The Development of SRCM and the Application at Home and Abroad

It is well-known that RCM is an advanced and effective maintenance strategy in practice. With the development of the automation and mechanization in modern industry, RCM method turns to be complex and consumes more resources in real production. However, the development and application of the Streamline RCM (SRCM) has injected new vitality for the new situation, especially in the nuclear power plants. This paper firstly introduces the background, the characteristics of the SRCM and the differences from RCM, and then shows the process in detail as well as the application status of the SRCM in country and abroad. It is proved that SRCM is a unique available method which saves the time and resource consumed, ensuring the integrity and correctness of the classical RCM. Finally, the weak points and the prospect are reviewed and prospected.

The Energy-Saving Diagnosis of PWR Nuclear Power Station Based on the Thermo-Economic Analysis Model

Exergy analysis model of PWR nuclear power station is developed in which signal flowing graph theory is introduced to set up the relation equations between input exergy flow and output exergy flow. Then, combining with resource distribution between different components, thermo-economic analysis model is obtained by setting up unit thermo-economic cost equations of different components with productive structure graph. Taking Daya Bay as an example, exergy analysis and thermal-economic analysis are put forward with detailed distribution of exergy and investment cost. Finally, aimed at energy-saving, static diagnosis is performed in two levels: energy conservation and cost reduction, and on this basis dynamic diagnosis is developed through sensitivity analysis considering different influence factors such as main steam temperature, fuel price, construction capital investment, post treatment cost and so on. The introduction of signal flow graph theory and thermal-economic structure theory is helpful to do performance estimation with high speed and good accuracy. It provides a new way for rapid optimization and offers an effective theoretical method for energy-saving of PWR nuclear power station including advanced reactor such as AP1000.

Progressive Evolution of the Core of the Fast Breeder Test Reactor

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.

Performance Evaluation of Japanese Nuclear Power Plant Based on Open Data and Information (3)

The excessive maintenance of the nuclear power plants (NPPs) may cause the early (infant) failure in Japan. An easy analysis; the Weibull analysis was applied to the evaluation of the failure mode. The Weibull analysis needs the hazard data. The maintenance information of the equipment which caused plant shutdown was required for the hazard calculation. However, maintenance information of the equipment was not open. Therefore, all equipment was assumed to be maintained during every shutdown. This assumption was based on renewal process. However, a repair after unplanned shutdown of NPP is generally a restoration of only failed function without system overhaul. The system must be considered to age continuously. The system was not renewed. The operation data must be regarded as one continuous data before and after unplanned shutdown. An improvement of the Weibull analysis was required for NPPs. The model of the Weibull analysis was investigated. The competitive model in which shutdown caused by other than focused equipment/cause may be supposed to be continuous data could not be applied for a comprehensive analysis. Furthermore, the calculation method of the Weibull analysis was investigated. The calculation method of the hazard was viewed. A denominator of the hazard is the number of data which is cut for every continuous data by renewal process. However, multiple considerations of operation periods before unplanned shutdowns might cause underestimation of the failure rate in case of restoration process. Therefore, a dominator of the hazard was not supposed to be the number of data but the number of survived equipments (plants) at each time according to the definition of the hazard. This improved method is for the restoration process. The performance of Japanese NPPs was evaluated by improved method. The failure modes of Japanese NPPs were early failure modes. Moreover, performances of U.S. NPPs was tried to be evaluated by improved method. Operation data was collected from “NRC Power Reactor Status Reports”. However, many “maintenance outage”s which are the shutdowns of unknown origin were found. Therefore, DOE information was supplemented to investigate the “maintenance outage”. Failure modes of U.S. NPPs were the early failure modes, and failure rates were larger than Japanese NPPs.

Integrated Power Recovery Using Markov Modeling

The solution to a Markov chain modeling electric power supply to critical equipment in a typical 4-loop pressurized water reactor following a Loss of offsite power event is compared with a convolution method. The standard “convolution integral” approach is described, and an alternative methodology based on a Markov model is illustrated.

The Development of a Generic Qualified Digital Platform

The nuclear industry and research institutes in Taiwan are conducting a joint effort project to establish a self-reliant nuclear Instrumentation and Control (I&C) system design and fabrication capabilities in Taiwan. The purposes of this project, as called Taiwan’s Nuclear I&C System (TaiNICS), are planned to support digital upgrade of the existing nuclear power plants and the new nuclear installations in Taiwan. The project will be a long term pursuit of several task branches, including establishment of a generic qualified digital platform, qualification and certification processes, nuclear I&C systems design, safety analyses for software common cause failure, licensing, and collaboration. The short term goal of this project is to submit the License Topical Report (LTR) of a generic digital platform for the review of Taiwan’s regulatory body in 2013.

Life Extension of the Fast Breeder Test Reactor

The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam, India. The reactor went critical in October, 1985 with a core of 23 unique high Plutonium carbide fuel subassemblies and the reactor power was rated for 10.5MWt with peak linear heat rating of fuel at 320W/cm. The extension of the target burn-up of this fuel based on Post Irradiation Examination at different stages enabled progressive expansion of the core and increase in reactor power. The reactor has been operated upto a power level of 18.6MWt/3MWe with a sodium temperature of 482°C max. The reactor has completed 24 years of operation and is currently under periodic safety review by the Atomic Energy Regulatory Board of India. As a part of the periodic safety review, equipment qualification status and ageing management studies have been presented to the regulators. Equipment qualification refers to the ability of the replaceable equipment to meet the functional requirements on demand, accomplished by periodic surveillance, maintenance and replacement. Ageing management addresses the residual life assessment of components which are passive, non-replaceable / replaceable with difficulty, taking into account their life degrading mechanisms. Over a period of time, based on the operational feedback, maintenance difficulties and obsolescence, several major components have been replaced. These include the Neutronic channels, UPS, computers of the Central Data Processing System, main boiler feed pumps, three control rod drive mechanisms, two control rods, central canal plug, deaerator lift pumps, reheaters of the steam water system, station batteries, DM plant and Nitrogen plant. The starting air system of the emergency diesel generators and isolation dampers of the reactor containment building have also been replaced. Regarding the non-replaceable components, residual life assessment has been carried out based on the operational history vis-a`-vis the design limits for each component. The life limiting mechanism of heat transport systems of FBTR are creep and fatigue. Since the reactor has operated only upto a temperature of 444°C till 2007, the creep effect is insignificant. The total number of thermal cycles seen by the reactor components as of 2007 was 163, as against the design cycle of 2000 for most of the components. Hence all the heat transport system components are as good as fresh ones. However, the major life limiting factor has been found to be the Neutronic fluence on the grid plate which supports the core. The fast flux at the grid plate location was measured using Np foils and the residual life of the reactor has been assessed to be 10.5 effective full power years. This paper details the life extension exercise being carried out for FBTR.

Evaluation on Radiological Consequence Induced by MSLB Accident With Alternative Source Term

The major threat that nuclear power plants (NPPs) pose to the safety of the public comes from the large amount radioactive material released during design-basis accidents (DBAs). Additionally, many aspects of Control Room Habitability, Environmental Reports, Facility Siting and Operation derive from the design analyses that incorporated the earlier accident source term and radiological consequence of NPPs. Depending on current applications, majority of Chinese NPPs adopt the method of TID-14844, which uses the whole body and thyroid dose criteria. However, alternative Source Term (AST) are commonly used in AP1000 and some LWRs (such as Beaver Valley Power Station, Units No. 1 and No. 2, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 And 2, Kewaunee Power Station and so on), so it is attempted to adopt AST in radiological consequence analysis of other nuclear power plants. By introducing and implementing the method of AST defined in RG 1.183 and using integral safety analysis code, a pressurized water reactor (PWR) of 900 MW nuclear power plant analysis model is constructed and the radiological consequence induced by Main Steam Line Break (MSLB) accident is evaluated. For DBA MSLB, the fractions of core inventory are assumed to be in the gap for various radionuclides and then the release from the fuel gap is assumed to occur instantaneously with the onset of assumed damage. According to the assumptions for evaluating the radiological consequences of PWR MSLB, dose calculation methodology is performed with total effective dose equivalent (TEDE) which is the criteria of dose evaluation. Compared with dose criteria of RG 1.183, the dose of control room, exclusion area boundary and outer boundary of low population zone are acceptable.

Study on a Methodology of Human Factor Engineering Operating Experience Review for Nuclear Power Plant

The reviewing of operating experience at nuclear power plants (NPP) is not only critically important to safe and reliable operations, but also useful to guide the design of new plants which are similar to the current one under review. How to identify and analyze the safety-related operating experience and then implement a more extensive review is a vital and challengeable issue. In this paper, a methodology of human factor engineering (HFE) operating experience review (OER) is proposed for NPP. The need for the application of HFE in the life cycle activities of NPP and other nuclear facilities has been demonstrated by plant operating histories and regulatory and industry reviews. As a very important element of HFE, the OER is performed from the beginning of the design process. The main purpose of performing an OER is to verify that the applicant has identified and analyzed HFE-related safety problems and issues in previous designs that are similar to the current one. In this way, negative features associated with predecessor designs may be avoided in the current NPP design while retaining positive features. The research of OER concentrates on the aspect of review criterion, scope and implementation procedure of the HFE-related operating experience. As the NRC requirement, the scope of operating experience can be divided into six types in accordance with sources of information. The implementation procedures of USA and China are introduced, respectively. The resolution of HFE OER issues involve function allocation, changes in automation, HSI equipment design, procedures, training, and so forth. The OER conclusions can contribute to other HFE activities and improve the safety, reliability and usability of the HSI design in NPP.

Steam Generator Asset Management Program

The Steam Generator Asset Management Program (SGAMP) is a long term program designed to maximize the performance and reliability of the steam generators. The SGAMP focuses on plant specific conditions and hence is applicable to the original or the replacement steam generators. It is recommended that the utility and the vendor form a joint steam generator management team (SGMT) to develop, monitor and implement a long-term plan to address steam generator operation, maintenance and life extension goals. The SGMT will consist of representatives from operations, chemistry, maintenance and engineering functions and will be responsible for making decisions related to the steam generators. The charter of the SGMT is to develop a steam generator strategic plan that will cost-effectively manage steam generator options. The strategic plan is consistent with the Steam Generator Program Guidelines (NEI 97-06 in the United States). The strategic plan is a living document and is revised periodically to incorporate inspection results, new technology developments, lessons learned and industry experience. Cost-benefit analyses of strategies may be performed to prolong steam generator operability through steam generator performance modeling (tube degradation, fouling, etc.), diagnostic tools, regulatory strategy, condition monitoring and operational assessment strategy, and maintenance strategy. The SGMT will provide input regarding potential maintenance of the steam generators with schedule and cost impacts for each outage. It will also recommend engineering evaluations to be performed in support of program goals and will develop short- and long-term recommendations. These recommendations will address action plans, performance measures and results. Secondary side inspection and cleaning strategy should be developed (techniques and frequency) to maximize performance cost-effectively. This paper is based on Westinghouse experience gained by working with several pressurized water reactor (PWR) plant operators in the United States (US).