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Morii, Tadashi*; *; Okamoto, Michiaki*; *
PNC TJ9216 98-003, 96 Pages, 1998/02
It is necessary for conforming the validity of the design based leak(DBL) of a FBR SG to evaluate the phenomena of sodium-water reaction. In this work, in order to analyze the temperature distribution in the sodium-water reaction area, we carried out the following work with using the multi-phase hydrodynamics analysis code 'CHAMPAGNE'. (1)The verification of validity of the code by the parameter analysis of the SWAT-3 Run-19 system. (2)The analyses of the actual SG. The following results have been obtained by these work. (a)The appropriate mesh division model of the helical type tubes and the boundary condition of the Run-19 system have been set. (b)The effects of the main model element (such as pressure drop in the tube group, the distributing method of reaction heat to liquid and gas phase and the reaction rate constant) on the analyses have been well understood. (c)The results of the analyses of the Run-19 system by the present optimal parameters have shown that maximum temperature level is proper. However, non-reaction steam area have been estimated more widely than originally. Therefor, it is necessary for improvement to improve the tube model and to expand the 2 dimension model to the 3 dimension model. (d)The result of the analysis which treated the cover-gas pressure as parameter have shown that the reaction temperature rises by about 50
C when the cover-gas pressure rises from 1.45 to 3.8kg/cm
. (e)The results of the analysis of the actual SG with the flow of sodium has shown that the flow of sodium causes the vertical vibration of the reaction area, and high-temperature area doesn't exist stably. It indicates that it is difficult for tubes overheating rupture to happen. It is necessary to ascertain whether or not the vertical vibration occurs under the other condition.
Chitose, Keiko*; Morii, Tadashi*
PNC TJ9214 90-004, 123 Pages, 1990/06
In the experimental fast reactor "JOYO", PNC(Power Reactor and Nuclear Fuel Development Corpolation) schedules to move two control rods from the third row of core to the fifth row, for the expansion of irradiation space, as the preparation for the shift to high power core, Firstly, only one control rod is moved. In this case, the reactivity worth of the control rod will decrease, because the control rod is shifted to outer region which has small reactivity worth. This study investigates the effect of the decrease of the reactivity worth of control rods on the transient of the accidents described in the present application of license. If the scram worth is about 0.067
K/K(decreasing from 0.074 K/K), the maximum temperatures of core increase. The change of the scram worth has an effect on the TOP(Transient overpower) accidents, but little effect on the LOF(Loss of Flow) and LOHS(Loss of Heat Sink) accidents. Then it proves that all of these maximum temperatures satisfy the safety criteria.
Morii, Tadashi*; *
PNC TJ9214 90-002, 93 Pages, 1990/04
In the experimental fast reactor "JOYO", PNC (Power Reactor and Nuclear Fuel Development Corporation) schedules to move one control rod from the inner third row of the core to the outer fifth row. Two topics have been studied in order to get a license for the shift of control rod. Firstly, the work energy generated from expansion of the disruptive core material after the hypothetical core disruptive accident have been calculated by the VENUS code. The results show that the work energy of the core after the shift of one control rod increase by about 5MJ to 78 MJ compared with that of the core before the shift, but is still smaller than 120 MJ of the work energy described in the present documentation for petition of a license. Secondary, the effect of the reactor scram under the condition of the two rods stuck has been analyzed to examine a decrease of the safety margin of the scram worth. The calculated results of the HARHO-IN code shows that the consequences of the representative 4 accidents which are described in the present documentation for petition of a license are acceptably small.
Morii, Tadashi*; Kinjo, Hidehito*
PNC TJ9214 89-008, 74 Pages, 1989/11
Effects of pin failure during Power-to-Melt test performed in the experimental fast reactor "JOYO" have been analyzed in the present study in order to obtain a license for the test. The accident scenario of the present study is that after a clad failure of a test pin, melted fuel is released and forms blockage of fuel and stainless steel debris in a coolant flow area of a compartment in the test assembly. The follwing subjects have been investigated in the present study. (1)Blockage in the coolant flow area. (2)Contact of melted fuel with compartment. Conclusions are summarized as follows. (1)Under a hypothetical assumption in which all 4 test pins are melted and form a blockage, if the outer surface of the compartment is cooled by Na flow of an about 2200kg/(m
s), the integrity of the compartment is assured. (2)If melted fuel released from the test pin directly come in contact with the inner surface of the compartment, the calculated results show that the fuel can not melt through the compartment.
Morii, Tadashi*; *; *; Chitose, Keiko*
PNC TJ9214 89-002, 193 Pages, 1989/03
In the experimental fast reactor "JOYO", PNC (Power Reactor and Nuclear Fuel Development Corporation) schedules two new tests programs in near future in order to verify high-power and high-burnup capability of LMFBR fuel pins : PTM (Power-to-Melt) test and RTCB (Run, to-Cladding Breach) test. The following safety analysis necessary to get a license for the PTM test has been performed in this study. (1)Caluculate test conditions of fuel, cladding and coolant temperature and the cladding strain of test pins. (2)Confirm that the cladding of test pin has not been broken in touch with molten fuel which oozes from central melting zone of the pin and also that taking account of the axial movement of liquid fuel strain of the cladding caused by thermal expansion of solid fuel is about 0.6%, Which is less than 3% of the cladding breaking strain. (3)Calculate temperature increases due to a mistaken setting up of coolant flow of test assembly and fing the condition which satisfies representative safety criteria.
*; *; Chitose, Keiko*; Morii, Tadashi*; *
PNC TJ9214 89-001, 159 Pages, 1989/03
In the experimental fast reactor "JOYO", PNC (Power Reactor and Nuclear Fuel Development Corporation) schedules two new tests programs in near future in order to verify high-power and high burn-up capability of LMFBR fuel pines : PTM (Power-to Melt) test and RTCB (Run-to-Cladding Breach) test. Firstly, in the present study, the temperature rises of the spare pine sadjacent to the RTCB test pin due to FP gas blanletting have been calculated with superposition of the anticipated transients and accident of the plant. The calculated results have shown that the fuel center and cladding middle temperature of the spare pines are lower than the limited values (fuel : 2630 C, cladding : 890C). Secondary, the radiation dose to population due to the accidents described in the present application of license has been calculated to be sufficiently lower than the limited values.
Himeno, Yoshiaki; ; Morii, Tadashi*; *
PNC TN9410 88-145, 11 Pages, 1988/10
Using sodium from 180kg to 3 metric tons, a series of tests has been conducted to develop the fire mitigation system and to study a design basis sodium leak accident and its consequences in the auxiliary building of the fast breeder reactor. In the test, flow pattern of a realistic leak from the sodium piping was investigated at first. Combustions of sodium in an open pool and a pool with a reduced opening were also studied together with combustion and flow of a sodium on an inclined steel liner and in a drain piping. Transient thermal conduction of a steel lined floor concrete during a sodium leak was tested and evaluated. In next, based on the results obtained, a fire mitigation system was developed and was mounted in a two storied concrete test rig. Then, a large scale test starting from a leak and ending by a self-extinguishment of a fire in the smothering tank was carried out. In final, consequences of sodium aerosols deposition on the reactor components and the electrical instruments have been experimentally studied.
; Morii, Tadashi*; Hiroi, Hiroshi*; Himeno, Yoshiaki
PNC TN9410 88-104, 110 Pages, 1988/08
[objectives] The detailed sodium fire analysis code, SOLFAS (Sodium Leak Fire and Aerosol Analysis Code System), is under development to make more accurate calculation for heat/mass transfer during a sodium fire. Objective of the present study is to investigate basic characteristics of version-0.1 of the code in regard to the solution scheme and the numerical algorithm. [Methods] Parametric calculations of single-phase one-component laminar natural convection heat transfer on a horizontal plane were conducted to investigate the dependence of some parameters on the calculated results as well as the CPU-time. [Results] Present study revealed that distribution of the calculational cells plays an important role to obtain results with high accuracy. The CPU-time strongly depended upon such parameters as time step, convergence condition, relaxation factors, over-relaxation factor, and so on. By selecting appropriate values for these parameters, reasonable results were obtained in the wide laminar flow region.
*; Morii, Tadashi*; *; *; *; *; Kinjo, Hidehito*
PNC TJ9214 88-003, 394 Pages, 1988/06
In the experimental fast reactor JOYO, two new tests programs are scheduled in near future in order to verify high-power and high-burnup capability of LMFBR fuel pins : PTM(Power-to-Melt) test and RTCB(Run-to-Cladding Breach) test. A study is performed for the license of the tests. In this study, the fuel assemblies for the PTM and RTCB tests is designed, Safety assessment for the possible events occurred in the course of the tests, namely melt of sample fuel (PTM test) and cladding breach (RTCB test) is performed. Lastly, safety assessment for the accidents occurred in the course of the tests is also performed to obtain a license. As for the PTM test, the fuel assembly is designed considering thermal expansion of fuel pins and designed in order to prevent local blockage caused by released sample fuel and hold the released fuel within the assembly in preparations for the worst case of pin failure. The effect of reactivity insert by fuel slumping and pellet-cladding mechanical interaction caused by fuel melt within the test pin occurred during the PTM test is studied to show an integrity of the test pins. Analytical results show an integrity of test pins under the conditions with superposition of anticipated transient whose possibility of occurrence is relatively high. Lastly, even if the test pin is breached to release sample fuel, the calculated radiation dose to population is shown to be sufficiently lower than the limited values. As for the RTCB test, the devices are proposed to lower the released rate of FP gas and prevent an adjacent pin failure caused by the released gas in the cource of the test. Analytical results for the event of gas impingement on the adjacent spare pins show no additional breach of those pins. Calculated results with superposition of the representative anticipated transients and accidents also show an integrity of the adjacent spare pins. The released radio-activities calculated for the above events are sufficient smaller ...
*; *; *; *; Morii, Tadashi*
PNC TJ9214 88-002, 423 Pages, 1988/06
This study investigates an applicability of the 'Guideline of LMFBR Safety', which was established in a licensing process for the prototype fast breeder reactor "MONJU" by Nuclear Safety Committee, to licensing for the facility change of "JOYO" in performing an efficiency test of FBR fuel. The philosophy for safety design which must be described in appendix 8 of the application of license for the facility change is proposed under the provisions of the 'Guideline of LMFBR Safety'. The following three accidents which was judged to require further analysis after reconsideration of the accident classification under the provisions of the 'Guideline of Assessment of LMFBR Safety' are evaluated. The calculated radiation dose to population is shown to be sufficiently lower than the limited values. (1)Fuel Handling Accident. (2)Sodium Leak Accident from Overflow Tank. (3)Sodium Leak Accident from Cold Trap. This report also includes the study on that the recriticality accident described in the present application of license can be classified in LPHC (Low Probability High Consequence) events. The appendix 10 of the application of license for the facility change is proposed in order to perform an efficiency test of FBR fuel.
Himeno, Yoshiaki; ; Morii, Tadashi*
PNC TN9410 88-094, 18 Pages, 1988/05
Flow pattern of a realistic sodium leak from the sodium piping equipped with jackets and thermal insulator was experimentally investigated. Then, based on this result, the fire mitigation system consisting of an inclined liner, a drain piping, and a smothering tank has been developed. The performance of the system was, in final, validated in the large-scale sodium leak and fire test in the SAPFIRE facility.
Morii, Tadashi*; *; *
PNC TN9410 87-088, 59 Pages, 1987/06
PNC-TN9410-87-088.pdf:3.33MBPNC-TN9410-87-088TR.pdf:3.29MB
A test, Run-E2, of sodium leak from a 1/3.5 scale simulated Na piping of the secondary circuit of Monju was conducted using the SOLFA-2 in the SAPFIRE facilities. In the simulated piping, a leak hole with an area reduced by (1/3.5) from an area of 1/4Dt of the actual piping was made on the upper wall of the piping in advance. In the test, a pressure equal to the system pressure of 3.8 kg/cmg of the hot leg piping in secondary circuit of the actual plant was applied, and sodium was spilled. Spill duration was approximately 13 minutes. The test results showed that the integrity of the insulation structure around the piping will not be broken by a Na Pressure and combustion heat during an accident, therefore, the spray leakage of Na can be fully prevented. Moreover, burning rate of Na during leakage was approximately 4% of the flow rate of the Na leak. Compared to about 30% obtained from the spray combustion test conducted previously using a spray nozzle, the combustion following a realistic Na Leak from the actual piping is found to be milder than that of spray combustion.
Morii, Tadashi*; Himeno, Yoshiaki
PNC TN9410 87-006, 51 Pages, 1987/01
In the study, the test results from SOLFA-2 of the SAPFIRE facility were used. But, the calculated results using the standard input data underestimated the test results. So, the other calculations by changing the following parameter were conducted. (1)Burning rate constant of a sodium droplet. (2)Gas buoyant force. (3)Pool burning effect. (4)E㎜isivity of aerosol containing gas. (5)Spray droplet size. Changing the above parameter from (1) to (4) did not make any significant improvement, in other words, the results still underestimated the test results. But, changing of the (5)parameter improved the calculated results. However, the best estimated results was with a spray droplet whose size is a half of a real one.
Morii, Tadashi*; *; *
PNC TN9410 86-124, 61 Pages, 1986/12
PNC-TN9410-86-124.pdf:3.08MBPNC-TN9410-86-124TR.pdf:3.23MB
On Sept. 27, 1985, a large scale sodium spray fire test (RUN-E1) has been conducted in an air atmosphere using the SOLFA-2 test vessel (100m
made from SUS) of the SAPFIRE facility. The major test conditions are as follows. (Spray Rate : 510 g/sec) (Spray Period : 1800 sec) (Spray Inlet Temperature : 505 C) (Spray Falling Height : 4 m) As a sodium spray started, the gas pressure and temperature rose rapidly and reached to the maximum values 1.24kg/cm-g and 700C, respectively, after about 1.2 minutes. The oxygen in the test vessel was consumed completely after 4 minutes. From oxygen consumption rate during this time, burning rate of sodium was calculated to be 160g-Na/sec that was equivalent to about 30% of the sodium spray rate (under the assumption of 100% Na
O production). Many thermo-couples installed in a spray corn region have been failed due to their exposure to the high temperature above 1000 C, which suggested the existence of a burning zone around the sodium droplets. No remarkable distribution of oxygen concentration was observed in the vertical direction of the vessel during a spray, indicating that the gas within the vessel was well mixed by natural convection due to gas temperature difference between the outside and the inside of a spray corn. Aerosol concentratian has reached the maximum value of 17.5g-Na/m after 5 min and decreased below 1 g-Na/m after 20 min.
Morii, Tadashi*; Himeno, Yoshiaki
PNC TN9410 86-066, 27 Pages, 1986/06
Sodium fire analysis has been performed for a large FBR to evaluate pressure and temperature transients and mass of burned sodium in case of a primary sodium leak accident. The major analytical conditions are as follows: [Position of sodium leak : Hot leg of primary coolant system] [Cross-sectional area of a leak hole : 1 cm] [Concrete cooling system : operated (just before failure), shut down (after sodium leak)] The most representative results gained through the present study are as follows: [Maximum Gas Pressure : 0.029 kg/cm2 -g (0.5 hr after a leak)] [Total Mass of Burned Sodium : 1.5 ton (3% of total leak sodium)] [Maximum Concrete Temperature (beneath sodium pool) : 140C (100 hr after a leak) These results indicate that a concrete cooling system to present abnormal temperature rise that may occure due to heat transfer from the hot primary coolant system was shown to be effective even in the accident conditions. However, further study will be needed to evaluate water release rate from the heated concrete.