中國核能科技“三步走”發(fā)展戰(zhàn)略的思考

蘇罡

推薦論文摘要

中國核能科技“三步走”發(fā)展戰(zhàn)略的思考

蘇罡

核能作為中國的戰(zhàn)略產(chǎn)業(yè)，通過堅(jiān)持創(chuàng)新驅(qū)動(dòng)戰(zhàn)略，正在不斷轉(zhuǎn)型升級(jí)。聚焦于發(fā)電方向，中國推出自主三代壓水堆技術(shù)，在快堆、高溫氣冷堆等具備第四代特征的核電技術(shù)方向?qū)崿F(xiàn)突破，作為核心成員參加國際ITER計(jì)劃并順利推進(jìn)采購包計(jì)劃，為“熱堆—快堆—聚變堆”三步走奠定了堅(jiān)實(shí)的基礎(chǔ)。本文在總結(jié)階段性核能產(chǎn)業(yè)科技發(fā)展基礎(chǔ)上，提出了“十三五”乃至更長時(shí)期，落實(shí)“三步走”戰(zhàn)略的發(fā)展趨勢和技術(shù)方向，特別是對如何實(shí)現(xiàn)“熱堆—快堆”第二步跨越、實(shí)現(xiàn)“快堆和聚變”第三步跨越進(jìn)行了思考，并展望了未來核能科技發(fā)展的前景。

自主三代核電技術(shù)；第四代核電技術(shù)；聚變技術(shù)；“三步走”戰(zhàn)略

來源出版物：科技導(dǎo)報(bào), 2016, 34(15): 33-41

聯(lián)系郵箱：蘇罡，gang_su@126.com

高溫氣冷堆核電技術(shù)產(chǎn)業(yè)化思考

高立本，石磊，上官子瑛，等

摘要：根據(jù)國外高溫氣冷堆核電技術(shù)發(fā)展歷程和我國高溫氣冷堆核電技術(shù)產(chǎn)業(yè)化的成就、現(xiàn)狀與發(fā)展方向，從有利于發(fā)揮高溫氣冷堆核電技術(shù)多用途特性和加快高溫氣冷堆核電技術(shù)“走出去”步伐的戰(zhàn)略角度考慮，本文提出了我國高溫氣冷堆核電技術(shù)產(chǎn)業(yè)化的相關(guān)建議，供政府有關(guān)部門及行業(yè)相關(guān)單位參考。

關(guān)鍵詞：高溫氣冷堆；四代核電；產(chǎn)業(yè)化進(jìn)展

來源出版物：中國核電, 2016, 9(1): 25-30

模塊式小型反應(yīng)堆研發(fā)現(xiàn)狀及前景分析

熊厚華，杜繼富，曾正魁，等

摘要：目前的核電站大多數(shù)為裝機(jī)容量較大的反應(yīng)堆，有些甚至達(dá)到了175萬kW的單機(jī)容量，大型堆核電站一次性投入成本高、建造周期長，且難以適應(yīng)小型電網(wǎng)的需求，而小型反應(yīng)堆恰好可以解決這一問題，全球掀起了模塊式小型核電機(jī)組的開發(fā)熱潮。文章主要介紹了模塊式小型反應(yīng)堆的研發(fā)現(xiàn)狀、用途，并分析了小型堆與常規(guī)堆的優(yōu)劣勢及其發(fā)展前景。

關(guān)鍵詞：模塊式；小型反應(yīng)堆；發(fā)展?fàn)顩r；前景

來源出版物：價(jià)值工程, 2015, 34(2): 30-31

海上核電船海水冷卻系統(tǒng)的防腐防污問題

蘭志剛，李新仲，肖鋼，等

摘要：本文通過介紹海上核電船的海水冷卻系統(tǒng)的海洋腐蝕及污損問題，論述其運(yùn)行中存在的各類腐蝕和污損類型，分析產(chǎn)生腐蝕和污損的原因，給出相應(yīng)的防腐防污控制方法，可供海上核電設(shè)計(jì)人員進(jìn)行參考。

關(guān)鍵詞：海上核電船；海水冷卻系統(tǒng)；防腐防污

來源出版物：全面腐蝕控制, 2015, 29(11): 55-57

核電用石墨密封墊片的試驗(yàn)方法研究

楊書益

摘要：核電用石墨密封墊片是一種新型結(jié)構(gòu)的密封墊片，目前還沒有相關(guān)測試方法的標(biāo)準(zhǔn)。本文通過對石墨密封墊片測試方法的研究，重點(diǎn)對墊片壓縮率、回彈率、應(yīng)力松弛率及密封泄露率的試驗(yàn)載荷進(jìn)行了試驗(yàn)分析，提出了試驗(yàn)載荷的要求。對于制訂該產(chǎn)品測試方法標(biāo)準(zhǔn)有重要的參考價(jià)值。

關(guān)鍵詞：石墨密封墊片；壓縮率；回彈率；密封

來源出版物：流體機(jī)械, 2015, 43(1): 47-50

AP1000與EPR儀控系統(tǒng)平臺(tái)對比分析

周曉寧

摘要：三代核電技術(shù)是目前在建機(jī)組安全性較高的技術(shù)，而儀控系統(tǒng)是核電站中重要系統(tǒng)之一。通過對AP1000和 EPR儀控系統(tǒng)的平臺(tái)總體結(jié)構(gòu)、軟硬件等方面進(jìn)行分析并做了對比，比較了三代核電儀控系統(tǒng)平臺(tái)的不同點(diǎn)，得出 AP1000儀控系統(tǒng)平臺(tái)更加安全、可靠。

關(guān)鍵詞：AP1000；EPR；儀控系統(tǒng)

來源出版物：電力與能源, 2014, 35(6): 757-760

釷與核能

宋旺旺，鄧海軍，甘霖，等

摘要：文章基于核電所面臨的安全性、核廢料處置及鈾儲(chǔ)量少的現(xiàn)狀，著重對一種新的、儲(chǔ)量更豐富、能代替鈾作核燃料且更安全、產(chǎn)生的輻射垃圾更少的元素釷進(jìn)行了介紹，著重對釷在各種堆型中的利用潛力進(jìn)行了分析，為未來核電能更持久的發(fā)展提供了一種新的選擇。

關(guān)鍵詞：釷；反應(yīng)堆；燃料

來源出版物：科技創(chuàng)新與應(yīng)用, 2014 (35): 33-33

第三代壓水堆核電站核島通風(fēng)空調(diào)系統(tǒng)核級(jí)冷卻器關(guān)鍵技術(shù)和工藝研究

劉自旺，劉靜

摘要：通過對第三代壓水堆核電站核島通風(fēng)空調(diào)系統(tǒng)用核級(jí)冷卻器關(guān)鍵技術(shù)和工藝進(jìn)行研究，論證適用于核島通風(fēng)空調(diào)系統(tǒng)核級(jí)冷卻器關(guān)鍵技術(shù)和工藝的可行性，并推廣至所有核電站的核級(jí)冷卻器。

關(guān)鍵詞：第三代壓水堆核電站；核島通風(fēng)空調(diào)系統(tǒng)；核級(jí)冷卻器；關(guān)鍵技術(shù)和工藝；抗震鑒定

來源出版物：制冷與空調(diào), 2015, 16(11): 43-46

來源出版物：Science and Technology of Nuclear

Installations, 2017, 3976049

聯(lián)系郵箱：Lee, SH; wing088@naver.com

Development and characterization of a fiber-optic monitoring system for the key environment variables of the spent nuclear fuel pool at a nuclear power plant

Kim, R; Park, CH; Yoo, WJ; et al.

Abstract:This study develops and characterizes a fiber-optic monitoring system for the key environment variables of the spent nuclear fuel pool (SNFP) at a nuclear power plant. The three key environmental variables indicating the SNFP status directly are its water temperature, water level, and radiation level. First, this study develops and characterizes the individual fiber-optic sensors for measuring the three key environmental variables and then assembles them into an integrated monitoring system. The individual fiber-optic sensors commonly use optical fibers to transmit the signals delivered from their sensing probes despite their different characteristics. For the fiber-optic temperature sensor (FOTS), two types of FOTS are developed: contact and non-contact types, which are distinguished by whether their sensing probes are in direct contact with water. The contact-type FOTS uses a copper metal cap as its sensing probe, and the non-contact-type FOTS uses an infrared optical fiber, whose peripheral surface is coated with an anti-fog solution as its sensing probe. The fiber-optic water level sensor (FOWS) consists of optical fibers with their ends connected to the sensing probes fabricated with a NaCl solution and stainless steel. The FOWS measures the water level using the Fresnel reflection phenomenon, i.e., reflection of a portion of incident light at a discrete interface between two media having different refractive indices. The FOWS identifies the water level by measuring the amount of light reflected at the interface between the sensing probe and its outside medium, which varies according to whether the sensing probe is in contact with water. The fiber-optic radiation sensor (FORS) measures the gamma radiation in the SNFP. The sensing probe of FORS is a cylindrical-shaped LYSO: Ce scintillator, whose peripheral is wrapped with aluminum foil as the reflector. After characterizing the three individual sensors developed in this study, they are assembled and tested at a model water pool, 500 mm × 500 mm × 500 mm in size. The performance test results shows that individual sensors can measure the changes in each environmental variable in real-time.

關(guān)鍵詞：fiber-optic monitoring system; spent nuclear fuelpool; environment variables; water temperature; water level; radiation level

來源出版物：Annals of Nuclear Energy, 2017, 99: 183-192

聯(lián)系郵箱：Moon, JH; jhmoon86@dongguk.ac.kr

Esearch on the attribution evaluating methods of dynamic effects of various parameter uncertainties on the in-structure floor response spectra of nuclear power plant

Li, JB; Lin, G; Liu, J; et al.

Abstract:Consideration of the dynamic effects of the site and structural parameter uncertainty is required by the standards for nuclear power plants (NPPs) in most countries. The anti-seismic standards provide two basic methods to analyze parameter uncertainty. Directly manually dealing with the calculated floor response spectra (FRS) values of deterministic approaches is the first method. The second method is to perform probability statistical analysis of the FRS results on the basis of the Monte Carlo method. The two methods can only reflect the overall effects of the uncertain parameters, and the results cannot be screened for a certain parameter’s influence and contribution. In this study, based on the dynamic analyses of the floor response spectra of NPPs, a comprehensive index of the assessed impact for various uncertain parameters is presented and recommended, including the correlation coefficient, the regression slope coefficient and Tornado swing. To compensate for the lack of guidance in the NPP seismic standards, the proposed method can effectively be used to evaluate the contributions of various parameters from the aspects of sensitivity, acuity and statistical swing correlations. Finally, examples are provided to verify the set of indicators from systematic and intuitive perspectives, such as the uncertainty of the impact of the structure parameters and the contribution to the FRS of NPPs. The index is sensitive to different types of parameters, which provides a new technique for evaluating the anti-seismic parameters required for NPPs.

關(guān)鍵詞：uncertain parameter; floor response spectra (FRS); soil-structure interaction (SSI); seismic analysis and structural design; nuclear power plant (NPP)

來源出版物：Earthquake Engineering and Engineering Vibration, 2017, 16(1): 47-54

Caesium-rich micro-particles: A window into the meltdown events at the Fukushima Daiichi Nuclear Power Plant

Furuki, G; Imoto, J; Ochiai, A; et al.

Abstract:The nuclear disaster at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011 caused partial meltdowns of three reactors. During the meltdowns, a type of condensed particle, a caesium-rich micro-particle (CsMP), formed inside the reactors via unknown processes. Here we report the chemical and physical processes of CsMP formation inside the reactors during the meltdowns based on atomic-resolution electron microscopy of CsMPs discovered near the FDNPP. All of the CsMPs (with sizes of 2.0–3.4μm) comprise SiO2glass matrices and～10-nm-sized Zn–Fe oxide nanoparticles associated with a wide range of Cs concentrations (1.1–19wt% Cs as Cs2O). Trace amounts of U are also associated with the Zn–Fe oxides. The nano-texture in the CsMPs records multiple reaction-process steps during meltdown in the severe FDNPP accident: Melted fuel (molten core)-concrete interactions (MCCIs), incorporating various airborne fission product nanoparticles, including CsOH and CsCl, proceeded via SiO2condensation over aggregates of Zn-Fe oxide nanoparticles originating from the failure of the reactor pressure vessels. Still, CsMPs provide a mechanism by which volatile and low-volatility radionuclides such as U can reach the environment and should be considered in the migration model of Cs and radionuclides in the current environment surrounding the FDNPP.

來源出版物：Scientific Reports, 2017, 42731

聯(lián)系郵箱：Utsunomiya, S;

utsunomiya.satoshi.998@m.kyushu-u.ac.jp

聯(lián)系郵箱：Li, JB; jianboli@dlut.edu.cn

Polymeric seal degradation in nuclear power plants: Effect of gamma radiation on sealing properties

Porter, CP; Edge, R; Ogden, MD

Abstract:An effort has been made to bridge the gap between academic knowledge of polymeric seal degradation and industrial practices. A series of physical and mechanical properties that can be related to the sealing behavior of three commercial samples of nitrile rubber have been studied for their degradation when exposed to gamma radiation. For all samples the glass transition temperature (Tg) and retention factor (RF) were found to increase with total dose whilst percentagechange in mass (ΔM%) was found to decrease. The ultimate uptake of carbon dioxide (CO2) did not appear to change with radiation dose but the kinetics of the absorption process were found to decrease, suggesting the formation of crosslinks. The crosslinks formed appear to be dependent on the original material composition and comparison against degradation of material properties supports the theory behind butadiene, BDN, content being linked to a propensity for crosslink clustering.

關(guān)鍵詞：CO2uptake; degradation; glass transition; mechanical properties; swelling

來源出版物：Journal of Applied Polymer Science, 2017, 134(12)

聯(lián)系郵箱：Ogden, MD; m.d.ogden@sheffield.ac.uk

Internal structure of cesium-bearing radioactive microparticles released from Fukushima nuclear Power Plant

Yamaguchi, N; Mitome, M; Kotone, AH; et al.

Abstract:Microparticles containing substantial amounts of radiocesium collected from the ground in Fukushima were investigated mainly by transmission electron microscopy (TEM) and X-ray microanalysis with scanning TEM (STEM). Particles of around 2μm in diameter are basically silicate glass containing Fe and Zn as transition metals, Cs, Rb and K as alkali ions, and Sn as substantial elements. These elements are homogeneously distributed in the glass except Cs which has a concentration gradient, increasing from center to surface. Nano-sized crystallites such as copper- zinc- and molybdenum sulfide, and silver telluride were found inside the microparticles, which probably resulted from the segregation of the silicate and sulfide (telluride) during molten-stage. An alkali-depleted layer of ca. 0.2μm thick exists at the outer side of the particle collected from cedar leaves 8 months after the nuclear accident, suggesting gradual leaching of radiocesium from the microparticles in the natural environment.

來源出版物：Scientific Reports, 2016, 6, 20548

聯(lián)系郵箱：Yamaguchi, N; nyamag@affrc.go.jp

Flow boiling heat transfer in a helically coiled steam generator for nuclear power applications

Santini, L; Cioncolini, A; Butel, MT; et al.

Abstract:Forced convection boiling of water was experimentally investigated in a 24 m long full-scale helically coiled steam generator tube, prototypical of the steam generators with in-tube boiling used in small modular nuclear reactor systems. Overall, 1575 axially local and peripherally averaged heat transfer coefficient measurements were taken, covering operating pressures in the range of 2–6 MPa, mass fluxes from 200 to

800 kg m?2s?1and heat fluxes from 40 to 230 kW m?2. The heat transfer coefficient was found to depend on the mass flux and on the heat flux, indicating that both nucleate boiling and convection are contributing to the heat transfer process. Seven widely quoted flow boiling correlations for straight tubes fitted the present helical coil databank with a mean absolute percentage error within 15%–20%, which was comparable with the experimental uncertainty of the measured heat transfer coefficient values, thus indicating that curvature effects on flow boiling are small and negligible in practical applications.

關(guān)鍵詞：helical coil; convective flow boiling; steam generator; curvature effect; small modular nuclear reactor

來源出版物：International Journal of Heat and Mass Transfer, 2016, 92: 91-99

聯(lián)系郵箱：Cioncolini, A; lorenzo.santini@enel.com

Numerical simulation and experimental verification of microstructure evolution in large forged pipe used for AP1000 nuclear power plants

Wang, SL; Yang, B; Zhang, MX; et al.

Abstract:AP1000 primary coolant pipe is a large special-shaped forged pipe made of 316LN stainless steel. Due to the non-uniform temperature and deformation during its forging, coarse and fine grains usually coexist in the forged pipe, resulting in the heterogeneous microstructure and anisotropic performance. To investigate the microstructure evolution during the entire forging process, in the present research, the database of the 316LN stainless steel was established and a numerical simulation was performed. The results indicate that the middle body section of the forged pipe has an extremely uniform average grain size with the value smaller than 30 μm. The grain sizes in the ends of body sections were ranged from 30 μm to 60 μm. Boss sections have relatively homogeneous microstructure with the average grain size 30 μm to 44 μm. Furthermore, a full-scale hot forging was carried out for verification. Comparison of theoretical and experimental results showed good agreement and hence demonstrated the capabilities of the numerical simulation presented here. It is noteworthy that all grains in theworkpiece were confirmed less than 180 μm, which meets the designer’s demands.

關(guān)鍵詞：microstructure evolution; numerical simulation; integral forging technology; 316LN stainless steel

來源出版物：Annals of Nuclear Energy, 2016, 87: 176-185

聯(lián)系郵箱：Yang, B; byang@ustb.edu.cn

Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

Katata, G; Chino, M; Kobayashi, T; et al.

Abstract:Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information), and simulations from the oceanic dispersion model SEA-GEARN-FDM, both developed by the authors. A sophisticated deposition scheme, which deals with dry and fogwater depositions, cloud condensation nuclei (CCN) activation and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The fallout to the ocean surface calculated by WSPEEDI-II was used as input data for the SEA-GEARN-FDM calculations. Reverse and inverse source-term estimation methods based on coupling the simulations from both models was adopted using air dose rates and concentrations, and sea surface concentrations. The results revealed that the major releases of radionuclides due to FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, the morning of 13 March after the venting event at Unit 3, midnight of 14 March when the SRV (Safely Relief Valve) at Unit 2 was opened three times, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates associated with reactor pressure changes in Units 2 and 3. The modified WSPEEDI-II simulation using the new source term reproduced local and regional patterns of cumulative surface deposition of total131I and137Cs and air dose rate obtained by airborne surveys. The new source term was also tested using three atmospheric dispersion models (MLDP0, HYSPLIT, and NAME) for regional and global calculations and showed good agreement between calculated and observed air concentration and surface deposition of137Cs in East Japan. Moreover, HYSPLIT model using the new source term also reproduced the plume arrivals at several countries abroad showing a good correlation with measured air concentration data. A large part of deposition pattern of total131I and137Cs in East Japan was explained by in-cloud particulate scavenging. However, for the regional scale contaminated areas, there were large uncertainties due to the overestimation of rainfall amounts and the underestimation of fogwater and drizzle depositions. The computations showed that approximately 27% of137Cs discharged from FNPS1 deposited to the land in East Japan, mostly in forest areas.

來源出版物：Atmospheric Chemistry and Physics,2015, 15(2): 1029-1070

聯(lián)系郵箱：Katata, G; katata.genki@jaea.go.jp

Assessment of thermal embrittlement in duplex stainless steels 2003 and 2205 for nuclear power applications

Tucker, JD; Miller, MK; Young, GA

Abstract:Duplex stainless steels are desirable for use in power generation systems because of their attractive combination of strength, corrosion resistance and cost. However, thermal embrittlement at intermediate homologous temperatures of ～475°C and below, limits upper service temperatures for many applications. New lean grade duplex alloys have improved thermal stability over standard grades and potentially increase the upper service temperature or the lifetime at a given temperature for this class of material. The present work compares the thermal stability of lean grade, alloy 2003, to standardgrade, alloy 2205, through a series of isothermal agings between 260°C and 482°C for times between 1 and 10000 h. Aged samples were characterized by changes in microhardness and impact toughness. Additionally, atom probe tomography was performed to illustrate the evolution of the α–α′ phase separation in both alloys at select conditions. Atom probe tomography confirmed that phase separation occurs via spinodal decomposition for both alloys, and identified the presence of Ni–Cu–Si–Mn–P clusters in alloy 2205, which may contribute to the embrittlement of this alloy. The impact toughness model predictions for the upper service temperature show that alloy 2003 may be viable for use in 288°C applications for 80-year service lifetimes based on a Charpy V-notch criteria of 47 J at room temperature. In comparison, alloy 2205 should be limited to 260°C applications for the same room temperature toughness of 47 J.

關(guān)鍵詞：duplex stainless steels; 475 degrees C embrittlement; spinodal decomposition; atom probe tomography

來源出版物：Acta Materialia, 2015, 87: 15-24

聯(lián)系郵箱：Tucker, JD; julie.tucker@oregonstate.edu

Nuclear power as a climate mitigation strategy: Technology and proliferation risk

Lehtveer M; Hedenus F

Abstract:Recent years have witnessed renewed interest in nuclear power in large extent due to the need to reduce carbon emissions to mitigate climate change. Most studies of cost and feasibility of stringent climate targets that include nuclear power focus on the currently available light water reactor (LWR) technology. Since climate mitigation requires a long-term commitment, the inclusion of other nuclear technologies such as mixed oxide-fuelled LWRs and fast breeder reactors may better describe the future energy supply options. These different options also entail different nuclear weapon proliferation risks stemming from uranium enrichment or reprocessing of spent fuel. To investigate this relation, we perform a scenario analysis using the global energy transition model. Our results indicate that meeting a scenario with a 430ppm CO2target for 2100 is feasible without the involvement of nuclear power; however the mitigation costs increase by around 20%. Furthermore, a lasting contribution by nuclear power to climate change mitigation can only be achieved by alternative fissile material production methods and global diffusion of nuclear technologies. This in turn bears important implications for the risk of nuclear proliferation for several reasons. First, knowledge and competence in nuclear technology becomes more accessible, leading to the risk of nuclear programmes emerging in states with weaker institutional capacity. Additionally, even if the reprocessing step in a fast breeder cycle proves to be essentially proliferation resistant, the build-up of breeder reactor systems necessitates a long transition period with large-scale use of enrichment technology and its related proliferation risks. Our study does not include the costs posed on society by nuclear accident risk and by the need to upscale safeguards and regulatory capacity to deal with increased proliferation risk.

關(guān)鍵詞：nuclear weapon proliferation; nuclear power; energy system model

來源出版物：Journal of Risk Research, 2015, 18(3): 273-290

聯(lián)系郵箱：Lehtveer, M; mariliis.lehtveer@chalmers.se

編輯：王微

Separation of transformers for class 1E Systems in nuclear power plants

Lee, SH; Chang, CK

In order to supply electric power to the safety related loads, safety and reliability of onsite power have to be ensured for the safety function performance in nuclear power plants. Even though the existing electric power system of APR1400 meets the requirements of codes regarding Class 1E system, there is a room for improvement in the design margin against the voltage drop and short circuit current. This paper discusses the amount that the voltage drop and short circuit current occur in the existing electric power system of APR1400. Additionally, this paper studies with regard to the improved model that has the extra margin against the high voltage drop and short circuit current by separation of unit auxiliary transformer (UAT) and standby auxiliary transformer (SAT) for the Class 1E loads. The improved model of the electric power system by separation of UAT and SAT has been suggested through this paper. Additionally, effects of reliability and cost caused by the electric power system modification are considered.