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"145" XA0102831 Paper presented in IAEA AGM on Long Term Needs for Nuclear Data Development, 28 November to 1 December, 2000 At IAEA, Vienna Indian Advanced Heavy Water Reactor for Thorium Utilisation and Nuclear Data Requirements and Status R. Srivenkatesan, Umasankari Kannan, Arvind Kumar, S. Ganesan and S.B. Degwekar Bhabha Atomic Research Centre Mumbai 400 085 INDIA Abstract BARC is embarking on thorium utilisation program in a concerted and consistent manner to achieve all round capabilities in the entire Thorium cycle under the Advanced Heavy Water Reactor (AHWR) development program. Upgrading our nuclear data capability for thorium cycle is one of the main tasks of this program. This paper gives a brief overview of the physics design features of the AHWR. The basic starting point of the analysis has been the lattice simulation of the fuel cluster employing the WIMS-D4 code package with 1986 version of 69 group library. For the analysis of thorium cycle, the present multi group version contains the three major isotopes viz., 232Th, 233U and 233Pa. To correctly evaluate the fuel cycle we require many more isotopes of the Th burnup chain. With the help of NDS, IAEA many other isotopes of interest in AHWR, actinides in the thorium burnup chain, burnable absorbers, etc., were generated. Some of them were added to the WIMS-D4 library and the results are discussed. The WIMSD-4 library is also being updated as part of the IAEA coordinated research project on Final Stage of WLUP with international cooperation. India is also taking part in CRP. The evaluation of AHWR lattice with this new library is presented. Some comments regarding the fission product data being used in WIMS libraries are given, which are tuned to U-Pu cycles. The measurements for 233U are rather old. Measurements in high energies are also very sparse. More attention by nuclear data community is required in this regard as well. India has also begun a modest program to assess the ADS concepts, with the aim of employing thermal reactor systems, such as AHWR. A one way coupled booster reactor concept is being analysed with available code systems and nuclear data. A brief summary of this concept is also being discussed in this paper. A general survey on the quality of the evaluated nuclear data of the major and minor isotopes of thorium cycle is also given. A major international effort is necessary to bring the nuclear database for the uranium-thorium cycle up to the reliability level of that for the uranium- plutonium fuel cycle. -146- 1. Introduction The long term objective of thorium utilisation on a large scale is one of the most distinguishing features of the Indian Nuclear Power Program. The inherent nuclear charcteristics of 233U, particularly the eta value in thermal energies make heavy water reactors as good option for thorium utilisation. Due to our extensive experience in PHWRs we would like to keep the technological advantages of reactor design and operation. At the same time one should build-in new safety features to be consistent with the modern perceptions of advanced reactors. Hence the AHWR is being developed in India for utilising thorium for power generation with passive safety features. [1,2] AHWR is a vertical pressure tube type reactor cooled by boiling light water and moderated by heavy water. The reactor has been designed to produce 750 MW(th) at a discharge burnup of about 25,000 MWd/te. Negative void coefficient has to be achieved, in spite of using boiling light water coolant. Another design objective is to be self- sustaining in 233U with most of the power from the conversion of thorium fuel while using plutonium as the external fissile feed. Also one has to minimise the external Pu feed in the equilibrium fuel cycle, such that major fraction of power is produced by the in situ conversion of Th and burning of 233U. AHWR incorporates several advanced passive safety features, e.g., heat removal through natural circulation, ECCS injection directly into the fuel cluster through a dedicated water channel inside the cluster, passive containment isolation etc. [2] In this paper after briefly summarising the physics design features of AHWR, the current methodologies used in the lattice physics simulations and nuclear data status are discussed. Then we comment on the nuclear data requirements for the thorium cycle in general. 2. Physics Design Features of AHWR The fuel design of AHWR has followed an evolutionary path ranging from a seed and blanket concept to a simplified composite cluster to achieve a good thermal hydraulic coupling. The composite cluster design also has changed from a square cluster to a circular cluster inside a circular pressure tube. [3] The aim was to design a single cluster, which is capable of achieving a negative void coefficient as well as satisfy the thermal hydraulic requirements. If all the fuel assemblies are similar and the core has a flatter power distribution, it would be possible to achieve optimised coolant conditions from the point of view of heat removal by natural circulation. The power density should also be low to get optimum heat flux ratios with the boiling coolant driven by natural circulation. With this in view, a composite cluster has been designed using two kinds of fuel namely, (Th-233U) MOX and (Th-Pu) MOX [3,4]. Plutonium has to be kept in the outer pins of a fuel cluster facing the moderator from reactivity and void coefficient considerations, while 233U can be kept inside in a relatively harder spectrum without losing reactivity. -147- The 233U content in the composite cluster has to be higher than the self-sustaining limit of about 1.5 % from reactivity as well as local peaking factor considerations. The overall self sustaining feature of 2 3U is achieved by its production in the outer (Th-Pu) MOX pins. The void reactivity is made negative in spite of using light water coolant by incorporating a large central (coolant displacer) rod with a slow burning absorber such as dysprosium - about 3 wt% dysprosium oxide in zirconium oxide. The cluster design is given in Fig. 1. The displacer rod contains a central hole through which the ECCS water is injected from the top and the water jets emerge all along the cluster. The use of such an absorber rod compromises the excellent neutron economy of a HWR, but offers inherent safety, which is the overriding principle. The core consists of 500 lattice locations, of which 452 will contain fuel, while the rest of the channels are used for Shut Down System -1 and control (adjuster) rods. The core has a height of about 5 meters, of which the fuel length is 3.5 m. The core diameter is about 8 m. The core layout is given in Fig. 2. The void reactivity of the core is nearly zero under nominal operating conditions, being within —3 mk for coolant density variation from inlet value of about 0.75 to 0.05 gm/cc. The coolant temperature coefficient variation of about +3 mk from cold to hot-stand-by condition, and a negative power coefficient of about 6 mk from hot-stand-by to hot operating condition, constitute the parameters that will dictate stability of the reactor. An important aspect of AHWR is also its Pu burning capabilit}'. The following table shows the initial and discharge vectors of Pu isotopes. The initial charge is from the reprocessed Pu from Indian PHWR employing natural uranium fuel (average exit burnup of about 6500 MWD/t). 239 241 242 Pu pu240 Pu Pu Initial 68.8 % 24.6 % 5.3% 1.3% Discharge 10.3 % 51.9 % 19.9 % 17.9 9, With each recycle of 233U, the amount of 232U will increase leading to inconvenient handling problems. Suitable development strategies are being evolved to cope with these problems. AHWR also offers an option to burn some minor actinides such as 231Pa, 237Np etc. Some results of our analysis are given in § 3.2. We are also planning a low power critical facility, which will validate our simulation strategies for the novel thorium fuel including nuclear data employed. 3. Calculational methodology and strategies for AHWR The WIMS D4 code and library has been used for lattice simulation of AHWR. This 69 group library has been extensively validated for the plutonium-uranium cycles. The thorium burnup chain is shown in Fig. 3. The 1986 version WIMS library we have used -148- has data tables only for three isotopes of the thorium cycle namely, 232Th, 233Pa and 233U. The WIMS library released this year as part of the IAEA-CRP also contains only these isotopes [5]. This is not sufficient to fully analyse the fuel cycle issues of thorium cycle or AHWR. Since AHWR will operate in a closed fuel cycle, we have to recycle the uranium isotopes generated in AHWR. This demands the assessment of shielding for the remote handling facilities for reprocessing of the discharged fuel and fabricating new clusters for future fuel loading. Although from the spent fuel point of view the generation of higher actinides is less than that of the uranium cycle, the isotopes having shorter half lives and significant absorption cross section cannot be ignored. For example, 232U is important from the point of view of spent fuel handling as the daughter products of 232U are highly radioactive. The concentrations of 2 2U has to be determined with at least 10 % accuracy [6]. 232Pa is another isotope which has got significant fission cross section for thermal neutrons ~700 barns. (There are some difference of opinions about the significance of this isotope.) Other isotopes of radioactivity significance in these processes are 228Th, 229Th, 23 Th, 231Pa,234Uetc.
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