I= STUDY OF REACTOR DESIGN PARAMETERS
Pakistan Research Repository Home
 

Title of Thesis
STUDY OF REACTOR DESIGN PARAMETERS

Author(s)
Liaquat Ali Khan
Institute/University/Department Details
Department of Physics/ University of the Punjab
Session
1999
Subject
Physics
Number of Pages
250
Keywords (Extracted from title, table of contents and abstract of thesis)
reactor design parameters, pakistan research reactor, low enriched uranium, leu, highly enriched( uranium, heu, onset of nucleate boiling, onb, onset of flow instability, ofi, departure from nucleate boiling, dnb

Abstract
Detailed reactor design calculations have been performed for a 10 MW swimming pool type Pakistan Research Reactor-l (P ARR-I) utilizing low enriched uranium (LEU) fuel and a 2' kW tank in pool type Pakistan Research Reactor-2 (P ARR-2) utilizing highly enriched( uranium (HEU) fuel. These calculations were aimed at: enhancing power level of P ARR-. from 9 to 10 MW, demonstrating inherent safety of P ARR-2, developing mathematical models for thermal hydraulic and accident analysis of P ARR-2 and their experimental validation, developing computer code for natural convection cooling of P ARR-l and it experimental validation, improving modeling procedures and validating methodology ant computer codes by experimental measurements.

Standard computer codes WIMS-D/4 and CITATION were employed to calculate con excess reactivity, reactivity loads due to temperatures, xenon and samarium, neutron fluxes power distribution and power peaking factors, reactivity feedback coefficients, control roc worth and shutdown margin, reactivity worth of fuel element and reflector. etc. The analytical predictions were validated by experimental measurements and. the agreement is found to b generally good.

The steady state thermal hydraulic analysis of P ARR-I and P ARR-2 was carried to compute coolant velocity, pressure drop, saturation temperature, temperature distribution in the core heat fluxes at Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB). Computer code DP and PARET were used for the analysis of forced convection cooling of PARR -1 core whereas a computer code FREECOM was developed for the analysis of natural convection cooling. Thermal hydraulic analysis 0 PARR-2 was carried out using standard correlations. The calculated core parameters, for the natural convection cooling, were validated by making experimental measurements on but the reactors and comparing analytical predictions with available experimental data for simile reactors. Good agreement was observed between theoretical and experimental values.

Computer code PARET was employed to investigate the transient response PARR-l core t several accident situations like uncontrolled withdrawal of control rods. flooding of bear tube, movement of core against thermal column and removal of an inpile experiment. The. reactivity limits imposed by clad melting temperature were also determined. In all of these transients, time histories of reactor power, energy release, peak fuel, clad and coolant temperatures were calculated. On the other hand, mathematical models were developed for the accident analysis of P ARR-2. The theoretical predictions for P ARR-2 were validated b) conducting reactivity insertion experiments. Reactivities of different magnitudes and durations were inserted into the core and transients were allowed to be terminated by intrinsic reactor behaviour without an external control or operators intervention. An external agreement was observed between analytical and experimental values.

The radiological consequence analysis of P ARR-I was also performed. Fission produce inventory of some important radioisotopes was calculated using standard computer code KORIGEN. The environmental impacts in case of design basis loss of coolant accident (LOCA) resulting in core melt down and release of fission product from reactor core to the containment building and eventually to atmosphere, after passing through various retention barriers, were analyzed. The atmospheric dispersion was modeled using a conservative, approach outlined in USNRC Regulatory Guide. The consequences were estimated in term of internal and external radiation doses to the workers and surrounding population. These dose estimates were then compared with the recommended dose limits given in 1 OCFRI 00 to determine the boundaries of exclusion and low population zones.

The isotopic composition and decay characteristics of an irradiated fuel element of PARR-were studied. Computer code KORIGEN was used to calculate the concentrations of uranium and plutonium isotopes, radioactivity, decay heat and spontaneous fission neutron source as function of operating cycle, fuel burnup and cooling time. The amount of fissile plutonium produced during irradiation and its contribution to total fissions occurring in the system was assessed. The contribution of gamma and alpha+beta to the total decay heat as a function ( cooling time was calculated. Similarly, the contribution of light elements, actinides an fission products to total activity and decay heat as a function of cooling time was also studied.

The fission product decay heat was calculated by KORIGEN and compared with the obtained from ANS 5.1 standard curves and other widely used semi-empirical correlations order to see how their predictions vary from each .other. The correlations used include Bors Wheeler formula or Way-Wigner type function, Casta-nedelik formula, SOTRAN code formula, Petterson-Schlitz Formula and EI-Wakil’s Formula.

Based on the preceding analyses, following conclusion are drawn:

The PARR-I can safely be upgraded to IO MW without compromising the safety of the reactor;

The P ARR-2 is an inherently safe reactor and can be licensed to operate without an operator on the console desk. It is therefore considered suitable for commissioning in densely populated areas;

The model developed for the steady state thermal hydraulic and accident analysis of P ARR-2 are quit adequate;

The theoretical predictions of FREECON are in good agreement with experimental measurements.

Good agreement between the anal:1ical and experimental values has validated the methodology and computer codes

Download Full Thesis
3244.94 KB
S. No. Chapter Title of the Chapters Page Size (KB)
1 0 Contents
161.13 KB
2 1 Introduction 1
58.97 KB
3 2 Description of Reactor 6
238.22 KB
  2.1 Introduction 6
  2.2 The IAEA Benchmark Reactor 6
  2.3 The Pakistan Research Reactor-1 9
  2.4 The Pakistan Research Reactor-2 15
4 3 Description of Computer Codes and Their Validation 22
152.25 KB
  3.1 Introduction 22
  3.2 Core Neutronics 23
  3.3 Thermal Hydraulics and Accident Analysis 25
  3.4 Isotopic Composition and Decay Characteristics 31
  3.5 Conclusions 32
5 4 Core Neutronics 33
515.32 KB
  4.1 Introduction 33
  4.2 Methodology 34
  4.3 Results and Discussion 42
  4.4 Conclusions 72
6 5 Thermal Hydraulics 74
486.08 KB
  5.1 Introduction 74
  5.2 Forced Convection Cooling 75
  5.3 Natural Convection Cooling 91
  5.4 Conclusions 109
7 6 Reactivity Induced Accidents 111
348.51 KB
  6.1 Introduction 111
  6.2 Transient Analysis of Parr-1 112
  6.3 Transient Analysis of Parr-2 122
  6.4 Conclusions 132
8 7 Radiological Consequence Analysis 135
325.67 KB
  7.1 Introduction 135
  7.2 Theory 136
  7.3 Behaviour of Radionuclides 138
  7.4 Multiple 142
  7.5 Engineered Safety Systems of Parr-1 143
  7.6 Radiological Consequence Analysis of PARR-1 144
  7.7 Result and Discussion 147
  7.8 Conclusions 153
9 8 Isotopic Composition and Decay Characteristics of Irradiated Fuel 154
244.74 KB
  8.1 Introduction 154
  8.2 Methodology 155
  8.3 Results and Discussion 157
  8.4 Conclusions 170
10 9 Decay Heat in Nuclear Reactors 172
107.34 KB
  9.1 Introduction 172
  9.2 Analytical Fitting Functions 174
  9.3 Point Depletion and Decay Calculations 177
  9.4 Results and Discussion 177
  9.5 Conclusions 179
11 10 Conclusions 181
56.35 KB
12 11 References 186
526.37 KB