Usha Pal - Academia.edu (original) (raw)

Papers by Usha Pal

Nuclear and Particle Physics Proceedings

Annals of Nuclear Energy, 2022

Nuclear Technology, 2001

A new reactor concept has been proposed for induction of thorium in an enriched uranium reactor. ... more A new reactor concept has been proposed for induction of thorium in an enriched uranium reactor. The neutronic characteristics of the fissile and fertile materials have been exploited to arrive at optimal fuel assembly and core configurations. Each fuel assembly consists of an enriched uranium seed zone and a thoria blanket zone. They are in the form of ring-type fuel clusters. The fuel is contained in vertical pressure tubes placed in a hexagonal lattice array in a D2O moderator. Boiling H2O coolant is used. The 235U enrichment is ~5.4%. The thoria rods contain the 233U bred in situ by irradiation of one batch load of mere thoria clusters (without the seed zone) for one fuel cycle in the same reactor. There is no need for external feed enrichment in thoria rods. Additionally, some moveable thoria clusters are used for the purpose of xenon override. The fissile production rate from the fertile material and the consumption rate of fissile inventory is judiciously balanced by the choice of U/Th fuel rod diameter and the number and location of thoria rods in the fuel assembly and in the core. During steady-state operation at rated power level, there is no need for any conventional control maneuvers such as change in soluble boron concentration or control rod movement as a function of burnup. Burnable poison rods are also not required. A very small reactivity fluctuation of ±2 mk in 300 effective full-power days of operation is achieved and can be nearly met by coolant inlet enthalpy changes or moveable thoria clusters. Control is required only for cold shutdown of the reactor. The uranium as well as thoria rods achieve a fairly high burnup of 30 to 35 GWd/tonne at the time of discharge. Since the excess reactivity for hot-full-power operation is nearly zero at all times during the fuel cycle and since the coefficients of reactivity due to temperature and density variations of coolant are nearly zero by design, there is hardly any possibility of severe accidents involving large reactivity excursions.

Annals of Nuclear Energy, 2016

Advances in reactor physics have led to the development of new computational technologies and upg... more Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group 'JEFF3.1' cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled ''Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3" submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication (k 1) factor, one group integrated flux, few group homogenized crosssections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo (MCU) values of the benchmark. Maximum deviation of 1.8% in k 1 is observed for variants with spent fuel and for the states with control rod whereas all the other results are in par with the results reported in the benchmark document. The few and multi-group macroscopic cross-sections and flux of all the nuclides also compare well with the benchmark results except for the 11 B macroscopic absorption cross section, which is further compared with the XNWLUP software. Inter-comparison of results with the generalized self-shielding model SHI of DRAGON code employing the traditional WIMSD formatted 172 group crosssection library has also been made to highlight the improvements made in computational schemes and cross-section library format.

Pramana, 2007

A 100 MWt reactor design has been conceived to support flux level of the order of 10 15 n/cm 2 /s... more A 100 MWt reactor design has been conceived to support flux level of the order of 10 15 n/cm 2 /s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D2O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium-aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of 8 × 10 14 n/cm 2 /s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.

Journal of the Korean Physical Society, 2011

WIMS Library Update Project (WLUP) was taken up by the IAEA for updating the nuclear cross sectio... more WIMS Library Update Project (WLUP) was taken up by the IAEA for updating the nuclear cross section data libraries. The 172 group WIMS libraries (45 fast, 47 resonance groups and 80 thermal) obtained under WLUP are used for reactor physics computations. These libraries have cross section data for 173 nuclides up to Cm. Resonance Integral Tabulation (RIT) data for 28 resonant nuclides are provided for a set of background cross sections and temperature values up to 1100 ◦K. In the reactor design computations, one requires simulation of reactor states with fuel temperatures reaching nearly up to melting point of 2800 ◦C for UO2 fuel. While using deterministic codes for high temperature calculation beyond 1100 ◦K, a linear extrapolation w.r.t. √ Tfuel is normally done. This is not quite satisfactory since even a small error in the slope near the highest temperature of 1100 ◦K data point could lead to significant error if the extrapolation is done up to very high temperatures. Recently an updated WIMS library has become available through WLUP follow up activities. This library contains RIT data for 48 resonant nuclides including several minor actinides and temperature extended up to 2500 ◦K and burnup chain has been extended up to Cf. Use of the new library has alleviated the problem of possible error in extrapolation. The new library called ‘HTEMPLIB’ has been tested for the design computations of VVER-1000 MWe reactor being constructed at Kudankulam, Tamilnadu, India. Two fuel types containing 4% and 3.6% enriched fuel were analyzed using the hexagonal lattice burnup code EXCEL. The results of the lattice analyses with the new WIMS library as well as the original WIMS library ‘JEFF31GX’, containing data up to a temperature of 1100 ◦K are presented in this paper.

… into the Next …, 2000

... 3. V. Jagannathan, Usha Pal, R. Karthikeyan, S. Ganesan, RP Jain and SU Kamat, 'ATBR... more ... 3. V. Jagannathan, Usha Pal, R. Karthikeyan, S. Ganesan, RP Jain and SU Kamat, 'ATBR A Thorium Breeder Reactor Concept for An Early Induction of Thorium in an Enriched Uranium Reactor', to appear in the Journal Nuclear Technology, (Also published as Report BARC ...

Energy Conversion and Management, 2006

Thorium does not have intrinsic fissile content unlike uranium. 232 Th has nearly three times the... more Thorium does not have intrinsic fissile content unlike uranium. 232 Th has nearly three times thermal absorption cross section compared to 238 U and hence requires much larger externally fed fissile content compared to uranium based fuel. These factors give a permanent economic competitive edge to uranium. Thus thorium is not inducted in any significant measure in present day power reactors, despite the fact that thorium is three times more abundant in the earth's crust than uranium. Uranium reserves vary from country to country and there is also difficulty in having equitable distribution of uranium. Thus when 235 U would get exhausted, perhaps much sooner in countries having limited uranium reserve, there will be a need to switch over from the today's open fuel cycle programme based on 235 U feed to closed fuel cycle based on Pu feed. At that stage thorium and (depleted) uranium would become equal candidates to form the fertile base. All economic considerations would have to be readdressed. The size and growth of the nuclear power programme based on closed fuel cycle would be dependent on maximizing the fissile conversion rate in those reactors. In this paper we reemphasize the principles and the details of the thermal reactor concept 'A Thorium Breeder Reactor' (ATBR), in which the use of PuO 2 seeded thoria fuel is found to give excellent core characteristics like two years cycle length with nearly zero control maneuvers, fairly high seed output to input ratio and intrinsically safe reactivity coefficients [Jagannathan V, Ganesan S, Karthikeyan R. Sensitivity studies for a thorium breeder reactor design with the nuclear data libraries of WIMS library update project. In:

Energy Conversion and Management, 2008

There are several types of fission reactors operating in the world adopting generally the open fu... more There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., 235 U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either depleted uranium or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia to come. This has to be done by innovating means of large scale fertile to fissile conversion and then using the man-made fissile material for sustenance as well as growth of fission nuclear power. This paper attempts to give a broad picture of the available options and the challenges in realizing the theoretical possibilities.

Annals of Nuclear Energy, 2008

Radionuclides content in the discharged fuel of the conceptual thorium breeder reactor ATBR-600 h... more Radionuclides content in the discharged fuel of the conceptual thorium breeder reactor ATBR-600 has been assessed and compared against other thermal power reactors considered in Indian nuclear power programme. The contribution of actinides and the fission products inventories in the discharged fuels are separately estimated and assessed. The ATBR-600 reactor is suggested for closed fuel cycle option. The relatively large presence of the unspent plutonium would in fact be recycled. Nonetheless, the data has been presented in the event of operating ATBR-600 like other present day power reactors in a once through fuel cycle mode.

Volume 1: Plant Operations, Maintenance, Installations and Life Cycle; Component Reliability and Materials Issues; Advanced Applications of Nuclear Technology; Codes, Standards, Licensing and Regulato, 2008

Loading of seedless thoria rods in internal blanket regions and using them later as part of seede... more Loading of seedless thoria rods in internal blanket regions and using them later as part of seeded fuel assemblies is the central theme of the thorium breeder reactor (ATBR) concept [1]. The fast reactors presently consider seedless blanket region surrounding the seeded core region. This results in slower fissile production rate in comparison to fissile depletion rate per unit volume. The overall breeding is achieved mainly by employing blanket core with more than double the volume of seeded core. The blanket fuel is discharged with fissile content of ∼30g/kg, which is much less than the asymptotic maximum possible fissile content of 100g/kg. This is due to smaller coolant flow provided for in the blanket regions. In a newly proposed fast thorium breeder reactor (FTBR) [2], the blanket region is brought in and distributed through out the core. By this the fissile depletion and production rates per unit volume become comparable. The core considered simultaneous breeding from both fer...