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

中國核電發展戰略研究

葉奇蓁

大型集成多功能中子學計算與分析系統Visual BUS的研究與發展

吳宜燦，李靜驚，李瑩，等

核能安全

·編者按·

原子的發現和核能的開發利用給人類社會發展帶來新的動力，極大增強人類認識世界和改造世界的能力。核能發展伴隨著核安全風險和挑戰。人類要更好地利用核能、實現更大發展，必須確保核安全、做好核應急。核安全是核能事業持續健康發展的生命線，核應急是核能事業持續健康發展的重要保障。

20世紀50年代中期，中國創建核工業。60多年來，中國致力于和平利用核能事業，發展推動核技術在工業、農業、醫學、環境、能源等領域廣泛應用。中國堅持發展與安全并重原則，執行安全高效發展核電政策，采用最先進的技術、最嚴格的標準發展核電。伴隨著核能事業的發展，核安全與核應急同步得到加強。中國的核設施、核活動始終保持安全穩定狀態，特別是核電安全水平不斷提高。面對核能事業發展新形勢新挑戰，中國核應急在技術、裝備、人才、能力、標準等方面還存在一定不足，這也是其他國家在開發利用核能進程中面臨的共同課題。中國將通過理念創新、科技創新、管理創新，不斷強化國家核應急管理，把核應急提高到新水平。

中國始終把核安全放在和平利用核能事業首要位置，堅持總體國家安全觀，倡導理性、協調、并進的核安全觀，秉持為發展求安全、以安全促發展的理念，始終追求發展和安全2個目標有機融合。半個多世紀以來，中國人民奮發圖強、歷盡艱辛，創建發展核能事業并取得輝煌成就。同時，不斷改進核安全技術，實施嚴格的核安全監管，加強核應急管理，核能事業始終保持良好安全記錄。

核事故影響無國界，核應急管理無小事。總結切爾諾貝利核事故、福島核事故的教訓，中國更加深刻認識到核應急的極端重要性，持續加強和改進核應急準備與響應工作，不斷提升中國核安全保障水平。中國在核應急法律法規標準建設、體制機制建設、基礎能力建設、專業人才培養、演習演練、公眾溝通、國際合作與交流等方面取得巨大進步，既為自身核能事業發展提供堅強保障，也為推動建立公平、開放、合作、共贏的國際核安全應急體系，促進人類共享核能發展成果作出積極貢獻。

本專題得到陳妍教授（環境保護部核與輻射安全中心）的大力支持。

·熱點數據排行·

截至2016年 8月 26日，中國知網（CNKI）和Web of Science（WOS）的數據報告顯示，以“核能（nuclear energy）”“核安全（nuclear safety）”“核安全文化（nuclear safety culture）”“核應急（Nuclear emergency）”為詞條可以檢索到的期刊文獻分別為1166條、9617條。本專題將相關數據按照：研究機構發文數、作者發文數、期刊發文數、被引用頻次進行排行，結果如下。

研究機構發文數量排名（CNKI）

研究機構發文數量排名（WOS）

作者發文數量排名（CNKI）

作者發文數量排名（WOS）

期刊發文數量排名（CNKI）

期刊發文數量排名（WOS）

根據中國知網（CNKI）數據報告，以“核能（nuclear energy）”“核安全（nuclear safety）”“核安全文化（nuclear safety culture）”“核應急（Nuclear emergency）”為詞條可以檢索到的高被引論文排行結果如下。

國內數據庫高被引論文排行

根據Web of Science統計數據，以“核能（nuclear energy）”“核安全（nuclear safety）”“核安全文化（nuclear safety culture）”“核應急（Nuclear emergency）”為詞條可以檢索到的高被引論文排行結果如下。

國外數據庫高被引論文排行

·經典文獻推薦·

基于Web of Science檢索結果，利用Histcite軟件選取LCS（Local Citation Score，本地引用次數）TOP50文獻作為節點進行分析，得到本領域推薦的經典文獻如下。

本領域經典文獻

來源出版物：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.

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

·推薦綜述·

中國核電發展戰略研究

葉奇蓁

1核能在中國能源可持續發展中的地位

1.1中國能源資源狀況分析

中國能源資源有3個基本特點。能源資源品種豐富，但人均占有量較少，在己探明儲量中煤炭占世界人均的56%、石油占11%，天然氣占4.6%。能源資源結構不盡合理，煤炭、水能相對豐富，而優質化石能源（石油）相對不足。能源資源分布與生產力布局不平衡，經濟發達地區在東南沿海，而水力資源在西部和西南部，煤炭主要在北方。

目前，我國能源發展面臨4個基本問題。即經濟社會發展中的能源供需總量平衡問題。長期以煤為主的能源結構，造成的環境、生態問題。西氣東運、北煤南運、西電東輸的能源輸運問題，我國煤炭運輸占鐵路運量的40%，占沿海和長江中下游水運1/3。對國外資源依存的能源供應安全問題。

核電的基本特性決定了在應對能源挑戰中有能力發揮無可替代的重要作用。核電不排放SO2等污染物和溫室氣體CO2，對環境后果實行嚴格管理，因此屬于清潔能源。而核電的安全可靠性正在不斷提高。核電對煤電具有較強經濟競爭力和替代能力，目前二代改進型核電站的電價大都與當地的標桿電價相當。核電燃料運輸量小。因此，我國在現階段發展核電是調整源布局的有效途徑。

1.2中國核能發展的技術路線

我國核能發展的技術路線是走熱堆、快堆、聚變堆三步發展的道路。在近期發展己經成熟的熱中子堆核電站，滿足當前和近期核電發展的需要。第二步發展快中子增殖堆核電站及配套的核燃料循環體系，充分利用鈾資源，實現裂變核能的可持續發展。第三步發展核聚變堆核電站，有望最終解決人類的能源供應問題。

目前，在熱堆核電發展階段，逐步實現由二代向三代過渡。在2020年以前，適度發展我國己經掌握技術的二代改進型壓水堆核電站。抓緊引進三代核電技術的消化吸收再創新，掌握技術、實現自主化，盡快實現三代核電的批量化建設。

1.3核電產業發展的目標

根據有關研究部門的預測，2020年我國電力總裝機將達到15億kW，核電總裝機容量將達到7000萬kW，核電容量占總容量的4.6%，占總發電量的7.0%左右。考慮能源結構調整的要求，2030年我國總發電裝機容量將達到20億kW，核電總裝機容量將達到2億kW，核電裝機容量占10%，占總發電量的15%。2050年我國將進入中等發達國家行列，以人均1.56 kW計算，總發電裝機容量將達到25億kW，核電總裝機容量將達到4億kW，核電占總裝容量的16%，占總發電量的22%。

2中國核電已形成規模化批量化發展格局

我國大陸投入商運的核電機組共有 11臺，總裝機容量為910萬kW，機組負荷因子達85%～92%，各項運行指標均高于世界平均水準，處于世界中上等水平以上。在全球441座核電站中，大多進入前50～60名。即將建成的嶺澳二期核電站和秦山核電二期擴建均進展良好，預期在2010—2011年將陸續投產發電。目前己有22臺二代改進型壓水堆核電站取得了路條，并己有7臺機組澆灌了第一罐混凝土。主設備己實現了批量采購，有的制造廠己簽訂了數臺或十余臺長周期設備。而核電站設計的標準化規范化工作也正在積極進行當中。

當前我國二代改進型壓水堆核電站己具備系列化規模化發展的有利條件。二代改進型壓水堆屬于成熟的堆型，設計經過驗證，自主化程度較高。有相當豐富的自主建設和自主運行經驗，平均建設周期小于5 a。設備國產化率超過70%，除主循環泵（目前己有3家制造廠在研制）外，主要的核電設備己具備堅實的國產化基礎。我國己建成的核電站的運行經驗表明，核電站的運行是安全的，沒有溫室氣體和有害氣體排放，放射性廢物的排放遠低于國家標準。

2.1二代改進型壓水堆核電站自主化能力分析

二代改進型壓水堆核電站隨著技術的發展和運行經驗的反饋，逐步引入新的成熟技術，使核電站的安全性得到進一步的提高。新設計建設的二代改進型壓水堆降低了堆芯功率密度，使熱工安全余量大于15%；加大穩壓器容量，增加了核電站運行的穩定性；增設附加應急柴油發電機系統，提高了供電的可靠性；增設安全殼過濾卸壓排放系統，防止安全殼超壓失效，并防止放射性外泄；應用概率安全分析技術以及風險管理技術，防止核電站出現嚴重事故；引入嚴重事故預防和緩解措施：如非能動氫復合系統防止氫爆、穩壓器卸壓排放系統防止高壓熔堆、田灣核電站還設計了堆芯捕集器用以在堆芯熔融時防止熔融物熔穿透安全殼底板；廣泛采用數字化儀控技術和先進控制室，改善了人機界面；汽輪發電機采用半速機組，提高了出力和熱效。

二代改進型壓水堆核電站在自主設計能力方面，形成了專業配套、結構合理的研究設計隊伍。

在項目管理能力方面，按國際通用項目管理模式管理，己基本與國際接軌。

在設備制造能力方面，3大集團都將基本具備每年提供2～3臺百萬千瓦級機組設備的能力。

在建設安裝能力方面，己經具有4個項目8臺機組的建設實踐。

在營運管理能力方面，根據世界核電運行者協會WANO的9項性能指標，3項進入前1/4的先進行列，有5項超過中值水平，只有1項略低于中值水平。

在安全監管能力方面，建立了與國際接軌的核安全管理和監督的法規制度，具備了全過程全方位監督管理的能力。

2.2大力堆進內陸核電建設

國際上大部分核電站建設在內陸。法國有65.1%的核電站建設在內陸，美國亦有75.7%的核電站建設在內陸。有些內陸國家，比如瑞士，5座核電站都在內陸的江河邊上，5座核電站總發電功率為3220 MW，占總發電量的37%，其他將近60%的發電量由水電提供。因此，國外其他國家的經驗表明，在內陸建核電站是完全可行的。

我國內陸地區經濟有了很大發展，電網容量亦有很大發展，但部分省份同樣存在缺乏煤炭和水力資源。2009年初南方各省發生了大面積、長時間的雪災，造成了廣大地區長時間的斷電，帶來了嚴重的后果。因此，僅依靠遠距離輸電和長途運煤是難以保障用電安全的。這樣，除提高電網的抗災害能力，建設緊急情況下不依賴燃料運輸的核電站是很有必要的。

從安全和環保要求看，內陸核電站和沿海核電站沒有本質的差別。目前成熟的核電站設計和建造技術完全可用到內陸核電。內陸江河流量多半不夠大，可采用冷卻塔閉式循環帶走余熱，以減輕溫排水對環境的影響。目前，百萬千瓦級核電站一機一塔要求塔高200 m，淋水面積16000 m2時，我國己能設計160 m，12000 m2冷卻塔，正在開展超大型冷卻塔的設計。因此按照核電規范選擇的廠址是能夠保證核電站的安全的。

2.2.1放射性液態流出物的排放控制

內陸廠址與沿海廠址相比，液態流出物中要考慮放射性物質到達人體的途徑及飲用水和灌溉等途徑。目前，我國江河、湖泊污染事件屢有發生，國家主管部門和公眾對于河流的排放控制均持高度關注和審慎的態度。核電廠環境輻射防護規定液態流出物排放的放射性總量每年≤200 GBq（不包括氚），URD文件中對先進壓水堆核電站規定每年≤1.85 GBq（不包括氚），EUR文件中對先進壓水堆核電站規定每年≤10 GBq（不包括氟）。從秦山二期2002—2006年統計的數據，年液態流出物排放的放射性總量為2～5 GBq。因此，目前設計的液態流出物處理系統完全能滿足國標要求，而實際運行水平遠低于國標要求，并與先進壓水堆核電站的要求相當。

2.2.2液態放射性流出物排放濃度控制

我國的《生活飲用水衛生標準》（GB57492006）中規定總β放射性小于1 Bq/L。《核動力廠環境輻射防護規定》（GB6249）提出核動力廠排放口下游1 km處受納水體中總月放射性濃度不得超過1 Bq/L，這就是要求在排放口下1 km處滿足生活飲用水標準。GB-14587—修訂版的征求意見稿，提出了100 Bq/L的排放罐出口濃度控制值。因此，經過適當的稀釋，核電廠液態放射性流出物排放濃度就可達到天然放射性本底水平。

內陸核電站由于采用冷卻塔閉式循環帶走余熱，沒有循環冷卻水對放射性廢液的稀釋。濱海壓水堆核電站液態流出物排放的內部實際控制值為≤1000～2000 Bq/L（不包括氚），經循環冷卻水對放射性廢液的稀釋 1000倍后，其濃度己相當低，一般≤1 Bq/L。俄羅斯濱河核電站要求液態流出物排放的濃度控制值為≤18 Bq/L（不包括氚）。所以，改進目前沿海核電站的液態放射性廢物的處理技術，是完全能滿足內陸核電站對液態放射性廢物處理和排放的要求的。

2.2.3液態放射性廢物處理技術

俄羅斯核電站放射性廢液處理采用了雙蒸發器處理系統，處理后的液體再經二級離子交換處理，凈化系數從10E3提高到10E5。美國采用反滲透廢液處理技術，實現廢水回用，以滿足“零液體排放”要求，并可針對某些元素進行高純度凈化或去除。美國 Comanch Peak核電站用于去除放射性，特別是Co膠體，CS和I到監測不到水平，凈化系數達 5.7×104。美國德賴斯登核電站用超級過濾+反滲透+去離子技術處理廢液。內陸核電站的含氟廢水，在廢水處理后，排入冷卻塔循環冷卻水中，通過蒸發向大氣排放。

3積極消化吸收第三代核電技術

3.1第三代核電發展的背景

1979年美國發生的三里島核電站事故和1986年前蘇聯發生的切爾諾貝利核電站事故，使公眾要求進一步提高核電的安全性。1990年EPRI根據主要電力公司意見出版了“電力公司要求文件（URD）”共 3卷。1994年歐洲聯盟出版了“歐洲電力公司要求（EUR）”共 4卷。這些文件對未來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濟要求，涉及各個技術和經濟領域。

第三代核電機組要有更高安全目標。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。第三代核電機組要有更好的經濟性，具體表現在機組額定功率為1000～1500 MWe，可利用因子＞87%，換料周期18～24月，電站壽命60 a，建設周期48～52月，電價要能與聯合循環的天然氣電廠相競爭。因此，第三代核電機組在技術上更先進。

3.2AP1000核電站的特點

AP1000核電站采用非能動安全系統。具體表現在采用非能動安注、多級非能動自動卸壓系統、非能動余熱排放系統和非能動安全殼冷卻系統。AP1000核電站引入了嚴重事故預防和緩解措施，如堆腔淹沒技術、安全殼內氫點火和氫復合系統、堆芯熔融物反應堆壓力容器內保持。同時，AP1000采用雙層安全殼和全數字化儀控系統。采用模塊化施工使建設工期縮短到 48個月。

AP1000核電站的反應堆冷卻劑系統（如圖1所示）采用屏蔽式電泵，取消了機械密封，采用在上部堆芯測量以及大容量穩壓器，焊接結構的堆內構件和壓力容器活性區及法蘭接管段大型整體鍛件。

AP1000核電站的非能動堆芯冷卻系統，不依賴外部電源，采用非能動余熱導出、非能動安全注入以及非能動安全殼冷卻可以保證長時間的安全停堆，還可以保證大于72 h不用操作員干預。

3.3ERP核電站的特點

EPR核電站的主要特點有以下幾個。EPR核電站功率高，達到1500～1700 MWe。采用4通道安全系統和雙層安全殼。引入了嚴重事故預防及緩解措施，如穩壓器卸壓、堆芯撲集器和非能動氫復合器。同時EPR核電站也采用全數字化儀控和模塊化施工。

3.4AP1000的關鍵技術

AP1000的關鍵技術是采用非能動安全系統，特別是非能動安全殼冷卻系統。AP1000核電站引入了嚴重事故的預防和緩解措施，包括自動卸壓系統久（ADS），抑制氫爆的氫復合系統（氫點火器和非能動氫催化復合），以及堆芯熔融物壓力容器內保持（IVR）等技術。同時AP1000核電站大容量屏蔽泵的設計和制造，爆破膜的設計和制造，以及大尺寸園柱形鋼安全殼的設計和建造也存在技術難點和需攻克的關鍵技術。

3.5重視三代核電引進中技術風險經濟風險的規避

AP1000的技術風險主要在于缺少首堆工程整體驗證的實踐證明，AP1000的設計認證尚未真正通過，而且還有一系列涉及安全的設計驗證工作未做，設計方案尚未固化，從美國條件的設計直接移植到中國，還需要作適應性修改。

AP1000核電站也存在一定的經濟風險。最近西屋公司與美國幾個電力公司簽訂在美國新建AP1000的總承包協議，比投資是我國自主建設核電的 2～3倍，也是招標引進時申報的2～3倍。

4鈾資源的保障

我國己探明一定數量鈾資源可以滿足近期核電發展的需要。國內鈾資源勘測有較好發展前景。理論預測鈾礦資源比較豐富，預測鈾資源總量超過幾百萬噸，加之我國相當范圍國土未經詳細勘查，因此擴大老礦區、加強深層勘查，開辟新基地前景看好。我國目前己探明儲量，加上海外采購和合作開采的天然鈾，足以保障2020年核電對天然鈾的需求。因此加大鈾資源的國內勘查力度，同時開拓國外鈾資源的供應，我國核電發展的鈾資源是一定能得到保證的。

從長期來看，到 2030—2050年我國的人口將達到頂峰16億，按平均每人消耗電力1.56 kW來計（相當于發達國家的中等水平），就需要25億kW的電力供應，其中16%為核電（相當于目前世界核電的平均份額），即4億kW的核電。到2050年我國對于天然鈾資源需求相當大，如果核電的比例比16%還要大，則對天然鈾資源的需求將更大。

5開發快中子增殖堆核電站、構建核燃料循環體系

5.1鈉冷快堆SFR

快中子增殖反應堆的主要特點在于它能增殖核燃料，即它每燃耗一個燃料原子，就可以生產出多于一個燃料原子，這樣一來，在理論上說，它可以將全部鈾資源都轉化為可燃燒的燃料并加以利用。采用適當增殖比的快中子堆，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。

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

5.2快中子反應堆在中國的發展

我國己在“十一五”期間建成實驗快中子堆。計劃2020年前后將建成原型快中子堆核電站，通過引進技術建設第一個快中子堆示范工程。2035年前后完成商用快中子堆核電站及核燃料循環系統的建設。此時，不僅可利用0.7% U—235，通過快中子堆增殖，還可利用大量的U—238（經快中子反應堆轉換的Pu）。

5.3加快商用后處理廠的建設和快堆燃料循環技術

的研發

近期目標主要是實現 2025年開式循環向閉式循環轉變，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鍵核心環節。建成年處理800 t重金屬乏燃料規模是適當的，但與2020年7000萬kW核電裝機規模相比還稍小。遠期目標主要是在2035年前后實現快堆核能系統的商化，快堆燃料制備和快堆乏燃料后處理的研究開發應與快堆同步進行。

5.4突破放射性廢物最小化和安全處置的關鍵技術

乏燃料管理和高放廢物處置仍然是核工業關鍵的挑戰。必須開展利用快堆進行放射性廢物擅變研究實現MA（次婀系核素）和LLFP（長壽命裂變產物）的徹底焚燒。要積極推進高放廢物安全處置的研究，我國高放廢物處置地下實驗室應于2020年建成，爭取在2040—2050年建成地質處置庫并投入運行。

【作者單位：中國核工業集團公司】

（摘自《電網與清潔能源》2010年1期）

·高被引論文摘要·

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大型集成多功能中子學計算與分析系統Visual BUS的研究與發展

吳宜燦，李靜驚，李瑩，等

中子學計算與分析是反應堆物理與輻射防護設計、燃料循環管理優化和核安全分析的基礎。在廣泛深入調研國內外中子學程序發展現狀和趨勢的基礎上，采用國際上先進的中子學模擬計算技術和現代計算機軟件技術，設計和研發了基于網絡的大型集成多功能中子學計算與分析軟件系統Visual BUS，可用于裂變、聚變和各類混合次臨界反應堆系統以及加速器等輻射裝置的計算與分析。一系列國際基準校驗計算和實際應用表明了該系統的正確性和有效性。本文重點介紹該系統的研發概況、技術特點和測試與應用情況。

中子學；計算；建模；可視化；Visual BUS

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核安全文化的發展與應用

張力

摘要：安全文化已對核能企業的安全性產生了重大影響。本文分析了核安全文化產生的背景，介紹了核安全文化在一些國家和組織應用發展的狀況，提出了推行安全文化過程中應注意的幾個問題，討論了評價安全文化績效的原則。

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關鍵詞：核能科學與工程；高溫氣冷堆

來源出版物：核科學與工程， 2000， 20（3）: 211-231

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人因失誤心理背景與核電站安全

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劉長欣

摘要：介紹國內外核電發展的最新情況，并就與核電有關的若干問題進行討論。中國是世界主要的核大國，但核電對我國的電力貢獻還很少，僅占全國發電量的1.43%。國家主管部門將于近期批準新的核電項目，并有可能就 2020年前的核電發展做出規劃，中國的核電發展即將步入快速、穩定的發展之路。

關鍵詞：核電；可用率；自主化；核安全；高放廢物；核擴散

來源出版物：中國電力， 2003， 36（9）: 27-33

被引頻次：23

基于BP神經網絡的核安全文化星級評價體系

焦曉佑，宋守信，吳俊勇

摘要：為了加強核電安全文化建設，本文提出了一種對核電安全文化進行科學、全面評價的方法。并根據核電安全文化的特點，以SMART準則為依據，從安全意識、安全價值觀、安全行為、安全現狀等方面確立了24項安全文化評價指標，提出了安全文化星級劃分標準，并在Visual Basic 6.0平臺上建立了基于BP神經網絡的安全文化星級評價體系。通過泛化能力測試，該體系能準確地評價出核電安全文化發展到了什么階段，具有良好的可行性和有效性，操作簡便，易于推廣。

關鍵詞：核電安全文化；星級評價；BP神經網絡

來源出版物：核動力工程， 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

摘要：核能是當今人類社會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口浪尖。本文從世界核能發展及中國能源需求出發，闡述了發展核能的重要性和必要性；對日本福島核事故基本情況進行了簡單介紹，并對事故原因作深入分析；從福島核事故對世界核電發展的影響、中國核電發展規劃、核電站選址、核電站設計運行、核電技術研發、核安全文化及核電人才培養等方面進行了分析思考，吸取經驗、總結教訓，切實把核安全擺在核電發展首位。

關鍵詞：核能；核安全；福島核事故；分析與思考

來源出版物：南京航空航天大學學報， 2012， 44（5）: 597-602

被引頻次：19

輻射防護劑研究現狀及其進展

趙斌，張軍帥，劉培勛

摘要：近年來，隨著世界核安全形勢的緊張以及放射治療的迅速發展，尤其是日本福島核電站發生核泄漏事故后，輻射防護劑的研究再一次引起人們的關注。輻射損傷防治藥物是救治與防護最為有效和直接的手段之一，在接觸放射性物質前使用，能預防射線對人體的損傷，受到照射后使用，能減輕放射病的臨床癥狀，促進早期恢復。自1949年Patt報道半胱氨酸能預防急性放射損傷以來的半個多世紀里，很多國家對輻射損傷的藥物預防進行了比較詳細的、深入的研究。本工作簡述了輻射防護劑的研究簡史、化學分類及其作用機理，并就其研究方向作了展望。

關鍵詞：輻射防護劑；研究簡史；化學分類；作用機制

來源出版物：核化學與放射化學， 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

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

·推薦論文摘要·

核電廠工程結構抗震研究進展

孔憲京，林皋

摘要：當前以及今后相當長一段時期，核電都將是中國積極發展的能源形式之一，保障核電安全是確保核電工程建設順利實施和安全運營的關鍵。然而，中國幅員廣闊，地質條件差異大，海域自然條件復雜；同時，中國地震活動范圍廣、強度大、頻度高，基于標準化設計的核電工程結構在建設過程中面臨著諸多問題。尤其是2011年日本大地震導致的福島核電事故的教訓，對核電工程的抗震安全提出了新的問題。結合大連理工大學十幾年來在解決我國核電工程結構抗震安全中的關鍵問題，以及在“地震作用下核電廠工程結構的功能失效機理及抗震安全評價”研究中所取得若干進展進行綜述性介紹，主要包括核島地基抗震適應性研究和核島安全相關工程結構抗震防災研究。

關鍵詞：核電廠；地基適應性；取排水構筑物；安全殼；抗震安全評價

來源出版物：中國工程科學， 2013， 15（4）: 62-74

福島核事故后核電廠安全改進行動分析

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家相繼開展的核電廠安全檢查、再評價行動，并得出相應的檢查和測試結論。法國、美國和中國等國家分別提出了福島核事故后改進核電廠安全的建議、要求和行動，并制定了具體工程措施：在極端外部事件的設防，嚴重事故預防和緩解，水、電、通風實體改進，限制嚴重事故下的放射性釋放和應急準備等主要方面開展的安全改進行動，將會提高核電廠的安全水平并提升緩解嚴重事故的能力。反思福島核事故，總結福島核事故對核電安全技術改進的促進作用，對未來核電安全技術的發展進行了展望。

關鍵詞：福島核事故；核電廠；核安全；改進行動

來源出版物：原子能科學技術， 2014， 48（3）: 486-491

我國內陸核電發展的環境風險可控性探析

潘自強，趙成昆，陳曉秋，等

摘要：闡述了內陸發展核電所關注的廠址安全、環境保護的幾個問題。分析表明，發展內陸核電是我國綠色低碳能源發展的重要戰略選擇，內陸核電的核安全是有保障的，環境風險可控，我國啟動內陸核電建設的條件已經成熟。

關鍵詞：內陸核電；核安全；環境風險；綠色低碳

來源出版物：環境保護， 2014， 42（1）: 10-14

核化學與放射化學的研究進展

張生棟，丁有錢，顧忠茂

摘要：在我國核能快速發展的新形勢下，新型核能資源的開發、乏燃料后處理、放射性廢物處理與處置等核燃料循環化學研究日益活躍。隨著科學技術的不斷發展，離子加速器、反應堆、各種類型的探測器和分析設備、以及計算機技術等的發展，核化學與放射化學研究的范圍和成果在不斷擴展和增加，如核安全、環境放射化學、放射分析化學、放射性藥物與標記化合物等，研究成果對于國防建設、核能發展、核技術應用等方面具有重要支撐作用。本文綜述了近年來國內在上述領域所取得的研究進展。

關鍵詞：核燃料循環化學；核化學；放射化學；環境放射化學；放射性藥物化學；核安全；核技術應用

來源出版物：化學通報， 2014， 77（7）: 660-669

以核安全文化引領核能與核技術利用事業安全、健康、可持續發展——《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》（以下簡稱《聲明》）是我國政府關于核安全文化的首個政策聲明。文章對新形勢下加強核安全文化建設的必要性、《聲明》對推動核安全文化建設的深遠意義、良好核安全文化的八大特性、全行業核安全文化建設的要求等進行了深入分析和解讀，并對持續推進核安全文化提出了相關倡議。

關鍵詞：核安全文化；核安全觀；核能；核技術利用

來源出版物：環境保護， 2015， 43（6）: 12-15

核電廠環境風險評價框架及方法

陳妍，鄭鵬，陳海英，等

摘要：目前核電廠風險評價技術分為核事故風險評價及非人類物種電離輻射防護評價。為發展一個包括非人類物種防護在內的核電廠輻射防護體系，本文借鑒環境風險評價的關鍵流程要素，提出包括公眾健康和非人類物種的核電廠環境風險評價框架。在這一框架的危害排序環節，對所選擇的各評價終點指標采用層次分析法，計算評價終點對核電廠環境風險的權重并進行排序，旨在發現對環境風險貢獻較大的評價終點并在風險管理中對其優先管理控制。

關鍵詞：環境風險評價；健康風險；生態風險

來源出版物：科技導報， 2015， 33（4）: 37-43

我國內陸核電的用水安全

張愛玲，陳曉秋，劉森林，等

摘要：在介紹我國擬建內陸核電機組的安全設計和廠用水系統的基礎上，分析了內陸核電的用水需求和保證率要求。結合我國水資源條件及水資源論證現狀，對如何保障內陸核電取水水源的可靠性與可行性進行了探討，并提出了內陸核電用水安全保障措施的建議。

關鍵詞：內陸核電；用水安全；廠用水系統；水源條件；水資源論證

來源出版物：水文， 2015， 35（3）: 69-73

核安全級數字化儀控系統軟件可靠性評估

劉盈，楊明

摘要：采用核電廠安全審查大綱技術的分支NUREG-0800 BTP7-14分別建立基于貝葉斯（Bayes）網絡的階段評估模型以及綜合評估模型。在階段評估模型中，確立8個階段，通過13個一級指標、74個二級指標、326個三級指標來完成對軟件階段性的實時評估。選用Hugin貝葉斯網絡分析工具，針對測試對象展開預測推理及敏感性分析。經過測試后得到該軟件在生命周期不同階段對標準的符合程度，經綜合評估模型推理，可得該軟件在標準層面的可靠性指標是98.84%。經敏感性分析，可以定性地發現軟件在生存周期中存在的薄弱環節，為評估核安全級數字化儀控系統的可靠性和安全性奠定基礎。

關鍵詞：核安全級；數字化儀控系統；軟件可靠性；標準；貝葉斯網絡

來源出版物：核動力工程， 2016， 37（1）: 143-147

加速器輻射安全評價常見問題探討

宋培峰，王曉峰，李恩敬

摘要：目的：探討核技術利用加速器項目輻射安全評價應關注的問題，并提出對策。方法：查閱 2014年度國家核安全局監督管理單位相關加速器輻射安全評價申報材料審查記錄，按照相關法規標準要求，對常見加速器項目申報材料中存在的問題進行整理與分析。結果：2014年度國家核安全局完成27個加速器相關項目輻射安全評價審查，約有9個項目存在執行限值模糊，12個項目存在屏蔽估算過程不完善，8個項目存在安全聯鎖措施描述遺漏、闡述不清晰或設置不當，5個項目存在放射性產物相關環節描述不充分等問題。結論：建議依據常見加速器項目應用類型不同采取可接受的執行限值、多方面完善屏蔽估算、全面合理評價安全聯鎖措施以及優化放射性產物的評價與管理，明確產物去向。

關鍵詞：加速器；輻射安全；評價；對策

來源出版物：中國職業醫學， 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.

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

編輯：衛夏雯

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