Fast Breeder Reactor Research Articles

Structural and thermodynamic study of dicesium molybdate Cs2Mo2O7: Implications for fast neutron reactors

The structure of α-Cs2Mo2O7 (monoclinic in space group P21/c), which can form during irradiation in fast breeder reactors in the space between nuclear fuel and cladding, has been refined in this work at room temperature from neutron diffraction data. Furthermore, the compounds' thermal expansion and polymorphism have been investigated using high temperature X-ray diffraction combined with high temperature Raman spectroscopy. A phase transition has been observed at Ttr(α→β)=(621.9±0.8)K using Differential Scanning Calorimetry, and the structure of the β-Cs2Mo2O7 phase, orthorhombic in space group Pbcm, has been solved ab initio from the high temperature X-ray diffraction data. Furthermore, the low temperature heat capacity of α-Cs2Mo2O7 has been measured in the temperature range T=(1.9–313.2)K using a Quantum Design PPMS (Physical Property Measurement System) calorimeter. The heat capacity and entropy values at T=298.15K have been derived as Cp,mo(Cs2Mo2O7,cr,298.15K)=(211.9±2.1)JK−1mol−1 and Smo(Cs2Mo2O7,cr,298.15K)=(317.4±4.3)JK−1mol−1. When combined with the enthalpy of formation reported in the literature, these data yield standard entropy and Gibbs energy of formation as ΔfSmo(Cs2Mo2O7,cr,298.15K)=−(628.2±4.4)JK−1mol−1 and ΔfGmo(Cs2Mo2O7,cr,298.15K)=−(2115.1±2.5)kJmol−1. Finally, the cesium partial pressure expected in the gap between fuel and cladding following the disproportionation reaction 2Cs2MoO4=Cs2Mo2O7+2Cs(g)+ 1/2 O2(g) has been calculated from the newly determined thermodynamic functions.

Investigations on the availability of second shutdown system of Indian FBR

Design and operating nuclear power plants with higher capacity factors is always desirable for improved economics and it is further more important in case of fast breeder reactors. It is well recognised that the availability of shutdown systems has an important bearing on the plant availability. In this paper, various investigations carried out on the availability of the second shutdown of Indian FBR are reported. The stability of the system performance against a range of anticipated disturbances was studied. Studies were designed to gain better technical insight into the system performance and to aid in finding out the performance bounds.Based on the design review of second shutdown system, in-house and international experiences, investigations related to the electromagnetic coupling, flow induced vibrations, subassembly bowing and the deterioration of electromagnet parting plane were carried out as part of the current experimental programme. Various innovative methods/schemes were adopted in the studies. These investigations helped in identifying the design vulnerabilities and thus in improving the system design, qualifying respective analytical/numerical studies, providing direction for further enhancement in system performance, demonstrating the margins available in the system design and more importantly much needed insights into various phenomenons to attend any unforeseen availability related issues of second shutdown system during their operation in reactor. Alternative design concepts were also discussed.

A Neutron Transport Characteristics Method for 3D Axially Extruded Geometries Coupled with a Fine Group Self-Shielding Environment

In this paper we describe some recent developments in the Method of Characteristics (MOC) for three-dimensional (3D) extruded geometries in the nuclear reactor analysis code APOLLO3®. We discuss the parallel strategies implemented for the transport sweep of the MOC solver in the OpenMP framework, and introduce the 3D version of the DPN operator that is customarily used in APOLLO2 to accelerate MOC convergence. In order to provide good physical results, we have also coupled the MOC with the self-shielding environment of APOLLO3®. We describe, in particular, the coupling techniques necessary to implement a full subgroup cross-section self-shielding method and a specialized version of the Tone self-shielding technique. In this framework, we use part of the tracking method used for the 3D calculation to provide the two-dimensional Collision Probability Method (CPM) coefficients necessary to produce the self-shielding calculations. We will show some important computational speedups also in the CPM of APOLLO3® with respect to the APOLLO2 CPM equivalent implementation, including the parallelization issue. Finally, we will compare our approach toward a Monte Carlo calculation of a fast breeder reactor hexagonal assembly representing a fertile-fissile interface.

Development of fast breeder reactor technology in India

India is pursuing a three stage nuclear power program employing its natural uranium and Thorium reserves. FBRs form the second stage of the program linking the first phase with natural U and third phase with Th fuels. The saga of FBR technology development in India is presented in this paper. The valuable experience in design and operation of Fast Breeder Test Reactor (FBTR) is recounted briefly. The R&D for the techno-economic demonstration of Prototype Fast Breeder Reactor (PFBR) is explained including technology development for addressing manufacturing challenges and engineering full-scale qualification. Finally a brief projection is made on the developments for FBR600 with improved safety and economics as well as metal fuel technology for future experimental and power reactors to follow.

Improvement of the “best representativity” method toward an industrial application

To judge the applicability of a critical experiment, it is necessary to confirm the similarities of the experiment with actual reactor conditions or equipment. The concept of the “representativity factor” has been well adopted since the late 1970’s, particularly for fast breeder reactors (FBRs) and future reactor studies. We extended this concept to the design of a light water reactor (LWR) system. In the first study, a new numerical evaluation method and a calculation system were developed and qualified. In the second study, an application calculation was conducted to validate the method. Based on the method, calculations were performed to correct the infinite neutron multiplication factor of a pressurized water reactor fuel assembly with using three kinds of critical experiments. The representativity factor became closer to unity. A correction of the infinite multiplication factor was well achieved.In this study, for aiming at an industrial application of the method, the method was improved to be more suitable for the single experiment case. In addition, a modification of the calculation procedures was proposed to prevent a very large increase of the final combined uncertainty. Furthermore, to validate the method, calculations were conducted according to ICSBEP 2010 data disk with utilizing critical experiments that were performed at different critical facilities. Simultaneously, with “SCALE 6.1.2” system and “ENDF/B-VII” 238G library, the applicability of the method was studied. All calculation results are shown and explained with physical considerations.

Reprocessing of spent nuclear fuel in India: Present challenges and future programme

Adoption of the closed nuclear fuel cycle is an imperative option for the Indian nuclear power programme in view of the limited resources of natural uranium and abundance of thorium. The issues associated with the long-term radiotoxicity of spent fuel can also be addressed in the closed fuel cycle with this option. The present paper focusses on the scientific and technological challenges associated with reprocessing of spent fuel from both the thermal and fast reactors. The experience gained in the operation of reprocessing plants in the separation of spent fuel of thermal reactors into uranium, plutonium and high level waste is described in this paper. Challenges resolved during the reprocessing of spent mixed carbide fuel from Fast Breeder Test Reactor (FBTR) in CORAL plant are described in detail. This experience enabled the design of a commercial scale reprocessing plant which is under construction to handle the spent fuel from 500 MWe Prototype Fast Breeder Reactor (PFBR). Recent developments in the separation of minor actinides from high level waste is also briefly discussed. The experience gained in the thorium fuel cycle with the reprocessing of spent thorium fuel is also covered.

Nuclear fuel cycle - developments and challenges in fuel fabrication technology in India

The paper gives a summary of the developments carried out for over four and a half decade in the field of nuclear fuel fabrication for power reactors in India. The facility at Nuclear Fuel Complex (NFC)is unique in the world with fully integrated manufacturing of fuel and core structural for the Pressurized Heavy Water Reactors (PHWRs), Boiling Water Reactor (BWR) and Fast Breeder Reactors (FBRs)operating in India. The integrated processing starts with the ore which is processed through several intermediate products to the finished fuel assemblies and several core structurals. The paper covers a summary of challenges encountered in technology development of materials and processes and their solutions arrived by carrying out detailed studies. The paper gives an overview of automation and advanced quality control methods used in the production line for capacity building to meet the enhanced fuel requirement. The paper also gives an outline of the future expansion program for fuel fabrication for enhancing the installed capacity of nuclear energy in India.

The evolution of the Indian nuclear power programme

The strategy for growth of nuclear power in India was planned nearly sixty years ago, noting the rather small uranium and large thorium reserves in the country. This has prompted India to adopt the well-known three stage programme. The first stage is primarily based on the Pressurised Heavy Water Reactors (PHWRs). The evolution of technology of PHWRs is discussed in the first part of this paper. India has constructed eighteen PHWRs, has achieved impressive availability factors and some of these reactors have achieved annual capacity factors of nearly hundred percent in the recent past. Having installed a nuclear power generation capacity of 6780 MWe (4460 MWe from PHWRs, 2000 MWe from PWRs and 320 MWe from BWRs), India is now poised to launch a major expansion programme. This will be based on the increased availability of uranium from import and from the augmented domestic supply. In the immediate future, the nuclear power capacity will grow by installing a series of indigenous PHWRs in addition to light water reactors built under international civil nuclear cooperation agreements. The growth of nuclear capacity in this period is aimed at increasing the share of nuclear power in meeting the base-load demand of non-carbon electricity required for the rapid economic growth in the country.India embarked on its second stage programme with the successful operation of a research reactor named Fast Breeder Test Reactor (FBTR). Based on the experience of FBTR and following the development of all the required enabling technologies in India, the Prototype Fast Breeder Reactor (PFBR) of 500 MWe (gross) capacity has been designed, constructed and is now at an advanced stage of commissioning. A large increase in the nuclear power generation capacity is envisaged through deployment of fast breeder reactors. These reactors will not only help in building up nuclear power capacity but also, in due course, enable conversion of thorium into fissile U233, which will fuel the reactors in the third stage of India's nuclear programme. The adoption of the closed fuel cycle for both thermal and fast reactors has dual objectives: multiplying the fissile inventory by fertile to fissile conversion and reducing the burden of long lived radioactive waste-both being essential for attaining near sustainability of nuclear power.India has designed an Advanced Heavy Water Reactor (AHWR). This reactor can use one of the following fuels:(i)Low enriched uranium dioxide-thorium dioxide based mixed oxide (MOX) fuel primarily in once through mode.(ii)Plutonium dioxide – thorium dioxide MOX fuel in the closed fuel cycle mode. Eventually when enough of U233 is generated the reactor can operate with the U233– thorium MOX fuel.A large number of passive safety features are incorporated in this design.As a part of the long term objective of the development of Accelerator Driven Sub-critical Systems (ADSS), work on the development of high power linear accelerator has been initiated.

Development of Oxide Dispersion Strengthened (ODS) Ferritic Steel Through Powder Forging

Oxide dispersion strengthened (ODS) ferritic steels are candidates for cladding tubes in fast breeder nuclear reactors. In this study, an 18%Cr ODS ferritic steel was prepared through powder forging route. Elemental powders with a nominal composition of Fe-18Cr-2 W-0.2Ti (composition in wt.%) with 0 and 0.35% yttria were prepared by mechanical alloying in a Simoloyer attritor under argon atmosphere. The alloyed powders were heated in a mild steel can to 1473 K under flowing hydrogen atmosphere. The can was then hot forged. Steps of sealing, degassing and evacuation are eliminated by using powder forging. Heating ODS powder in hydrogen atmosphere ensures good bonding between alloy powders. A dense ODS alloy with an attractive combination of strength and ductility was obtained after re-forging. On testing at 973 K, a loss in ductility was observed in yttria-containing alloy. The strength and ductility increased with increase in strain rate at 973 K. Reasons for this are discussed. The ODS alloy exhibited a recrystallized microstructure which is difficult to achieve by extrusion. No prior particle boundaries were observed after forging. The forged compacts exhibited isotropic mechanical properties. It is suggested that powder forging may offer several advantages over the traditional extrusion/HIP routes for fabrication of ODS alloys.

Experiments in LEENA facility with modified wire type leak detector layout in large sodium pipelines

Sodium cooled Fast Breeder Reactors (SFRs) are envisaged in the second phase of Indian nuclear power programme. Liquid sodium is used as the coolant in the SFRs due to its favourable nuclear properties and excellent heat transfer properties. Leaks in sodium systems have the potential of being exceptionally hazardous due to the reaction of liquid sodium with oxygen and water vapour in the air. When a sodium leak occurs, the sodium leak rate, the total quantity of sodium leaked and leak detector layout governs the detection time. Other factors to be considered are insulation material packing condition, distance between the leak point and detector, heater layout, pipe geometry, temperature etc. Potential regions of leakage in Fast Breeder Reactor (FBR) sodium circuits are near welds, high stress areas and regions subjected to thermal striping. Early detection of leak is needed for minimizing the quantity of sodium leaked to outside and consequent damages. Three wire type leak detectors (WLDs positioned at 90°, 180° and 270°) working on conductivity principle are used for detecting sodium leak in the large horizontal secondary sodium pipelines of Prototype Fast Breeder Reactor (PFBR). It was found from the upper boundary curve based on LEENA (LEak Experiments in NAtrium) facility experimental results with present PFBR leak detector layout, that a leak rate of 200g/h and above can be detected within 6h. To improve the detection capability, modified leak detector layout was proposed for FBR 800 NB size horizontal secondary circuit pipelines of FBR with seven WLDs positioned at 45°, 90°, 135°, 180°, 225°, 270° and 315° by taking top of pipeline as 0°. It is required to assess the performance of wire type leak detectors along with its layout on large horizontal sodium pipelines and confirm that they are meeting the requirements. LEENA facility was re-commissioned and five sodium leak experiments with modified leak detector layout on large horizontal pipeline were conducted with different sodium leak rates. From the experimental data it was observed that there was considerable reduction in detection time with modified WLD layout. This is due to the reduction in distance between leak point and leak detector (Reduced from 0.63m to 0.31m). From the experiments it was found that a leak rate of 200g/h can be detected in one hour.

Vitreous and crystalline phosphate high level waste matrices: Present status and future challenges

Vitrification is currently the most effective process for immobilization of nuclear waste. However, ubiquitous borosilicate glass is not suitable for immobilization of nuclear waste from advanced reactors such as Fast Breeder Reactors (FBR) because solubility of many compounds/elements existing in the spent fuel in borosilicate glasses is quite poor. In order to possess a viable immobilization strategy for wastes arising from advanced reactors, alternatives to borosilicate glasses such as phosphate glasses, glass-ceramics and crystalline waste forms are being investigated. This review aims to provide an overview of nuclear waste immobilization employing phosphate-based glasses, glass-ceramics and crystalline ceramic hosts, focusing on structure and properties that make these new matrices suitable for the challenging task of waste immobilization.

Corrosion Behavior of Yttria-Stabilized Zirconia-Coated 9Cr-1Mo Steel in Molten UCl3-LiCl-KCl Salt

For the electrorefining step in the pyrochemical reprocessing of spent metallic fuels of future sodium cooled fast breeder reactors, 9Cr-1Mo steel has been proposed as the container material. The electrorefining process is carried out using 5-6 wt.% UCl3 in LiCl-KCl molten salt as the electrolyte at 500 °C under argon atmosphere. In the present study, to protect the container vessel from hot corrosion by the molten salt, 8-9% yttria-stabilized zirconia (YSZ) ceramic coating was deposited on 9Cr-1Mo steel by atmospheric plasma spray process. The hot corrosion behavior of YSZ-coated 9Cr-1Mo steel specimen was investigated in molten UCl3-LiCl-KCl salt at 600 °C for 100-, 500-, 1000- and 2000-h duration. The results revealed that the weight change in the YSZ-coated specimen was insignificant even after exposure to molten salt for 2000 h, and delamination of coating did not occur. SEM examination showed the lamellar morphology of the YSZ coating after the corrosion test with occluded molten salt. The XRD analysis confirmed the presence of tetragonal and cubic phases of ZrO2, without any phase change. Formation of UO2 in some regions of the samples was evident from XRD results.

Hydraulic characteristics of a fast reactor fuel subassembly: An experimental investigation

Fuel subassemblies of a fast reactor consist of fuel pin bundle with helically wound spacer wires, arranged in a triangular pitch within a hexagonal wrapper. The fuel pins are located within the subassembly. Further the subassembly comprises of a diffuser where the cross section changes from cylindrical to hexagonal, mixing plenum before the exit of pin bundle and a specially designed blockage adapter. Accurate assessment of the pressure drop in the fuel subassembly is essential to ensure adequate core cooling and design of sodium pump. Experimental determination of pressure drop characteristics in the subassembly by simulating the hydraulic condition in the subassemblies of the reactor core is considered essential as a better choice as correlations reported in the literature cannot be directly used for all the complex regions present in the subassembly. This is due to the fact that flows in the interconnecting sections are highly under developed. Further, the flow regime in a fuel subassembly varies from laminar (during shutdown heat removal under natural convection) to completely turbulent under full power condition.To understand the hydraulic characteristics of the 500MWe Proto type Fast Breeder Reactor (PFBR) fuel subassembly, an experimental facility has been commissioned. Experiments on full scale subassembly with dummy fuel pins have been performed using water as simulant. Experiments have been conducted covering a wide range of Reynolds number encompassing laminar, transition and turbulent regimes. In the rod bundle, no abrupt changes in friction factor were observed when the flow changes from laminar to turbulent. The experimental results have been transposed using Euler number similarity, to determine the pressure drop for sodium flow in the reactor. Possible uncertainties in the experimental data have been quantified. Useful pressure drop correlations have been proposed for various section of the subassembly of a fast breeder reactor. Finally the results for fuel pin bundle region are compared with the data reported in the literature and a satisfactory agreement has been observed.

Experimental and Computational Forensics Characterization of Weapons-Grade Plutonium Produced in a Fast Reactor Neutron Environment

A terrorist attack using an improvised nuclear device is one of the most serious dangers facing the United States. The work presented here is part of an effort to improve nuclear deterrence by developing a methodology to attribute weapons-grade plutonium to a source reactor by measuring the intrinsic physical characteristics of the interdicted plutonium. In order to demonstrate the developed methodology, plutonium samples were produced from depleted uranium dioxide (DUO2) surrogates irradiated in a fast-neutron environment. In order to replicate the neutron flux in a fast-neutron-spectrum reactor and obtain experimental samples emulating weapons-grade plutonium produced in the blanket of a fast breeder reactor, DUO2 samples were placed in a gadolinium sheath and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Previous computational work on this topic identified several fission products that could be used to distinguish between reactor types (fast and thermal reactors), specifically: 137Cs, 134Cs, 154Eu, 125Sb, 144Ce, 85Rb, 147Pm, and 150Sm along with the plutonium isotopes. Simulations of the fast neutron irradiation of the DUO2 fuel surrogates in the HFIR were carried out using the Monte Carlo radiation transport code MCNPX 2.7. Comparisons of the predicted values of plutonium and fission product concentrations to destructive and nondestructive assay measurements of neutron-irradiated DUO2 surrogates are presented here. The agreement between the predictions and gamma spectroscopic measurements in general were within 10% for 134Cs, 137Cs, 154Eu, and 144Ce. Additional experimental results (mass spectroscopy) agreed to within 5% for the following isotopes: 85Rb, 147Pm, 150Sm, 154Eu, 148Nd, 144Ce, and 239Pu. Two indicator isotopes previously suggested to differentiate between the reactor types were ruled out for use in the attribution methodology; 125Sb was ruled out due to the difficulty in accurately predicting its concentration, and 242Pu was ruled out because of its low content in weapons-grade plutonium.

Studies on Dissolution Behavior of the Surface Layer of Sodium-Exposed SS 316LN in Decontaminating Formulation Using PEMHS

The Prototype Fast Breeder Reactor (PFBR) is nearing completion at Kalpakkam, India. Sodium is the heat transfer medium for PFBR, and austenitic steel SS 316LN is the material of construction for the sodium circuits of the reactor. During reactor service, the inner surfaces of the sodium circuit pipelines undergo corrosion by interacting with liquid sodium, forming ferritic layers. Radioactive nuclides formed by the activation of corrosion products are deposited on the ferritic surface, resulting in a radioactive burden on maintenance personnel. Chemical decontamination is generally carried out by dissolving the surface ferritic layer on the inside surface of the sodium circuit. In this context, a study of the dissolution behavior of the ferritic layer on SS 316LN samples formed by exposure to liquid sodium at 823 K was carried out by monitoring the H2 released during the chemical interaction with decontamination formulation. The decontamination chemical formulation was a mixture of sulfuric acid and phosphoric acid. This paper discusses the sample preparation, formation of the ferritic layer, and studies carried out on its dissolution behavior in decontamination formulation by monitoring the hydrogen released during the reaction using a proton exchange membrane–based hydrogen sensor.

Comparative Analysis on Nuclear Fuel Sustainability Aspect of FBR

Recycle program of spent nuclear fuel (SNF) will have some challanges in term of fuel cycle capability and its facilities as well as nuclear non-proliferation concern of special nuclear materials. A different analysis approach as a comparative study have been analyzed based on breeding ratio and heavy metal inventory ratio concepts in fast breeder reactor (FBR) type. Breeding ratio and heavy metal inventory obtain higher than unity which shows breeding gain or surplus inventory of heavy metals are obtained. Breeding ratio indicates the fuel conversion capability from conversion process of fertile materials into fissile material such as fertile materials of U-238, Pu-238, Pu-240 and fissile materials of Pu-239 and Pu-241. Inventory ratio approaches are appropriate to estimate some selected actinide as a mass inventory production such as plutonium inventory ratio which estimate the surplus mass inventory from the ratio of produced plutonium at the net of operation to the initial inventory ratio.

High temperature corrosion of cold worked YUS409D bellows of bellow-sealed valve in LBE

Lead-bismuth eutectic (LBE) loop test is highly contributes to the lead-alloy-cooled fast breeder reactor (LFR) and accelerator driven system (ADS) research and development by providing comprehensive results of both corrosion and erosion phenomenon. Bellows-sealed valve is a crucial part in the LBE loop test apparatus, due to its capability of preventing corrosion on valve spring, thus improves the operation time of the system. LBE is very corrosive to stainless steel by formation of oxide layer or elemental dissolution, e.g. Ni. Thus, new type of bellows for bellows-sealed valve made of nickel free material, i.e. YUS409D, is proposed to be used in the LBE. Bellows material undergo heat treatments for mechanical improvement including cold working and annealing. The thickness reduction by the heat treatments is about 90% of initial condition. Corrosion behavior of the bellows has been studied in stagnant LBE at 500 and 600 °C for 500 hours. The oxygen concentration was controlled at about 10−7 wt%. Typical oxide layers were developed on the surface. Oxidation rate was sharply increased at 600°C.

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

A Roadmap of Innovative Nuclear Energy System

Nuclear is a dense energy without CO2 emission. It can be used for more than 100,000 years using fast breeder reactors with uranium from the sea. However, it raises difficult problems associated with severe accidents, spent fuel waste and nuclear threats, which should be solved with acceptable costs. Some innovative reactors have attracted interest, and many designs have been proposed for small reactors. These reactors are considered much safer than conventional large reactors and have fewer technical obstructions. Breed-and-burn reactors have high potential to solve all inherent problems for peaceful use of nuclear energy. However, they have some technical problems with materials. A roadmap for innovative reactors is presented herein.

Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium–Plutonium Carbide based fuel and 300 MWt Power

A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.