On a non-detailed reading, I came across the following statements in Mr T. W. Venkat Krishna's PhD thesis quoted above, which to me, appear to be inaccurate.NRao wrote:A thesis from TX A&M 2007:
NUCLEAR FUEL CYCLE ASSESSMENT OF INDIA: A TECHNICAL STUDY FOR U.S.–INDIA COOPERATION
Geek value:
In at least 4 places (pages 40, 51, 54, and 67) it has been indicated that computer simulations with ORIGEN2.2 and HELIOS-1.4 lattice code have been used, using a 37-pincell CANDU fuel bundle, [for the 220 MWe PHWRs such as RAPS1 etc] for assessment of Pu production from beginning till 1998, power production and Uranium consumed, fuel cycle analysis for future projections, and meeting the fuel needs of future fast breeder reactors.
On the contrary, India's 220 MWe reactors are based on Canada's Douglas Point Reactor which had a 19 - pencil fuel rod design. Please see here (html), here (jpg). and here (jpg). Also see Slide 44 of this presentation by NPCIL at IAEA (pdf document).
In Pages 36, while on the topic of "Assessment of Plutonium Production and Use: Beginning till 1974" and again in page 70 while discussing "Future of Weapons Program", Mr Venkat Krishna asserts, "The assumptions and conclusions have been reached on the basis of reactor operator experience and knowledge of design details." {I presume that this is applicable to his entire thesis and refers to his own experience and knowledge. If so, I imagine that he might be (or was) a DAE employee.}
Except for the two 540 Mwe reactors Tarapur 3&4, all the other unsafeguarded reactors are 220 MWe plants, and hence the assumptions cited above seem non-applicable. Modelling using a 37-pincell fuel rod assembly for the proposed Thermal Breeder might be OK since at present it is only at a preliminary conceptual design stage.
At page 35 Mr Venkat Krishna indicates that the quantity of depleted U required for RAPS 1 operation was calculated on the assumption of "92mk excess reactivity". He has indicated what would happen if it were to be lower than 92mk but has not said how his conclusions might change if the actual excess reactivity maintained during plant operations was more than 92 mk.
In "Chapter III Indian Nuclear Facilities", he states at page 36:
Quote :
. . .
Mining is followed by milling for recovering U3O8 from the ore. There is on average a 15% loss in processing activity (ore to U3O8) and 20% losses in conversion from U3O8 to UF6. The UF6 is sent to the NFC for fabricating into uranium oxide pellets for PHWR fuel bundles. UF6 is also used as a feed for the uranium enrichment plant.
Unquote
However, at NFC's web site we see the following:
Quote:
PHWR Fuel
The raw material for the production of PHWR fuel in NFC is Magnesium Di-Uranate (MDU), popularly known as yellow cake. The MDU concentrate is obtained from the uranium mine and mill at Jaduguda, Jharkhand. The impure MDU is subjected to nitric acid dissolution followed by solvent extraction and precipitation with ammonia to get Ammonium Di-Uranate (ADU). Further steps of controlled calcination and reduction form sinterable uranium dioxide (UO2) powder.
The UO2 Powder is converted to high-density cylindrical pellets by various operations like; pre compaction, granulation, binder addition, final compaction and sintering at high temperature (1700 degree centigrade) in hydrogen atmosphere. The sintered pellets are then centreless ground to desired dimensions and dried.
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In other words, UF6 is not the input material to NFC for the NatU fuel of Indian PHWRs.
Before ending I must confess that I have no idea as to how the above aspects may affect the final conclusions reached by Mr. Venkat Krishna in his Doctoral dissertation. Perhaps members of this Forum having expertise in Reactor Physics aspects of India's PHWRs may comment.
TIA.