Application of Fusion Neutron Source for Denaturing of Plutonium

Potential of DT fusion neutron source to enhance proliferation resistance properties of plutonium by means of its isotopic denaturing is addressed. The approach is exemplified by denaturing of pure ²³⁹Pu and plutonium of typical LWR spent fuel through transmutation of neptunium. The essential feature of a fusion driven system proposed in the study is a zero mass balance of plutonium: total plutonium inventory is constant during irradiation. The system is capable to convert pure ²³⁹Pu into plutonium composition with more than 20% fraction of key ²³⁸Pu isotope during 1,000 d of irradiation under initial neutron loading of 1 MW.m^?2. Denaturing of LWR spent fuel plutonium under the same conditions would increase its ²³⁸Pu content up to 10-12%.     

高温气冷堆和轻水堆超铀元素管理特性研究

乏燃料中长寿命锕系元素对环境造成长期潜在危害,本文研究球床高温气冷堆不同燃料循环中超铀元素的产生和焚烧特性。在250 MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环的乏燃料中提取铀和钚作为核燃料,设计了PuO₂和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM相同结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。采用高温气冷堆物理设计程序VSOP,研究了高温气冷堆一次通过燃料循环和不同闭式燃料循环的超铀元素焚烧特性,并与轻水堆燃料循环结果进行比较和分析。结果表明：高温气冷堆一次通过燃料循环超铀元素生成率约为轻水堆的1/2;高温气冷堆闭式燃料循环能有效嬗变超铀元素。     

Core Burnup Calculation of Plutonium Burning Inert-Matrix Fueled High Temperature Gas Cooled Reactor

For the efficient reduction of excess plutonium amount, Japan Atomic Energy Research Institute (JAERI, now Japan Atomic Energy Agency) has studied a concept of rock-like oxide (ROX) fuel as a kind of inert matrix fuel (IMF). In the JAERI study, ROX fuel is burnt in existing light water reactors (LWRs), while in this study, pebble bed type high temperature gas cooled reactor (HTGR) is studied, mainly because of its high neutron economy and softer neutron spectrum than LWRs. Here, PuO₂-yttria stabilized zirconia (YSZ: (Zr,Y)O_2-x) particles are dispersed in graphite matrix. In the ROX fueled LWR study, it was necessary to improve fuel temperature reactivity coefficients by adding such additives as ²³⁸U and Er. Here in HTGR, although the negative temperature coefficient is much larger than that in LWR without any improvements, temperature coefficient was improved as large as possible to the level of UO₂ HTGR case by adding Er in the fuel. Burnup calculations on PuO_2-YSZ fueled HTGR, by simulating the continuous refueling of fuel pebbles with the batch fuel loading, showed almost complete transmutation for ²³⁹Pu and more than 80% for the total plutonium. As the maximum power density of the fuel pebble obtained by the core burnup calculation was very large in comparison with the UO₂ HTGR, the maximum temperature in YSZ fuel particle was also evaluated. Despite the low thermal conductivity of YSZ, the evaluated YSZ temperature was well below the melting point, thanks to the high thermal conductivity of graphite and small YSZ particle size. Here, the high power density in the Pu-YSZ fueled core might become a problem, but is possible to be reduced by adjusting the initial plutonium enrichment.     

Advanced fuel cycle scenarios for High Temperature Gas Reactors

This study evaluates nuclear fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. Three fuel cycle scenarios, the Light Water Reactor (LWR)–Advanced Recycling Reactor (ARR) recycle, the LWR–High Temperature Gas Reactor (HTGR)–ARR recycle, and the HTGR partial recycling fuel cycle, are assessed for their mass flow and electricity generation costs and the results are compared to those of the LWR once-through fuel cycle. The spent fuels are recycled in both the Consolidated Fuel Treatment Center and the Actinide Management Island, which are capable of reprocessing spent fuels by Uranium Extraction and Pyrochemical processes, respectively. The mass flow calculations show that the Transuranics (TRU) which have a long-term radiation effect can be completely burned in the recycling fuel cycles, resulting in 350, 450 and 6 times reduction of TRU inventory for the LWR–ARR, LWR–HTGR–ARR and HTGR partial recycling fuel cycles, respectively, when compared to the once-through fuel cycle. The electricity generation costs of these fuel cycle scenarios were estimated to be 39.1, 34.9 and 25.7 USD/MW h(e), which are comparable to or smaller than that of the once-through fuel cycle. Although the candidate fuel cycles adopt reprocessing options which raise fuel cycle cost, increase in uranium cost and the advanced design of the HTGR can further reduce the advanced fuel cycle costs of the HTGR.     

CANDU堆先进燃料循环的展望

介绍ＣＡＮＤＵ堆的天然铀燃料循环以及最近开发的适合未来近期的先进燃料循环。高中子经济性,不停堆换料以及简单的燃料束设计,使得ＣＡＮＤＵ堆具有非常优良的燃料循环灵活性和多样性。     

Fuel cycle analysis of GFR using advanced fuels

The potential benefits of a synergistic light-water reactor (LWR) and gas-cooled fast reactor (GFR) fuel cycle were evaluated for its impact on the front-end and back-end of the fuel cycle. Comparisons were made with conventional once-through cycle (OTC) and MOX fuel cycle. Variations in the synergistic LWR/GFR fuel cycles were based on the degree of recycle in the LWR including both plutonium and reprocessed uranium with concomitant impact on used LWR fuel inventory. This provided a wide range in composition of GFR feed from low to high plutonium content with impact on transmutation/incineration within the GFR fuel cycle. Self-recycle of all actinides was modeled for the GFR with analyses demonstrating that the GFR can be sustained on and consequently accept a wide range of feed materials. Analyses were done using Monteburns along with MCNP and Origen2.2 to model a 60-year period corresponding to the anticipated lifetime of supposed contemporary LWRs and GFRs. All cycles were evaluated based on actinide total mass and isotopic inventory, radiotoxicity, heatload, and resource requirements including natural uranium and SWU. For comparison, all fuel cycles were normalized based on 1 TWHe output. Improvements in fuel cycle performance are dictated by the production and incineration of minor actinides in the GFR and their continued recycle may not be feasible due to the buildup of troublesome isotopes such as ²⁴⁴Cm and ²⁵²Cf. But where uranium and plutonium continue to be recycled beyond the 60-year period, the LWR/GFR cycles demonstrated order of magnitude reductions in used fuel inventories, heatload, and radiotoxicity on a per TWHe basis over LWR only cycles. The full details of the advanced fuel cycle methodology and results are presented.     

Feasibility of Reprocessed Uranium in LWR Fuel Cycle for Protected Plutonium Production

Protected plutonium production (PPP) is an intrinsic measure to enhance the proliferation resistance of Pu by raising the ²³⁸Pu isotopic concentration, which denatures Pu on account of the high spontaneous fission neutron (SFN) rate and large decay heat (DH). This study is aimed at examining the feasibility of reprocessed uranium (RepU) with or without the addition of minor actinide (MA) in LWR fuel cycle for PPP and to make a tentative economic assessment of RepU possessing the PPP feature. It was analytically clarified that RepU enriched to 5% ²³⁵U by centrifugation produced denatured Pu at higher burnup than about 40GWd/t. By the addition of more than 0.5% MA to RepU and natural uranium both enriched to 5%, Pu generated in the uranium fuel with MA added could be denatured up to 40 GWd/t at least. A diagram designed with functions of SFN rate and DH explicated the PPP features of re-enriched RepU and enriched natural uranium with or without MA addition. The economic assessment indicated that the cost of fuel cycle applying re-enriched RepU would be comparable to that of the conventional fuel cycle, if the cost of the source RepU is low. In addition, the LWR fuel cycle applying RepU for PPP was discussed.     

Current status of scientific activity in Russia on high level radioactive waste transmutation and use of weapon grade and power plutonium in subcritical systems driven by proton accelerator

Features of neutron fuel cycles with the accelerator-driven system (ADS) as well as fission product and actinide transmutation in the ADS are analyzed in this paper: fuel type, fertile materials, neutron consumption, secondary radioactivity, change in radiotoxicity of actinides. The use of weapon-grade and power plutonium in the ADS is also considered. Information on various design versions of the ADS blanket including study of the sectioned blanket with neutron valves, its performance and R&D programme, including a neutron source driven by the 56.MeV “Istra” proton linac, is given.     

Actinide reformer concept

The partitioning and transmutation technology is effective to reduce the environmental impact from disposition of high-level radioactive wastes and improve the efficiency of geological disposal. However, Am and Cm and their daughter nuclides are difficult to handle in the fuel cycle because of their high decay heat and radioactivity. These nuclides also give the chemical instability which harms the soundness of fuel pellet properties.

We propose a new system concept “actinide reformer”, which reforms Am and Cm into Pu by neutron capture reactions and decay. Am and Cm are separated from the PUREX reprocessing process and converted to chloride molten-salt fuel. Using liquid-type fuel, above mentioned defects can be compensated. Actinide reformer is an accelerator-driven system which is composed of a 10 MW-class cyclotron, a tungsten target and a subcritical core. Spent molten-salt fuel is reprocessed as an on-line fuel exchange manner to extract fission products and recover Pu to send back to a power generation cycle. The decay heat and neutron emission from the fuel with recovered Pu are smaller than those of MOX fuel with 5% of minor actinide addition. It expects no significant engineering difficulties and cost increase for construction of MOX fuel based reprocessing/fabrication plant and power reactors.     

Neutronics and material attractiveness for PWR thorium systems using monte carlo techniques

Thorium (Th) oxide fuel offers a significant advantage over traditional low-enriched uranium and mixed uranium/plutonium oxide (MOX) fuel irradiated in a Light Water Reactor. The benefits of using thorium include the following: 1) unlike depleted uranium, thorium does not produce plutonium, 2) thorium is a more stable fuel material chemically than LEU and may withstand higher burnups, 3) the materials attractiveness of plutonium in Th/Pu fuel at high burnups is lower than in MOX at currently achievable burnups, and 4) thorium is three to four times more abundant than uranium. This paper quantifies the irradiation of thorium fuel in existing Light Water Reactors in terms of: 1) the percentage of plutonium destroyed, 2) reactivity safety parameters, and 3) material attractiveness of the final uranium and plutonium products. The Monte Carlo codes MCNP/X and the linkage code Monteburns were used for the calculations in this document, which is one of the first applications of full core Monte Carlo burnup calculations. Results of reactivity safety parameters are compared to deterministic solutions that are more traditionally used for full core computations.Thorium is fertile and leads to production of the fissile isotope ²³³U, but it must be mixed with enriched uranium or reactor-/weapons-grade plutonium initially to provide power until enough ²³³U builds in. One proposed fuel type, a thorium-plutonium mixture, is advantageous because it would destroy a significant fraction of existing plutonium while avoiding the creation of new plutonium. ²³³U has a lower delayed neutron fraction than ²³⁵U and acts kinetically similar to ²³⁹Pu built in from ²³⁸U. However, as with MOX fuel, some design changes may be required for our current LWR fleet to burn more than one-third a core of Th/Pu fuel and satisfy reactivity safety limits. The calculations performed in this research show that thorium/plutonium fuel can destroy up to 70% of the original plutonium per pass at 47 GWd/MTU, whereas only about 30% can be destroyed using MOX. Additionally, the materials attractiveness of the final plutonium product of irradiated plutonium/thorium fuel is significantly reduced if high burnups (∼94 GWD/MTU) of the fuel can be attained.     

Some aspects of risk reduction strategy by multiple recycling in fast burner reactors of the plutonium and minor actinide inventories

This paper shows the impact of recycling light water reactor (LWR) mixed oxide (MOX) fuel in a fast burner reactor on the plutonium (Pu) and minor actinide (MA) inventories and on the related radioactivities. Reprocessing of the targets for multiple recycling will become increasingly difficult as the burnup increases. Multiple recycling of Pu + MA in fast reactors is a feasible option which has to be studied very carefully: the Pu (except the isotopes Pu-238 and Pu-240), Am and Np levels decrease as a function of the recycle number, while the Cm-244 level accumulates and gradually transforms into Cm-245. Long cooling times (10 + 2 years) are necessary with aqueous processing. The paper discusses the problems associated with multiple reprocessing of highly active fuel types and particularly the impact of Pu-238, Am-241 and Cm-244 on the fuel cycle operations. The calculations were performed with the zero-dimensional ORIGEN-2 code. The validity of the results depends on that of the code and its cross-section library. The time span to reduce the initial inventory of Pu + MA by a factor of 10 amounts to 255 years when average burnups are limited to 150 GW · d t⁻¹ (tonne).     

A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3-5 years to construct.     

Safety design

JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs.This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R&D needs for establishing the safety philosophy for the future HTGRs are reported.     

Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc…). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called “BUCAL1”. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k_∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.     

Build-Up and Decay of Actinide Nuclides in Fuel Cycle of Nuclear Reactors

In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1,000MW. The following items were examined;

1. quantities of actinide nuclides build-up in the reactor

2. build-up and decay of activities of actinides in the spent fuel

3. build-up and decay of activities of actinides after reprocessing, and

4. variation of isotopie composition of plutonium with high burn-up.

It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.     

Dependence of the Parameters of an Electronuclear System on the Composition of Mixed Uranium–Plutonium Fuel

Mathematical simulation is used to study the dependence of the parameters of an electronuclear system with uranium–plutonium mixed fuel on the amount of plutonium and ²⁴⁰Pu admixture in the fuel. As an example, a 10–20 kW(t) experimental system built at Dubna on the basis of a synchrocyclotron with 660 MeV protons is examined. A 2% admixture of ²⁴⁰Pu decreases the neutron multiplication factor from 0.95 to 0.9 and decreases the neutron yield and heat released by almost a factor of 2. This decrease can be compensated by increasing the plutonium content in the mixed fuel from 25 to 27%.     

Transmutation of long-lived radioactive waste based on double-strata concept

A considerable attention is directed toward the reduction in the long-term potential hazard by partitioning and transmutation (P-T): separating long-lived nuclides from the waste stream and converting them into either shorter-lived or non-radioactive ones. The effects of higher Pu and minor actinide (MA) compositions on the transmutation rates have been studied for a typical mixed oxide (MOX)-fuel fast breeder reactor (FBR) core with 2600 MWt. The calculations showed that the transmutation rate for (Pu, MA) compositions from MOX -LWR becomes one half than that from UO₂-light water reactor (LWR). Furthermore, MA accumulation and transmutation based on Double-Strata Scenario have been investigated for introducing the accelerator driven transmutation system (ADS) with 800 MWt. It was shown that in the scenario of nuclear plant capacities for maximum 140 GWe, which consists of LWRs and FBRs, the introduction of ADS can play a significant role as “Transmuter” in the back-end of fuel cycle.     

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li₂₀Sn₈₀ coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.     

Nuclide separation from spent nuclear fuels by using tertiary pyridine resin

The newly nuclide separation system from spent nuclear fuels is proposed. The proposed separation system consists of recovery of nuclear fuel elements, separation of trivalent minor actinide from lanthanide, and separation of some fission products such as strontium. This separation system is based on the chromatographic technique using the tertiary pyridine resin. Separation experiments using mixed oxide fuel highly irradiated in fast reactor “Joyo” were carried out. The recovery of plutonium, the separation of minor actinide from fission products including lanthanides, and the separation of americium and curium were achieved. The recovery or removal of platinum group elements and technetium was also investigated, and the removal of these elements prior to the main reprocessing process has been proposed.