International Nuclear Information System (INIS)

Uranyl nitrate ("2"3"3U) was retained mainly in lung, skeleton, pelt and kidney 60 days after inhalation exposure. The concentration in lung exceeded that in other tissues. Preliminary results with "2"3"2U indicate retention in lung similar to that of "2"3"3U, at total deposited mass levels of 0.1 #mu#g or less.

International Nuclear Information System (INIS)

Absorption of ''organically bound'' "2"3"3U and "2"4"1Am from gastrointestinal tracts of adult rats was twice that of the inorganic nitrate form. There was no difference between transport of "2"3"2U and "2"3"3U by adult rats, but "2"3"2U absorption by newborn rats was four times that of "2"3"3U. Absorption of "2"3"8Pu nitrate through the gut of the newborn is relatively insensitive to dose but incorporation in the gut mucosa saturates at high dose levels. Absorption of "2"3"8Pu is greater at 4 hr of age than at 24 hr.

Energy Technology Data Exchange (ETDEWEB)

The delayed neutrons from neutron-induced fission in /sup 232/U, /sup 237/Np, /sup 238/Pu, /sup 241/Am, /sup 242/Am/sup m/, /sup 245/Cm, and /sup 249/Cf were studied for the first time; those from /sup 232/Th, /sup 233/U, /sup 235/U, /sup 238/U, /sup 239/Pu, /sup 241/Pu, and /sup 242/Pu were measured again. The data were used to develop an expression for the prediction of the absolute delayed neutron yield, and the prediction of delayed neutron emission with time. This approach accurately predicts observed delayed neutron yields and decay characteristics. A fission product yield model was used in conjunction with delayed neutron emission probability to analytically predict delayed neutron characteristics. The results of this analysis are in excellent agreement with experimental values.

International Nuclear Information System (INIS)

High resolution Monte-Carlo simulations show that the neutron spectrum, fuel burnup and fuel temperature feedback effect of a PWR core loaded with Thoria-based fuel (Th/Pu-O_2) do not significantly differ from the MOX fuelled one due to the similar neutronic characteristics of both fertile materials (Th-232, U-238). The core physics of this fuel variant is characterized by an enhanced moderator/void temperature coefficient (by factor 2.4) and high incineration rate for Pu (approx. 60 %). A PWR core loaded with the Molybdenum-based inert matrix fuel (IMF) - in contrast to MOX-, shows a harder spectrum, resulting in small temperature coefficients of reactivity and particularly in a higher fuel depletion rate as well as an enhanced TRU reduction performance. The incineration of Pu amounts to 46 % resulting, in turn, in generation of minor actinides of about 10 % of the total Pu consumption. The higher excess reactivity resulting from the initial ...

International Nuclear Information System (INIS)

1 - Description or function: Continuous energy cross-section data library for the Monte Carlo program MCNP based on the JEF-2.2 evaluated nuclear data library (ACE Format). Format: ACE Number of groups: Continuous energy Nuclides (107): H-1, H-2, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, O-17, F-19, Na-23, Mg-nat, Al-27, Si-nat, Cl-nat, Ti-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Zr-90, Zr-91, Zr-92, Zr-94, Zr-96, Zr-nat, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-100, Mo-nat, Tc-99, Ru-101, Ru-102, Ru-104, Rh-103, Pd-105, Pd-107, Ag-109, I-129, Xe-131, Cs-133, Pr-141, Nd-143, Nd-145, Pm-147, Sm-147, Sm-149, Sm-150, Sm-151, Sm-152, Eu-153, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Pb-nat, Bi-209, Th-232, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-239bis, Pu-240, Pu-241, Pu-242, ...

International Nuclear Information System (INIS)

1 - Description of program or function: MCB-JEF2.2 is a continuous-energy cross section libraries in ACE Format suitable for the MCB-1C and MCNP codes. Libraries for various materials were generated at six different Temperatures, and cover the energy range up to 20 MeV. Format: ACE. Number of groups: Continuous energy. Nuclides: H-1, H-2, H-3, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat., N-14, N-15, O-16, O-17, Na-23, F-19, Mg-nat., Al-27, Si-nat., P-31, S-32, S-33, S-34, S-36, Cl-nat, K-nat, Ca-nat., Ti-nat, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-59, Ni-60, Ni-61, Ni-62, Ni-64, Cu-nat, Ga-nat, Ge-72, Ge-73, Ge-74, Ge-76, As-75, Se-74, Se-76, Se-77, Se-78, Se-80, Se-82, Br-79, Br-81, Kr-78, Kr-80, Kr-82, Kr-83, Kr-84, Kr-85, Kr-86, Rb-85, Rb-86, Rb-87, Sr-84, Sr-86, Sr-87, Sr-88, Sr-89, Sr-90, Y-89, Y-90, Y-91, Zr-nat, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Nb-94, Nb-95, Mo-nat, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, ...