核能安全

ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology

Chadwick， M. B.； Oblozinsky， P； Herman， M； et al.

Design and development of the AHWR: The Indian thorium fuelled innovative nuclear reactor

Sinha RK； Kakodkar A

Deliberately small reactors and the second nuclear era

Ingersoll， D. T.

Economic potential of modular reactor nuclear power plants based on the Chinese HTR-PM project

Zhang， Zuoyi； Sun， Yuhang

中國(guó)核電發(fā)展戰(zhàn)略研究

葉奇蓁

大型集成多功能中子學(xué)計(jì)算與分析系統(tǒng)Visual BUS的研究與發(fā)展

吳宜燦，李靜驚，李瑩，等

核能安全

·編者按·

原子的發(fā)現(xiàn)和核能的開(kāi)發(fā)利用給人類社會(huì)發(fā)展帶來(lái)新的動(dòng)力，極大增強(qiáng)人類認(rèn)識(shí)世界和改造世界的能力。核能發(fā)展伴隨著核安全風(fēng)險(xiǎn)和挑戰(zhàn)。人類要更好地利用核能、實(shí)現(xiàn)更大發(fā)展，必須確保核安全、做好核應(yīng)急。核安全是核能事業(yè)持續(xù)健康發(fā)展的生命線，核應(yīng)急是核能事業(yè)持續(xù)健康發(fā)展的重要保障。

20世紀(jì)50年代中期，中國(guó)創(chuàng)建核工業(yè)。60多年來(lái)，中國(guó)致力于和平利用核能事業(yè)，發(fā)展推動(dòng)核技術(shù)在工業(yè)、農(nóng)業(yè)、醫(yī)學(xué)、環(huán)境、能源等領(lǐng)域廣泛應(yīng)用。中國(guó)堅(jiān)持發(fā)展與安全并重原則，執(zhí)行安全高效發(fā)展核電政策，采用最先進(jìn)的技術(shù)、最嚴(yán)格的標(biāo)準(zhǔn)發(fā)展核電。伴隨著核能事業(yè)的發(fā)展，核安全與核應(yīng)急同步得到加強(qiáng)。中國(guó)的核設(shè)施、核活動(dòng)始終保持安全穩(wěn)定狀態(tài)，特別是核電安全水平不斷提高。面對(duì)核能事業(yè)發(fā)展新形勢(shì)新挑戰(zhàn)，中國(guó)核應(yīng)急在技術(shù)、裝備、人才、能力、標(biāo)準(zhǔn)等方面還存在一定不足，這也是其他國(guó)家在開(kāi)發(fā)利用核能進(jìn)程中面臨的共同課題。中國(guó)將通過(guò)理念創(chuàng)新、科技創(chuàng)新、管理創(chuàng)新，不斷強(qiáng)化國(guó)家核應(yīng)急管理，把核應(yīng)急提高到新水平。

中國(guó)始終把核安全放在和平利用核能事業(yè)首要位置，堅(jiān)持總體國(guó)家安全觀，倡導(dǎo)理性、協(xié)調(diào)、并進(jìn)的核安全觀，秉持為發(fā)展求安全、以安全促發(fā)展的理念，始終追求發(fā)展和安全2個(gè)目標(biāo)有機(jī)融合。半個(gè)多世紀(jì)以來(lái)，中國(guó)人民奮發(fā)圖強(qiáng)、歷盡艱辛，創(chuàng)建發(fā)展核能事業(yè)并取得輝煌成就。同時(shí)，不斷改進(jìn)核安全技術(shù)，實(shí)施嚴(yán)格的核安全監(jiān)管，加強(qiáng)核應(yīng)急管理，核能事業(yè)始終保持良好安全記錄。

核事故影響無(wú)國(guó)界，核應(yīng)急管理無(wú)小事?？偨Y(jié)切爾諾貝利核事故、福島核事故的教訓(xùn)，中國(guó)更加深刻認(rèn)識(shí)到核應(yīng)急的極端重要性，持續(xù)加強(qiáng)和改進(jìn)核應(yīng)急準(zhǔn)備與響應(yīng)工作，不斷提升中國(guó)核安全保障水平。中國(guó)在核應(yīng)急法律法規(guī)標(biāo)準(zhǔn)建設(shè)、體制機(jī)制建設(shè)、基礎(chǔ)能力建設(shè)、專業(yè)人才培養(yǎng)、演習(xí)演練、公眾溝通、國(guó)際合作與交流等方面取得巨大進(jìn)步，既為自身核能事業(yè)發(fā)展提供堅(jiān)強(qiáng)保障，也為推動(dòng)建立公平、開(kāi)放、合作、共贏的國(guó)際核安全應(yīng)急體系，促進(jìn)人類共享核能發(fā)展成果作出積極貢獻(xiàn)。

本專題得到陳妍教授（環(huán)境保護(hù)部核與輻射安全中心）的大力支持。

·熱點(diǎn)數(shù)據(jù)排行·

截至2016年 8月 26日，中國(guó)知網(wǎng)（CNKI）和Web of Science（WOS）的數(shù)據(jù)報(bào)告顯示，以“核能（nuclear energy）”“核安全（nuclear safety）”“核安全文化（nuclear safety culture）”“核應(yīng)急（Nuclear emergency）”為詞條可以檢索到的期刊文獻(xiàn)分別為1166條、9617條。本專題將相關(guān)數(shù)據(jù)按照：研究機(jī)構(gòu)發(fā)文數(shù)、作者發(fā)文數(shù)、期刊發(fā)文數(shù)、被引用頻次進(jìn)行排行，結(jié)果如下。

研究機(jī)構(gòu)發(fā)文數(shù)量排名（CNKI）

研究機(jī)構(gòu)發(fā)文數(shù)量排名（WOS）

作者發(fā)文數(shù)量排名（CNKI）

作者發(fā)文數(shù)量排名（WOS）

期刊發(fā)文數(shù)量排名（CNKI）

期刊發(fā)文數(shù)量排名（WOS）

根據(jù)中國(guó)知網(wǎng)（CNKI）數(shù)據(jù)報(bào)告，以“核能（nuclear energy）”“核安全（nuclear safety）”“核安全文化（nuclear safety culture）”“核應(yīng)急（Nuclear emergency）”為詞條可以檢索到的高被引論文排行結(jié)果如下。

國(guó)內(nèi)數(shù)據(jù)庫(kù)高被引論文排行

根據(jù)Web of Science統(tǒng)計(jì)數(shù)據(jù)，以“核能（nuclear energy）”“核安全（nuclear safety）”“核安全文化（nuclear safety culture）”“核應(yīng)急（Nuclear emergency）”為詞條可以檢索到的高被引論文排行結(jié)果如下。

國(guó)外數(shù)據(jù)庫(kù)高被引論文排行

·經(jīng)典文獻(xiàn)推薦·

基于Web of Science檢索結(jié)果，利用Histcite軟件選取LCS（Local Citation Score，本地引用次數(shù)）TOP50文獻(xiàn)作為節(jié)點(diǎn)進(jìn)行分析，得到本領(lǐng)域推薦的經(jīng)典文獻(xiàn)如下。

本領(lǐng)域經(jīng)典文獻(xiàn)

來(lái)源出版物：Reliability Engineering & System Safety，2004， 83（1）: 57-77

ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology

Chadwick， M. B.； Oblozinsky， P； Herman， M； et al.

Abstract: We describe the next generation general purpose Evaluated Nuclear Data File， ENDF/B-VIL0， of recommended nuclear data for advanced nuclear science and technology applications. The library， released by the U.S. Cross Section Evaluation Working Group（CSEWG）in December 2006， contains data primarily for reactions with incident neutrons， protons， and photons on almost 400 isotopes， based on experimental data and theory predictions. The principal advances over the previous ENDF/B-VI library are the following:（1） New cross sections for U， Pu， Th； Np and Am actinide isotopes， with improved performance in integral validation criticality andneutron transmission benchmark tests；（2） More precise standard cross sections for neutron reactions on H，6Li，10B， An and for235，238U fission， developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation（WPEC）:（3） Improved thermal neutron scattering:，（4） An extensive set of neutron cross sections on fission products developed through a WPEG collaboration；（5） A large suite of photonuclear reactions；（6） Extension of many neutron-and protoninduced evaluations up to 150 MeV:（7） Many new light nucleus neutron and proton reactions；（8） Post-fission beta-delayed photon decay spectra:，（9） New radioactive decay data:，（10） New methods for uncertainties and covariances， together with covariance evaluations for some sample cases； and（11） New actinide fission energy deposition. The paper provides an overview of this library，consisting of 14 sublibraries in the same ENDF-6 format as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries， focusing on neutron reactions. Extensive validation， using radiation transport codes to simulate measured critical assemblies， show major improvements:（a） The Ion-standing underprediction of low enriched uranium thermal assemblies is removed；（b） The238U and208Pb and9Be reflector biases in fast systems are largely removed；（c） ENDF/B-VI.8 good agreement for simulations of thermal high-enriched uranium assemblies is preserved；（d） The underprediction of fast criticality of235，238U and239Pu assemblies is removed； and（e） The intermediate spectrum critical assemblies are predicted more accurately. We anticipate that the new library will play an importanrole in nuclear technology applications，including transport simulations supporting national security， nonproliferation， advanced reactor and fuel cycle concepts， criticality safety， fusion， medicine， space applications， nuclear astrophysics， and nuclear physics facility design. The ENDF/B-VII.0 library is archived at the National Nuclear Data Center， BNL， and can be retrieved from www.nndc.bnl.gov.

來(lái)源出版物：Nuclear Data Sheets， 2006， 107（12）: 2931-3059

Design and development of the AHWR: The Indian thorium fuelled innovative nuclear reactor

Sinha RK； Kakodkar A

Abstract: India has chalked out a nuclear power program based on its domestic resource position of uranium and thorium. The first stage started with setting up the Pressurized Heavy Water Reactors（PHWR） based on natural uranium and pressure tube technology. In the second phase， the fissile material base will be multiplied in Fast Breeder Reactors using the plutonium obtained from the PHWRs. Considering the large thorium reserves in India， the future nuclear power program will be based on thorium-233U fuel cycle. However， there is a need for the timely development of thorium-based technologies for the entire fuel cycle. The Advanced Heavy Water Reactor（AHWR） has been designed to fulfill this need. The AHWR is it 300 MW， vertical， pressure tube type， heavy water moderated， boiling light water cooled natural circulation reactor. The fuel consists of（Th-Pu）O2and（Th-233U）O2pins. The fuel cluster is designed to generate maximum energy out of233U， which is bred in situ from thorium and has a slightly negative void coefficient of reactivity. For the AHWR， the well-proven pressure tube technology has been adopted and many passive safety features， consistent with the international trend， have been incorporated. A distinguishing feature which makes this reactor unique，from other conventional nuclear power reactors is the fact that it is designed to remove core heat by natural circulation. under normal operating conditions， eliminating the need of pumps. In addition to this passive feature，several innovative passive safety systems have been incorporated in the design， for decay heat removal under shut down condition and mitigation of postulated accident conditions. The design of the reactor has progressively undergone modifications and improvements based on the feedbacks from the analytical and the experimental R&D. This paper gives the details of the current design of the AHWR.

來(lái)源出版物： Nuclear Engineering and Design， 2006， 236（7-8）: 683-700

Deliberately small reactors and the second nuclear era

Ingersoll， D. T.

Abstract: Smaller sized nuclear reactors were instrumental during the pioneering days of commercial nuclear power to facilitate the development and demonstration of early reactor technologies and to establish operational experience for the fledgling nuclear power industry. As the U.S. embarks on its “second nuclear era，” the question becomes: Will smaller sized plants have a significant role in meeting the nation’s needs for electricity and other energy demands？A brief review of our nuclear history is presented relativeto plant size considerations， followed by a review of several commonly cited benefits of small reactors. Several“deliberately small” designs currently being developed in the U.S. are briefly described， as well as some of the technical and institutional challenges faced by these designs. Deliberately small reactors offer substantial benefits in safety. security， operational flexibilities and economics，and they are well positioned to figure prominently in the second nuclear era.

Keywords: small medium reactors； deliberately small reactors； second nuclear era； nuclear renaissance； new reactor designs

來(lái)源出版物：Progress in Nuclear Energy， 2009， 51（4-5）: 589-603

Economic potential of modular reactor nuclear power plants based on the Chinese HTR-PM project

Zhang， Zuoyi； Sun， Yuhang

Abstract: Modular reactors with improved safety features have been developed after the Three-Mile Island accident. Economics of small modular reactors compared to large light water reactors whose power output is 10 times higher is the major issue for these kind of reactors to be introduced into the market. Based on the Chinese high temperature gas-cooled reactor pebble-bed module（HTRPM） project， this paper analyzes economical potentials of modular reactor nuclear power plants. The reactor plant equipments are divided into 6 categories such as RPV and reactor internals， other NSSS components and so on. The economic impact of these equipments is analyzed. It is found that the major difference between an HTR-PM plant and a PWR is the capital costs of the RPV and the reactor internals. The fact， however， that RPV and reactor internals costs account for only 2% of the total plant costs in PWR plants demonstrates the limited influence of this difference. On the premise of multiple NSSS modules forming a nuclear power plant with a plant capacity equivalent to a typical PWR plant， an upper value and a target value of the total plant capital costs are estimated. A comparison is made for two design proposals of the Chinese HTR-PM project. It is estimated that the specific costs of a ready-to-build 2 × 250 MWth modular plant will be only 5% higher than the specific costs of one 458 MWth plant. When considering the technical uncertainties of the latter， a 2 × 250 MWth modular plant seems to be more attractive. Finally， four main points are listed for MHTGRs to achieve economic viability.

來(lái)源出版物：Nuclear Engineering and Design， 2007， 237（23）: 2265-2274

·推薦綜述·

中國(guó)核電發(fā)展戰(zhàn)略研究

葉奇蓁

1核能在中國(guó)能源可持續(xù)發(fā)展中的地位

1.1中國(guó)能源資源狀況分析

中國(guó)能源資源有3個(gè)基本特點(diǎn)。能源資源品種豐富，但人均占有量較少，在己探明儲(chǔ)量中煤炭占世界人均的56%、石油占11%，天然氣占4.6%。能源資源結(jié)構(gòu)不盡合理，煤炭、水能相對(duì)豐富，而優(yōu)質(zhì)化石能源（石油）相對(duì)不足。能源資源分布與生產(chǎn)力布局不平衡，經(jīng)濟(jì)發(fā)達(dá)地區(qū)在東南沿海，而水力資源在西部和西南部，煤炭主要在北方。

目前，我國(guó)能源發(fā)展面臨4個(gè)基本問(wèn)題。即經(jīng)濟(jì)社會(huì)發(fā)展中的能源供需總量平衡問(wèn)題。長(zhǎng)期以煤為主的能源結(jié)構(gòu)，造成的環(huán)境、生態(tài)問(wèn)題。西氣東運(yùn)、北煤南運(yùn)、西電東輸?shù)哪茉摧斶\(yùn)問(wèn)題，我國(guó)煤炭運(yùn)輸占鐵路運(yùn)量的40%，占沿海和長(zhǎng)江中下游水運(yùn)1/3。對(duì)國(guó)外資源依存的能源供應(yīng)安全問(wèn)題。

核電的基本特性決定了在應(yīng)對(duì)能源挑戰(zhàn)中有能力發(fā)揮無(wú)可替代的重要作用。核電不排放SO2等污染物和溫室氣體CO2，對(duì)環(huán)境后果實(shí)行嚴(yán)格管理，因此屬于清潔能源。而核電的安全可靠性正在不斷提高。核電對(duì)煤電具有較強(qiáng)經(jīng)濟(jì)競(jìng)爭(zhēng)力和替代能力，目前二代改進(jìn)型核電站的電價(jià)大都與當(dāng)?shù)氐臉?biāo)桿電價(jià)相當(dāng)。核電燃料運(yùn)輸量小。因此，我國(guó)在現(xiàn)階段發(fā)展核電是調(diào)整源布局的有效途徑。

1.2中國(guó)核能發(fā)展的技術(shù)路線

我國(guó)核能發(fā)展的技術(shù)路線是走熱堆、快堆、聚變堆三步發(fā)展的道路。在近期發(fā)展己經(jīng)成熟的熱中子堆核電站，滿足當(dāng)前和近期核電發(fā)展的需要。第二步發(fā)展快中子增殖堆核電站及配套的核燃料循環(huán)體系，充分利用鈾資源，實(shí)現(xiàn)裂變核能的可持續(xù)發(fā)展。第三步發(fā)展核聚變堆核電站，有望最終解決人類的能源供應(yīng)問(wèn)題。

目前，在熱堆核電發(fā)展階段，逐步實(shí)現(xiàn)由二代向三代過(guò)渡。在2020年以前，適度發(fā)展我國(guó)己經(jīng)掌握技術(shù)的二代改進(jìn)型壓水堆核電站。抓緊引進(jìn)三代核電技術(shù)的消化吸收再創(chuàng)新，掌握技術(shù)、實(shí)現(xiàn)自主化，盡快實(shí)現(xiàn)三代核電的批量化建設(shè)。

1.3核電產(chǎn)業(yè)發(fā)展的目標(biāo)

根據(jù)有關(guān)研究部門的預(yù)測(cè)，2020年我國(guó)電力總裝機(jī)將達(dá)到15億kW，核電總裝機(jī)容量將達(dá)到7000萬(wàn)kW，核電容量占總?cè)萘康?.6%，占總發(fā)電量的7.0%左右?？紤]能源結(jié)構(gòu)調(diào)整的要求，2030年我國(guó)總發(fā)電裝機(jī)容量將達(dá)到20億kW，核電總裝機(jī)容量將達(dá)到2億kW，核電裝機(jī)容量占10%，占總發(fā)電量的15%。2050年我國(guó)將進(jìn)入中等發(fā)達(dá)國(guó)家行列，以人均1.56 kW計(jì)算，總發(fā)電裝機(jī)容量將達(dá)到25億kW，核電總裝機(jī)容量將達(dá)到4億kW，核電占總裝容量的16%，占總發(fā)電量的22%。

2中國(guó)核電已形成規(guī)?；炕l(fā)展格局

我國(guó)大陸投入商運(yùn)的核電機(jī)組共有 11臺(tái)，總裝機(jī)容量為910萬(wàn)kW，機(jī)組負(fù)荷因子達(dá)85%～92%，各項(xiàng)運(yùn)行指標(biāo)均高于世界平均水準(zhǔn)，處于世界中上等水平以上。在全球441座核電站中，大多進(jìn)入前50～60名。即將建成的嶺澳二期核電站和秦山核電二期擴(kuò)建均進(jìn)展良好，預(yù)期在2010—2011年將陸續(xù)投產(chǎn)發(fā)電。目前己有22臺(tái)二代改進(jìn)型壓水堆核電站取得了路條，并己有7臺(tái)機(jī)組澆灌了第一罐混凝土。主設(shè)備己實(shí)現(xiàn)了批量采購(gòu)，有的制造廠己簽訂了數(shù)臺(tái)或十余臺(tái)長(zhǎng)周期設(shè)備。而核電站設(shè)計(jì)的標(biāo)準(zhǔn)化規(guī)范化工作也正在積極進(jìn)行當(dāng)中。

當(dāng)前我國(guó)二代改進(jìn)型壓水堆核電站己具備系列化規(guī)?；l(fā)展的有利條件。二代改進(jìn)型壓水堆屬于成熟的堆型，設(shè)計(jì)經(jīng)過(guò)驗(yàn)證，自主化程度較高。有相當(dāng)豐富的自主建設(shè)和自主運(yùn)行經(jīng)驗(yàn)，平均建設(shè)周期小于5 a。設(shè)備國(guó)產(chǎn)化率超過(guò)70%，除主循環(huán)泵（目前己有3家制造廠在研制）外，主要的核電設(shè)備己具備堅(jiān)實(shí)的國(guó)產(chǎn)化基礎(chǔ)。我國(guó)己建成的核電站的運(yùn)行經(jīng)驗(yàn)表明，核電站的運(yùn)行是安全的，沒(méi)有溫室氣體和有害氣體排放，放射性廢物的排放遠(yuǎn)低于國(guó)家標(biāo)準(zhǔn)。

2.1二代改進(jìn)型壓水堆核電站自主化能力分析

二代改進(jìn)型壓水堆核電站隨著技術(shù)的發(fā)展和運(yùn)行經(jīng)驗(yàn)的反饋，逐步引入新的成熟技術(shù)，使核電站的安全性得到進(jìn)一步的提高。新設(shè)計(jì)建設(shè)的二代改進(jìn)型壓水堆降低了堆芯功率密度，使熱工安全余量大于15%；加大穩(wěn)壓器容量，增加了核電站運(yùn)行的穩(wěn)定性；增設(shè)附加應(yīng)急柴油發(fā)電機(jī)系統(tǒng)，提高了供電的可靠性；增設(shè)安全殼過(guò)濾卸壓排放系統(tǒng)，防止安全殼超壓失效，并防止放射性外泄；應(yīng)用概率安全分析技術(shù)以及風(fēng)險(xiǎn)管理技術(shù)，防止核電站出現(xiàn)嚴(yán)重事故；引入嚴(yán)重事故預(yù)防和緩解措施：如非能動(dòng)氫復(fù)合系統(tǒng)防止氫爆、穩(wěn)壓器卸壓排放系統(tǒng)防止高壓熔堆、田灣核電站還設(shè)計(jì)了堆芯捕集器用以在堆芯熔融時(shí)防止熔融物熔穿透安全殼底板；廣泛采用數(shù)字化儀控技術(shù)和先進(jìn)控制室，改善了人機(jī)界面；汽輪發(fā)電機(jī)采用半速機(jī)組，提高了出力和熱效。

二代改進(jìn)型壓水堆核電站在自主設(shè)計(jì)能力方面，形成了專業(yè)配套、結(jié)構(gòu)合理的研究設(shè)計(jì)隊(duì)伍。

在項(xiàng)目管理能力方面，按國(guó)際通用項(xiàng)目管理模式管理，己基本與國(guó)際接軌。

在設(shè)備制造能力方面，3大集團(tuán)都將基本具備每年提供2～3臺(tái)百萬(wàn)千瓦級(jí)機(jī)組設(shè)備的能力。

在建設(shè)安裝能力方面，己經(jīng)具有4個(gè)項(xiàng)目8臺(tái)機(jī)組的建設(shè)實(shí)踐。

在營(yíng)運(yùn)管理能力方面，根據(jù)世界核電運(yùn)行者協(xié)會(huì)WANO的9項(xiàng)性能指標(biāo)，3項(xiàng)進(jìn)入前1/4的先進(jìn)行列，有5項(xiàng)超過(guò)中值水平，只有1項(xiàng)略低于中值水平。

在安全監(jiān)管能力方面，建立了與國(guó)際接軌的核安全管理和監(jiān)督的法規(guī)制度，具備了全過(guò)程全方位監(jiān)督管理的能力。

2.2大力堆進(jìn)內(nèi)陸核電建設(shè)

國(guó)際上大部分核電站建設(shè)在內(nèi)陸。法國(guó)有65.1%的核電站建設(shè)在內(nèi)陸，美國(guó)亦有75.7%的核電站建設(shè)在內(nèi)陸。有些內(nèi)陸國(guó)家，比如瑞士，5座核電站都在內(nèi)陸的江河邊上，5座核電站總發(fā)電功率為3220 MW，占總發(fā)電量的37%，其他將近60%的發(fā)電量由水電提供。因此，國(guó)外其他國(guó)家的經(jīng)驗(yàn)表明，在內(nèi)陸建核電站是完全可行的。

我國(guó)內(nèi)陸地區(qū)經(jīng)濟(jì)有了很大發(fā)展，電網(wǎng)容量亦有很大發(fā)展，但部分省份同樣存在缺乏煤炭和水力資源。2009年初南方各省發(fā)生了大面積、長(zhǎng)時(shí)間的雪災(zāi)，造成了廣大地區(qū)長(zhǎng)時(shí)間的斷電，帶來(lái)了嚴(yán)重的后果。因此，僅依靠遠(yuǎn)距離輸電和長(zhǎng)途運(yùn)煤是難以保障用電安全的。這樣，除提高電網(wǎng)的抗災(zāi)害能力，建設(shè)緊急情況下不依賴燃料運(yùn)輸?shù)暮穗娬臼呛苡斜匾摹?/p>

從安全和環(huán)保要求看，內(nèi)陸核電站和沿海核電站沒(méi)有本質(zhì)的差別。目前成熟的核電站設(shè)計(jì)和建造技術(shù)完全可用到內(nèi)陸核電。內(nèi)陸江河流量多半不夠大，可采用冷卻塔閉式循環(huán)帶走余熱，以減輕溫排水對(duì)環(huán)境的影響。目前，百萬(wàn)千瓦級(jí)核電站一機(jī)一塔要求塔高200 m，淋水面積16000 m2時(shí)，我國(guó)己能設(shè)計(jì)160 m，12000 m2冷卻塔，正在開(kāi)展超大型冷卻塔的設(shè)計(jì)。因此按照核電規(guī)范選擇的廠址是能夠保證核電站的安全的。

2.2.1放射性液態(tài)流出物的排放控制

內(nèi)陸廠址與沿海廠址相比，液態(tài)流出物中要考慮放射性物質(zhì)到達(dá)人體的途徑及飲用水和灌溉等途徑。目前，我國(guó)江河、湖泊污染事件屢有發(fā)生，國(guó)家主管部門和公眾對(duì)于河流的排放控制均持高度關(guān)注和審慎的態(tài)度。核電廠環(huán)境輻射防護(hù)規(guī)定液態(tài)流出物排放的放射性總量每年≤200 GBq（不包括氚），URD文件中對(duì)先進(jìn)壓水堆核電站規(guī)定每年≤1.85 GBq（不包括氚），EUR文件中對(duì)先進(jìn)壓水堆核電站規(guī)定每年≤10 GBq（不包括氟）。從秦山二期2002—2006年統(tǒng)計(jì)的數(shù)據(jù)，年液態(tài)流出物排放的放射性總量為2～5 GBq。因此，目前設(shè)計(jì)的液態(tài)流出物處理系統(tǒng)完全能滿足國(guó)標(biāo)要求，而實(shí)際運(yùn)行水平遠(yuǎn)低于國(guó)標(biāo)要求，并與先進(jìn)壓水堆核電站的要求相當(dāng)。

2.2.2液態(tài)放射性流出物排放濃度控制

我國(guó)的《生活飲用水衛(wèi)生標(biāo)準(zhǔn)》（GB57492006）中規(guī)定總β放射性小于1 Bq/L?！逗藙?dòng)力廠環(huán)境輻射防護(hù)規(guī)定》（GB6249）提出核動(dòng)力廠排放口下游1 km處受納水體中總月放射性濃度不得超過(guò)1 Bq/L，這就是要求在排放口下1 km處滿足生活飲用水標(biāo)準(zhǔn)。GB-14587—修訂版的征求意見(jiàn)稿，提出了100 Bq/L的排放罐出口濃度控制值。因此，經(jīng)過(guò)適當(dāng)?shù)南♂專穗姀S液態(tài)放射性流出物排放濃度就可達(dá)到天然放射性本底水平。

內(nèi)陸核電站由于采用冷卻塔閉式循環(huán)帶走余熱，沒(méi)有循環(huán)冷卻水對(duì)放射性廢液的稀釋。濱海壓水堆核電站液態(tài)流出物排放的內(nèi)部實(shí)際控制值為≤1000～2000 Bq/L（不包括氚），經(jīng)循環(huán)冷卻水對(duì)放射性廢液的稀釋 1000倍后，其濃度己相當(dāng)?shù)?，一般? Bq/L。俄羅斯濱河核電站要求液態(tài)流出物排放的濃度控制值為≤18 Bq/L（不包括氚）。所以，改進(jìn)目前沿海核電站的液態(tài)放射性廢物的處理技術(shù)，是完全能滿足內(nèi)陸核電站對(duì)液態(tài)放射性廢物處理和排放的要求的。

2.2.3液態(tài)放射性廢物處理技術(shù)

俄羅斯核電站放射性廢液處理采用了雙蒸發(fā)器處理系統(tǒng)，處理后的液體再經(jīng)二級(jí)離子交換處理，凈化系數(shù)從10E3提高到10E5。美國(guó)采用反滲透廢液處理技術(shù)，實(shí)現(xiàn)廢水回用，以滿足“零液體排放”要求，并可針對(duì)某些元素進(jìn)行高純度凈化或去除。美國(guó) Comanch Peak核電站用于去除放射性，特別是Co膠體，CS和I到監(jiān)測(cè)不到水平，凈化系數(shù)達(dá) 5.7×104。美國(guó)德賴斯登核電站用超級(jí)過(guò)濾+反滲透+去離子技術(shù)處理廢液。內(nèi)陸核電站的含氟廢水，在廢水處理后，排入冷卻塔循環(huán)冷卻水中，通過(guò)蒸發(fā)向大氣排放。

3積極消化吸收第三代核電技術(shù)

3.1第三代核電發(fā)展的背景

1979年美國(guó)發(fā)生的三里島核電站事故和1986年前蘇聯(lián)發(fā)生的切爾諾貝利核電站事故，使公眾要求進(jìn)一步提高核電的安全性。1990年EPRI根據(jù)主要電力公司意見(jiàn)出版了“電力公司要求文件（URD）”共 3卷。1994年歐洲聯(lián)盟出版了“歐洲電力公司要求（EUR）”共 4卷。這些文件對(duì)未來(lái)壓水堆和沸水堆核電站提出了電力公司明確和完整的要求，更高的安全要求和經(jīng)濟(jì)要求，涉及各個(gè)技術(shù)和經(jīng)濟(jì)領(lǐng)域。

第三代核電機(jī)組要有更高安全目標(biāo)。即堆芯熱工安全裕量＞15%，堆芯損壞概率＜10-5/堆年，大量放射性外泄＜10-6/堆年。第三代核電機(jī)組要有更好的經(jīng)濟(jì)性，具體表現(xiàn)在機(jī)組額定功率為1000～1500 MWe，可利用因子＞87%，換料周期18～24月，電站壽命60 a，建設(shè)周期48～52月，電價(jià)要能與聯(lián)合循環(huán)的天然氣電廠相競(jìng)爭(zhēng)。因此，第三代核電機(jī)組在技術(shù)上更先進(jìn)。

3.2AP1000核電站的特點(diǎn)

AP1000核電站采用非能動(dòng)安全系統(tǒng)。具體表現(xiàn)在采用非能動(dòng)安注、多級(jí)非能動(dòng)自動(dòng)卸壓系統(tǒng)、非能動(dòng)余熱排放系統(tǒng)和非能動(dòng)安全殼冷卻系統(tǒng)。AP1000核電站引入了嚴(yán)重事故預(yù)防和緩解措施，如堆腔淹沒(méi)技術(shù)、安全殼內(nèi)氫點(diǎn)火和氫復(fù)合系統(tǒng)、堆芯熔融物反應(yīng)堆壓力容器內(nèi)保持。同時(shí)，AP1000采用雙層安全殼和全數(shù)字化儀控系統(tǒng)。采用模塊化施工使建設(shè)工期縮短到 48個(gè)月。

AP1000核電站的反應(yīng)堆冷卻劑系統(tǒng)（如圖1所示）采用屏蔽式電泵，取消了機(jī)械密封，采用在上部堆芯測(cè)量以及大容量穩(wěn)壓器，焊接結(jié)構(gòu)的堆內(nèi)構(gòu)件和壓力容器活性區(qū)及法蘭接管段大型整體鍛件。

AP1000核電站的非能動(dòng)堆芯冷卻系統(tǒng)，不依賴外部電源，采用非能動(dòng)余熱導(dǎo)出、非能動(dòng)安全注入以及非能動(dòng)安全殼冷卻可以保證長(zhǎng)時(shí)間的安全停堆，還可以保證大于72 h不用操作員干預(yù)。

3.3ERP核電站的特點(diǎn)

EPR核電站的主要特點(diǎn)有以下幾個(gè)。EPR核電站功率高，達(dá)到1500～1700 MWe。采用4通道安全系統(tǒng)和雙層安全殼。引入了嚴(yán)重事故預(yù)防及緩解措施，如穩(wěn)壓器卸壓、堆芯撲集器和非能動(dòng)氫復(fù)合器。同時(shí)EPR核電站也采用全數(shù)字化儀控和模塊化施工。

3.4AP1000的關(guān)鍵技術(shù)

AP1000的關(guān)鍵技術(shù)是采用非能動(dòng)安全系統(tǒng)，特別是非能動(dòng)安全殼冷卻系統(tǒng)。AP1000核電站引入了嚴(yán)重事故的預(yù)防和緩解措施，包括自動(dòng)卸壓系統(tǒng)久（ADS），抑制氫爆的氫復(fù)合系統(tǒng)（氫點(diǎn)火器和非能動(dòng)氫催化復(fù)合），以及堆芯熔融物壓力容器內(nèi)保持（IVR）等技術(shù)。同時(shí)AP1000核電站大容量屏蔽泵的設(shè)計(jì)和制造，爆破膜的設(shè)計(jì)和制造，以及大尺寸園柱形鋼安全殼的設(shè)計(jì)和建造也存在技術(shù)難點(diǎn)和需攻克的關(guān)鍵技術(shù)。

3.5重視三代核電引進(jìn)中技術(shù)風(fēng)險(xiǎn)經(jīng)濟(jì)風(fēng)險(xiǎn)的規(guī)避

AP1000的技術(shù)風(fēng)險(xiǎn)主要在于缺少首堆工程整體驗(yàn)證的實(shí)踐證明，AP1000的設(shè)計(jì)認(rèn)證尚未真正通過(guò)，而且還有一系列涉及安全的設(shè)計(jì)驗(yàn)證工作未做，設(shè)計(jì)方案尚未固化，從美國(guó)條件的設(shè)計(jì)直接移植到中國(guó)，還需要作適應(yīng)性修改。

AP1000核電站也存在一定的經(jīng)濟(jì)風(fēng)險(xiǎn)。最近西屋公司與美國(guó)幾個(gè)電力公司簽訂在美國(guó)新建AP1000的總承包協(xié)議，比投資是我國(guó)自主建設(shè)核電的 2～3倍，也是招標(biāo)引進(jìn)時(shí)申報(bào)的2～3倍。

4鈾資源的保障

我國(guó)己探明一定數(shù)量鈾資源可以滿足近期核電發(fā)展的需要。國(guó)內(nèi)鈾資源勘測(cè)有較好發(fā)展前景。理論預(yù)測(cè)鈾礦資源比較豐富，預(yù)測(cè)鈾資源總量超過(guò)幾百萬(wàn)噸，加之我國(guó)相當(dāng)范圍國(guó)土未經(jīng)詳細(xì)勘查，因此擴(kuò)大老礦區(qū)、加強(qiáng)深層勘查，開(kāi)辟新基地前景看好。我國(guó)目前己探明儲(chǔ)量，加上海外采購(gòu)和合作開(kāi)采的天然鈾，足以保障2020年核電對(duì)天然鈾的需求。因此加大鈾資源的國(guó)內(nèi)勘查力度，同時(shí)開(kāi)拓國(guó)外鈾資源的供應(yīng)，我國(guó)核電發(fā)展的鈾資源是一定能得到保證的。

從長(zhǎng)期來(lái)看，到 2030—2050年我國(guó)的人口將達(dá)到頂峰16億，按平均每人消耗電力1.56 kW來(lái)計(jì)（相當(dāng)于發(fā)達(dá)國(guó)家的中等水平），就需要25億kW的電力供應(yīng)，其中16%為核電（相當(dāng)于目前世界核電的平均份額），即4億kW的核電。到2050年我國(guó)對(duì)于天然鈾資源需求相當(dāng)大，如果核電的比例比16%還要大，則對(duì)天然鈾資源的需求將更大。

5開(kāi)發(fā)快中子增殖堆核電站、構(gòu)建核燃料循環(huán)體系

5.1鈉冷快堆SFR

快中子增殖反應(yīng)堆的主要特點(diǎn)在于它能增殖核燃料，即它每燃耗一個(gè)燃料原子，就可以生產(chǎn)出多于一個(gè)燃料原子，這樣一來(lái)，在理論上說(shuō)，它可以將全部鈾資源都轉(zhuǎn)化為可燃燒的燃料并加以利用。采用適當(dāng)增殖比的快中子堆，可以將鈾資源的利用率由普通的熱堆的不足 1%，提高到 60%～70%，從而有效防止鈾資源枯竭的威脅。

快中子增殖反應(yīng)堆中等規(guī)模的電功率為 150～500 MWe，一般采用熱冶金金屬燃料后處理循環(huán)。大型規(guī)模的電功率為500～1500 MWe，一般采用先進(jìn)水法氧化燃料后處理循環(huán)。堆出口溫度可達(dá) 550℃?？熘凶釉鲋撤磻?yīng)堆用鈉作為冷卻劑，主要分為池式或環(huán)路式2種。

5.2快中子反應(yīng)堆在中國(guó)的發(fā)展

我國(guó)己在“十一五”期間建成實(shí)驗(yàn)快中子堆。計(jì)劃2020年前后將建成原型快中子堆核電站，通過(guò)引進(jìn)技術(shù)建設(shè)第一個(gè)快中子堆示范工程。2035年前后完成商用快中子堆核電站及核燃料循環(huán)系統(tǒng)的建設(shè)。此時(shí)，不僅可利用0.7% U—235，通過(guò)快中子堆增殖，還可利用大量的U—238（經(jīng)快中子反應(yīng)堆轉(zhuǎn)換的Pu）。

5.3加快商用后處理廠的建設(shè)和快堆燃料循環(huán)技術(shù)

的研發(fā)

近期目標(biāo)主要是實(shí)現(xiàn) 2025年開(kāi)式循環(huán)向閉式循環(huán)轉(zhuǎn)變，減緩天然鈾資源的消耗，并為快中子堆提供核燃料，在 2020年前后建成大型商用后處理廠是關(guān)鍵核心環(huán)節(jié)。建成年處理800 t重金屬乏燃料規(guī)模是適當(dāng)?shù)?，但與2020年7000萬(wàn)kW核電裝機(jī)規(guī)模相比還稍小。遠(yuǎn)期目標(biāo)主要是在2035年前后實(shí)現(xiàn)快堆核能系統(tǒng)的商化，快堆燃料制備和快堆乏燃料后處理的研究開(kāi)發(fā)應(yīng)與快堆同步進(jìn)行。

5.4突破放射性廢物最小化和安全處置的關(guān)鍵技術(shù)

乏燃料管理和高放廢物處置仍然是核工業(yè)關(guān)鍵的挑戰(zhàn)。必須開(kāi)展利用快堆進(jìn)行放射性廢物擅變研究實(shí)現(xiàn)MA（次婀系核素）和LLFP（長(zhǎng)壽命裂變產(chǎn)物）的徹底焚燒。要積極推進(jìn)高放廢物安全處置的研究，我國(guó)高放廢物處置地下實(shí)驗(yàn)室應(yīng)于2020年建成，爭(zhēng)取在2040—2050年建成地質(zhì)處置庫(kù)并投入運(yùn)行。

【作者單位：中國(guó)核工業(yè)集團(tuán)公司】

（摘自《電網(wǎng)與清潔能源》2010年1期）

·高被引論文摘要·

被引頻次：89

大型集成多功能中子學(xué)計(jì)算與分析系統(tǒng)Visual BUS的研究與發(fā)展

吳宜燦，李靜驚，李瑩，等

中子學(xué)計(jì)算與分析是反應(yīng)堆物理與輻射防護(hù)設(shè)計(jì)、燃料循環(huán)管理優(yōu)化和核安全分析的基礎(chǔ)。在廣泛深入調(diào)研國(guó)內(nèi)外中子學(xué)程序發(fā)展現(xiàn)狀和趨勢(shì)的基礎(chǔ)上，采用國(guó)際上先進(jìn)的中子學(xué)模擬計(jì)算技術(shù)和現(xiàn)代計(jì)算機(jī)軟件技術(shù)，設(shè)計(jì)和研發(fā)了基于網(wǎng)絡(luò)的大型集成多功能中子學(xué)計(jì)算與分析軟件系統(tǒng)Visual BUS，可用于裂變、聚變和各類混合次臨界反應(yīng)堆系統(tǒng)以及加速器等輻射裝置的計(jì)算與分析。一系列國(guó)際基準(zhǔn)校驗(yàn)計(jì)算和實(shí)際應(yīng)用表明了該系統(tǒng)的正確性和有效性。本文重點(diǎn)介紹該系統(tǒng)的研發(fā)概況、技術(shù)特點(diǎn)和測(cè)試與應(yīng)用情況。

中子學(xué)；計(jì)算；建模；可視化；Visual BUS

來(lái)源出版物：核科學(xué)與工程， 2007， 27（4）: 365-373

被引頻次：41

核安全文化的發(fā)展與應(yīng)用

張力

摘要：安全文化已對(duì)核能企業(yè)的安全性產(chǎn)生了重大影響。本文分析了核安全文化產(chǎn)生的背景，介紹了核安全文化在一些國(guó)家和組織應(yīng)用發(fā)展的狀況，提出了推行安全文化過(guò)程中應(yīng)注意的幾個(gè)問(wèn)題，討論了評(píng)價(jià)安全文化績(jī)效的原則。

關(guān)鍵詞：安全文化；核電站；核安全

來(lái)源出版物：核動(dòng)力工程， 1995， 16（5）: 443-446

被引頻次：40

世界核電發(fā)展趨勢(shì)與高溫氣冷堆

吳宗鑫，張作義

摘要：核能的發(fā)展面臨經(jīng)濟(jì)競(jìng)爭(zhēng)力、核安全、核廢物的最終處置及防止核武器材料擴(kuò)散的挑戰(zhàn)。為改善公眾的可接受性，核電廠的安全性進(jìn)一步改進(jìn)。電力市場(chǎng)體制的非管制化改革加劇了電力技術(shù)的競(jìng)爭(zhēng)。環(huán)境保護(hù)意識(shí)增強(qiáng)使核廢物的處置倍受關(guān)注。80年代中期以來(lái)發(fā)展的先進(jìn)輕水堆核電廠如ABWR，System 80+，EPR，AP600等是今后一段時(shí)期內(nèi)商用核電的主力堆型。進(jìn)入2000年之際，美國(guó)能源部正在規(guī)劃發(fā)展第四代先進(jìn)核能系統(tǒng)，目標(biāo)是在2020年或之前，向市場(chǎng)提供經(jīng)過(guò)驗(yàn)證的成熟的第四代核電廠技術(shù)，以替代美國(guó)退役的核電容量。球床高溫氣冷堆被認(rèn)為是第四代先進(jìn)核能系統(tǒng)的優(yōu)選技術(shù)。南非ESKOM電力公司選擇了球床高溫氣冷堆作為今后核電發(fā)展的堆型。清華大學(xué)承擔(dān)設(shè)計(jì)和建設(shè)的10 MW高溫氣冷實(shí)驗(yàn)堆計(jì)劃在2000年內(nèi)臨界。通過(guò)10 MW高溫氣冷堆的建造，我國(guó)已形成了高溫氣冷堆技術(shù)的自主知識(shí)產(chǎn)權(quán)，初步具備了自主設(shè)計(jì)、制造和建造的能力。

關(guān)鍵詞：核能科學(xué)與工程；高溫氣冷堆

來(lái)源出版物：核科學(xué)與工程， 2000， 20（3）: 211-231

被引頻次：38

人因失誤心理背景與核電站安全

張力

摘要：人因失誤是造成核電站事故的主要因素之一，而現(xiàn)場(chǎng)的心理背景在誘發(fā)人因失誤的過(guò)程中起著十分重要的作用。本文分析了人行為時(shí)心理背景的結(jié)構(gòu)，總結(jié)了幾種典型的人誤心理背景。最后指出，消除不利于安全的心理背景之根本途徑是建立企業(yè)安全文化，并提出了核安全文化的基本特征。

關(guān)鍵詞：人的行為；人因失誤；心理因素

來(lái)源出版物：核動(dòng)力工程， 1992， 13（5）: 27-30

被引頻次：33

切爾諾貝利事故及其影響與教訓(xùn)

胡遵素

摘要：本文從核安全與輻射防護(hù)的角度出發(fā)，根據(jù)幾年來(lái)國(guó)際的研究與報(bào)道以及現(xiàn)場(chǎng)訪問(wèn)所了解的情況，對(duì)前蘇聯(lián)切爾諾貝利核電站事故發(fā)生的原因、影響及其教訓(xùn)進(jìn)行了簡(jiǎn)要回顧。內(nèi)容包括對(duì)切爾諾貝利核電站的簡(jiǎn)單描述、事故發(fā)生的過(guò)程、事故后的應(yīng)急行動(dòng)與防護(hù)措施、健康與環(huán)境影響以及事故的原因與經(jīng)驗(yàn)教訓(xùn)。從安全角度看，該電站的型反應(yīng)堆的空泡正反應(yīng)性系數(shù)、反應(yīng)性余量不足、控制棒從最高位置開(kāi)始下落時(shí)有一個(gè)反應(yīng)性增長(zhǎng)區(qū)以及沒(méi)有有效的圍封等是在設(shè)計(jì)上使此次事故得以發(fā)生并釀成災(zāi)難性后果的根本原因。操作人員把幾個(gè)“極不可能事件”組合在一起，是引發(fā)事故的直接“導(dǎo)火線”。這次事故暴露的最大問(wèn)題是前蘇聯(lián)在核安全管理方面的缺陷。筆者認(rèn)為，提高核能安全性的關(guān)鍵在于健全管理體制和提高安全文化水平。

關(guān)鍵詞：核電站；事故；切爾諾貝利；核安全；設(shè)計(jì)；管理；安全文化

來(lái)源出版物：輻射防護(hù)， 1994， 14（5）: 321-335

被引頻次：30

大亞灣核電站的核安全文化建設(shè)探討

陸瑋，唐炎釗

摘要：論述了核電站管理中安全文化的概念及安全文化的發(fā)展階段。重點(diǎn)分析了大亞灣核安全文化形成的背景及過(guò)程，闡述了大亞灣核安全文化的核心理念，提出了核電站安全文化指標(biāo)，總結(jié)了大亞灣核電站實(shí)施核安全文化的主要措施 ；描述了透明度的普及，并對(duì)大亞灣核電站核安全文化實(shí)施的效果進(jìn)行了系統(tǒng)分析。

關(guān)鍵詞：大亞灣；核電站；核安全文化；企業(yè)文化建設(shè)

來(lái)源出版物：核科學(xué)與工程， 2004， 24（3）: 205-210

被引頻次：25

國(guó)內(nèi)外核電發(fā)展?fàn)顩r及相關(guān)問(wèn)題

劉長(zhǎng)欣

摘要：介紹國(guó)內(nèi)外核電發(fā)展的最新情況，并就與核電有關(guān)的若干問(wèn)題進(jìn)行討論。中國(guó)是世界主要的核大國(guó)，但核電對(duì)我國(guó)的電力貢獻(xiàn)還很少，僅占全國(guó)發(fā)電量的1.43%。國(guó)家主管部門將于近期批準(zhǔn)新的核電項(xiàng)目，并有可能就 2020年前的核電發(fā)展做出規(guī)劃，中國(guó)的核電發(fā)展即將步入快速、穩(wěn)定的發(fā)展之路。

關(guān)鍵詞：核電；可用率；自主化；核安全；高放廢物；核擴(kuò)散

來(lái)源出版物：中國(guó)電力， 2003， 36（9）: 27-33

被引頻次：23

基于BP神經(jīng)網(wǎng)絡(luò)的核安全文化星級(jí)評(píng)價(jià)體系

焦曉佑，宋守信，吳俊勇

摘要：為了加強(qiáng)核電安全文化建設(shè)，本文提出了一種對(duì)核電安全文化進(jìn)行科學(xué)、全面評(píng)價(jià)的方法。并根據(jù)核電安全文化的特點(diǎn)，以SMART準(zhǔn)則為依據(jù)，從安全意識(shí)、安全價(jià)值觀、安全行為、安全現(xiàn)狀等方面確立了24項(xiàng)安全文化評(píng)價(jià)指標(biāo)，提出了安全文化星級(jí)劃分標(biāo)準(zhǔn)，并在Visual Basic 6.0平臺(tái)上建立了基于BP神經(jīng)網(wǎng)絡(luò)的安全文化星級(jí)評(píng)價(jià)體系。通過(guò)泛化能力測(cè)試，該體系能準(zhǔn)確地評(píng)價(jià)出核電安全文化發(fā)展到了什么階段，具有良好的可行性和有效性，操作簡(jiǎn)便，易于推廣。

關(guān)鍵詞：核電安全文化；星級(jí)評(píng)價(jià)；BP神經(jīng)網(wǎng)絡(luò)

來(lái)源出版物：核動(dòng)力工程， 2007， 28（1）: 105-114

被引頻次：22

核能與核安全：日本福島核事故分析與思考

陳達(dá)

摘要：核能是當(dāng)今人類社會(huì)不可或缺的重要能源，日本福島核事故危害巨大并再次將核能利用推向風(fēng)口浪尖。本文從世界核能發(fā)展及中國(guó)能源需求出發(fā)，闡述了發(fā)展核能的重要性和必要性；對(duì)日本福島核事故基本情況進(jìn)行了簡(jiǎn)單介紹，并對(duì)事故原因作深入分析；從福島核事故對(duì)世界核電發(fā)展的影響、中國(guó)核電發(fā)展規(guī)劃、核電站選址、核電站設(shè)計(jì)運(yùn)行、核電技術(shù)研發(fā)、核安全文化及核電人才培養(yǎng)等方面進(jìn)行了分析思考，吸取經(jīng)驗(yàn)、總結(jié)教訓(xùn)，切實(shí)把核安全擺在核電發(fā)展首位。

關(guān)鍵詞：核能；核安全；福島核事故；分析與思考

來(lái)源出版物：南京航空航天大學(xué)學(xué)報(bào)， 2012， 44（5）: 597-602

被引頻次：19

輻射防護(hù)劑研究現(xiàn)狀及其進(jìn)展

趙斌，張軍帥，劉培勛

摘要：近年來(lái)，隨著世界核安全形勢(shì)的緊張以及放射治療的迅速發(fā)展，尤其是日本福島核電站發(fā)生核泄漏事故后，輻射防護(hù)劑的研究再一次引起人們的關(guān)注。輻射損傷防治藥物是救治與防護(hù)最為有效和直接的手段之一，在接觸放射性物質(zhì)前使用，能預(yù)防射線對(duì)人體的損傷，受到照射后使用，能減輕放射病的臨床癥狀，促進(jìn)早期恢復(fù)。自1949年P(guān)att報(bào)道半胱氨酸能預(yù)防急性放射損傷以來(lái)的半個(gè)多世紀(jì)里，很多國(guó)家對(duì)輻射損傷的藥物預(yù)防進(jìn)行了比較詳細(xì)的、深入的研究。本工作簡(jiǎn)述了輻射防護(hù)劑的研究簡(jiǎn)史、化學(xué)分類及其作用機(jī)理，并就其研究方向作了展望。

關(guān)鍵詞：輻射防護(hù)劑；研究簡(jiǎn)史；化學(xué)分類；作用機(jī)制

來(lái)源出版物：核化學(xué)與放射化學(xué)， 2012， 34（1）: 8-13

被引頻次：990

ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology

Chadwick， M. B.； Oblozinsky， P； Herman， M； et al.

Abstract: We describe the next generation general purpose Evaluated Nuclear Data File， ENDF/B-VIL.0， of recommended nuclear data for advanced nuclear science and technology applications. The library， released by the U.S. Cross Section Evaluation Working Group（CSEWG）in December 2006， contains data primarily for reactions with incident neutrons， protons， and photons on almost 400 isotopes， based on experimental data and theory predictions. The principal advances over the previous ENDF/B-VI library are the following:（1） New cross sections for U， Pu， Th； Np and Am actinide isotopes， with improved performance in integral validation criticality and neutron transmission benchmark tests；（2） More precise standard cross sections for neutron reactions on H，6Li，10B， An and for235，238U fission， developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation（WPEC）:（3） Improved thermal neutron scattering:（4） An extensive set of neutron cross sections on fission products developed through a WPEG collaboration；（5） A large suite of photonuclear reactions；（6） Extension of many neutron-and protoninduced evaluations up to 150 MeV:（7） Many new light nucleus neutron and proton reactions；（8） Post-fission beta-delayed photon decay spectra:（9） New radioactive decay data:（10） New methods for uncertainties and covariances， together with covariance evaluations for some sample cases； and（11） New actinide fission energy deposition. The paper provides an overview of this library；consisting of 14 sublibraries in the same ENDF-6 format as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries， focusing on neutron reactions. Extensive validation， using radiation transport codes to simulate measured critical assemblies， show major improvements:（a） The Ion-standing underprediction of low enriched uranium thermal assemblies is removed；（b） The238U and208Pb and9Be reflector biases in fast systems are largely removed；（c） ENDF/B-VI.8 good agreement for simulations of thermal high-enriched uranium assemblies is preserved；（d） The underprediction of fast criticality of235，238U and239Pu assemblies is removed； and（e） The intermediate spectrum critical assemblies are predicted more accurately. We anticipate that the new library will play an importanrole in nuclear technology applications， including transport simulations supporting national security， nonproliferation， advanced reactor and fuel cycle concepts， criticality safety， fusion， medicine， space applications， nuclear astrophysics， and nuclear physics facility design. The ENDF/B-VII.0 library is archived at the National Nuclear Data Center， BNL， and can be retrieved from www.nndc.bnl.gov.

來(lái)源出版物：Nuclear Data Sheets， 2006， 107（12）: 2931-3059

被引頻次：166

Materials challenges in nuclear energy

Zinkle， S.J； Was， GS

Abstract: Nuclear power currently provides about 13% of electrical power worldwide， and has emerged as a reliable baseload source of electricity. A number of materials challenges must be successfully resolved for nuclear energy to continue to make further improvements in reliability， safety and economics. The operating environment for materials in current and proposed future nuclear energy systems is summarized， along with a description of materials used for the main operating components. Materials challenges associated with power uprates and extensions of the operating lifetimes of reactors are described. The three major materials challenges for the current and next generation of water-cooled fission reactors are centered on two structural materials aging degradation issues（corrosion and stress corrosion cracking of structural materials and neutron-induced embrittlement of reactor pressure vessels）， along with improved fuel system reliability and accident tolerance issues. The major corrosion and stress corrosion cracking degradation mechanisms for light-water reactors are reviewed. The materials degradation issues for the Zr alloy-clad UO2 fuel system currently utilized in the majority of commercial nuclear power plants are discussed for normal and off-normal operating conditions. Looking to proposed future（Generation IV） fission and fusion energy systems，there are five key bulk radiation degradation effects（low temperature radiation hardening and embrittlement；radiation-induced and -modified solute segregation and phase stability； irradiation creep； void swelling； and high-temperature helium embrittlement） and a multitude of corrosion and stress corrosion cracking effects（including irradiation-assisted phenomena） that can have a major impact on the performance of structural materials.

Keywords: nuclear materials； radiation effects； stress corrosion cracking； structural alloys（steels and nickelbase）； nuclear fuels

來(lái)源出版物：ACTA Materialia， 2013， 61（3）: 735-758

被引頻次：118

On the relative importance of input factors in mathematical models: Safety assessment for nuclear waste disposal

Saltelli， A； Tarantola， S

Abstract: This article deals with global quantitative sensitivity analysis of the Level E model， a computer code used in safety assessment for nuclear waste disposal. The Level E code has been the Subject of two international benchmarks of risk assessment codes and Monte Carlo methods and is well known in the literature. We discuss the Level E model with reference to two different settings. In the first setting， the objective is to find the input factor that drives most of the output variance. In the second setting，we strive to achieve a preestablished reduction in the variance of the model output by fixing the smallest number of factors. The emphasis of this work is on how to define the concept of importance in an unambiguous way and how to assess it in the simultaneous occurrence of correlated input factors and non-additive models.

Keywords: analysis of variance； correlated input；nonadditive model； sensitivity analysis

來(lái)源出版物：Journal of the American Statistical Association， 2002， 97（459）: 702-709

被引頻次：95

Assessing safety culture in nuclear power stations

Lee， T； Harrison， K

Abstract: Definitions of safety culture abound， but they variously refer to the safety-related values， attitudes，beliefs， risk perceptions and behaviours of all employees. This assembly may seem too inclusive to be meaningful，but each represents a different level of processing and the choice for measurement（or intervention） is more pragmatic than theoretical. The present study addresses mainly attitudes， but also reported behaviours. This is done using a 120-item questionnaire covering eight domains of safety in three nuclear power stations. Principal components analysis yields 28 factors - all but four of which are correlated with one or more of nine criteria of accident history. Differences by gender， age， shifts/days and work areas are revealed， but these are confounded by type of job and ANOVAS are applied to clarify the main sources of variation. The effects on safety culture of a number of organisational components are also explored. For example the role of safety in team briefings，management style， work pressure versus safety， etc. It is concluded that personnel safety surveys can usefully be applied to deliver a multi-perspective. comprehensive and economical assessment of the current state of a safety culture and also to explore the: dynamic inter-relationships of its working parts.

Keywords: safety culture； nuclear accidents； nuclear employees； nuclear power stations； safety attitudes

來(lái)源出版物：Safety Science， 2000， 34（1-3）: 61-97

被引頻次：88

Optimization of standby safety system maintenance schedules in nuclear power plants

Harunuzzaman， M； Aldemir， T

Abstract: A methodology and a computational scheme are developed based on dynamic programming（DP） to find the minimum cost maintenance schedule for nuclear power plant standby safety systems. Surveillance and testing are assumed to return the component to as-good-as-new condition whether accompanied by restorative maintenance only or full repair or replacement. The methodology defines component state as the number of unsurveilled and untested maintenance intervals or stages， and the optimization process is decomposed into（a） feasibility screening and（b） DP search. This approach achieves a significant reduction in the state space over which the DP search is to be performed. The application of the scheme is demonstrated on the ten-component high-pressure injection system of a pressurized water reactor. This demonstration indicates that the scheme is viable and efficient and particularly suited to exploit any economies of scale and scope that may be present.

Keywords:dynamicprogramming；maintenance optimization； reliability-centered maintenance

來(lái)源出版物：Nuclear Technology， 1996， 113（3）: 354-367

被引頻次：81

Deliberately small reactors and the second nuclear era

Ingersoll， D. T.

Abstract: Smaller sized nuclear reactors were instrumental during the pioneering days of commercial nuclear power to facilitate the development and demonstration of early reactor technologies and to establish operational experience for the fledgling nuclear power industry. As the U.S.embarks on its “second nuclear era，” the question becomes: Will smaller sized plants have a significant role in meeting the nation’s needs for electricity and other energy demands？ A brief review of our nuclear history is presented relative to plant size considerations， followed by a review of several commonly cited benefits of small reactors. Several “deliberately small” designs currently being developed in the U.S. are briefly described， as well as some of the technical and institutional challenges faced by these designs. Deliberately small reactors offer substantial benefits in safety. security， operational flexibilities and economics， and they are well positioned to figure prominently in the second nuclear era.

Keywords: small medium reactors； deliberately small reactors； second nuclear era； nuclear renaissance； new reactor designs

來(lái)源出版物：Progress in Nuclear Energy， 2009， 51（4-5）: 589-603

被引頻次：48

Assessment of safety-critical software in nuclear-power-plants

Parnas， DL； Asmis， GJK； Madey， J

Abstract: This article outlines an approach to the design，documentation， and evaluation of computer systems. We believe that this approach allows the use of software in many safety-critical applications because it enables the systematic comparison of the program behavior with the engineering specifications of the computer system. Many of the ideas in this article have been used by the Atomic Energy Control Board of Canada（AECB） in its safety assessment of the software for the shutdown systems of the Darlington Nuclear Power Generating Station. The four main elements of this approach follow:（1） Formal Documentation of Software Requirements: Most of the details of a complex environment can be ignored by system implementers and reviewers if they are given a complete and precise statement of the behavioral requirements for the computer system. We describe five mathematical relations that specify the requirements for the software in a computerized control system.（2） Design and Documentation of the Module Structure: Complexity caused by interactions between separately written components can be reduced by applying “Information Hiding”（also known as Data Abstraction， Abstract Data Types， and Object-Oriented Programming） if the interfaces are precisely and completely documented.（3） Program Function Documentation: Software executions are lengthy sequences of state changes described by complex algorithms. The effects of these execution sequences can be precisely specified and documented with tabular representations of the program functions discussed by Mills and others. Also，large programs can be decomposed and presented as a collection of well- documented smaller programs.（4）“Tripod Approach” to Assessment: There are three basic approaches to the assessment of complex software products:（i） testing，（ii） systematic inspection， and（iii） certification of people and processes. Assessment of a complex system cannot depend on any one of these alone. The approach used on the Darlington shutdown software， which included systematic inspection as well as both planned and statistically designed random testing， is outlined. Certification of software engineers remains a difficult issue.

來(lái)源出版物：Nuclear Safety， 1991， 32（3）: 189-198

被引頻次：35

Nanofluids for enhanced economics and safety of nuclear reactors: An evaluation of the potential features， issues， and research gaps

Buongiorno， Jacopo； Hu， Lin-Wen； Kim， Sung Joong

Abstract: Nanofluids are engineered colloidal suspensions of nanoparticles in water and exhibit a very significant enhancement（up to 200%） of the boiling critical heat flux（CHF） at modest nanoparticle concentrations（＜= 0.1% by volume）. Since CHF is the upper limit of nucleate boiling，such enhancement offers the potential for major performance improvement in many practical applications that use nucleate boiling as their prevalent heat transfer mode. The Massachusetts Institute of Technology is exploring the nuclear applications of nanofluids， specifically the following three: 1. main reactor coolant for pressurized water reactors（PWRs）. 2. coolant for the emergency core cooling system（ECCS） of both PWRs and boiling water reactors. 3. coolant for in-vessel retention of the molten core during severe accidents in high-power-density light water reactors. The main features and potential issues of these applications are discussed. The first application could enable significant power uprates in current and future PWRs， thus enhancing their economic performance. Specifically， the use of nanofluids with at least 32% higher CHF could enable a 20% power density uprate in current plants without changing the fuel assembly design and without reducing the margin to CHF The nanoparticles would not alter the neutronic performance of the systemsignificantly. A RELAP5 analysis of the large-break loss-of-coolant accident in PWRs has shown that the use of a nanofluid in the ECCS accumulators and safety injection can increase the peak-cladding-temperature margins（in the nominal-power core） or maintain them in uprated cores if the nanofluid has a higher post-CHF heat transfer rate. The third application can increase the margin to vessel breach by 40% during severe accidents in high-power density systems such as Westing house AP1000 and the Korean APR1400. In summary， the use of nanofluids in nuclear systems seems promising； however， several significant gaps are evident， including， most notably， demonstration of thenanofluidthermal-hydraulicperformanceat prototypical reactor conditions and the compatibility of the nanofluid chemistry with the reactor materials. These gaps must be closed before any of the aforementioned applications can be implemented in a nuclear power plant.

Keywords: nanofluids； reactor coolant； critical heat flux

來(lái)源出版物：Nuclear Technology， 2008， 162（1）: 80-91

被引頻次：33

Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties

Nutt， WT； Wallis， GB

Abstract: We apply methods from order statistics to the problem of satisfying regulations that specify individual criteria to be met by each of a number of outputs， k， from a computer code simulating nuclear accidents. The regulations are assumed to apply to an ‘extent’， gamma（k），（such as 95%） of the cumulative probability distribution of each output， k， that is obtained by randomly varying the inputs to the code over their ranges of uncertainty. We use a 'bracketing' approach to obtain expressions for the confidence， 6， or probability that these desired extents will be covered in N runs of the code. Detailed results are obtained for k = 1， 2， 3， with equal extents， gamma， and are shown to depend on the degree of correlation of the outputs. They reduce to the proper expressions in limiting cases. These limiting cases are also analyzed for an arbitrary number of outputs， k. The bracketing methodology is contrasted with the traditional ‘coverage’approach in which the objective is to obtain a range of outputs that enclose a total fraction， gamma， of all possible outputs， without regard to the extent of individual outputs. For the case of two outputs we develop an alternate formulation and show that the confidence， 6， depends on the degree of correlation between outputs. The alternate formulation reduces to the single output case when the outputs are so well correlated that the coverage criterion is always met in a single run of the code if either output lies beyond an extent gamma， it reduces to Wilks’ expression for un-correlated variables when the outputs are independent， and it reduces to Wald’s result when the outputs are so negatively correlated that the coverage criterion could never be met by the two outputs of a single run of the code. The predictions of both formulations are validated by comparison with Monte Carlo simulations.

Keywords: nuclear safety； outputs of codes； regulations；non-parametric methods； bracketing； coverage； confidence來(lái)源出版物：Reliability Engineering & System Safety，

2004， 83（1）: 57-77

被引頻次：32

Scale 6: Comprehensive nuclear safety analysis code system

Bowman， SM

Abstract: Version 6 of the Standardized Computer Analyses for Licensing Evaluation（SCALE） computer software system developed at Oak Ridge National Laboratory， released in February 2009， contains significant new capabilities and data for nuclear safety analysis and marks an important update for this software package， which is used worldwide. This paper highlights the capabilities of the SCALE system， including continuous-energy flux calculations for processing multigroup problem-dependent cross sections，ENDF/B-VII continuous-energy and multigroup nuclear cross-section data， continuous-energy Monte Carlo criticality safety calculations， Monte Carlo radiation shielding analyses with automated three-dimensional variance reduction techniques， one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations， twoand three-dimensional lattice physics depletion analyses，fast and accurate source terms and decay heat calculations，automated burnup credit analyses with loading curve search， and integrated three-dimensional criticality accident alarm system analyses using coupled Monte Carlo criticality and shielding calculations.

Keywords: reactor physics； sensitivity； uncertainty；criticality safety

來(lái)源出版物：Nuclear Technology， 2011， 174（2）: 126-148

·推薦論文摘要·

核電廠工程結(jié)構(gòu)抗震研究進(jìn)展

孔憲京，林皋

摘要：當(dāng)前以及今后相當(dāng)長(zhǎng)一段時(shí)期，核電都將是中國(guó)積極發(fā)展的能源形式之一，保障核電安全是確保核電工程建設(shè)順利實(shí)施和安全運(yùn)營(yíng)的關(guān)鍵。然而，中國(guó)幅員廣闊，地質(zhì)條件差異大，海域自然條件復(fù)雜；同時(shí)，中國(guó)地震活動(dòng)范圍廣、強(qiáng)度大、頻度高，基于標(biāo)準(zhǔn)化設(shè)計(jì)的核電工程結(jié)構(gòu)在建設(shè)過(guò)程中面臨著諸多問(wèn)題。尤其是2011年日本大地震導(dǎo)致的福島核電事故的教訓(xùn)，對(duì)核電工程的抗震安全提出了新的問(wèn)題。結(jié)合大連理工大學(xué)十幾年來(lái)在解決我國(guó)核電工程結(jié)構(gòu)抗震安全中的關(guān)鍵問(wèn)題，以及在“地震作用下核電廠工程結(jié)構(gòu)的功能失效機(jī)理及抗震安全評(píng)價(jià)”研究中所取得若干進(jìn)展進(jìn)行綜述性介紹，主要包括核島地基抗震適應(yīng)性研究和核島安全相關(guān)工程結(jié)構(gòu)抗震防災(zāi)研究。

關(guān)鍵詞：核電廠；地基適應(yīng)性；取排水構(gòu)筑物；安全殼；抗震安全評(píng)價(jià)

來(lái)源出版物：中國(guó)工程科學(xué)， 2013， 15（4）: 62-74

福島核事故后核電廠安全改進(jìn)行動(dòng)分析

張琳，李文宏，楊紅義

摘要：介紹了福島核事故后世界上主要核電國(guó)家相繼開(kāi)展的核電廠安全檢查、再評(píng)價(jià)行動(dòng)，并得出相應(yīng)的檢查和測(cè)試結(jié)論。法國(guó)、美國(guó)和中國(guó)等國(guó)家分別提出了福島核事故后改進(jìn)核電廠安全的建議、要求和行動(dòng)，并制定了具體工程措施：在極端外部事件的設(shè)防，嚴(yán)重事故預(yù)防和緩解，水、電、通風(fēng)實(shí)體改進(jìn)，限制嚴(yán)重事故下的放射性釋放和應(yīng)急準(zhǔn)備等主要方面開(kāi)展的安全改進(jìn)行動(dòng)，將會(huì)提高核電廠的安全水平并提升緩解嚴(yán)重事故的能力。反思福島核事故，總結(jié)福島核事故對(duì)核電安全技術(shù)改進(jìn)的促進(jìn)作用，對(duì)未來(lái)核電安全技術(shù)的發(fā)展進(jìn)行了展望。

關(guān)鍵詞：福島核事故；核電廠；核安全；改進(jìn)行動(dòng)

來(lái)源出版物：原子能科學(xué)技術(shù)， 2014， 48（3）: 486-491

我國(guó)內(nèi)陸核電發(fā)展的環(huán)境風(fēng)險(xiǎn)可控性探析

潘自強(qiáng)，趙成昆，陳曉秋，等

摘要：闡述了內(nèi)陸發(fā)展核電所關(guān)注的廠址安全、環(huán)境保護(hù)的幾個(gè)問(wèn)題。分析表明，發(fā)展內(nèi)陸核電是我國(guó)綠色低碳能源發(fā)展的重要戰(zhàn)略選擇，內(nèi)陸核電的核安全是有保障的，環(huán)境風(fēng)險(xiǎn)可控，我國(guó)啟動(dòng)內(nèi)陸核電建設(shè)的條件已經(jīng)成熟。

關(guān)鍵詞：內(nèi)陸核電；核安全；環(huán)境風(fēng)險(xiǎn)；綠色低碳

來(lái)源出版物：環(huán)境保護(hù)， 2014， 42（1）: 10-14

核化學(xué)與放射化學(xué)的研究進(jìn)展

張生棟，丁有錢，顧忠茂

摘要：在我國(guó)核能快速發(fā)展的新形勢(shì)下，新型核能資源的開(kāi)發(fā)、乏燃料后處理、放射性廢物處理與處置等核燃料循環(huán)化學(xué)研究日益活躍。隨著科學(xué)技術(shù)的不斷發(fā)展，離子加速器、反應(yīng)堆、各種類型的探測(cè)器和分析設(shè)備、以及計(jì)算機(jī)技術(shù)等的發(fā)展，核化學(xué)與放射化學(xué)研究的范圍和成果在不斷擴(kuò)展和增加，如核安全、環(huán)境放射化學(xué)、放射分析化學(xué)、放射性藥物與標(biāo)記化合物等，研究成果對(duì)于國(guó)防建設(shè)、核能發(fā)展、核技術(shù)應(yīng)用等方面具有重要支撐作用。本文綜述了近年來(lái)國(guó)內(nèi)在上述領(lǐng)域所取得的研究進(jìn)展。

關(guān)鍵詞：核燃料循環(huán)化學(xué)；核化學(xué)；放射化學(xué)；環(huán)境放射化學(xué)；放射性藥物化學(xué)；核安全；核技術(shù)應(yīng)用

來(lái)源出版物：化學(xué)通報(bào)， 2014， 77（7）: 660-669

以核安全文化引領(lǐng)核能與核技術(shù)利用事業(yè)安全、健康、可持續(xù)發(fā)展——《核安全文化政策聲明》解讀

郭承站

摘要：《核安全文化政策聲明》（以下簡(jiǎn)稱《聲明》）是我國(guó)政府關(guān)于核安全文化的首個(gè)政策聲明。文章對(duì)新形勢(shì)下加強(qiáng)核安全文化建設(shè)的必要性、《聲明》對(duì)推動(dòng)核安全文化建設(shè)的深遠(yuǎn)意義、良好核安全文化的八大特性、全行業(yè)核安全文化建設(shè)的要求等進(jìn)行了深入分析和解讀，并對(duì)持續(xù)推進(jìn)核安全文化提出了相關(guān)倡議。

關(guān)鍵詞：核安全文化；核安全觀；核能；核技術(shù)利用

來(lái)源出版物：環(huán)境保護(hù)， 2015， 43（6）: 12-15

核電廠環(huán)境風(fēng)險(xiǎn)評(píng)價(jià)框架及方法

陳妍，鄭鵬，陳海英，等

摘要：目前核電廠風(fēng)險(xiǎn)評(píng)價(jià)技術(shù)分為核事故風(fēng)險(xiǎn)評(píng)價(jià)及非人類物種電離輻射防護(hù)評(píng)價(jià)。為發(fā)展一個(gè)包括非人類物種防護(hù)在內(nèi)的核電廠輻射防護(hù)體系，本文借鑒環(huán)境風(fēng)險(xiǎn)評(píng)價(jià)的關(guān)鍵流程要素，提出包括公眾健康和非人類物種的核電廠環(huán)境風(fēng)險(xiǎn)評(píng)價(jià)框架。在這一框架的危害排序環(huán)節(jié)，對(duì)所選擇的各評(píng)價(jià)終點(diǎn)指標(biāo)采用層次分析法，計(jì)算評(píng)價(jià)終點(diǎn)對(duì)核電廠環(huán)境風(fēng)險(xiǎn)的權(quán)重并進(jìn)行排序，旨在發(fā)現(xiàn)對(duì)環(huán)境風(fēng)險(xiǎn)貢獻(xiàn)較大的評(píng)價(jià)終點(diǎn)并在風(fēng)險(xiǎn)管理中對(duì)其優(yōu)先管理控制。

關(guān)鍵詞：環(huán)境風(fēng)險(xiǎn)評(píng)價(jià)；健康風(fēng)險(xiǎn)；生態(tài)風(fēng)險(xiǎn)

來(lái)源出版物：科技導(dǎo)報(bào)， 2015， 33（4）: 37-43

我國(guó)內(nèi)陸核電的用水安全

張愛(ài)玲，陳曉秋，劉森林，等

摘要：在介紹我國(guó)擬建內(nèi)陸核電機(jī)組的安全設(shè)計(jì)和廠用水系統(tǒng)的基礎(chǔ)上，分析了內(nèi)陸核電的用水需求和保證率要求。結(jié)合我國(guó)水資源條件及水資源論證現(xiàn)狀，對(duì)如何保障內(nèi)陸核電取水水源的可靠性與可行性進(jìn)行了探討，并提出了內(nèi)陸核電用水安全保障措施的建議。

關(guān)鍵詞：內(nèi)陸核電；用水安全；廠用水系統(tǒng)；水源條件；水資源論證

來(lái)源出版物：水文， 2015， 35（3）: 69-73

核安全級(jí)數(shù)字化儀控系統(tǒng)軟件可靠性評(píng)估

劉盈，楊明

摘要：采用核電廠安全審查大綱技術(shù)的分支NUREG-0800 BTP7-14分別建立基于貝葉斯（Bayes）網(wǎng)絡(luò)的階段評(píng)估模型以及綜合評(píng)估模型。在階段評(píng)估模型中，確立8個(gè)階段，通過(guò)13個(gè)一級(jí)指標(biāo)、74個(gè)二級(jí)指標(biāo)、326個(gè)三級(jí)指標(biāo)來(lái)完成對(duì)軟件階段性的實(shí)時(shí)評(píng)估。選用Hugin貝葉斯網(wǎng)絡(luò)分析工具，針對(duì)測(cè)試對(duì)象展開(kāi)預(yù)測(cè)推理及敏感性分析。經(jīng)過(guò)測(cè)試后得到該軟件在生命周期不同階段對(duì)標(biāo)準(zhǔn)的符合程度，經(jīng)綜合評(píng)估模型推理，可得該軟件在標(biāo)準(zhǔn)層面的可靠性指標(biāo)是98.84%。經(jīng)敏感性分析，可以定性地發(fā)現(xiàn)軟件在生存周期中存在的薄弱環(huán)節(jié)，為評(píng)估核安全級(jí)數(shù)字化儀控系統(tǒng)的可靠性和安全性奠定基礎(chǔ)。

關(guān)鍵詞：核安全級(jí)；數(shù)字化儀控系統(tǒng)；軟件可靠性；標(biāo)準(zhǔn)；貝葉斯網(wǎng)絡(luò)

來(lái)源出版物：核動(dòng)力工程， 2016， 37（1）: 143-147

加速器輻射安全評(píng)價(jià)常見(jiàn)問(wèn)題探討

宋培峰，王曉峰，李恩敬

摘要：目的：探討核技術(shù)利用加速器項(xiàng)目輻射安全評(píng)價(jià)應(yīng)關(guān)注的問(wèn)題，并提出對(duì)策。方法：查閱 2014年度國(guó)家核安全局監(jiān)督管理單位相關(guān)加速器輻射安全評(píng)價(jià)申報(bào)材料審查記錄，按照相關(guān)法規(guī)標(biāo)準(zhǔn)要求，對(duì)常見(jiàn)加速器項(xiàng)目申報(bào)材料中存在的問(wèn)題進(jìn)行整理與分析。結(jié)果：2014年度國(guó)家核安全局完成27個(gè)加速器相關(guān)項(xiàng)目輻射安全評(píng)價(jià)審查，約有9個(gè)項(xiàng)目存在執(zhí)行限值模糊，12個(gè)項(xiàng)目存在屏蔽估算過(guò)程不完善，8個(gè)項(xiàng)目存在安全聯(lián)鎖措施描述遺漏、闡述不清晰或設(shè)置不當(dāng)，5個(gè)項(xiàng)目存在放射性產(chǎn)物相關(guān)環(huán)節(jié)描述不充分等問(wèn)題。結(jié)論：建議依據(jù)常見(jiàn)加速器項(xiàng)目應(yīng)用類型不同采取可接受的執(zhí)行限值、多方面完善屏蔽估算、全面合理評(píng)價(jià)安全聯(lián)鎖措施以及優(yōu)化放射性產(chǎn)物的評(píng)價(jià)與管理，明確產(chǎn)物去向。

關(guān)鍵詞：加速器；輻射安全；評(píng)價(jià)；對(duì)策

來(lái)源出版物：中國(guó)職業(yè)醫(yī)學(xué)， 2016， 43（3）: 361-364

A combined deterministic and probabilistic procedure for safety assessment of beyond design basis accidents in nuclear power plant: Application to ECCS performance assessment for design basis LOCA redefinition Kang， Dong Gu； Ahn， Seung-Hoon；

Chang， Soon Heung； et al.

Abstract: The concept and assessment approach of nuclear safety in nuclear power plants（NPPs） have been evolved with the technological progress and the lessons learned from the major events. Recently， studies on the integrated approach of deterministic and probabilistic method have been done. In this study， a combined deterministic and probabilistic procedure（CDPP） is proposed for safety assessment of the beyond design basis accidents（BDBAs）. In the CDPP， the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments， resulting in more reliable values of core damage frequency and conditional core damage probability. To verify applicability of the methodology，performance of the APR-1400 emergency core cooling system is assessed against large break loss of coolant accident（LOCA）， under the premise that LOCAs for any breaks larger than transition break size would be regarded as BDBA. In addition， discussions are made for analysis results for allowable NPP changes of emergency diesel generator start time extension and power uprating. It is concluded that the proposed CDPP is applicable to safety assessment of BDBAs in NPPs without significant erosionof the safety margin.

來(lái)源出版物：Nuclear Engineering and Design， 2013， 260: 165-174

A survey of dynamic methodologies for probabilistic safety assessment of nuclear power plants

Aldemir， Tunc

Abstract: Dynamic methodologies for probabilistic safety assessment（PSA） are defined as those which use a time-dependent phenomenological model of system evolution along with its stochastic behavior to account for possible dependencies between failure events. Over the past 30 years， numerous concerns have been raised in the literature regarding the capability of the traditional static modeling approaches such as the event-tree/fault-tree methodology to adequately account for the impact of process/hardware/software/firmware/human interactions on the stochastic system behavior. A survey of the types of dynamic PSA methodologies proposed to date is presented，as well as a brief summary of an example application for the PSA modeling of a digital feedwater control system of an operating pressurized water reactor. The use of dynamic methodologies for PSA modeling of passive components and phenomenological uncertainties are also discussed.

Keywords: probabilistic safety assessment； probabilistic risk assessment； epistemic uncertainties

來(lái)源出版物：Annals of Nuclear Energy， 2013， 52（S1）: 113-124

Extension of station blackout coping capability and implications on nuclear safety

Volkanovski， Andrija； Prosek， Andrej； et al.

Abstract: The safety of the nuclear power plant depends on the availability of the continuous and reliable sources of electrical energy during all modes of operation of the plant. The station blackout corresponds to a total loss of all alternate current（AC） power as a result of complete failure of both offsite and on-site AC power sources. The electricity for the essential systems during station blackout is provided from the batteries installed in the nuclear power plant. The results of the probabilistic safety assessment show that station blackout is one of the main and frequently the dominant contributor to the core damage frequency. The accident in Fukushima Daiichi nuclear power plants demonstrates the vulnerability of the currently operating nuclear power plants during the extended station blackout events. The objective of this paper is， considering the identified importance of the station blackout initiating event， to assess the implications of the strengthening of the SBO mitigation capability on safety of the NPP. The assessment is done with state-of-art deterministic and probabilistic methods and tolls with application on reference models of nuclear power plants. The U.S. NRC Station Blackout Rule describes procedure for the assessment of the size and capacity of the batteries in the nuclear power plant. The description of the procedure with the application on the reference plant and identified deficiencies in the procedure is presented. The safety analysis is done on reference model of the nuclear power plant. Obtained results show large decrease of the core damage frequency with strengthening of the station blackout mitigation capability. The time extension of blackout coping capability results in the delay of the core heat up for at least the extension time interval. Availability and operation of the steam driven auxiliary feedwater system maintains core integrity up to 72 h after the successful shutdown， even in the presence of the reactor coolant pumps seal leakage. The largest weighted decrease of the core damage frequency considering the costs for the modification is obtained for the modification resulting in extension of the station blackout coping capability. The importance of the common cause failures of the emergency diesel generators for the obtained decrease of the core damage frequency and overall safety of the plant is identified in the obtained results. The results of the analysis support the latest recommendations and expected revisions to the corresponding regulatory requirement by the U.S. Regulatory Commission considering the station blackout mitigation capability.

來(lái)源出版物：Nuclear Engineering and Design， 2013， 255: 16-27

Design of integrated passive safety system（IPSS）for ultimate passive safety of nuclear power plants

Chang， Soon Heung； Kim， Sang Ho； Choi， Jae Young

Abstract: The design concept of integrated passive safety system（IPSS） which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system， passive safety injection system， passive containment cooling system， passive in-vessel retention and cavity flooding system， and filtered venting system with containmentpressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently， all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents（BDBAs） focused on a station black out（SBO）. The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also， the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high， as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore， the operation time of the IPSS can be increased by refilling coolant from the containment outside into integrated passive safety tanks（IPSTs）. The coolant in the IPSTs is used for various functions in accident scenarios. Also， potential problems for the realistic installation of the IPSS are proposed and the solutions to these problems are schematically described. IPSS is the design for the passive safety enhancement in preparation for a loss of AC power. Consequently， it is designed for the supplementation and enhancement of current nuclear power plants， not as a replacement. The specific optimization design for each current or future reactor will be studied as further works.

來(lái)源出版物：Nuclear Engineering and Design， 2013， 260: 104-120

Theory and implementation of nuclear safety system codes - Part I: Conservation equations，flow regimes， numerics and significant assumptions

Roth， Glenn A.； Aydogan， Fatih

Abstract: The design and analysis of the thermal/hydraulic systems of nuclear power plants necessitates system codes that can be used in the analysis of steady-state and transient conditions. Due to the dispersed development of system codes over many laboratories and universities， there are several system codes available for use. Many of the available codes have multiple similar versions developed for specific user needs. The code comparisons provided in the two parts of this article series allow users to select the appropriate system code for their specific problems. In this comparison， the governing equations for mass， momentum and energy conservation are evaluated. It will be shown that the governing equations do riot vary substantially between the codes considered. Most of them utilize a lumped approach with only two fields to represent two phase flow. Two-phase flows are divided into flow regimes based on their appearance and the flow structure. The regimes are used to select appropriate closure relationships to model heat transfer， interfacial drag， and other flow conditions. In addition， major assumptions about the governing and closure equations in these codes are compared and discussed. The most significant of the assumptions is that the governing equations can be discretized in time. The numerical approach of the codes is compared to one another since the numerical approach not only affects the speed of the system codes but also the accuracy of the results. In the second part of this article， the closure relations， their major assumptions， experimental verification and validation are discussed. The results of these articles also guide the development of these system codes， the underlying thermal/hydraulic models， and indicate areas where models must be improved to adequately address issues with new reactor design and development activities.

Keywords: system code comparison； nuclear plant analysis； relap； trace； cathare； athlet

來(lái)源出版物：Progress in Nuclear Energy， 2014， 76: 160-182:

Theory and implementation of nuclear safety system codes - Part II: System code closure relations， validation， and limitations

Roth， Glenn A.； Aydogan， Fatih

Abstract: This is Part II of two articles describing the details of thermal-hydraulic system codes. In this second part of the article series， the system code closure relationships（used to model thermal and mechanical non-equilibrium and the coupling of the phases） for the governing equations are discussed and evaluated. These include several thermal and hydraulic models， such as heat transfer coefficients for various flow regimes， two phase pressure correlations， two phase friction correlations， drag coefficients and interfacial models between the fields.These models are often developed from experimental data. The experiment conditions should be understood to evaluate the efficacy of the closure models. Code verification and validation， including Separate Effects Tests（SETs） and Integral effects tests（IETs） is also assessed. It can be shown from the assessments that the test cases cover a significant section of the system code capabilities， but some of the more advanced reactor designs will push the limits of validation for the codes. Lastly， the limitations of the codes are discussed by considering next generation power plants， such as Small Modular Reactors（SMRs）， analyzing not only existing nuclear power plants，but also next generation nuclear power plants. The nuclear industry is developing new， innovative reactor designs，such as Small Modular Reactors（SMRs）， High-Temperature Gas-cooled Reactors（HTGRs） and others. Sub-types of these reactor designs utilize pebbles，prismatic graphite moderators， helical steam generators，innovative fuel types， liquid metal coolants， and many other design features that may not be fully analyzed by current system codes. This second part completes the series on the comparison and evaluation of the selected reactor system codes by discussing the closure relations， validation and limitations. These two articles indicate areas where the models can be improved to adequately address issues with new reactor design and development.

Keywords: system code comparison； nuclear plant analysis； relap； trace； cathare； athlet

來(lái)源出版物：Progress in Nuclear Energy， 2014， 76: 55-72

Accurate fission data for nuclear safety

Solders， A； Gorelov， D； Jokinen， A； et al.

Abstract: The accurate fission data for nuclear safety（AIFONS） project aims at high precision measurements of fission yields， using the renewed IGISOL mass separator facility in combination with a new high current light ion cyclotron at the University of Jyvaskyla. The 30 MeV proton beam will be used to create fast and thermal neutron spectra for the study of neutron induced fission yields. Thanks to a series of mass separating elements，culminating with the JYFLTRAP Penning trap， it is possible to achieve a mass resolving power in the order of a few hundred thousands. In this paper we present the experimental setup and the design of a neutron converter target for IGISOL. The goal is to have a flexible design. For studies of exotic nuclei far from stability a high neutron flux（1012） neutrons/s） at energies 1-30 MeV is desired while for reactor applications neutron spectra that resembles those of thermal and fast nuclear reactors are preferred. It is also desirable to be able to produce（semi-）monoenergetic neutrons for benchmarking and to study the energy dependence of fission yields. The scientific program is extensive and is planed to start in 2013 with a measurement of isomeric yield ratios of proton induced fission in uranium. This will be followed by studies of independent yields of thermal and fast neutron induced fission of various actinides.

來(lái)源出版物：Nuclear Data Sheets， 2014， 119: 338-341

China’s approach to nuclear safety: From the perspective of policy and institutional system

Mu， Ruimin； Zuo， Jian； Yuan， Xueliang

Abstract: Nuclear energy plays an important role in the energy sector in the world. It has achieved a rapid development during the past six decades and contributes to over 11% of the world’s electricity supply. On the other side， nuclear accidents have triggered substantial debates with a growing public concern on nuclear facilities. Followed by the Fukushima nuclear accident， some developed countries decided to shut down the existing nuclear power plants or to abandon plans to build new ones. Given this background， accelerating the development of nuclear power on the basis of safety in China will make it a bellwether for other countries. China assigns the top priority to the nuclear safety in nuclear energy development and has maintained a good record in this field. The policy and institutional system provide the necessary guarantee for the nuclear energy development and safety management. Furthermore， China’s approach to nuclear safety provides a benchmark for the safe development and utilization of nuclear power. This research draws an overall picture of the nuclear energy development and nuclear safety in China from the policy and institutional perspective.

Keywords: nuclear energy； safety； policy； institution；China

來(lái)源出版物：Energy Policy， 2015， 76: 161-172

Coupling a CFD code with neutron kinetics and pin thermal models for nuclear reactor safety analyses

Chen， Zhao； Chen， Xue-Nong； Rineiski， Andrei； et al.

Abstract: Most system codes are based on the onedimensional lumped-parameter method， which is unsuitabletosimulatemulti-dimensionalthermal-hydraulics problems. CFD method is a good tool to simulate multidimensional thermal-hydraulics phenomena in the nuclear reactor， which can increase the accuracy of analysis results. However， since there is no neutron kinetics model and pin thermal model in current CFD codes， the application of the CFD method in the area of nuclear reactor safety analyses is still limited. Coupling a CFD code with the neutron kinetics model（PKM） and the pin thermal model（PTM） is a good way to use CFD code to simulate multi-dimensional thermal-hydraulics problems of nuclear reactors. The motivation for this work is to develop a CFD/neutron kinetics coupled code named FLUENT/PK for nuclear reactor safety analyses by coupling the commercial CFD code named FLUENT with the point kinetics model（PKM） and the pin thermal model（PTM）. The mathematical models and the coupling method are described and the unprotected transient overpower（UTOP）accident of a liquid metal cooled fast reactor（LMFR） is chosen as an application case. As a general validation， the calculated results are used to compare with that of another multi-physics coupled code named SIMMER-Ill and good agreements are achieved for various characteristic parameters.

Keywords: CFD； neutron kinetics； pin thermal model；safety analysis

來(lái)源出版物：Annals of Nuclear Energy， 2015， 83: 41-49

Exploring the relationship between safety culture and safety performance in US nuclear power operations

Morrow， Stephanie L； Koves， G. Kenneth；Barnes， Valerie E

Abstract: How do nuclear power plant workers， within a single national culture， perceive safety culture within their organizations？ What is the relationship between safety culture and other indicators of safety？ Is the construct of safety culture useful for predicting future plant performance？ These questions were addressed in the current study using a survey administered to a sample of personnel at 97% of the nuclear power plants in the United States， resulting in 2876 responses from 63 nuclear power plant sites. Exploratory and confirmatory factor analysis revealed a multi-factor structure to the safety culture survey. For each nuclear power plant， the mean score for the total survey results and the factor means were correlated with organization-level performance indicators both concurrently and one year following the survey administration.Correlationssuggestedmeaningful，statistically significant relationships between safety culture， as measured by the survey， and multiple nuclear power plant performance indicators. This study presents a unique look at safety culture across the United States nuclear power industry and takes a critical step toward establishing that safety culture is empirically related to safety performance.

Keywords: safety culture； nuclear power； safety performance；humanandorganizationalfactors；organizational culture； safety climate

來(lái)源出版物：Safety Science， 2014， 69（S1）: 37-47

A strategy for the qualification of multi-fluid approaches for nuclear reactor safety

Lucas， D； Rzehak， R； Krepper， E； et al.

Abstract: CFD-simulations for two-phase flows applying the multi -fluid approach are not yet qualified to provide reliable predictions for unknown flows. Among others， one important reason is the missing agreement within the community on closure models to be used. Considering specific phenomena or not， using different models and adjustable constants， most papers presenting a model validation end up with a good agreement with experimental data. However a case by case selection of models and constants does not help to improve the predictive capabilities of such models. For this reason the definition of baseline models considering all known phenomena that could be important is proposed. In such baseline models all parameter have to be defined， i.e.， there are no tuning parameters by definition. Therefore these baseline models have to be applied to many experiments with different complexity. Shortcomings of the models and our physical understanding of the complex flow phenomena have to be identified by detailed analyses on the deviations between experimental data and simulation results. A modification of the baseline model will only be done if it bases on physical considerations and improves the overall performance of the model. This requires a huge effort， but seems to be the only way to improve the situation. In particular more complete experimental data are required. Joint activities on the development of such baseline models are desirable. The paper illustrates this strategy by a baseline model for polydisperse bubbly flows which is presently developed at HZDR.

來(lái)源出版物：Nuclear Engineering and Design， 2016， 266: 2-11

Contributing to the nuclear 3S’s via a methodology aiming at enhancing the synergies between nuclear security and safety

Cipollaro， Antonio； Lomonaco， Guglielmo

Abstract: Nuclear safety， nuclear security and nuclear safeguards regimes have not historically developed at the same pace and surely have not reached the same level of maturity. Nevertheless， these aspects are of special relevance in the current global nuclear energy context when considering the numerous countries that have and will have the legitimate ambition to start a nuclear energy production programme without any or scarce previous nuclear safety， security and safeguards cultures. The future development of nuclear energy exploitation will depend more and more on the convergence of decisions from governments， from the nuclear industry， from utilities， from private and institutional investors as well as from the level of acceptance by the public opinion. Following an in-depth state-of-the-artanalysisandliteraturesearch，a methodological approach focussed on the safety and security connections is presented， as it seems a field where more commonalities and operational aspects could be possibly found and exploited.

Keywords: nuclear security； nuclear safety； 3S；vulnerability； terrorism and sabotage； critical infrastructures

來(lái)源出版物：Progress in Nuclear Energy， 2016， 86: 31-39

Molecular data of mixed metal oxides with importance in nuclear safety

Kovacs， Attila； Konings， Rudy J. M

Abstract: The gas-phase structural and spectroscopic properties of selected mixed metal oxides（Cs2CrO4， Cs2MnO4， Cs2MoO4， Cs2RuO4， BaMoO4， BaMoO3） have been calculated using Density Functional Theory（DFT）. The possible structural isomers have been analyzed and for the found global minima the vibrational（IR， Raman）spectra have been predicted taking into account also anharmonic corrections. The bonding properties have been characterized by means of the Natural Bond Orbital analysis model while the low-lying excited electronic states have been calculated using time-dependent DFT. In order to assess the stability of the target species the dissociation enthalpies have been evaluated.

來(lái)源出版物：Journal of Nuclear Materials， 2016， 477: 134-138

Progress of experimental research on nuclear safety in NPIC

Gong， H； Zan， Y； Peng， C； et al.

Abstract: Two kinds of Generation III commercial nuclear power plants have been developed in CNNC（China National Nuclear Corporation）， one is a small modular reactor ACP100 having an equivalent electric power 100 MW， and the other is HPR1000（once named ACP1000）having an equivalent electric power1000 MW. Both NPPs widely adopted the design philosophy of advanced passive safety systems and considered the lessons from Fukushima Daichi nuclear accident. As the backbone of the R&D of ACP100 and HPR1000， NPIC（nuclear power Institute of China） has finished the engineering verification test of main safety systems， including passive residual heat removal experiments， reactor cavity injection experiments，hydrogencombustionexperiments，andpassive autocatalytic recombiner experiments. Above experimental work conducted in NPIC and further research plan of nuclear safety are introduced in this paper.

來(lái)源出版物：Kerntechnik， 2016， 81（2）: 125-133

編輯：衛(wèi)夏雯

Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties

Nutt， WT； Wallis， GB

We apply methods from order statistics to the problem of satisfying regulations that specify individual criteria to be met by each of a number of outputs， k， from a computer code simulating nuclear accidents. The regulations are assumed to apply to an ‘extent’， gamma（k），（such as 95%） of the cumulative probability distribution of each output， k， that is obtained by randomly varying the inputs to the code over their ranges of uncertainty. We use a ‘bracketing’ approach to obtain expressions for the confidence， 6， or probability that these desired extents will be covered in N runs of the code. Detailed results are obtained for k = 1， 2， 3， with equal extents， gamma， and are shown to depend on the degree of correlation of the outputs. They reduce to the proper expressions in limiting cases. These limiting cases are also analyzed for an arbitrary number of outputs， k. The bracketing methodology is contrasted with the traditional ‘coverage’approach in which the objective is to obtain a range of outputs that enclose a total fraction， gamma， of all possible outputs， without regard to the extent of individual outputs. For the case of two outputs we develop an alternate formulation and show that the confidence， 6， depends on the degree of correlation between outputs. The alternate formulation reduces to the single output case when the outputs are so well correlated that the coverage criterion is always met in a single run of the code if either output lies beyond an extent gamma， it reduces to Wilks’ expression for un-correlated variables when the outputs are independent， and it reduces to Wald’s result when the outputs are so negatively correlated that the coverage criterion could never be met by the two outputs of a single run of the code. The predictions of both formulations are validated by comparison with Monte Carlo simulations.

nuclear safety； outputs of codes； regulations；non-parametric methods； bracketing； coverage； confidence