In 1994 it was shown, by a computational method, at the All-Union Scientific-Research Institute of Technical Physics that the ratio of the thermal-neutron fluence in the central experimental channel of a pulse uranium-graphite reactor [ 1] to the energy release increases by approximately 30-40% as the core heats up from 300 to -1200 K. This effect, called for brevity "bleaching" of graphite, is caused by a decrease of the effective trapping and fission cross sections with increasing energy of the thermal neutrons as a result of their thermalization in graphite at a higher temperature. On account of the bleaching, as the uranium-graphite fuel is heated not only is a nonlinear relation between the energy release and fluence observed, but the mean-free path length of thermal neutrons and therefore their leakage from the core increase, which gives rise to intense self-quenching of the reactivity and determines the reactor dynamics, the parameters of the fission pulses, the required efficiency of the control rods, the excess reactivity introduced, and the consequences of accidents. Computational substantiation of the multichannel pulsed uranium-graphite reactor [2] is in progress at the All-Russia Scientific-Research Institute of Technical Physics, using the PRIZMA-D [3] and MCNP [4] programs, which have been verified on experimental data [5, 6]. Calculations have shown that the graphite bleaching effect is stronger in a multichannel pulsed graphite reactor than in pulse graphite reactor, and the indicated ratio of the neutron fluence to energy release for the MIRG reactor increases by 100% as the core heats up from room temperature to 2000 K. To verify the programs, experiments were performed on the pulse graphite reactor measuring the fluence of thermal and fast neutrons in bursts with small energy release at a different initial core temperature and with the fuel heated up to 1020 K in 1995 and up to 1220 K in 1997. Measurement Procedure. To measure the fluence of thermal and fast neutrons 176Lu, 197Au, 186W, 139La, 59Co, 63Cu, 23Na, 45Sc, 5~ 27A1, 58Ni, and 47Ti pure-metal foil activation detectors consisting based on aluminum alloys or epoxy resinswere irradiated in open form and in cadmium screens -15 mm in diameter and 7 mm high cylindrical containers with 1 mm thick walls. During irradiation the open detectors and the cadmium-containing detectors were separated from one another by at least 5 cm in the horizontal direction. Three sets of detectors were irradiated in each run: one set was placed 100 nun below the geometric center of the central experimental channel in empty space, the second set was placed 100 nun above the center in a 80 mm in diameter and 80 mm high steel container with a 9.3 mm thick wall, and the third set was placed at the geometric center of the side experimental channel in empty space. The temperature of the active zone was measured with a standard thermocouple, placed at the center of the column "g-9," and its indications were close to the average-volume temperature of the core.
