Comparison of RELAP5 calculations of VVER-1000 coolant transient benchmark phase 1 at different power

This paper provides comparisons between experimental data of “MCP switching on when the other three MCPs are in operation” and RELAP5 calculations with different initial levels of the reactor power 29.45% and 27.47% from the nominal.

The reference power plant for this analysis is Unit 6 at the Kozloduy nuclear power plant (NPP) site. RELAP5/MOD3.2 computer code has been used to simulate the investigated transient. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation.

This investigation was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

温排水效应对秦山核电厂1号机组的影响及应对措施

随着秦山核电基地全面建成,由于杭州湾海水潮流及电厂取、排水口位置等因素所致,方家山核电厂1号、2号机组产生的温排水对秦山核电厂1号机组(以下称秦一厂机组)的运行造成了较大影响.导致机组在夏季期间被迫频繁调整出力,同时造成部分系统和设备的运行参数偏离正常范围,给电厂造成了较大的运行负担.本文介绍了温排水效应对秦一厂机组运...     

High power accelerators in fundamental physics and nuclear technology

The trends of the high power accelerators development are outlined. The natural examples of their applications in nuclear physics and technology are discussed: muon physics, physics of rare decays, intense 14 MeV — neutron source based on muon catalyzed fusion (INS — MCF) and accelerator driven system (ADS) for nuclear waste incineration. The accelerator with power ˜ 10 MW and particle energy ˜ 1 GeV/nucl is considered as the best candidate for these purposes.     

红沿河核电厂2号机组首循环寿期中启动物理试验结果分析

因电网调峰能力不足,红沿河核电厂2号机组首循环运行过程中,于2014年1~3月进行了卸料不换料停机检修,再启动阶段进行了临界、零功率和升功率物理试验,验证了循环寿期中反应堆重要堆芯设计参数。该文叙述了红沿河2号机组反应堆首循环寿期中卸料不换料后启动物理试验理论计算与现场试验,验证了寿期中启动物理试验理论计算值与实测结果的符合程度,分析了反应堆相关参数在寿期初与寿期中随燃耗变化特性。试验结果表明,理论预计值与实测结果符合良好,偏差满足验收准则。     

Calculation of risk-based detection limits for radionuclides in the liquid effluents from Korean nuclear power plants

In order to review if present detection limits of radionuclides in liquid effluent from nuclear power plants are effective enough to warrant compliance with regulatory discharge limits, a risk-based approach is developed to derive a new detection limit for each radionuclide based on radiological criteria. Equations and adjustment factors are also proposed to discriminate the validity of the detection limits for multiple radionuclides in the liquid effluent with or without consideration of the nuclide composition. From case studies to three nuclear power plants in Korea with actual operation data from 2006 to 2015, the present detection limits have turned out to be effective for Hanul Unit 1 but may not be sensitive enough for Kori Unit 1 (8 out of 14 radionuclides) and Wolsong Unit 1 (9 out of 42 radionuclides). However, it is shown that the present detection limits for the latter two nuclear power plants can be justified, if credit is given to the radionuclide composition. Otherwise, consideration should be given to adjustment of the present detection limits. The risk-based approach of this study can be used to determine the validity of established detection limits of a specific nuclear power plant.     

Condition telemonitoring and diagnosis of power plants using web technology

The monitoring and diagnostic systems currently installed in power plants generally supply information for control room displays and for on-site personnel. Telemonitoring is also frequently used. In this case, relevant diagnostic data are transmitted remotely to a special laboratory for analysis using highly specialized equipment and software.

The appearance of the terms “Monitoring” and “Diagnosis” alongside the term “Web Technology” in the title of this paper does not mean that remote access to diagnostic systems over the Internet is being presented here as a simple extension of the existing situation.

Condition telemonitoring and diagnosis based on Web technology is a new departure in diagnostic system design philosophy. It is the technology used to integrate diagnostic systems into a customer's IT infrastructure (Intranet or Internet).

Siemens has started to use Web-based condition telemonitoring and diagnosis in some power plants (nuclear and fossil-fueled) to provide a global source of specialist support.     

Experiments on the load following behaviour of PWR fuel rods

KWU had studied the effects of load following operation on fuel performance from the beginning of commercial operation of nuclear power plants: The first power cycling experiments were started in 1970 in the nuclear power plant Obrigheim (KWO) and in the High Flux Reactor (HFR) Petten. These power cycling tests performed at various power levels and burnups of up to 25 GWd/t(U) showed that the fuel rod cycling performance compares well with the performance of fuel rods operated under essentially constant load at comparable power levels.Two additional cycling tests as described in this paper were performed in the HFR Petten with preirradiated PWR fuel rods having burnups of up to 40 GWd/t(U). These experiments comprised up to 60 cycles between 250/360 W/cm and 215/320 W/cm with 10% power overshoot (400, 370 W/cm) after each cycle. Also, these experiments ended up with sound fuel rods. Moreover, detailed investigations before and between power cycles and after experiment termination showed clearly that the fuel performance corresponds to a single ramp to peak power and that the cycling effects are indeed very small. This confirmed earlier findings that due to crack reversal in the UO₂ the cyclic dimensional changes mainly occur in the UO₂ itself. Altogether the experiments show that power cycling does not lead to fuel rod failures, which is also confirmed by successful load follow operation in commercial power plants.     

Safety aspects and environmental impact of an inertial confinement fusion reactor

Releases into the environment of radioactive materials contained in heavy ion fusion (HIF) reactor plants must be prevented by similar safety design concepts as they are applied to present fission converter (e.g. LWR's) and breeder reactors (LMFBR's). This study identifies significant safety aspects of inertial confinement fusion power plant concepts and relates them to the more familiar basis of knowledge about the safety and the hazards of other advanced nuclear power reactor systems such as the LMFBR. Assessments of doses to be expected after the release of tritium from HIF reactor plants — normally and accidentally — are performed and compared with dose limits and with doses resulting from facilities of the fission fuel cycle. Needs for safety related research and development specifically for inertial confinement fusion as well as for the modelling of the various exposure pathways due to released tritium are pointed out.     

Late-phase melt conditions affecting the potential for in-vessel retention in high power reactors

If cooling is inadequate during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 accident. In such a case, concerns about containment failure and associated risks can be eliminated if it is possible to ensure that the lower head remains intact so that relocated core materials are retained within the vessel. Accordingly, in-vessel retention (IVR) of core melt as a key severe accident management strategy has been adopted by some operating nuclear power plants and planned for some advanced light water reactors. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements can provide sufficient heat removal to assure IVR for high power reactors (i.e., reactors with power levels up to 1500 MWe). Consequently, a joint United States/Korean International Nuclear Energy Research Initiative (I-NERI) has been launched to develop recommendations to improve the margin of success for in-vessel retention in high power reactors. This program is initially focussed on the Korean Advanced Power Reactor—1400 MWe (APR1400) design. However, recommendations will be developed that can be applied to a wide range of existing and advanced reactor designs. The recommendations will focus on modifications to enhance ERVC and modifications to enhance in-vessel debris coolability. In this paper, late-phase melt conditions affecting the potential for IVR of core melt in the APR1400 were established as a basis for developing the I-NERI recommendations. The selection of ‘bounding’ reactor accidents, simulation of those accidents using the SCDAP/RELAP5-3D^© code, and resulting late-phase melt conditions are presented. Results from this effort indicate that bounding late-phase melt conditions could include large melt masses (>120,000 kg) relocating at high temperatures (3400 K). Estimated lower head heat fluxes associated with this melt could exceed the maximum critical heat flux, indicating additional measures such as the use of a core catcher and/or modifications to enhance external reactor vessel cooling may be necessary to ensure in-vessel retention of core melt.     

Power flattening options for the ENHS (encapsulated nuclear heat source) core

The feasibility of power flattening while maintaining a nearly constant k_eff over the core life is assessed for the Encapsulated Nuclear Heat Source (ENHS). A couple of approaches are considered — using different fuel dimensions and using different enrichment levels across the core. Three new cores with flattened power distribution are successfully designed: Design-I uses different fuel rod diameters but uniform fuel composition; Design-II uses different fuel enrichment in the radial direction but uniform fuel rod dimensions; Design-III is similar to Design-II but uses enrichment splitting also in the axial direction. Relative to the reference ENHS core, the BOL peak-to-average channel power ratio is reduced from 1.50 to 1.15, 1.22 and 1.15 and the average discharge burnup increases by 8.5%, 27.9% and 41.2% for, respectively, Design-I, -II and -III. The corresponding burnup reactivity swings over 20 years of full power operation are 0.37%, 0.52% and 0.60% relative to 0.22% of the reference design. Design-II and -III have a negative coolant expansion reactivity defect while in the reference design this defect is positive. The radial power flattening increases the reactivity worth of the peripheral absorbers of the three new designs while the central absorber reactivity worth is reduced but their sum is nearly maintained. The newly designed cores have slightly more positive coolant void reactivity worth than the reference ENHS core.     

RELAP5/MOD3.2 investigation of a VVER-1000 MCP switching on Benchmark problem

This paper provides comparisons between experimental data of Kozloduy NPP “MCP switching on when the other three MCP are in operation”, with Relap5 calculations. The investigated thermal-hydraulic driven transient is characterized by spatially dependant non-symmetric processes. RELAP5/MOD3.2 computer code has been used to simulate the investigated transient. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. The event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which leads to insertion of positive reactivity due to the modeled feedback mechanisms. The main purpose of this investigation was to improve the discrepancy between the calculations and the plant data. The sensitivity calculation investigates the mixing in reactor vessel and influence of heat structure on the hot legs temperature. The areas of improvements to the Relap5 model are:

• •The non-symmetrical mixing in downcomer and reactor vessel annular exit.
• •The influence of heat structure temperature on the time delay for equipments measurements.
• •Investigation of pressurizer water level – using the hot legs temperature correction.

The RELAP5/MOD3.2 model of Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios have been developed and validated in the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS) Sofia, and Kozloduy NPP. The model provides a significant analytical capability for the specialists working in the field of NPP safety.This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of the VVER-1000 against the experimental transient data received from the Kozloduy NPP Unit 6. The comparisons indicate good agreement between the RELAP5 results and the experimental data. The sensitivity investigation improves the discrepancy between the calculation and the plant data.This investigation was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Plant life management activities of LWR plants in Japan

Plant life management activities of Japanese LWR plants have been conducted since the early 1990s by the utilities and MITI (Ministry of International Trade and Industry) cooperatively. In Japan, where the regulatory practices are different from those in the US, there is neither law nor regulation that prescribes a licensed plant life for nuclear power plants. When an annual inspection is completed without any problem, the next cycle of operation would be permitted and this cycle can be repeated. However, it is generally known that mechanical components and structures deteriorate as they get older. So, we consider it very important to evaluate the long-term integrity of major systems, structures and components of old nuclear power plants. Japanese plant life management study consists of two parts. Both parts of the study were carried out confirming the integrity for the long-term operation of the three oldest Japanese LWR plants: Tsuruga Power Station Unit No.1 (BWR), Mihama Power Station Unit No.1 (PWR) and Fukushima Dai-ichi Nuclear Power Station Unit No.1 (BWR). The Part 1 study was conducted for the purpose of obtaining an outlook for long-term safety operation and was completed in 1996. The Part 2 study was conducted ensuring the plant integrity for the long-term operation in terms of, not only safety, but also reliability. The results of the Part 2 study were made public in February, 1999. Then, the recommended maintenance items were to be added to the existing maintenance programs of the three LWR plants.     

The encapsulated nuclear heat source reactor for low-waste proliferation-resistant nuclear energy

The encapsulated nuclear heat source (ENHS) is a new Pb-Bi cooled modular reactor concept that features a combination of the following useful features that may make nuclear energy more attractive: (1) 20 years of full power operation without refueling. (2) Nearly constant fissile fuel contents and k_eff. (3) No on-site refueling and fueling hardware. (4) The ENHS modules are factory manufactured and transported already fueled to the site. (5) No access to neutrons. (6) No mechanical connections between the ENHS module and the energy conversion plant (The ENHS module has the function of a nuclear battery — with 20 years of full power operation at 125 MW_th). (7) At end of life, the ENHS module serves as a spent fuel storage cask and, later, as a spent fuel shipping cask. That is, the fuel is locked inside the ENHS from “cradle to grave”. (8) 100% natural circulation resulting in passive load following capability and autonomous control. This combination of features offers a highly safe nuclear energy system that is characterized by low waste, high proliferation resistance and high uranium utilization. The low waste and high uranium ore utilization are achieved by recycling the Pu and MA many times using a proliferation-resistant dry process; only fission products are to be extracted between cycles. Spent LWR fuel can provide for the HM make-up. The high level of proliferation resistance is obtained by restricting access to the fuel and neutrons and by eliminating the economic incentive of the client country to invest in sensitive technologies or infrastructure that can be used for clandestine production of strategic nuclear materials.     

Experience and plans with PCI protection in KWU LWR plants

The experience with early operational guidelines to eliminate PCI failures in LWR fuel is briefly discussed. For future applications a more detailed PCI surveillance and protection model is proposed. It is designed for the use in administrative guidelines as well as in automatic power density surveillance and limitation systems. Important model parameters are directly derivable form experimental data by using the ‘RSST Approach’ that—in order to prevent PCI failures—at least one out of four ‘predictors’ (i.e. power range, power step, speed of power increase, or time at transient overpower) has to be below a critical value at any operating time. An algorithm is provided for defining and monitoring an adequate ‘conditioned power’ as a reference power for acceptable power ramps.The operational consequences of the new surveillance model are discussed and show, that expected power losses are similar or less than from early guidelines.Finally, relevant features of mechanistic PCI fuel rod models are discussed. Some of the PCI failure prediction models, which have been proposed in the literature, seem to be unnecessarily conservative and—if strictly applied to LWR core surveillance—lead to unduely severe restrictions in plant operation.     

Seismic instrumentation for nuclear power plants: An interpretative review of current practice and the related standard in Germany

The German nuclear safety standard KTA 2201: “Design of nuclear power plants against seismic events”, consists of the following parts: 1. basic principles; 2. characteristics of seismic excitation; 3. design of structural components; 4. design of mechanical and electrical parts; 5. seismic instrumentation; and 6. measures subsequent to earthquakes.While Part 1 was published in June 1975, Part 5 was approved by the Nuclear Safety Standards Commission — Kerntechnischer Ausschuss (KTA) — in June 1977. The other parts are still under development. The requirements of the safety standard KTA 2201.5 deal with

1. (a) number of location (number and location of acceleration recording systems for different sites, single-block plants and multi-block plants);

2. (b) characteristics of instruments (readiness and operation of instruments, margin or errors, dynamic and operation characteristics, duration of records, seismic switch);

3. (c) triggering and information (loss of electric power, start of the acceleration recording systems, threshold of acceleration for triggers and seismic switches, optical and acoustic information); and

4. (d) documentation (results of recordings, inspection and tests).

The purpose of this paper is to present some detailed requirements of the safety standard KTA 2201.5, with its philosophy, and compare these with corresponding requirements in the US. It will be shown that with relatively few instruments, which are very reliable in operation and in triggering, an optimum of data may be available after an earthquake.     

APA-H程序应用于田湾核电站1号机组堆芯物理参数计算验证

为验证APA-H(ALPHA-H/PHOENIX-H/ANC-H)程序系统应用于田湾核电站1号机组(VVER1000)堆芯物理参数计算的可行性,针对田湾核电站1号机组第6~9燃料循环的燃料管理开展计算研究。对临界硼酸浓度、组件相对功率分布以及启动物理试验进行模拟计算,并与试验测量数据进行比对。结果表明,计算值与试验测量数据符合良好,满足验收准则。APA-H程序系统可用于田湾核电站1号机组的堆芯物理参数计算。     

Threshold oxygen levels in sodium necessary for the formation of NaCrO2 in sodium-steel systems

A knowledge of the threshold oxygen level in liquid sodium necessary for the formation of NaCrO₂ in sodium-steel systems is useful in the operation of fast breeder reactors. There is considerable discrepancy in the data reported in the literature. In order to resolve this, the problem was approached from two sides. Direct measurement of oxygen potential in the Na(l)-Cr(s)-NaCrO₂(s) phase field using the galvanic cell In, In₂O₃/YDT/Na, Cr, NaCrO₂ yielded: _o2 = −800847 + 147.85 T J/mol O₂ (657–825 K). Knudsen cell-mass spectrometric measurements were carried out in the phase field NaCrO₂(s)-Cr₂O₃(s)-Cr(s) to obtain the Gibbs energy of formation of NaCrO₂ as: ΔG^o_f,T(NaCrO₂) = −870773 + 193.171 T J/mol (825–1025 K). The threshold oxygen levels deduced from G^o_f,T (NaCrO₂) data were an order of magnitude lower than the directly measured values. The difference between the two sets of data as well as differing experimental observations from operating liquid sodium systems are explained on the basis of the influence of dissolved carbon.     

利用核电厂模拟器进行操纵员可靠性研究

为提高核电厂运动可靠性,需要对核电厂操纵员进行可靠性研究。本文结合我国核电厂操纵员可靠性研究的状况,并参考国际上流行的核电厂操纵员可靠性研究方法,利用两参数威布尔分布的理论在核电厂模拟器上对我国核电厂操纵员进行认知可靠性研究,将该方法得到的结果与其他理论模型的结果进行了比较和讨论,得到了一致的认知。本文的研究方法可为真实核电厂运行提供参考。