr. Valentin Fuster

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