Nonproliferation improvements and challenges presented by small modular reactors
Small modular reactors (SMRs) may provide an energy option that will not emit greenhouse gases. From a commercial point-of-view, SMRs will be suitable to serve smaller energy markets with less developed infrastructure, to replace existing old nuclear and coal power plants, and to provide process heat in various industrial applications. In this paper, we examine how SMRs might challenge and improve the existing nonproliferation regime. To motivate our discussion, we first present the opinions gathered from an international group of nuclear experts at an SMR workshop. Next, various aspects of SMR designs such as: fissile material inventory, core-life, refueling, burnup, digital instrumentation and controls, under- ground designs, sealed designs, enrichment, breeders, excess reactivity, fuel element size, coolant opacity, and sea-based nuclear plants are discussed in the context of proliferation concerns. In doing this, we have used publicly available design information about a number of SMR designs (B&W mPower, SVBR-100, KLT-40S, Toshiba 4S, and General Atomics EM 2 ). Finally, a number of recommendations are offered to help alleviate proliferation concerns that may arise due to SMR design features.
THERMAL-HYDRAULIC ANALYSIS OF SMR WITH NATURALLY CIRCULATING PRIMARY SYSTEM DURING LOSS OF FEED WATER ACCIDENT
ABSTRACT Small Modular Reactors (SMRs) have several advantages over conventional large reactors. With integral and simplified design, application of natural laws for safety system, and lower capital cost this reactor is very suitable to be deployed in Indonesia.  One of SMR designs being developed implements natural driving force for its primary cooling system. With such innovative approach, it is important to understand safety implication of the design for all operating circumstances. One of conditions need to be investigated is the loss of feed-water (LoFW) accident. In this study, thermal-hydraulic performance of the SMR with naturally circulating primary system during LoFW accident is analysed. The purpose is to investigate the characteristics of flow in primary system during the accident and to clarify whether the naturally circulating coolant  is adequately capable to transfer the heat from core in order to maintain safe condition under considered scenario. The method used is by representing the reactor system into RELAP5 code generic models and performing numerical simulation. Calculation result shows that following the initiating event and reactor trip, primary system flow becomes significantly fluctuated and coolant temperature decreases gradually, while in secondary side steam quality descends into saturated. The primary flow slows down from ~711 kg/s to ~263 kg/s and starts to increase up again at t= ~46 seconds. At the slowest point, fuel centerline and coolant temperatures were ~565 K and ~554 K, showing that temperatures of  the fuel and coolant are still below its design limit and saturation point, respectively. This fact reveals that throughout transient the two main thermal hydraulic parameters stay in acceptable values so it could be concluded that under LoFW accident the SMR with naturally circulating primary system is in safe condition. Keywords: SMR, loss of feed water, natural circulation, reactor safety, RELAP5 ABSTRAK Reaktor daya kecil modular (SMR) memiliki beberapa keunggulan dibanding reaktor daya besar konvensional. Dengan disain yang lebih sederhana dan terintegrasi, penerapan hukum alamiah untuk sistem keselamatannya dan biaya modal yang rendah, reaktor ini sangat cocok untuk dibangun di Indonesia. Salah satunya disain SMR yang sedang dikembangkan menerapkan gaya penggerak alami untuk sistim pendingin primernya. Dengan disain seperti itu, adalah sangat penting untuk memahami implikasinya terhadap aspek keselamatan pada seluruh kondisi operasi. Salah satu yang perlu diinvestigasi adalah kecelakaan kehilangan air umpan (LoFW). Pada studi ini, dilakukan analisis kinerja thermal hidrolik SMR yang menggunakan sistim pendinginan primer sirkulasi alam saat kecelakaan LoFW. Tujuannya adalah untuk menginvestigasi karakteristik aliran sistem primer saat kecelakaan LoFW dan untuk memastikan apakah aliran sirkulasi alam cukup untuk memindahkan panas dari teras guna menjaga kondisi tetap aman selama kecelakaan tersebut. Metoda yang digunakan adalah dengan merepresentasikan sistem reaktor ke dalam model-model generik program RELAP5 dan melakukan simulasi numerik. Hasil perhitungan menunjukkan bahwa setelah kejadian pemicu dan trip reaktor, pada sisi primer laju alirnya berfluktuasi secara signifikan dan temperatur pendinginnya menurun secara bertahap sedangkan  pada sisi sekunder kondisi uap berubah menjadi uap jenuh. Laju alir turun dari ~711 kg/detik menjadi ~263 kg/detik sebelum kembali naik lagi pada t=~46 detik. Saat laju alir di titik terendah, temperatur pusat bahan bakar dan fluida pendingin adalah sekitar  ~565 K dan  ~554 K, yang menujukkan bahwa temperatur bahan bakar masih jauh di bawah batas disain dan temperatur fluidanya juga berada di bawah titik saturasi. Keadaan ini menunjukkan bahwa saat transien kedua parameter utama termohidrolik reaktor tetap dalam kondisi yang dapat diterima sehingga dapat disimpulkan  bahwa saat  kecelakaan kehilangan air umpan, SMR dengan sistim primer sirkulasi alam tetap dalam kondisi aman. Kata kunci: SMR, kehilangan air umpan, sirkulasi alamiah, keselamatan reaktor, RELAP5
Lead-alloy coolant technology and materials – technology readiness level evaluation
Abstract Lead and lead–bismuth eutectic heavy liquid metal coolants are under wide-ranging international investigation and development for advanced nuclear systems for energy production and waste transmutation (reactor-based or accelerator-driven). This report reviews the major supporting international R&D programs, the key advances in the main areas of coolant technology and materials, the state of technology, and the strategic directions for further development. Based on this review, we conservatively evaluate the technological readiness level (TRL) for programmatic and industrial applications in high-temperature advanced reactors to be 7, “one-dimensional engineering-scale demonstration”, or the first level in the proof-of-performance category. A 3-D engineering-scale integral test and demonstration facility should be the next step toward the realization of a test and demonstration nuclear system (reactor or accelerator-driven). The recent success of MEGAPIE, a 1 MW class lead–bismuth eutectic spallation target operating at the Paul Scherrer Institute signals that for such applications of short to intermediate durations at moderate temperatures, the TRL is close to 9, meaning the technology is nearly ready for deployment.