P. Mohanakrishnan

Theoretical and experimental investigations of reactor parameters in a U-233 fuelled research reactor

The Kalpakkam Mini Reactor (KAMINI) has the unique distinction of being the only operating pool type research reactor in the world at present, that utilizes U-233 as the fissile material. It is a 30 kW pool type reactor, operated with an alloy of U-233 and aluminium as fuel, light water as moderator and beryllium oxide as reflector. It is designed to operate at a maximum power level of 30 kW and mostly used for neutron radiography and neutron irradiation experiments. In the present work core calculations are carried out to estimate the worths of safety control plates (SCPs) and one adjustable reflector block of KAMINI by using the lattice homogenization code SMAXY and 3-D transport code TRITAC along with the WIMS cross-section data library. The worth of the SCPs is determined experimentally by rod drop method. A comparison of the calculated worth values, is done with the experimentally determined ones. The neutron spectra measurements carried out at two-sample irradiation locations, called thimble irradiation locations in KAMINI are also presented in this work. The neutron spectra at the thimble irradiation locations are generated by, the help of core calculations performed by SMAXY, TRITAC and WIMS combination. The measured total fluxes, fluxes in various neutron energy ranges and the gold reaction rates at both the thimble locations are compared with the calculated values. It is found that the calculated values of worths and fluxes match well with the measured values.

Moderator temperature effect on reactivity in light water moderated experimental reactors

Abstract The moderator temperature coefficient of reactivity has been measured in a U-233 fuelled plate type light water moderated reactor (KAMINI). Using neutron cross sections based on WIMS library, lattice homogenization code SMAXY and 3D core calculation code COMESH the moderator temperature coefficient of reactivity is predicted very well (within 0.6 pcm per °C). For U-235 and Pu-239 fuelled KRITZ experiments also, moderator temperature coefficients of reactivity have been predicted and compared with Studsvik results using lattice code MURLI and 2D core calculations. Our predictions of the moderator temperature coefficient of reactivity is fairly good for these cores. For U-233 fuelled KAMINI, the close agreement on moderator temperature coefficient of reactivity was not found using more recent U-233 cross-sections derived from ENDF/B-IV (WIMKAL). By comparing the α (capture to fission ratio) of U-233 at low energies between WIMS and ENDF/B-IV and VI, the better comparison with WIMS is attributed to the flatter α behavior with energy. There appears to be a need for measurement of α of U-233 below 0.1 eV.

Effects of cross section sets and quadrature orders on neutron fluxes and on secondary 24Na activation rate of a pool type 500 MWe FBR

In the early stages of shield design for the 500 MWe pool type Prototype Fast Breeder Reactor (PFBR), the DLC-37 coupled (100n, 21 γ) cross section set, extended to include data for the plutonium isotopes, only was used for the transport calculations. Several limitations were recognised in this set. Two new sets IGC-S2 (100n, 21 γ) and IGC-S3 (175n, 42 γ) have been created and validated at IGCAR, based on ENDF/B-VI (5). Comparison of the results of two-dimensional transport calculations using DLC-37, IGC-S2 and IGC-S3 is presented in this paper, for PFBR. Since the activation of secondary sodium in the Intermediate Heat Exchanger (IHX) depends significantly but differently on the axial and radial neutron fluxes leaking from the core, 2-D transport calculations with reasonable orders for the anisotropy and angular quadrature are necessary. The effects of anisotropy in the scattering cross sections and the SN order, on the shield parameters are studied with IGC-S2. It is found that 175-group calculations clearly predict the finer details of the neutron spectrum in the sodium pool and would facilitate more accurate shield design of PFBR.

Generation and validation of a new 121 group coupled (n, γ) cross section library for fast reactor applications

Abstract A new 121 group (100 neutron group, 21 gamma group) coupled cross section library with anisotropic scattering has been generated at IGCAR for 25 nuclides by using the 1985 version of NJOY code system. The basic data used are ENDF/B-IV for neutrons and DLC-99 for photons. A set of six neutron source benchmarks, one gamma source benchmark and one gamma production benchmark has been used in validating this new library. Of these, the first six are computational benchmarks and the last two are experimental benchmarks. Reasonably good predictions have been obtained with this new set relative to the reference solutions. Thus, capability has been developed at IGCAR for generating new multigroup cross section libraries for shielding applications.

Reliability comparison of computer based core temperature monitoring system with two and three thermocouples per sub-assembly for Fast Breeder Reactors

Prototype Fast Breeder Reactor (PFBR) is a mixed oxide fuelled, sodium cooled, 500 MWe, pool type fast breeder reactor under construction at Kalpakkam, India. The reactor core consists of fuel pins assembled in a number of hexagonal shaped, vertically stacked SubAssemblies (SA). Sodium flows from the bottom of the SAs, takes heat from the fission reaction, comes out through the top. Reactor protection systems are provided to trip the reactor in case of design basis events which may cause the safety parameters (like clad, fuel and coolant temperature) to cross their limits. Computer based Core temperature monitoring system (CTMS) is one of the protection systems. CTMS for PFBR has two thermocouples (TC) at the outlet of each SA(other than central SA) to measure coolant outlet temperature, three TC at central SA outlet and six thermocouples to measure coolant inlet temperature. Each thermocouple at SA outlet is electronically triplicated and fed to three computer systems for further processing and generate reactor trip signal whenever necessary. Since the system has two sensors per SA and three processing units the redundancy provided is not independent. A study is done to analyze the reliability implications of providing three thermocouples at the outlet of each SA and thereby feed independent thermocouple signals to three computer systems. Failure data derived from fast reactor experiences and from reliability prediction methods provided by handbooks are used. Fault trees are built for the existing CTMS system with two TC per SA and for the proposed system with three TC per SA. Failure probability upon demand and spurious trip rates are estimated as reliability indicators. Since the computer systems have software intelligence to sense invalid field inputs, not all sensor failures would directly affect the system probability to fail upon a demand. For instance, the coolant outlet temperature cannot be lower than the coolant inlet temperature. This intelligence is taken into account by assuming different “fault coverage percentage” and comparing the results. A 100% fault coverage means the software algorithm could detect all of the possible thermocouple faults. It was found that the system probability to fail upon demand is reduced in the new independent system but the spurious trip rate is slightly worse. The diagnostic capability is marginally affected due to complete independence. The paper highlights how an intelligent computer based safety system poses difficulties in modeling and the checks and balances between an interlinked and independent redundancy.

Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs

Abstract Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed.