bluuebunny/crossref_metadata_2025_split huggingface dataset:
Fast On-Line Source Term Estimation of Non-Constant Releases in Nuclear Accident Scenario Using Extended Kalman Filter
Authors: Tricard Pierre , Fang Sheng , Wang Jianlong , Li Hong , Qu Jingyuan , Tong Jiejuan , Fang Dong 
Published: February 2014

In nuclear accidents, source estimation methods can be used to assess the amount and status of radioactive materials emission, which is important for nuclear emergency response. Current studies primarily focus on source estimation of a constant release mode, which may not match real nuclear accident scenario.
To solve this problem, a fast on-line source term estimation method for non-constant release mode was proposed in this study. Based on the well-established RIMPUFF dispersion model, the extended Kalman filter algorithm was utilized to perform an online estimation of the total amount and height of the release source simultaneously. To simulate the accidental radioactive release mode, a non-constant function was used as the state description of the release source. Both air dispersion model error and measurement error were considered. To demonstrate the feasibility of the proposed method, twin experiments were performed with release profiles. A sensitivity study was also performed to assess the robustness of the algorithm.
The experiment results demonstrate that the proposed method can provide robust source term estimation for both constant and non-constant release. The estimation accuracy depends on the steepness of the curve in the growth of the release mode and the distribution of simulated monitors.
Hydrogen Permeability of Rubber Gaskets
Authors: Nakano Mitsuyuki , Sasaki Haruko , Watanabe Tomokazu , Maezawa Ayumi 
Published: February 2014

At Fukushima Dai-ichi nuclear power plant zircaloy that was used for cladding tube was exposed to high temperature steam and Hydrogen was produced by rapid oxidation. Essentially produced oxygen had to be sealed in containment vessel and released to the air by venting. However Hydrogen released from containment vessel were stored in reactor building and exploded at the NPP.
From the experience engineers recognized that countermeasure to prevent Hydrogen leak from containment vessel was required. Unfortunately there are few studies relevant to Hydrogen permeability at high temperature and demanded Hydrogen permeability coefficient of sealing material in the case of high temperature exposure.
In this study tests were conducted to predict Hydrogen permeability through rubber gasket at ACCIDENT. Experimental conditions were 1) deteriorated by heat, 2) deteriorated by heat and steam, 3) at high temperature.
Tests were conducted for Silicone rubber, EPDM, NBR and Fluoro-rubber and we found that there was no significant difference between the rubbers Hydrogen permeability coefficient under conditions from 1) to 3) without the case of losing its elasticity.
Currently we can present basic data to predict Hydrogen leak volume from containment vessel to reactor building at ACCIDENT. Based on the data we calculated Hydrogen leak volume from containment vessel through top flange of Fukushima Dai-ichi nuclear power plant in the case that zircaloy was melted completely as example.
Scaling Assessment for the Integral Effect Tests: 2 — Open System Depressurization
Authors: Cai Xiaoyu 
Published: February 2014

In this paper a method is described for using NOTRUMP models to corroborate the Hierarchical Two-Tiered Scaling (H2TS) methodology that has been used for design of the APEX and SPES-2 test facilities. These facilities were built for the Westinghouse Electric Corporation to obtain data on the performance of the passive safety systems of the advanced pressurized water reactors.
Similarity between the prototype system and the scaled test facilities is investigated for the open system depressurization phenomena in the postulated small break loss of coolant accident transient. The objective of this analysis is to provide a basis that an integral test for the passive safety system will provide valid experimental data for the high-ranked phenomena that may occur during the hypothetical SBLOCA transients. In this way, the experiment will capture the phenomena that would be expected to occur in the plant, and hence provide data that can be used to validate computer code.
The Status on Design Collaboration in ITER Chinese Domestic Agency
Authors: Wu Shuqin , Song Yuntao , Xing Chao , Luo Delong 
Published: February 2014

Being an international cooperative engineering project, integration and compatibility of ITER design from numbers of different contributors worldwide is one of the crucial challenges. In order to fully integrate analysis, design, integration, and simulation of mechanical and plant design, and minimize interfaces between tools and methodologies, ITER Project developed a Computer Assisted Design (CAD) collaborative Platform for ITER Organization (IO) and all the parties to implement design collaborations. And also, ITER project standardized design activity via mandatory rules and recommendations for the design lifecycle and tool selections which are documented in the CAD Manual and other related supporting satellite documents.
ITER Chinese Domestic Agency (CN DA) as one of seven parties of ITER Project, undertakes twelve Procurement Packages to design and manufacture several ITER TOKAMAK Machine components and systems, including Magnet Support, Corrective Coils, Magnet Feeders, AC/DC Converters, Diagnostic Systems, Gas Injection System, Glow Discharge Cleaning and so on. Most of them comprise plenty of design work with different design maturities.CN DA accepts ITER design rules and standards, and performs proper deign collaboration schemes according to corresponding PA Design Collaboration Implementation Forms (DCIF).
This paper will present the status of CN DA design collaboration with IO and other DAs. It will include an overall description of design collaboration strategy and infrastructure in CN DA; the introduction of design tasks CN DA involved; and the associated organization and implementation of design collaboration schemes in specific PAs.
Hollow-Fiber Membrane Based Micro-Filtration Technology for the Treatment of Low Level Nuclear Laundry Waste and Floor Drain Water
Authors: Zhang Lei , Farooq Khalid , Liao Zhiyuan , Pang Kewen 
Published: February 2014

A commonly used treatment for laundry waste and floor drain water in a nuclear power plant (NPP) in China involves filtration, followed by passing through a mixed ion exchange bed or evaporator. Due to the presence of foam and suspended solid in the laundry waste, the treatment systems require frequent maintenance and have a short filter service life. Pall ARIA™ platform, a hollow-fiber membrane based Micro-Filtration (MF) technology, was recently used at a German nuclear power plant to upgrade the low level radioactive laundry waste treatment system. The MF system with 0.1 μm average pore size and totally automated backwashable design was able to successfully concentrate the low level radioactive waste before it was sent to the evaporator.
An 18-month long pilot test of the MF system, with a flow capacity of 3–6 m3/hr, using 3 modules, was performed at the 1300 MW, Pressurized Water Reactor (PWR) nuclear power plant, at Phillipsburg, Germany, to substitute its legacy centrifuge separators in waste water treatment system. During the long pilot testing, the MF technology amply demonstrated its capability for consistent high particulate removal efficiency, safe & reliable operation and high availability with average recovery rate of 95%. An average activity of 17900 Bq/m3 was reported in the filtrate. The 20X concentrated stream from the MF system was processed by the evaporator. The paper discusses the current liquid radwaste treatment practices in China, and the details of 18-month long trial of the MF system at Phillipsburg NPP.
Study on Neutronic Characteristics and Neutronic-Thermohydraulic Coupling Behavior of SCWR With Thorium-Based Fuel
Authors: Liu Shichang , Cai Jiejin 
Published: February 2014

A novel thorium-based fuel assembly for SCWR has been introduced in this paper, taking advantage of the neutron spectrum of SCWR. Single channel thermohydraulic model was integrated with calculation codes “dragon” and “donjon”, developing a computation platform for neutronics assessment and thermohydraulic analysis of the proposed fuel assembly.
Results show that the thorium-based fuel assembly meets design rules of SCWR. Meanwhile, it has considerable advantages on both efficient fuel utilization and lower minor actinide generation.
Reliability Analysis for Power Supply System in a Typical Reprocessing Facility Based on GO Methodology
Authors: Wang Renze , Feng Zongyang , Li Pengbo 
Published: February 2014

Since the nuclear energy industry in China has developed rapidly, the demand for spent fuel reprocessing is becoming more and more distinct. In a reprocessing facility, power supply system (PSS) is important for operation-stability and safety-guarantee of the whole facility. Therefore, the reliability of the PSS deserves in-depth investigation and analysis.
Fault tree (FT) methodology is the most accustomed methodology for system-reliability analysis and it has been standardized and applied widely. GO methodology is another effective methodology for system-reliability analysis but it has not been applied widely.
In this work, GO methodology was applied to analyze the reliability of the PSS in a typical reprocessing facility. First, for modeling expediently, based on the fact that tie breakers are set in the system, tie breaker operator was defined by myself. Then, GO methodology modeling and quantitative analysis were performed, minimal cut sets (MCSs) and average unavailability of the system were obtained. Finally, parallel analysis between GO methodology and fault tree methodology was also performed. The results from the two methodologies are completely coincident.
The results of this work show the following two points:
1. Setup of tie breakers in the PSS of the reprocessing facility is rational and necessary;
2. For reliability analysis of the PSS of the reprocessing facility, parallel with fault tree methodology, GO methodology has two distinct advantages: Its modeling is much easier and its chart is much more succinct.
Fangjiashan Nuclear Power Plant Unit1 Fuel Handling and Storage System (PMC) V&V Overview
Authors: Ma Gang 
Published: February 2014

The product of Fangjiashan Nuclear Power Plant (NPP) Unit1 Fuel Handling and Storage (PMC) System is the first Chinese PMC system, with its own intellectual property rights. During the development and design lifecycle for the product, China Nuclear Control System Engineering Co., Ltd (CNCS) was responsible for Verification and Validation (V&V) processes of the PMC system, taking a significant role for ensuring the reliability and safety of the software in PMC system. The PMC V&V project is also the first ever independent V&V project of the China nuclear industry.
V&V is a technical discipline of systems engineering. The purpose of V&V is to help the development team build quality into the system during the life cycle. The V&V processes determine whether the product developed satisfies its intended use and user’s needs. V&V provides an objective assessment of products and processes throughout the life cycle. This assessment demonstrates whether the requirements are correct, complete, accurate, consistent, and testable.
CNCS V&V PMC project team performed PMC V&V activities in compliance with IEEE Std 1012-1998, which is endorsed by NRC Regulatory Guide 1.168-2004. After critical analyses by V&V team, the Software Integrity Level (SIL) of PMC is SIL3, so associated V&V activities must conform to the minimum tasks list in IEEE Std 1012-1998 Table 1. After 2 years of hard working, Fangjiashan NPP Unit1 PMC system V&V activities have been completed. The whole activities and methodology was passed the final evaluation by the customer and was highly compliment by the international V&V expert committee at year of 2012.
This paper discusses how to establish V&V organization and infrastructure, and then how to perform V&V activities in high quality, and the specific new and best practice methods, approach and tools for executing V&V tasks. Because the PMC V&V project is the first ever independent V&V project of the China nuclear industry, the methods, approach and tools which was executed in the whole lifecycle and depicted in this paper could help to guide and be referred by the future V&V activities and projects.
The Study of Cavitation Characteristics of a Heavy Liquid-Metal Coolant
Authors: Bokov P. A. , Beznosov A. V. , Lvov A. V. , Iarmonov M. V. , Bokova T. A. 
Published: February 2014

Adequate design solution and maintenance of circuits with fast reactors cooled by lead and lead-bismuth coolants require taking into account the peculiarities of hydrodynamics of these coolant flows. The design pressure of saturated vapors of lead and its alloys at temperatures of 400–550 °C is 10 −18−10 −10 atm, which is significantly less than that of sodium or water. Processes of traditional cavitation cannot occur in a flow of heavy liquid-metal coolants because of their specific character. The main circulation pumps of reactor circuits are one of their basic elements. In fact, the flow-type parts of these pumps and other vane pumps operating in lead and its alloys cannot be calculated by traditional methods in terms of cavitation characteristics; appropriate calculation formulas are not currently available. To study cavitation processes in a heavy liquid-metal coolant flow, the authors have carried out the experiments aimed at:
- determining the conditions of disconnection of liquid lead and lead-bismuth eutectic column;
- determining the cavitation characteristics of a centrifugal pump transferring lead at a temperature of 500 °C;
- studying the characteristics of ejector (Venturi nozzle) in a liquid metal;
- studying the cavitation erosion effect of the lead coolant on impeller vanes of an axial-flow pump in a limited volume in the FT-4-A stand at a lead flow rate of up to 1200 t/h;
- studying the cavitation characteristics of an axial-flow pump in the FT-4 stand at shaft speeds of 13.34–25 Hz.
These studies are performed with the lead coolant at temperatures of 450°−550°C, oxygen in lead from 10−4−10−5 to 100, flow rates from 20 to 1800 t/h, which corresponds to velocities of the lead coolant flow from 1.0 to 26 m/s.
The experiments have shown that as distinct from water, traditional cavitation processes in a heavy liquid-metal coolant (HLMC) flow are not recorded. The probable cavitation mechanism is gas cavitation. The allowance for the specific character of hydrodynamics of HLMC flows is necessary for adequate design engineering and maintenance of some elements of the reactor circuit.
Reactivity Hold-Down Technique for a Soluble Boron Free PWR Using TRISO Particle Fuel
Authors: Dai Xiang , Cao Xinrong 
Published: February 2014

TRISO coated particle, developed for HTGR initially, has advantages of nuclear proliferation-resistance and fuel integrity against the release of fission products. In this paper, a 350MWt small sized PWR core design utilizing TRISO fuel concept is presented. TRISO particles are dispersed in graphite matrix to form the fuel compact, and then the fuel compact is clad by Zircaloy-4 cladding to form a fuel rod. The graphite matrix increases thermal conductivity of fuel compact, so that the fuel average temperature would be well below conventional PWRs’.
In order to simplify reactor design, operation and maintenance, soluble boron free concept while operation is introduced. The emphasis of the study is put on the reactivity hold-down technique for the 350MWt PWR core. Excess reactivity is suppressed through a combination of Pu-240 adding with Gd2O3 loading. Pu-240 is added into UO2 fuel kernel of some assemblies, and Gd2O3 rods are loaded in other assemblies. The non-fissile plutonium isotope Pu-240 has a considerably high thermal neutron capture cross section compared to U-238, so that the Pu-240 added fuel can greatly suppress excess reactivity over burnup. Besides, reactor core life would be extended by adding proper amount of Pu-240 for its converting into Pu-241 which is a fissile isotope. Combining Pu-240 adding with Gd2O3 loading, the designed core reaches an average core burnup of approximately 58GWD/t, as well as a core life of nearly 6EFPY.
The Study of Qinshan III PHWR Thermosyphoning Capability
Authors: Sheng Xin 
Published: February 2014

Thermosyphoning capability, also known as the natural circulation is an important nuclear safety indicator in NPP. In the absence of any forced circulation of the NPP primary system, the cooling of the fuel to remove the decay heat can be provided by thermosyphoning to the steam generator (SG).
Based on the theory of natural circulation and simulated test, this paper will further identify thermosyphoning capability of Qinshan III PHWR (CANDU6). Analysis shows that, the ultimate reactor heat sinks are SGs and related safety valves in case of loss main pumps. Thermosyphoning capability was confirmed as effective and acceptable.
Furthermore, loss of Power accident happened during Qinshan III Unit#1 normal operating condition. We will investigate the impact of reactor core cooling and validate thermosyphoning as the capability of Qinshan III PHWR heat sink on accident.
Finally, this paper will give some suggestions to NPP operator’s actions issues with reactor core temperature difference, reactor inlet headers sub-cooling margin, reactor outlet header temperatures, and others. Meanwhile, paper also discuss on the improvement of CANDU future design.
About Harmonization of Deterministic and Probabilistic Approaches to the NPPs Safety Substantiation
Authors: Ponomarenko Grigory L. 
Published: February 2014

Severe accident on the Fukushima NPP lowered the public trust for nuclear power engineering and must lead to the revision of the NPPs safety substantiation in the trend of the more strict regulatory body’s requirements. This relates to the deterministic (DSA) and probabilistic (PSA) safety analysis. In the author’s opinion, the most effective increase in the safety should be reached due to convergence or harmonization of deterministic and probabilistic methods. In particular, at fulfillment of DBAs in a number of cases one should consciously move in the direction of more realistic analysis of the actually BDBAs emergencies (in the traditional understanding) with the imposition of more than one Initiating Event (IE) or more than one Single Failure. In turn, when the PSA is fulfilled one should move from the realistic analysis without taking into account uncertainties in the direction of their reasonably conservative calculation. This work presents the practical realization of harmonious interaction of methods DSA and PSA based on the example of the estimation of the minimally sufficient CRs CPS for WWERs. A new conceptual approach to the definition of a minimum required quantity of CRs as part of the Spatial Effects Methodology is presented here and in papers [15, 16]. This approach combines the neutron-physical, thermal-hydraulic and probabilistic aspects, using DBC’s Fuel Criteria generally accepted for CD realization. Analysis is performed with the up-to-date coupled code KORSAR/GP [1–3] on example of the mode MSLB, which has the most dependence on EP with respect to any other accident.
This paper presents an approach that allows:
- to assess the EP success criteria with multiple stuck CRs for the purposes of PSA. In other words — assess the maximum number of stuck CRs when CD does not yet occur;
- justify the low (P_EPfail ≤ 10−5, in accordance with Russian State Standard [4]) probability of EP failure on demand, for different amounts of CRs in a range from 49 till 121 pcs. Furthermore, this probability allows substantiate the low partial CD probability P_CDpartial < ∼10−10 for concrete scenarios with multiple stuck CRs as a negligible contribution in the total CD probability which should be P_CDtotal < 10−5;
- a joint optimization for positioning and quantities of CRs and power monitoring detectors (SPND) in central cell of FAs. Indeed, if the total number of CRs in WWER-1000 becomes more than 79–85 pcs, it creates some inconvenience, particularly the need to shift the power monitoring detectors from the central cell of fuel assembly into another noncentral cell.
The Thermal Balance Calculation of AP1000 Turbine Roll-Up by Non-Nuclear Steam
Authors: Wang Shunda , Li Xiuli 
Published: February 2014

The initial turbine non-nuclear steam roll-up is an important test during nuclear power plant startup. The purpose of the test is to discover and solve turbine problem before reactor achieve critical. The benefit of this test is milestone may be shorter and save time. Use AP1000 equipments design parameters to calculate the basement of this test, by the way, we can find a safe method to startup.
In this test, the heat is produced by reactor coolant pumps and PZR heaters. During roll-up, to avoid critical zones of turbine and reduce thermal stress, we hold on some speed points. At low speed period we can make up heat to make sure we can achieve the purpose of this test. So the calorimetric balance between primary side and second side is very important.
Non-nuclear steam has many benefits for nuclear power plant startup. First, the secondary loop completes build before the primary system, but as one of the most important equipments, turbine must be examined to make sure it can roll-up and operation without no disease, so before primary system equipments installation finished we can save lots of time to solve turbine problems; the second one, during this duration, all of equipments of primary system can perform their function in the first time, we also make sure system function can perform correctly and timely. So, as a big picture, the milestone of startup can be shorter by replace nuclear steam rolling-up the turbine.
Preliminary Design for the Back Plate of Sub-Module of CN HCCB TBM
Authors: Zhao Zhou , Hu Gang , Wang Qijie , Cao Qixiang , Feng Kaiming , Zhang Guoshu 
Published: February 2014

The latest design of CN HCCB TBM is 1×4 configuration, which is consisted of 4 sub-modules and a big back plate. For the sub-module, the back plate is the most key and complicated part that is a multi pipes and plates structure, the space between the plates are to form the functional cavities that are used to distribute and collect the coolant and purge gas, the cavities are connected by the pipes. The cavity 1 is used to distribute the coolant (from the big back plate) into the branch cooling channels of caps and rib, the cavity 2 is to collect the coolant from the branch channels of caps and rib and to distribute it into lots of branch channels of cooling plates, the cavity 3 is to mix and collect the coolant from the branch channels of cooling plates that will flow into the back plate of TBM finally.
The preliminary analysis results show that, the mass flow rate is between 2.20×10−2 and 2.33×10−2 kg/s for the coolant flowing into branch channels of caps, and it is between 1.82×10−3 kg/s and 1.84×10−3kg/s for that of rib, which indicates the cavity 1 will lead a good uniformity of mass flow rate; although the mass flow rate is between in a large range from 5.29×10−4 kg/s to 9.19×10−4kg/s for that of cooling plates, the thermal hydraulic analysis of sub-module shows that the minimum value (5.29 × 10−4 kg/s) can also meet the temperature requirements of materials.