ISSN:0029-5450    

DOI:10.13182/NT87-A33916

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

Boiling water reactor fuel channels bulge and bow because of pressure, temperature, fast neutron flux, and their gradients. Channel deformations can be calculated by means of the finite element technique. Calculated bulge and bow results for WNP-2 fuel channels in different core locations and at different power levels have been obtained as functions of core residence time. In general, channel bulge is largest at the core center and decreases toward the core periphery. Bulge increases with the power level and the core residence time. Channel bow is largest at the core periphery and decreases for the next two rows of channels radially inward. Bow rate is highest in the first reactor cycle and then decreases. After an initial period, bow ceases to increase with residence time and may even decrease. The analytical results are being used by the Channel Management Program at Washington Public Power Supply System to optimize the utilization of fuel channels.

ISSN:0029-5450    

DOI:10.13182/NT87-A33917

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

An unprotected loss-of-flow (LOF) event has been analyzed for a 1000-MW(electric) axially heterogeneous core (AHC) at the end of an equilibrium cycle, using a realistic model to evaluate the AHC safety potential. The SAS3D code was used for the initiating phase analysis, while a phenomenological approach was employed for the transition phase. The SAS3D results showed that the system rapidly approached subcriticality after experiencing a benign power burst, because of axially flattened fuel worth distribution and reduced sodium-void worth particularly around the core center. During the transition phase, fuel-steel discharge into the interassembly gaps, coupled with engagement of the upper axial blanket material in the core region, was found to result in permanent subcriticality and a nonenergetic termination of the LOF event.

ISSN:0029-5450    

DOI:10.13182/NT87-A33918

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

A method was developed for faster than real-time liquid-metal fast breeder reactor intermediate heat exchanger (IHX) analysis for purposes of continuous on-line data validation, plant state verification, and fault identification. The basic feature of this method is the utilization of spatial nodes whose sizes vary with time. The use of time-variant node sizes leads to adequately accurate solutions with a few nodes and at short computation times. Applications of this methodology to reference IHX problems with the IBM 3033 machine showed that the computation time for steady-state analysis was∼6 ms. For transient analysis, a computation time that was one-sixteenth of the real transient time was achieved. This time can be further reduced if the special sparse structure of the system Jacobian matrix is exploited. The analysis of the Experimental Breeder Reactor-II test 8A showed that the maximum difference between temperatures predicted by this methodology and measurements was∼6K.

ISSN:0029-5450    

DOI:10.13182/NT87-A33919

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

During the events of a reactor scram, the control room operators play a vital role in the diagnosis of the causes and control of the plant. It is critically important that the operators immediately detect an abnormal scram situation related to the plant protection system and take necessary actions to shut down the nuclear reaction safely. The present study develops a proof-of-principle prototype of a postscram analyzer. It is an operator aid information system designed to assist the operators in the recognition of possible abnormal scram situations immediately after a scram and to facilitate postscram analysis for diagnosis of root causes and for speedy plant restart. The resultant displays for man-machine interface demonstrate that a postscram analyzer can provide vital and concise information in the control room to enhance the productivity of the plant operators.

ISSN:0029-5450    

DOI:10.13182/NT87-A33920

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

To investigate the coolability of a uniformly heated tube by free convection of atmospheric air, heat transfer experiments were conducted using vertical open an-nuli with adiabatic outer walls. To examine the effect of the annulus ratio on the coolability of the heated tube, the experiments employed four annuli (diameter ratios of 1.155, 1.33, 1.63, and 12.0). The operating parameters included heat fluxes up to 1.38 W/cm2with a corresponding surface temperature of 856 K.The results, extrapolated to 1200 K, were used to provide a qualitative estimate of the coolability of multirod bundles, as a function of the equilibrium surface temperature and the pitch-to-diameter (P/D) ratio. Although the decay heat removal rate for P/D values<1.5 increased rapidly with P/D ratio, for larger P/D values the decay heat removal rate was insensitive to either the P/D value or the rod arrangement in the bundle. These results suggest that in TRIGA-type reactors at a typical P/D ratio of 1.12, the maximum decay heat removal level is∼1 kW/m. This maximum corresponds to an initial decay power following sustained operation at∼12.5 kW/m.

ISSN:0029-5450    

DOI:10.13182/NT87-A33921

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

A method to evaluate the effect of human factors on probabilistic safety analysis (PSA) is developed. The main features of the method are as follows:1. A time-dependent multibranch tree is constructed to treat time dependency of human error probability.2. A sensitivity analysis is done to determine uncertainty in the PSA due to branch time of human error occurrence, human error data source, extraneous act probability, and human recovery probability.The method is applied to a large-break, loss-of-coolant accident of a boiling water reactor-5. As a result, core melt probability and risk do not depend on the number of time branches, which means that a small number of branches are sufficient. These values depend on the first branch time and the human error probability.

ISSN:0029-5450    

DOI:10.13182/NT87-A33922

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

Chemical and radiochemical analyses of reactor coolant samples taken during defueling of the Three Mile Island Unit-2 (TMI-2) reactor provide relevant data to assist in understanding the solution chemistry of the radionuclides retained within the TMI-2 reactor coolant system. Hydrogen peroxide was added to various plant systems to provide disinfection for microbial contamination and has provided the opportunity to observe radionuclide release under different oxygen chemical potentials. A comparison of the radionuclide release rates with and without hydrogen peroxide has been made for these separate but related cases, i.e., the fuel transfer canal and connecting spent-fuel pool A with the TMI-2 reactor plenum in the fuel transfer canal, core debris grab sample laboratory experiments, and the reactor vessel fluid and associated core debris. Correlation and comparison of these data indicate a physical parameter dependence (surface-to-volume ratio) affecting all radionuclide release; however, selected radionuclides also demonstrate a chemical dependence release under the different oxygen chemical potentials.

ISSN:0029-5450    

DOI:10.13182/NT87-A33923

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

The optimization problem of the in-core fuel management of a thorium-fueled CANDU pressurized heavy water reactor (PHWR) consists of several component actions: the number of fresh fuel bundles inserted in the channels, the choice of the channels to be refueled next, the refueling rate, and the composition of the fresh fuel bundles (the latter relevant to advanced fuel cycles). Several fresh fuel compositions of232Th and233U were investigated and compared to the self-sufficient equilibrium thorium (SSET) cycle fuel, in terms of the objective function of an optimal fuel management problem. This optimization problem consisted of minimizing the total refueling rate at equilibrium with respect to criticality and power peaking constraints. The maximum acceptable value of the form factor was equal to 1.20, the form factor defined as the maximum-to-average power density ratio in the reactor core.The reactor core was divided into two refueling zones, each characterized by a uniform refueling rate for its channels. The control variables of the optimization problem were the average fluences (irradiations) of the bundles discharged from the channels of each of the zones, these variables being directly related to the refueling rates. A computer code, ASTERIX, was written to solve the optimization problem, using a steepest descent method, which required only a moderate number of diffusion calculations.Simulation was performed on simple models of a 600-MW CANDU PHWR. Because of the presence of233Pa in the fuel, the diffusion calculations are nonlinear, needing a more complex solution technique. The cell parameters used were calculated by the Atomic Energy of Canada Limited code LATREP for a two energy-group model. This optimization technique gave optimal results that represent substantial savings in the refueling rates (up to 14%) when compared to nonoptimal feasible cases. The comparison of the various fuel compositions studied revealed that the sensitivity of the refueling rate (and the burnup) to the fresh fuel content is quite large for the SSET fuel and the low enrichment fuels.

ISSN:0029-5450    

DOI:10.13182/NT87-A33924

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

A need for characterization of the iodine source term used in safety calculations for hypothesized light water reactor core disruptive accidents has motivated a study in iodine volatility. Previous experimental studies have been directed at evaluating volatility of iodine at a single time shortly (1 to 12 h) after introduction into the aqueous phase. The very important variables of time in solution and gamma radiation dose rate for a range of iodine concentrations (10−8to 10−5gI/ml) and pHs(5, 9, and 11) are explored. All experiments were performed at∼25°C, first in the absence of a significant radiation field and later with a gamma radiation dose rate ranging from 0.003 to 0.06 Mrad/h. Iodine was introduced as either molecular I2or Nal with131I(8.04-day half-life) as a tracer.Results of experiments with nonirradiated systems indicated very little volatility with Nal-initiated studies. The I2-initiated systems at pH 5 were the most volatile whereas experiments at pH 9 and 11 showed decreasing iodine volatility with time. From the experiments at pH 9, it is inferred that the partition coefficient of HOI is∼1000.A pronounced radiation-induced reduction in iodine volatility in pH 5 iodide solutions has been demonstrated as well as a dose rate dependence in the transient phase. As with nonirradiated systems, irradiated alkaline solutions exhibit low volatility.A computer-based model incorporating water radiolysis and iodine radiolytic chemical reactions has been formulated and tested. The model successfully predicts radiation-induced volatility changes in pH 5 iodide systems. The experimentally observed dose rate dependence is also verified. At pH 9, the agreement between experimental results and predicted results is not good.

ISSN:0029-5450    

DOI:10.13182/NT87-A33925

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor