The Breed and Burn (B&B) reactors come in handy to deal with the large stock piles of depleted uranium, using it as fuel to power the reactor. The discharge fuelΒ ...
LWR/TWR synergetic 2 tier fuel cycle by
Sai Raja Gopal VADLAMUDI in partial fulfilment of the requirements for the degree of Master of Nuclear Energy specialization in Nuclear Reactor Physics and Engineering at INSTN
Supervised by Dr. Massimiliano FRATONI UC Berkeley
Abstract
Abstract The Breed and Burn (B&B) reactors come in handy to deal with the large stock piles of depleted uranium, using it as fuel to power the reactor. The discharge fuel of B&B contains 10% of fissile plutonium. This study analyses the possible options in generating more energy while reducing the plutonium content by using a reconditioned B&B discharge to fuel Pressurized Water Reactor (PWR). Three reprocessing processes with high proliferation resistance are considered namely, AIROX, Melt Refining and FLURO-OX process. Using AIROX processed fuel in a PWR core, no additional burnup can be achieved without compromising on the power density, and whereas when the PWR is fuelled with Melt refined fuel an additional burnup of 66 GWd/MTIHM can be achieved using the nominal power density. Melt refined fuel mixed with different types of fuels provides various options to incinerate fissile plutonium. Using smaller diameter pins an additional burnup of 95.2 GWd/MTIHM can be achieved. Using pure FLURO-OX processed fuel in a PWR core is not practical due to the positive reactivity coefficients. However with a slight modification in the FLURO-OX process uranium can be retained in the fuel, which can compensate for the positive reactivity coefficients. Fuelling PWR with such fuel an additional burnup of about 108.1 GWd/MTIHM can be achieved at the nominal power density. Instead of once through cycle, using twice burning cycle in PWR an additional burnup of about 120 GWd/MTIHM can be achieved. .
i
Contents
Contents Abstract ...................................................................................................................................... i Contents ....................................................................................................................................ii List of Tables ........................................................................................................................... iv List of Figures ........................................................................................................................... v List of Abbreviations .............................................................................................................. vi Acknowledgment ....................................................................................................................vii 1
Introduction ...................................................................................................................... 1 1.1 Motivation and Background .......................................................................................... 1 1.2 Travelling wave reactor ................................................................................................. 2 1.3 Spent fuel composition of B&B reactor......................................................................... 3 1.4 Methodology to calculate density, burnup and cycle length.......................................... 4 1.5 PWR core representation ............................................................................................... 5 1.6 Simulation Tools ............................................................................................................ 6 1.7 Outline............................................................................................................................ 6
2
AIROX processed fuel ..................................................................................................... 8 2.1 AIROX process .............................................................................................................. 8 2.2 Objective ........................................................................................................................ 9 2.3 Density ........................................................................................................................... 9 2.4 AIROX reference calculation ........................................................................................ 9 2.5 Zirconium dioxide fuels ............................................................................................... 11
3
Melt refined fuel ............................................................................................................. 12 3.1 Objective ...................................................................................................................... 12 3.2 Density ......................................................................................................................... 12 3.3 Melt refining reference calculation .............................................................................. 13 3.4 Zirconium Oxide fuels ................................................................................................. 15 3.5 More options to increase moderation ........................................................................... 15
4
FLURO-OX processed fuel ........................................................................................... 20 4.1 Objective ...................................................................................................................... 21 ii
Contents 4.2 Density ......................................................................................................................... 21 4.3 Problem with FLURO-OX processed fuel .................................................................. 21 4.4 FLURO-OX fuel mixed with Natural uranium Oxide ................................................ 21 4.5 Modified Fluro-Ox process (MFO) ............................................................................. 23 5
Assembly model .............................................................................................................. 25 5.1 Geometry...................................................................................................................... 25 5.2 π€β evolution................................................................................................................ 26 5.3 Burnup &Cycle length ................................................................................................. 27 5.4 More assembly models ................................................................................................ 27
6
Reactivity coefficients .................................................................................................... 30 6.1 Excess reactivity and critical boron concentration ...................................................... 30 6.2 Methodology ................................................................................................................ 32 6.3 Void coefficients .......................................................................................................... 32
7
Discharge fuel composition ........................................................................................... 34
8
Multiple recycling .......................................................................................................... 36
9
Conclusion ...................................................................................................................... 38
10
Bibliography ................................................................................................................... 40
iii
List of Tables
List of Tables 1.1 1.2 1.3 2.1 2.2 2.3 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 4.1 4.2 4.3 4.4 4.5 5.1 5.2 6.1 6.2 7.1 8.1
Composition of the spent fuel of TWR at 1000 days after discharge. ........................... 3 Oxide densities ............................................................................................................... 4 PWR unit cell dimensions .............................................................................................. 6 Removal fractions in AIROX process ........................................................................... 8 Density of AIROX fuel and UO2 fuel ............................................................................ 9 Discharge Burnup and Cycle length of PWR cores fueled with AIROX processed fuel and with UO2 fuel. ............................................................................................... 10 Removal fractions in MR process ................................................................................ 12 Density of the MR processed fuel and UO2 fuel.......................................................... 12 Discharge burnup and cycle length of PWR cores fueled with MR processed B&B fuel and with UO2 fuel ................................................................................................. 14 Discharge burnup and cycle length of PWR cores fueled with MR processed B&B fuel mixed with ZrO2 in different mass fractions. ....................................................... 15 Discharge burnup and cycle length of PWR cores fueled with MR processed B&B fuel for different pin diameters .................................................................................... 17 Parameters of the South Texas Project ........................................................................ 17 Thermal hydraulic calculations for the small pin and the standard pin ....................... 18 Discharge burnup and cycle length of PWR cores fueled with annular fuel pins fueled with MR processed B&B fuel. .......................................................................... 19 Removal fractions in the FO process ........................................................................... 20 Density of FO processed fuel along with UO2 fuel ..................................................... 21 Discharge burnup and cycle length for PWR cores fueled with FO processed fuel mixed with natural UO2 ............................................................................................... 22 Density of the MFO processed and FO processed fuel................................................ 24 Discharge burnup and cycle length for PWR core fueled with Modified Fluro-Ox process.......................................................................................................................... 24 Discharge burnup and cycle length for assembly and pin cell model representation of PWR core fueled with MR processed fuel .............................................................. 27 Discharge burnup and cycle length for different assembly designs with different fuel compositions ......................................................................................................... 29 Critical boron concentration for different cases at BOC ............................................. 30 Reactivity coefficient values at BOC for PWR cores fueled with MR and FO fuels. . 32 Characteristics of discharge fuel .................................................................................. 34 Achievable average discharge burnup and discharge plutonium values for the 2 tier system with twice burning cycle in PWR .......................................................... 37
iv
List of Figures
List of Figures 1.1 1.2 1.3 2.1 2.2 2.3 3.1 3.2 3.3 3.4 4.1 4.2 4.3 4.4 5.1 5.2 5.3 5.4 5.5 6.1 6.2 6.3 8.1
PWR/TWR synergetic 2 tier fuel cycle ......................................................................... 2 Fuel pin and core description of TWR .......................................................................... 2 Pin cell ........................................................................................................................... 6 Unit operations in AIROX process ................................................................................ 8 Evolution of πβ with burnup for the PWR fueled with AIROX fuel ......................... 10 Evolution of πβ with burnup for PWR cores fueled with AIROX fuel mixed zirconium dioxide ........................................................................................................ 11 Evolution of πβ with burnup for PWR core fueled with MR processed fuel for different M/F ratios in comparison with PWR core fueled with UO2 fuel .................. 13 Evolution of πβ with burnup for PWR cores fueled with MR processed fuel mixed with zirconium dioxide ..................................................................................... 14 Evolution of πβ with burnup for PWR cores with small and standard fuel pins fueled with MR fuel ..................................................................................................... 16 Evolution of πβ with burnup for annular pins fueled with MR fuel with ZO2 and SiC, and reference case. ........................................................................................ 18 Flow sheet of the βFluor-Oxβ Process Applicable to the Fuel Processing of the ABWR with a MOX Core. [11] ................................................................................... 20 Evolution of πβ with burnup for PWR cores fueled with FO processed fuel mixed with NatUO2...................................................................................................... 22 Modified Fluor-Ox process .......................................................................................... 23 Evolution of πβwith burnup for PWR core fuelled with MFO processed fuel .......... 24 Average discharge burnup evolution with M/F ratio for PWR fuelled with MR fuel ............................................................................................................................... 25 PWR assembly model .................................................................................................. 26 Evolution of πβ with burnup for assembly and pin cell both fuelled with MR processed fuel............................................................................................................... 26 (a) & (b) shows assembly designs with 4 and 8 extra water rods respectively ........... 28 Evolution of πβ with burnup for different assembly designs with different fuels ..... 28 Normalized neutron flux per unit lethargy at BOC for PWR cores fueled with MR, MFO and UO2 fuels ............................................................................................. 31 Excess reactivity vs cycle length for PWR cores fueled with MFO, MR and UO2 .............................................................................................................................. 31 Evolution of πβ with void percentage at BOC for PWR cores fueled with MR and FO fuels ................................................................................................................. 33 2-tier system with twice burning cycle in PWR .......................................................... 36
v
List of Abbreviations
List of Abbreviations
B&B
Advanced Boiling Water Reactor Atomics International Reduction Oxidation process Breed and Burn reactor
BOC
Beginning of Cycle
BOL
Beginning of Life
CANDLE
EOC
Constant Axial shape of Neutron flux, nuclides densities and power shape During Life of Energy production reactor CANada Deuterium Uranium reactor Experimental Breeder Reactor End of Cycle
EOL
End of Life
FHM
Final Heavy Metal
FIMA FO
Fissions per Initial Metal Atom Fluro-Ox process
IHM
Initial Heavy Metal
MDNBR
Minimum Departure from Nucleate Boiling Ratio Modified Fluro-Ox Process
ABWR AIROX
CANDU EBR
MFO
MOX
Moderator to Fuel volume ratio Mixed Oxide fuel
MR
Melt Refined fuel
PWR
Pressurized Water Reactor
TWR
Travelling Wave Reactor
M/F
vi
Acknowledgment
Acknowledgment I would like to thank my supervisor Dr. Massimiliano Fratoni and Prof. Ehud Greenspan for their guidance and valuable suggestions during the planning and development of this research work. I would like to thank Christian Di Sanzo for providing information related to B&B discharge fuel and for helping me with my research work. I would like to thank Mr. Yasuo Hirose for providing the information regarding the FLURO-OX process.
vii
Introduction
1 Introduction 1.1 Motivation and Background
In recent years the nuclear industry became interested in Breed and Burn reactors as uranium utilization can be increased significantly compared to the current operating nuclear reactors without need for fuel recycling. In order to produce 1 ton of 4.5% enriched uranium, it takes about 8 to 10 tons of natural uranium, the rest is discarded as waste known as depleted uranium containing about 0.2% to 0.3% of U-235. Using fast reactors it is possible to fission close to 100% of the depleted uranium, but the neutron induced damage effects constrain the burnup of the fuel to about 10% to 15% FIMA (fissions per initial metal atom) depending on the neutron spectrum. Consequently, the fuel has to be recycled multiple times in order to increase uranium utilization. Although it is technically feasible, due to economic viability and proliferation concerns there is significant objection in U.S and many other countries. Using special class of fast reactors known as Breed and Burn (B&B) reactors it is possible to increase the uranium utilization without any need for reprocessing. The Breed and Burn reactor refers to breeding plutonium in the depleted uranium and burning the bred plutonium. The first proposal of a fast reactor that could operate on breed and burn condition was suggested in 1958 by Feinberg [1]. In B&B reactors to initiate the chain reaction at first the reactor is fed with some enriched uranium or the TRU extracted from PWR, which is known as starter fuel and later on the reactor could operate continuously by only being fed by depleted uranium. The minimum required average discharge burnup using metallic fuel and low leakage core is 1820% FIMA (Fissions per Initial Metal Atom) [2]. The spent fuel of the B&B reactor contains relatively high amount of fissile plutonium about 10% of the heavy metal. The main objective of this study is to increase the utilization of the depleted uranium while maintaining high proliferation resistance and reduce the total amount of plutonium (TRU) discharged per unit of electricity generated as well as reduce the fissile-to-total plutonium ratio by reusing the fuel in Pressurized water reactor (PWR) after limited reprocessing. This dual tier system of TWR (Travelling Wave Reactor, see section 1.2)-PWR is shown in the Figure 1.1. Reprocessing the B&B discharge fuel using conventional reprocessing techniques is not acceptable in US and many other countries due to proliferation concerns, so there is need to look out for the processes which have high proliferation resistance. Based upon this criterion three reprocessing processes are considered in this study, AIROX, Melt Refining and FLURO-OX process. The preliminary feasibility analysis of fuelling PWRs with discharge B&B fuel was done by Christian [3] and it was concluded that in order to maximize the reduction of the fissile plutonium content, higher Moderator to Fuel volume ratio (M/F) is required. However reaching the higher M/F ratio of about 5 is not practically possible, as it reduces the power 1
Introduction density significantly. The main objective of this study is to find a practical solution to reduce the plutonium content and the fissile plutonium fraction without having penalty on the power density.
Figure 1.1 PWR/TWR synergetic 2 tier fuel cycle
1.2 Travelling wave reactor In 2006 TerraPower was established in order to commercialize the Breed and Burn reactors. TerraPower called their reactor as Traveling wave reactor (TWR) which is similar to that of CANDLE1 reactor. In 2008 the design was changed to standing wave reactor with cylindrical core geometry [4]. The design of the TerraPowerβs core is shown in the Figure 1.2. The discharge fuel of TWR is considered for this study.
Figure 1.2 Fuel pin and core description of TWR [4] 1
CANDLE stands for Constant Axial shape of Neutron flux, nuclides densities and power shape During Life of Energy production; it is a type of B&B reactor.
2
Introduction
1.3 Spent fuel composition of B&B reactor The discharge fuel of B&B reactor contains relatively high amount of plutonium about 12.28% of the overall heavy metal, out of which 81.8% is fissile plutonium. The major contribution to heavy metal mass in the discharge fuel comes from uranium and plutonium. The initial fuel is alloyed with 5% zirconium, no transmutation is considered for this value from the BOL2 to EOL3. Table 1 summarizes the list of all elements of the discharge fuel of B&B reactor [5]. Table 1.1 Composition of the spent fuel of TWR at 1000 days after discharge.
Fission Product Concentration
2 3
Mass of Actinide and Daughter nuclides
Fission Product by Element
Mass (g) at EOL + 1000 days
Fission Product by Element
Mass (g) at EOL + 1000 days
Fission Product by Element
Mass (g) at EOL + 1000 days
Ge As Se Br Kr Rb Sr Y
5.31E-02 1.70E-02 3.02E+00 1.17E+00 1.56E+01 1.52E+01 2.81E+01 1.85E+01
Te I Xe Cs Ba La Ce Pr
4.48E+01 2.43E+01 3.18E+02 2.56E+02 1.19E+02 8.20E+01 1.50E+02 7.64E+01
Ac Am At Bi Cm Fr He Np
7.40E-12 1.78E+01 2.06E-22 8.68E-11 4.43E-02 2.06E-18 3.96E-02 9.70E+00
Zr Nb Mo Tc Ru Rh Pd Ag Cd In Sn Sb
1.95E+02 3.02E-04 2.25E+02 5.90E+01 2.04E+02 6.80E+01 1.57E+02 1.70E+01 9.37E+00 7.50E-01 8.61E+00 1.75E+00
Nd Pm Sm Eu Gd Tb Dy Ho Er Tm Yb
2.51E+02 4.18E-01 7.64E+01 7.92E+00 8.92E+00 5.33E-01 3.63E-01 1.41E-02 4.22E-03 0.00E+00 0.00E+00
Pa Pb Po Pu Ra Rn Sf Th Tl U
3.86E-07 4.82E-06 3.97E-14 1.52E+03 9.40E-09 1.38E-12 8.83E-08 3.14E-05 1.54E-12 1.08E+04
Total
2.44E+03
Total
1.23E+04
BOL-Beginning of Life EOL-End of Life
3
Mass of Cladding Elements and initial fuel alloyed with Zr (5%) Mass (g) Fission at EOL* Product by Element C Cr Fe Mn Si V W Zr
0.388231 24.58784 173.2427 3.57E-03 0.431365 0.485287 1.08E+00 7.50E+02
*These values assume no transmutation (mass at End of Life (EOL) = mass at Beginning of Life (BOL)).
Total
9.50E+02
Introduction
1.4 Methodology to calculate density, burnup and cycle length It should be noted that as discharge burnup increases, the Pu discharged per unit electricity generated decreases. In order to determine the maximum achievable average discharge burnup the cycle length has to be calculated first. Moreover, for all the simulations the density of the fuel is required. This section is devoted to the methodology to calculate the density, burnup and cycle length. 1.4.1 Density In order to determine the density of the reprocessed fuel the following equation is used [3], 1 π= ( 1.1 ) π€ βππ=1 π ππ π€π indicates the weight fraction of the ith oxide, ππ indicates the density of ith oxide. It should be noted that only one oxide form is considered for each element and some elements do not form oxides, in that case pure element density is used. Table 1.2 summarizes the oxides considered for different elements and their respective densities. Table 1.2 Oxide densities [3]
Actinide and daughters oxides
Fission product oxides oxide
g/cc
Oxide
g/cc
oxide
g/cc
GeO2
4.228
TeO2
5.67
Ac2O3
-
As2O3
3.74
I2O5
4.98
AmO2
11.68
SeO2
3.954
XeO3
4.55
At
-
Br
-
Cs2O
4.65
Bi2O3
8.9
Kr
-
BaO2
5.68
Cm
-
Rb2O
4
La2O
6.51
Fr
-
SrO
4.7
Ce2O
6.2
He
-
Y2O3
5.01
Pr2O
6.9
NpO2
11.1
ZrO2
5.68
Nd2O
7.24
Pa
-
NbO
7.3
Pm2O
6.85
PbO
9.53
MoO2
6.47
Sm2O
8.347
PoO2
8.9
TcO2
6.9
Eu2O
7.4
PuO2
11.5
RuO2
6.97
Gd2O
7.407
Ra
-
Rh2O3
8.2
Tb4O
7.3
Rn
-
PdO
8.3
Dy2O
7.8
ThO2
10
Ag2O
7.14
Ho2O
8.41
Tl2O
10.45
CdO
6.95
Er2O
8.64
UO2
10.97
In2O3
7.179
Tm2O
8.6
SnO
6.45
Yb2O
9.17
Sb2O3
5.67
4
Introduction 1.4.2 Burnup and Cycle length The maximum achievable average burnup and cycle length values are required to find out the best reprocessing process to be applied to discharge B&B fuel. The maximum achievable average discharge burnup can be estimated using the methodology described in the paper by Christian [3]. For the convenience the methodology is quoted again, The average discharge burnup can be estimated from the unit cell πβ evolution with burnup assuming n fuel batches and a 2.5% neutron leakage probability. The core average reactivity can be estimated from equation ( 1.2 )
πππππ =
where, πβ,π =
πβ β1 πβ
βππ=1 ππ πβ,π β 0.025 π
( 1.2 )
is the reactivity calculated from unit cell of batch π,ππ is the fraction of
total power generated by batch π, for this analysis ππ is assumed to be 1βπ. Reactivity at the beginning of cycle (BOC) and reactivity at the end of cycle (EOC) for a 5 batch core with fuel cycle length ECL are estimated from equation ( 1.3 ) and ( 1.4 ) respectively,
π0 + π1βπΉπΆπΏ + π2βπΉπΆπΏ + π3βπΉπΆπΏ + π4βπΉπΆπΏ β 0.025 5
( 1.3 )
π1βπΉπΆπΏ + π2βπΉπΆπΏ + π3βπΉπΆπΏ + π4βπΉπΆπΏ + π5βπΉπΆπΏ β 0.025 5
( 1.4 )
ππ΅ππΆ =
ππΈππΆ =
where, π0 is the initial reactivity of the fresh fuel, π1βπΉπΆπΏ is the reactivity of the fuel which has resided in the core for one cycle, π2βπΉπΆπΏ is the reactivity of the fuel which has resided in the core for two cycles and so on. Both ππ΅ππΆ and ππΈππΆ values should be greater than zero, in order to have practical core with certain burnup and cycle length. The fuel cycle length is determined by searching iteratively and then the maximum achievable average burnup value can be estimated from that cycle length.
1.5 PWR core representation PWR core is represented by unit pin cell as shown in the Figure 1.3 with reflective boundary conditions applied to it. The outer fuel rod diameter is 9.5 mm, the zircaloy clad thickness is 0.571 mm and the gap thick is 0.082 mm [6]. The linear power generation rate is considered to be 175.2 W/cm [7]. The density of the water used is 0.7 g/cm3 5
Introduction
Figure 1.3 Pin cell
Table 1.3 PWR unit cell dimensions
outer rod diameter, D (mm) clad thickness (mm) gap thickness (mm) lattice pitch, P (square) (mm) P/D Moderator/Fuel volume ratio (M/F)
9.5 0.57146 0.08243 12.6 1.326 1.67
1.6 Simulation Tools For all neutronic and depletion analysis SERPENT is used. SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code using ENDF VII library for the cross sections. Although the methodology to predict the burnup and cycle length is pretty straight forward, MATLAB is used for the programming purpose, in order to find out the iteratively the cycle length.
1.7 Outline This study is organised as follows: Chapter 1: Introduction Chapter 2: The feasibility of using AIROX process in processing the discharge B&B fuel is analysed.
6
Introduction Chapter 3: The feasibility of using Melt Refining process in processing the discharge B&B fuel is analysed. Chapter 4: The feasibility of using FLURO-OX process in processing the discharge B&B fuel is analysed. Chapter 5: This chapter describes the benefit of using an assembly model instead of pin cell to represent PWR core and different assembly designs are analysed. Chapter 6: This chapter is devoted for the calculation of reactivity coefficients for different options studied in Chapter 5. Chapter 7: This chapter assess the characteristics of the spent fuel produced in B&B reactors in comparison with the different options studied in Chapter 5. Chapter 8: Instead of using once through cycle in PWR, twice burning cycle is suggested in this chapter.
7
AIROX processed fuel
2 AIROX processed fuel 2.1 AIROX process Atomics International Reduction Oxidation process known as AIROX is a dry processing technique where the volatiles and cladding are separated from the spent fuel. It was initially developed to reuse the spent fuel of LWR in CANDU. This process was developed only for oxide fuels. In this work, it is assumed that the fuel is first converted to oxide and then undergoes AIROX process. This process removes only few fission products using reductionoxidation process. At the end of the process, 100% Kr, Xe, T, I, C and Ar, 90% of Cs, Ru and 75% of Te, Cd are removed in gaseous form. Operations involved in the AIROX process are summarized in the Figure 2.1 [8]. The removal fractions in AIROX process are tabulated in Table 2.1.
Figure 2.1 Unit operations in AIROX process
Table 2.1 Removal fractions in AIROX process
Elements
Removal fractions in AIROX process 0% 0% 0% 100% H, He, N, O, F, Ne, Cl, Ar, Kr, Xe, Rn, I, T, C 90% Cs, Ru 75% Te, Cd
Th, Am U Other HM FPs
8
AIROX processed fuel
2.2 Objective
The objective of this chapter is to study the feasibility of AIROX process in processing the discharge B&B fuel, as the AIROX processed fuel B&B fuel still contains many fission products the macroscopic absorption cross section is much higher than the 4-5% enriched UO2 fuel, so higher M/F (Moderator to Fuel volume) ratio is required to achieve appreciable burnup. The M/F ratio was increased in the previous study [3] by increasing the unit cell pitch, however increasing the cell pitch decreases the power density. In this chapter other options like inert matrix fuels are explored in order to utilize the AIROX processed fuel without compromising on power density.
2.3 Density The density of the fuel is calculated using the same methodology as in the section 1.4.1. The density of the AIROX processed B&B fuel is less than UO2 fuel, even the heavy metal density is less than the uranium oxide fuel due to the presence of fission products. Table 2.2 Density of AIROX fuel and UO2 fuel
Type of fuel
HM mass fraction (%)
fuel density (g/cm3)
HM density (g/cm3)
UO2
88.1
10.4
9.16
AIROX processed B&B fuel
72.6
9.70
7.04
2.4 AIROX reference calculation In this section, the two PWR cores with different M/F ratios 1.67 and 5, fuelled with AIROX processed B&B discharge fuel are analysed in comparison with a PWR pore fuelled with 5% enriched UO2. 2.4.1 πβ evolution Figure 2.2 shows the πβ (infinite multiplication factor) evolution with burnup for the PWR core fuelled with AIROX processed B&B discharge fuel and 5% enriched UO2 fuel and for a modified PWR core with M/F ratio of 5, fuelled with AIROX process fuel. The πβ value is never greater than 1 from BOL to EOL for the case with AIROX processed B&B fuel in a PWR core, whereas it exceeds 1 when the M/F ratio is increased to 5, which is consistent with the initial hypothesis. The evolution of πβ with burnup is parabolic in shape; the large increase in reactivity at the beginning is due to the fission products. Even for the case with 9
AIROX processed fuel M/F ratio 5 the initial πβ value is not greater than 1, which implies to establish initial criticality some fissile material has to be added to compensate for the neutron capture in the fission products.
1.5
AIROX (M/F~1.67) AIROX (M/F~5) 5% enriched UO2 (M/F~1.67)
1.4 1.3 1.2
kο₯
1.1 1.0 0.9 0.8 0.7 0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 2.2 Evolution of πβ with burnup for the PWR fueled with AIROX fuel
2.4.2 Burnup and cycle length Using the methodology described in the section 1.4.2, the maximum attainable average burnup and cycle length values for the modified PWR core (M/F~5) fuelled with AIROX processed B&B discharge fuel are calculated for 5 batch core. The resulting values are tabulated in the Table 2.3. When the M/F ratio is 5, the AIROX processed fuel can achieve similar burnup values as the 5% enriched UO2 fuel in standard PWR reference core with M/F ratio 1.67, but with shorter cycle length due to lower heavy metal density. Table 2.3 Discharge Burnup and Cycle length of PWR cores fueled with AIROX processed fuel and with UO2 fuel
Type of fuel
M/F
Average discharge burnup (GWd/MTIHM)
Cycle length (days)
5% enriched UO2
1.67
63.6
350.7
5
64.6
275.1
AIROX processed B&B fuel
10
AIROX processed fuel
2.5
Zirconium dioxide fuels
As increasing the pitch to increase the moderator has a huge penalty on the power density, the idea of using inert matrix fuels is investigated in this section. In this case zirconium dioxide is mixed with AIROX processed fuel. The zirconium dioxide is mixed with AIROX processed fuel in different mass fractions, from 0% to 20% of the total mass of the fuel.
2.5.1 πβ evolution Evolution of πβ with burnup for PWR cores fueled with mixed fuels of zirconium dioxide and AIROX in different mass fractions are plotted in the Figure 2.3, it can be observed that the πβ increases with increase in the amount of zirconium oxide in the fuel but πβ is never greater than 1, reactor can never be critical. Increasing further the amount of zirconium dioxide in the fuel would increase the πβ value; however the amount of heavy metal in the fuel pin decreases at the same time, which results in increase specific power of the fuel if the overall power is kept constant. With the increase of the specific power the length of the cycle becomes shorter. Hereby it can be concluded that without compromising on the power density, the AIROX process is not suitable for the dual tier system of TWR/PWR.
1.00
20% ZrO2 + 80% AIROX 10% ZrO2 + 90% AIROX
0.95
100% AIROX
kο₯
0.90
0.85
0.80
0.75
0.70 0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 2.3 Evolution of πβ with burnup for PWR cores fueled with AIROX fuel mixed zirconium dioxide
11
Melt refined fuel
3 Melt refined fuel The Melt refining process was developed in the EBR-II project. In melt refining process the cladding is removed and fuel is loaded onto zirconia crucibles and the mixture is melted at 1300β. In this process the volatile and gaseous fission products are removed and some solid fission products are removed by oxidation with zirconia of the crucible. For this study meltrefining process is assumed to remove 100% of Br, Kr, Rb, Cd, I, Xe and Cs; 95% of Sr, Y, Te, Ba and lanthanides and 95% of Th and Am. The removal fractions in melt refining process are tabulated in the Table 3.1. It is assumed that after reprocessing the fuel is converted into oxide form. For convenience the melt refining process is further referred as MR process. Table 3.1 Removal fractions in MR process
Element Th, Am U Other HM FPs
Removal fraction 95% 0% 0% 100% H, He, N, O, F, Ne, Cl, Ar, Kr, Xe, Rn, Br, Rb, Cd, I, Cs 95% Sr, Y, Te, Ba, La-Lu
3.1 Objective The main objective of this chapter is to study the feasibility of MR process. To reach higher burnup values using MR processed fuel in the PWR core, higher M/F ratio is required [3]. Instead of increasing the cell pitch other ideas have to be investigated, to make sure that the power density is not compromised.
3.2 Density The density of the MR processed B&B discharge fuel is calculated using the same methodology as in the section 1.4.1. The calculated values of density, heavy metal density and heavy metal mass fraction of MR processed fuel in comparison with values of UO2 fuel are tabulated in the Table 3.2. The density of the MR processed fuel is less compared to UO2 fuel, due to presence of fission products. Table 3.2 Density of the MR processed fuel and UO2 fuel
Type of fuel UO2 MR processed B&B discharge fuel
HM mass fraction (%) 88.1 75.4
Fuel density (g/cm3) 10.40 9.95
12
HM density (g/cm3) 9.16 7.50
Melt refined fuel
3.3 Melt refining reference calculation In this section, the two PWR cores with different M/F ratios 1.67 and 5, fuelled with MR processed B&B discharge fuel are analysed in comparison with a PWR core fuelled with 5% enriched UO2
3.3.1 πβ evolution Figure 3.1 Shows the πβ evolution with burnup for the PWR reference core fuelled with MR processed B&B discharge fuel. Moreover the Figure 3.1.shows the πβ evolution with burnup for the modified PWR core with M/F ratio of 5 fuelled with MR processed fuel. The πβ value is greater than 1 at the BOL even for the case with MR processed B&B fuel in standard PWR reference core which is not the same with the AIROX processed fuel as mentioned in the section 2.4.1, which proves in neutronic point of view MR processed fuel is better than the AIROX processed fuel and the πβ value is much higher than 1 when the M/F ratio is increased to 5, which is consistent with the initial hypothesis.
1.5
5% enriched UO2 (M/F~1.67)
1.4
MR (M/F~5) MR (M/F~1.67)
1.3 1.2
kο₯
1.1 1.0 0.9 0.8 0.7 0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 3.1 Evolution of πβ with burnup for PWR core fueled with MR processed fuel for different M/F ratios in comparison with PWR core fueled with UO2 fuel
13
Melt refined fuel 3.3.2 Burnup and cycle length
In order to see the clear impact of the M/F ratio, maximum attainable average burnup and cycle length values are calculated, using the same methodology as in section 1.4.2. The burnup and the cycle length values are calculated for the PWR core fuelled with MR processed for both M/F ratios 1.67 and 5, using batch fraction 1/5 and the values are tabulated in the Table 3.3 in comparison with the PWR fuelled with 5% enriched UO2 fuel. The maximum achievable average discharge burnup when MR processed fuel is used in the PWR core is about 49.3 GWd/MTIHM, which is less than the burnup value that is attainable using 5% enriched UO2 fuel. However when the M/F ratio is increased to 5, the achievable burnup value increases to 104.5 GWd/MTIHM. Table 3.3 Discharge burnup and cycle length of PWR cores fueled with MR processed B&B fuel and with UO2 fuel
Type of fuel
M/F ratio
MR processed B&B fuel MR processed B&B fuel 5% enriched UO2
1.67 5 1.67
Average discharge Burnup (GWd/MTIHM) 49.3 104.5 63.6
Cycle length (days) 224.1 475.0 350.7
1.10
100% MR 10% ZrO2 + 90% MR
1.05
20% ZrO2 + 80% MR
kο₯
1.00
0.95
0.90
0.85
0.80 0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 3.2 Evolution of πβ with burnup for PWR cores fueled with MR processed fuel mixed with zirconium dioxide
14
Melt refined fuel
3.4 Zirconium Oxide fuels Even though the burnup value achievable in PWR core using MR processed fuel is not too low, to gain additional burnup the idea of inert matrix fuel is investigated in this section. Instead of increasing the pitch size of the reference unit cell to increase moderation, zirconium dioxide is introduced in the fuel. The zirconium dioxide is introduced at different mass fractions in the fuel from 0% to 20% of the total mass of the fuel. Figure 3.2 Evolution of πβ with burnup for PWR cores fueled with MR processed fuel mixed with zirconium dioxide. It can be observed from the plot that there is an increase in the moderation with the introduction of zirconium dioxide. But with the increase of zirconium dioxide the amount of fuel in the pin reduces resulting in higher specific power as the total power is constant, which indeed results in shorter cycle lengths.
3.4.1 Burnup and Cycle length
In order to prove the hypothesis that shorter lengths will result from mixing zirconium dioxide with MR fuel, the burnup and the cycle length values are determined for the two different PWR cores fuelled with 90% MR fuel mixed with 10% ZrO2 and 80% MR fuel mixed with 20% ZrO2respectively, using the same methodology as in the section 1.4.2 considering batch fraction 1/5. Table 3.4 shows the burnup and cycle length values, it can be observed that with the increase in the zirconium dioxide the maximum attainable average burnup increases from 49.3 to 58 GWd/MTIHM for the PWR core with 80% MR mixed with 20% ZrO2, but the cycle length decreased which is consistent with the initial hypothesis. Overall this idea doesnβt look really interesting due to the shorter cycle lengths. Table 3.4 Discharge burnup and cycle length of PWR cores fueled with MR processed B&B fuel mixed with ZrO2 in different mass fractions.
Type of fuel 20% ZrO2 + 80% MR 10% ZrO2 + 90% MR 100% MR
Average Burnup (GWd/MTIHM) 58 52.6 49.3
Cycle length (days) 189.1 206.0 224.1
3.5 More options to increase moderation As using inert matrix fuel did not give encouraging result, further ideas are investigated in order to increase the moderation without increasing the cell pitch. Two other options that are investigated in this section in order to increase the moderation are, 1) Reduction of Pin diameter 2) Annular pin with moderating material in the annulus. 15
Melt refined fuel 3.5.1 Reduction in pin diameter
It was found in the work done by Christian [3] when M/F ratio is increased to 2.8 the maximum attainable average burnup is about 94.7 GWD/MTIHM. So in order to maintain the moderator to fuel volume ratio 2.8 the radius of the fuel should be decreased from 4.11 mm to 3.46 mm. If the fuel pin diameter is reduced keeping the total power the same, the linear power remains the same, according to the equation ( 3.1 ) that would result in the increase of heat flux and the specific power increases as well and results in shorter cycle lengths. π β² = 2ππ β π β²β² ππ πππ =
( 3.1 )
πβ² π β ππ 2
( 3.2 )
where π β² is linear power, π β²β² is heat flux and ππ πππ is specific power unit mass of fuel.
3.5.1.1 π β evolution Figure 3.3 shows the evolution of πβ with burnup for the PWR cores with smaller diameter pins fuelled with MR processed fuel in comparison with the PWR core with standard pins fuelled with MR processed fuel, significant difference can be observed between the two cases. The value of πβ at BOL increases from 1.08 to 1.16 for the PWR core with smaller diameter pins. This observation is consistent with the initial hypothesis.
1.20
MR standard pin MR small pin
1.15 1.10
kο₯
1.05 1.00 0.95 0.90 0.85 0.80 0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 3.3 Evolution of πβ with burnup for PWR cores with small and standard fuel pins fueled with MR fuel
16
Melt refined fuel 3.5.1.2 Burnup and Cycle length In order to see the clearly the impact of the using smaller pin, the maximum attainable average burnup and cycle length values are calculated using the same methodology as in the section 1.4.2 and are tabulated in the Table 3.5. Even though the specific power has increased due to the decrease in the diameter, the cycle length did not decrease due to the large increase in the burnup value from 49.3 to 87.2 GWd/MTIHM. This result shows a clear neutronic advantage in using smaller pin. All the calculations are done using batch fraction 1/5
Table 3.5 Discharge burnup and cycle length of PWR cores fueled with MR processed B&B fuel for different pin diameters
Radius of the fuel (cm) 0.3463(M/F-2.8) 0.4096(M/F-1.67)
Average discharge Burnup (GWd/MTIHM) 87.2 49.3
Cycle length (days) 291.8 224.1
3.5.1.3 Thermal Hydraulic Problems Even though in neutronics point of view it is feasible it could face some thermal hydraulic problems due to the increase in the heat flux, increase in the hydraulic diameter which could impact the DNBR ratio, so further thermal hydraulic analysis is performed in the section 3.5.1.4. 3.5.1.4 Thermal Hydraulic Analysis In order to calculate the MDNBR, minimum departure from nucleate boiling ratio W3 correlation [9] is used. The dimensions and properties used for the reference fuel assembly are taken from the South Texas Project [7].Table 4.2 shows the relevant parameters used in the calculation. Table 3.6 Parameters of the South Texas Project
Parameter
Value
Pressure (MPa) Number of fuel rods
15.5 50952
Coolant mass flow rate (kg/s) Coolant inlet temperature (β) Active length of the fuel rod (m)
18627 294 4.26
Linear heat rate (kW/m) Width of the fuel assembly (m)
17.52 0.216
Radial peak to average power
1.65
Table 3.7 shows thermal hydraulic calculations for the small pin and the standard pin. All the calculations of pressure drop and MDNBR do not assume any grid spacers or any sort obstructions in the path of the coolant which could indeed increase the pressure drop and for
17
Melt refined fuel the temperature calculations, Dittus-Boelterβs correlation is used to calculated the heat transfer coefficients. The MDNBR for the standard fuel pin is 2.17 [7] and is considered as reference value in this calculation and for the small pin fuel the MDNBR is 1.45 if the flow rate is kept constant. But for the case with small fuel pin the pressure drop decreases, so the flow rate is increased to match the pressure drop of the case with standard pin diameter, i.e. the mass flow is increased from 18627 kg/s to 27000 kg/s. For the case with increased mass flow, the MDNBR value is 2.14 for the small pin, which is close to the reference value. But there is a slight increase in the fuel centre line temperature all the values are tabulated in the Table 3.7. The peak fuel centreline temperature for the small diameter pin is 2185β and this value is still less than 2280β, which is typical value for 3400 MW PWR [10].
Table 3.7 Thermal hydraulic calculations for the small pin and the standard pin
Type of fuel pin
Power (MW)
MDNBR
Pressure Drop (MPa)
Mass Flow (kg/sec)
Peak Fuel centreline temperature (β)
Average fuelcentre line temperature (β)
Standard
3800
2.16
0.08
18627
2087
1024
Small pin
3800
1.45
0.05
18627
2199
1024
Small pin
3800
2.14
0.08
27000
2185
1064
MR standard pin MR annular pin with ZrO2 in the annulus
1.10
MR annular pin with SiC in the annulus
1.05
kο₯
1.00
0.95
0.90
0.85
0.80 0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 3.4 Evolution of πβ with burnup for annular pins fueled with MR fuel with ZO2 and SiC, and reference case.
18
Melt refined fuel 3.5.2 Annular pin with moderating material in the annulus The other way to increase the moderation is by using annular pins with some moderating material in the annulus. Two different moderating materials are tested in this section, zirconium dioxide (ZrO2) and silicon carbide (SiC). As the fuel mass is reduced to accommodate moderating material at the centre the specific power increases, which results in shorter cycle lengths unless high burnups are reached. Figure 3.4 shows the evolution of πβ with burnup for the PWR cores with annular pins with ZrO2 and SiC in the annulus and fuelled with MR fuel, and for the reference case The πβ value at BOL for annular fuel pin fuelled with MR fuel and ZrO2 in the annulus is higher compared to the 100% MR fuel, and with SiC in the annulus the πβ is much higher.
3.5.2.1 Burnup and Cycle length The values of the maximum achievable average burnup and cycle length for two PWR cores with annular pins fuelled with MR fuelled processed B&B fuel with ZrO2 and SiC in the annulus are calculated using the same methodology as in the section 1.4.2. The calculated values are tabulated in the Table 3.8. Having SiC in the annulus higher value burnup can be achieved compared to ZrO2 in the annulus due to higher moderation capacity of SiC. Using these types of annular fuel pins higher average discharge burnup can be achieved compared to standard fuel pin with MR fuel, but the cycle length decreases due to decrease of heavy metal content in the fuel pin. It is not economically feasible to operate reactors on such a short cycle lengths, so using annular pins is not be best way to increase the burnup. All the calculations are done using batch fraction 1/5
Table 3.8 Discharge burnup and cycle length of PWR cores fueled with annular fuel pins fueled with MR processed B&B fuel.
Type of pin
Average discharge burnup (GWd/MTIHM)
Cycle length (days)
Annular fuel pin with ZrO2 in the annulus
53.8
176.9
Annular fuel pin with SiC in the annulus
59.6
195.7
Standard fuel pin
49.3
224.1
19
FLURO-OX processed fuel
4 FLURO-OX processed fuel The FLURO-OX process is developed by Yasuo Hirosea [11] is applicable for the fuel processing of ABWR with a MOX core. In this process, the cladding is removed during voloxidation; NF3 is used as thermal sensitive reagent, it reacts with different compounds at different temperatures. At 300 β NF3 reacts with Tc and Mo oxide, at 400β with Ru, Rh Pd, and Te and at 500β with Uranium oxides to form volatile fluorides. Then oxide precipitation process is applied to separate TRU elements from other residual materials of fluorination process effectively. After oxide precipitation process Sr, Cs, Ba, Ln and Ca are removed in liquid phase. The removal fractions in the FLURO-OX process is tabulated in the Table 4.1 The detail flow sheet of the FLURO-OX process applicable to the fuel processing of the ABWR with a MOX Core is shown in the Figure 4.1. For further discussions this process is referred as FO process.
Figure 4.1 Flow sheet of the βFluor-Oxβ Process Applicable to the Fuel Processing of the ABWR with a MOX Core. [11] Table 4.1 Removal fractions in the FO process
Element Th, Am U Other HM FPs
FO process 0% 100% 0% 100% H , He ,N ,O ,F ,Ne ,Cl ,Ar ,Kr Xe ,Rn ,Se ,Sr ,Cs ,Ba ,Ln ,Ca ,Ru , Rh ,Pd ,Te ,Tc ,Mo
20
FLURO-OX processed fuel
4.1 Objective The main objective of this chapter is to study the feasibility of FO process, in order to be applied to B&B discharge fuel. When compared to MR process and AIROX process, FO process looks more promising in neutronic point of view as it removes more fission products.
4.2 Density The density of the FO processed fuel is determined using the methodology as in the section 1.4.1. The evaluated density, heavy metal density and the heavy metal mass fraction values for the FO fuel are tabulated in the Table 4.2 in comparison with the 5% enriched UO2 fuel
Table 4.2 Density of FO processed fuel along with UO2 fuel
Type of fuel
HM mass fraction (%)
fuel density (g/cm3)
HM density (g/cm3)
UO2 fuel
88.1
10.40
9.16
FO processed B&B fuel
49.0
8.11
3.96
4.3 Problem with FLURO-OX processed fuel As most of the heavy metal in the FO processed B&B discharge fuel is plutonium, using only FO processed fuel the reactor becomes completely unpractical which always stays super critical. In order to avoid such a problem inert matrix materials like zirconium dioxide can be introduced but still one more problem pertain i.e. the positive reactivity coefficients due to the lack of resonant materials. This is an important problem concerning the safety and control of the reactor. To counterbalance the positive reactivity coefficients, resonant materials like U-238, Th-232 and Er should be added.
4.4 FLURO-OX fuel mixed with Natural uranium Oxide As mentioned in the section 4.3, some resonant material has to be added to the FO processed fuel; in this study U-238 was chosen. Adding natural uranium which has 99.3% U-238 the problem of positive reactivity coefficients can be managed. Natural uranium dioxide (NatUO2) is added to the FO processed fuel in different mass fractions to find a good practical solution.
21
FLURO-OX processed fuel FO processed fuel is mixed with NatUO2 in different mass fraction from 0% to 20%. Below 10% the achievable burnup is very low. The Figure 4.2 shows the evolution of πβ value with burnup for three different cases with 10% FO, 15% FO and 20% FO mixed with 90% NatUO2, 85% NatUO2 and 80% NatUO2 respectively.
10% FO + 90% UO2
1.2
15% FO + 85% UO2 20% FO + 80% UO2
kο₯
1.1
1.0
0.9
0.8
0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 4.2 Evolution of πβ with burnup for PWR cores fueled with FO processed fuel mixed with NatUO2
4.4.1 Burnup and Cycle length The maximum attainable average discharge burnup and the cycle length values for the three different PWR cores fuelled with 10% FO fuel mixed with 90% NatUO2 fuel, 15% FO fuel mixed with 85% NatUO2 fuel and 20% FO fuel mixed with 80% NatUO2 fuel respectively are calculated using the same methodology as in the section 1.4.2 and are tabulated in the Table 4.2. It can be observed that with increase in the FO% the achievable burnup increases, with the increase in FO mass fraction in the fuel from 10% to 20% the burnup is increased from 51 to 106 GWd/MTIHM. All the calculations are done using batch fraction 1/5 Table 4.3 Discharge burnup and cycle length for PWR cores fueled with FO processed fuel mixed with natural UO2
Type of Fuel
Average discharge burnup (GWd/MTIHM)
Cycle length (days)
FO-10% and 90% Nat U
51.1
283.0
FO-15% and 85% Nat U
77.8
413.5
FO-20% and 80% Nat U
106.0
539.6
22
FLURO-OX processed fuel
4.5 Modified Fluro-Ox process (MFO) A modification in the FLURO-OX process is suggested in this section. Instead of adding natural uranium oxide the FLURO-OX process is modified, so that uranium is not removed in the first place. This modified process is shown in the Figure 4.3. The final product in this process would be Uranium oxide mixed with TRU oxide and zirconium oxide. This process is similar to the original process with only difference being the exclusion of heating the fuel to about 550β in the fluidized bed reactor. In this study no loss of uranium and plutonium are considered. Without the removal of uranium this process becomes more proliferation resistant compared to the original FLURO-OX process. For the convenience in further discussion this process is referred as MFO which stands for Modified Fluro-Ox process.
B&B fuel assembly
Fuel rod cropping
H-3, Kr-85, Xe, I, Sr
Air/oxygen
Voloxidation
Hull, Spring, Cladding
U3O8,FPs, TRU
Fluidized Bed Reactor
400 ΛC
Ru, Rh, Pd, Te
300 ΛC
Tc, Mo
U, TRU, Zr, Sr, Cs, Ba, Ln
NF3/Ar
CaO
Sr, Cs, Ba, Ln, Ca Liquid phase
LiF-NaF solution
750 ΛC Ppt. vessel TRUO2+UO2+ZrO2 Ppt.-seperation
LnO2 filtered Waste salt Sr, Ba, Cs, Ca
Figure 4.3 Modified FLURO-OX process
4.5.1 Density
The density of the discharge B&B fuel processed from the MFO process is determined using the same methodology as in the section 1.4.1. The value of the density, heavy metal density and heavy metal mass fraction are tabulated in the in comparison with the discharge B&B fuel processed from the original FO process. 23
FLURO-OX processed fuel Table 4.4 Density of the MFO processed and FO processed fuel
Type of fuel
HM mass fraction (%) 49.0 80.2
FO MFO
Fuel density (g/cm3) 8.11 10.27
HM density (g/cm3) 3.96 8.24
1.20
MFO 1.15
1.10
kο₯
1.05
1.00
0.95
0.90
0.85 0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 4.4 Evolution of πβ with burnup for PWR core fuelled with MFO processed fuel
4.5.2 πβ evolution, Burnup and Cycle length Figure 4.4 shows the evolution of πβ with burnup for the PWR core with MFO processed fuel. To see clearly the impact of the new process, the maximum achievable average discharge burnup and cycle length values for PWR core fuelled with MFO processed fuel are calculated using similar methodology as mentioned in the section 1.4.2. The values for both 3 and 5 batch cores are tabulated in the Table 4.5. The achievable burnup is much higher compared to the PWR fuelled with 5% enriched UO2 fuel. Table 4.5 Discharge burnup and cycle length for PWR core fueled with Modified Fluro-Ox process
MFO processed fuel Parameter 3 batch core
5 batch core
Burnup (GWd/MTIHM)
89.6
98.8
Cycle length (days)
740.4
489.9
24
Assembly model
5 Assembly model
Average disharge burnup (MWd/MTIHM)
In this chapter instead of representing the PWR core using infinite unit pin cell with reflective boundary conditions, core is represented using an assembly model with infinite length and reflective boundary conditions. The interest in going for the assembly model is to take advantage of the water in the control rod guide tubes and instrumentation thimble. In the normal pin cell the M/F ratio is 1.67, whereas in the assembly model the M/F ratio is 1.95. With this increase in M/F ratio, the gain in the maximum achievable average discharge burnup when PWR is fuelled with 5% enriched UO2 fuel is very less, however it has significant impact on the achievable burnup when PWR is fuelled with MR processed fuel. It can be observed from the Figure 5.1 that the average discharge burnup increases significantly with the increase in the M/F ratio from 1.67 to 4.5. In this chapter the pin cell calculation is compared with the assembly calculation and different assembly designs with different types of fuels are studied.
110000
MR fuel
100000 90000 80000 70000 60000 50000 40000 2
4
6
8
10
M/F ratio
Figure 5.1 Average discharge burnup evolution with M/F ratio for PWR fuelled with MR fuel
5.1 Geometry
The PWR assembly consists of 264 fuel rods, and 24 control rod guide tubes and one instrumentation thimble. In general when the control rods are out the control rod guide tubes they are filled with water. 25
Assembly model
Figure 5.2 PWR assembly model
5.2 πβ evolution Figure 5.3 shows the evolution of πβ with burnup for both assembly and pin cell models fuelled with MR processed fuel. It can be observed that the πβ value for the assembly model is higher than the pin cell model representation of PWR core, which is consistent with the initial hypothesis.
1.15
MR pin cell MR assembly model
1.10
1.05
kο₯
1.00
0.95
0.90
0.85
0.80 0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 5.3 Evolution of πβ with burnup for assembly and pin cell both fuelled with MR processed fuel
26
Assembly model
5.3 Burnup &Cycle length
The maximum achievable average discharge burnup and cycle length values for both assembly and pin cell models fuelled with MR processed fuel are calculated using the methodology same as in section 1.4.2 and are tabulated in the Table 5.1. It can be observed from the Table 5.1 that the burnup increases from 49.3 GWd/MTIHM to 65.3 GWd/MTIHM when assembly model is used, the increase in quite significant. Therefore assembly model is used for further analysis to represent the PWR core. All the calculations are done for a five batch core.
Table 5.1 Discharge burnup and cycle length for assembly and pin cell model representation of PWR core fueled with MR processed fuel
Type of calculation
M/F ratio
Average discharge burnup (GWd/MTIHM)
Cycle length (days)
Pin cell model
1.67
49.3
224.1
Assembly model
1.95
66.0
299.9
5.4 More assembly models In this section eight different cases are modelled, some using standard assembly model and others with slight modification in the assembly model. In cases 1 and 2, standard assembly model is fuelled with MR processed fuel, with standard fuel pins (fuel radius ~ 4.1 mm) in case-1 and smaller fuel pins (fuel radius ~ 3.46 mm) in case-2. The M/F ratio is higher in case-2 compared to case-1. In case-3 and case-4, standard assembly model with annular fuel pins fuelled with MR processed fuel and SiC in the annulus. The radius of the annulus is 2.2 mm and 1.9 mm for case-3 and case-4 respectively. The M/F ratio is higher in case-3 compared to case-2, due to more SiC volume. Cases 5 and 6 use different fuel assembly models with 4 extra water rods and 8 extra water rods respectively, both these assembly models are fuelled with MR-processed fuel. Cases 7 and 8 both use standard assembly model and are fuelled with MFO processed fuel and 5% enriched uranium UO2 respectively. A linear power of 46252.8 W/cm is assumed for the depletion analysis for the cases 1,2,3,4 and 7, this corresponds to a typical linear power, 17.52 kW/m per fuel rod. As the linear power per fuel rod is kept constant at 17.52 kW/m for both the cases 5 and 6, the resulting linear powers are 45552.0 W/cm and 44851.2 W/cm respectively, the reduction in the linear power is due to the less number of fuel rods. 27
Assembly model
Figure 5.4 (a) & (b) shows assembly designs with 4 and 8 extra water rods respectively
5.4.1 πβ evolution Figure 5.5 shows the evolution of πβ for all the eight cases. The case 1 is taken as reference where the MR processed fuel is used in standard assembly; with smaller pins the πβ value increases from 1.11 to 1.19 due to the increase in M/F ratio i.e. in case 2. There is a very slight increase in πβ value with the increase of water rods i.e. in case 5 and case 6 compared to case 1 but not significant. Using annular fuel pins the πβ values at BOL are 1.13 and 1.12 for cases 3 and 4 respectively. Whereas, when FO processed fuel is used in the standard assembly model the πβ value at BOL increased from 1.11 to 1.2.
1.5
case 1 case 2 case 3 case 4 case 5 case 6 case 7 case 8
1.4 1.3 1.2
kο₯
1.1 1.0 1.13
0.9 kο₯
1.12
0.8
1.11
1.10
0.0
0.7
0.5
1.0
1.5
Burnup (GWd/MTIHM)
0
20
40
60
80
100
120
Burnup (GWd/MTIHM)
Figure 5.5 Evolution of πβ with burnup for different assembly designs with different fuels
28
Assembly model 5.4.2 Burnup and cycle length In order to clearly see the impact of different assembly models, the maximum achievable average discharge burnup and cycle length values are calculated using the same methodology mentioned in the section 1.4.2. The burnup and cycle length values calculated for both 3 batch and 5 batch cores are tabulated in the Table 5.2 Table 5.2 Discharge burnup and cycle length for different assembly designs with different fuel compositions
Average linear power per assembly (W/cm)
Average burnup Cycle length (GWd/MTIHM) (days) 5 batches 5 batch
Average burnup (GWd/MTIHM)
Cycle length (days)
3 batch
Case-1
46252.8
66.1
294.9
59.7
451.9
Case-2
46252.8
95.2
309.2
86.7
469.1
Case-3
46252.8
75.9
240.5
68.8
363.5
Case-4
46252.8
72.1
255.3
65.2
385.0
Case-5
45552.0
68.1
309.6
61.8
467.9
Case-6
44851.2
70.2
318.8
63.5
480.8
Case-7
46252.8
108.2
536.2
97.9
808.8
Case-8
46252.8
67.2
371.0
60.4
555.2
There is a slight increase in the attainable burnup value in the cases 5 and 6 with the introduction of the water rods, but the increase is not really significant and there is slight decrease in the overall power produced per assembly as well. There is a 45% increase in the burnup value when the smaller pins are used i.e. in case 2. Even with annular fuel pins (cases 3 and 4) the increase in the burnup is quite significant but due to the less heavy metal in the fuel pin the cycle lengths are shorter compared to the other cases. So only the two cases using MR processed fuel are considered for further analysis, one with standard fuel pin (case-1) and other with small fuel pin (case-2). Whereas, in case-7 where MFO processed fuel is used in standard assembly model an additional burnup of 108.2 GWd/MTIHM can be achieved so this case is also considered for further analysis. For all further analysis assembly model representation of core is used.
29
Reactivity coefficients
6 Reactivity coefficients This chapter is devoted for the calculation of reactivity coefficients, as they play a vital role in the reactivity control and safety. The reactivity coefficients are calculated for the following three cases, I. II. III.
PWR assembly model (M/F~1.95) fuelled with MR processed B&B discharge fuel PWR assembly model (M/F~1.95) fuelled with MFO processed B&B discharge fuel PWR assembly model with reduced pin diameter i.e. with M/F ratio 3.2 fuelled with MR processed B&B discharge fuel.
Five batch core is considered, for the calculation of the reactivity coefficients at any particular time in the cycle, simulation is done using an assembly model with average fuel composition which is the average of the fuel composition of all the five batches at that particular time.
6.1 Excess reactivity and critical boron concentration Before calculating the reactivity coefficients, the evolution of excess reactivity with the cycle length is calculated for all the three cases. The excess reactivity values are given by the equation ( 1.2 ). In PWRs the excess reactivity is compensated by introducing soluble boron. The maximum excess reactivity occurs at the beginning of the cycle as shown in the Figure 6.2. So, the maximum critical boron concentration appears at the beginning of the cycle. The critical boron concentration that is required to compensate the excess reactivity is calculated for each case and values are tabulated in the Table 6.1. Even though the excess reactivity is higher in case of 5% enriched UO2 fuel compared to MFO processed fuel, the critical boron concentration is higher in the as of MFO processed fuel due to the difference in the spectrum. Figure 6.1, shows that the spectrum is harder when MFO processed fuel or MR processed fuel is used compared 5% enriched UO2 fuel. Due to the less neutrons in thermal region the efficiency of boron reduces in the case of MFO processed fuel and MR processed fuel. As it can be observed from the plot that the spectrum get softer with the reduction is pin diameter when MR processed fuel is used, due to the increase in moderation. It should be noted that the burnup reactivity swing is significantly smaller than the conventional PWRβs fuelled with UO2 fuel. Table 6.1Critical boron concentration for different cases at BOC
Critical Boron concentration (ppm) 2445 1240 1420 1400
Type of fuel MFO fuel MR fuel MR fuel in small pin 5% enriched UO2
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Reactivity coefficients
Normalized neutron flux / lethargy
0.25
MFO fuel MR fuel MR fuel (small diameter pin) 5% enriched uranium
0.20
0.15
0.10
0.05
0.00 1E-9
1E-8
1E-7
1E-6
1E-5
1E-4
1E-3
0.01
0.1
1
10
Energy (MeV)
Figure 6.1 Normalized neutron flux per unit lethargy at BOC for PWR cores fueled with MR, MFO and UO2 fuels
10000
MFO fuel MR fuel MR fuel (small pin) 5% enriched uranium
Excess reactivity (pcm)
8000
6000
4000
2000
0 0
100
200
300
400
500
600
Cycle length (days)
Figure 6.2 Excess reactivity vs cycle length for PWR cores fueled with MFO, MR and UO2
31
Reactivity coefficients
6.2 Methodology In order to calculate the coolant temperature coefficients, the thermal scattering cross section library and the density of the coolant has to be changed with respect to the temperature of the coolant. At BOC the coefficients are calculated, where there is maximum boron concentration. The temperature of the coolant is changed over a range of 50β from 277β to 327β, the densities and libraries are changed accordingly to calculate the coolant reactivity coefficient. Similarly, for the Doppler coefficient the temperature of the fuel is changed from 627β to 677β. The reactivity coefficients for the variation from state 1 to 2 are determined by the following equation,
πΌ=
π1 β π2 105 πππ Γ π1 π2 π1 β π2 β
( 6.1 )
where k1 and k2 are the multiplication factors of the state 1 and 2 respectively, and T1 and T2 are the temperatures of the state 1 and 2 respectively. The values of the calculated coolant temperature and Doppler coefficients are tabulated in the Table 6.2. It should be noted that both Doppler and coolant temperature coefficients are negative for all the three cases.
Table 6.2 Reactivity coefficient values at BOC for PWR cores fueled with MR and FO fuels.
Fuel
Pin diameter (mm)
Assembly type
Moderator Temperature coefficient (pcm/β)
Doppler Coefficient (pcm/β)
4.75
17 x 17
-39
-2.09
4.117
17 x 17
-27
-3.09
4.75
17 x 17
-22
-3.63
MR
MFO
6.3 Void coefficients Another important reactivity coefficient is the void coefficient; Figure 6.3 shows the evolution of πβ with void fraction at BOC. When MFO processed fuel is used the πβ value decreases up to 70% void but increases thereafter, however the void coefficient is still negative. The same trend is observed with MR processed fuel, but with small fuel pin the coefficient is more negative compared to the standard fuel pin.
32
Reactivity coefficients
1.04
MFO MR MR (~small pin)
1.02 1.00 0.98
kο₯
0.96 0.94 0.92 0.90 0.88 0.86 0.84 0
20
40
60
80
100
Void Percentage (%)
Figure 6.3 Evolution of πβ with void percentage at BOC for PWR cores fueled with MR and FO fuels
It can be remarked from this chapter that all reactivity coefficients are negative for these three cases.
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Discharge fuel composition
7 Discharge fuel composition In order to find out the best possible way to incinerate the fissile plutonium of the B&B discharge fuel, the amount of plutonium discharged from each case has to be analysed. The same three cases are considered in this chapter as in the chapter 6. The three cases are mentioned here again for the convenience, I. II. III.
PWR core represented with an assembly model (M/F~1.95) fuelled with MR processed B&B discharge fuel PWR core represented with an assembly model (M/F~1.95) fuelled with MFO processed B&B discharge fuel PWR core represented with an assembly model with smaller pins i.e. with M/F ratio 3.2 fuelled with MR processed B&B discharge fuel.
Using mass balance [12] the amount of plutonium discharged per unit electricity generated is calculated. The thermal efficiency of B&B reactor is taken as 40% and for PWR as 33%. Table 7.1 shows the characteristics of the discharge fuel for all the three cases. Table 7.1 Characteristics of discharge fuel
Characteristic
Breed and Burn reactor
5% enriched UO2 (M/F~1.95)
MR fuel (M/F ~1.95)
MFO fuel (M/F~1.95)
MR fuel with small pin (M/F~3.2)
3 batch
5 batch
3 batch
5 batch
3 batch
5 batch
3 batch
5 batch
Burnup (GWD/MTIHM)
171
60.5
67.3
59.7
66.1
97.9
108.2
86.7
95.2
Fuel density (g/cc)
17.3
10.4
10.4
9.95
9.95
10.27
10.27
9.95
9.95
Heavy Metal density (g/cc)
16.4
9.16
9.16
7.50
7.50
8.24
8.24
7.50
7.50
Specific HM loading (kg/GWeY)
5339.9
18294.5
16446.0
4145.8
4048.8
3626.9
3508.4
3765.0
3659.2
Uranium utilization (%)
18.00
0.60
0.60
23.50
23.90
26.20
27.10
24.70
25.50
Fissile/ Total plutonium (%)
81.8
68.3
62.7
69.3
67.8
59.9
57.8
61.1
57.8
Total Pu/ HM (%)
12.28
1.16
1.25
9.91
9.61
8.53
8.09
7.98
7.48
kg of Pu/GWeY
537.7
210.9
204.3
314.3
296.1
228.3
206.9
226.2
203.9
Kg of Fissile Pu /GWeY
439.8
144.1
128.1
217.8
200.8
136.8
119.6
138.2
117.9
34
Discharge fuel composition The uranium utilization in a B&B (TWR) reactor is 18%. The amount of discharged plutonium per unit electricity generated is 537.7 kg/ GWeY. It can be observed that there is a significant decrease in the amount of plutonium due to the additional irradiation. Using MR fuel in PWR core, the amount of plutonium discharged per unit electricity produced is 296. kg versus 537.7 kg/ GWeY. The discharge plutonium content can be decreased further when MR fuel is used in the PWR core with smaller diameter pins to about 203.9 versus 204.3 kg/ GWeY when 5% enriched UO2 is used. When MFO fuel is used in the PWR core the amount of discharge plutonium is 206.9 versus 204.3 kg/ GWeY when 5% enriched UO2 fuel is used. It can be remarked that the reduction of plutonium in the final discharge fuel can be maximized using MFO processed B&B discharge fuel. Moreover the utilization of uranium can be increased further from 18% to 27% when MFO processed discharge B&B fuel is used in PWR core. Overall the dual-tier system of TWR-PWR is effective in increasing the uranium utilization and reducing the fissile plutonium content.
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Multiple recycling
8 Multiple recycling Using MFO processed B&B discharge fuel in a PWR core an additional burnup of about 108 GWd/MTIHM can be achieved. But the average discharge burnup of the currently working PWRβs is about 50 GWd/MTIHM [13]. The discharge burnup is supposed to increase to about 55 GWd/MTIHM [14] in near future. Some studies [15] suggest that the discharge burnup can be increased further to about 60 GWd/MTIHM or more, thanks to the newly developed zirconium alloys. The highest average fuel burn-up attainable within the 5.0% enrichment limit is approximately 65GWd/MTIHM [13]. As going to higher burnup in a PWR core might not be practical at the present time, taking the limit as 65 GWd/MTIHM three different cases are studied in this chapter; the fuel is discharged in the first cycle at 50, 60 and 65 GWd/MTIHM in cases 1, 2 and 3 respectively and then the discharge fuel is processed again using MFO process and loaded back in the PWR for the second cycle as shown in the Figure 8.1.
Figure 8.1 2-tier system with twice burning cycle in PWR
The maximum attainable average discharge burnup for the second cycle is tabulated in the Table 8.1. For cases 1, 2 and 3, the maximum attainable discharge burnup for the second cycle are 70.1, 61.8 and 57.7 GWd/MTIHM respectively. It should be noted that the cycle length of the first and second cycle are different due to the difference in burnup and the difference in the initial heavy metal content. In case 2 the discharge burnup of the first and second cycle are almost the same. The amount of plutonium discharged per unit electricity 36
Multiple recycling generated for the case 2 is about 188.3 versus 203.9 kg / GWeY for once through cycle in PWR, moreover the amount of fissile plutonium discharged is 108.2 versus 128.1 kg / GWeY of PWR fuelled with 5% enriched UO2 fuel.
Table 8.1 Achievable average discharge burnup and discharge plutonium values for the 2 tier system with twice burning cycle in PWR
Case number
Discharge burnup (GWd/MTIHM) First cycle Second cycle
Kg of Pu/GWeY
Fissile Pu/Total Pu
Kg of Fissile Pu/ GWeY
Case 1
50
70.1
189.9
57.1%
108.4
Case 2 Case 3
60 65
61.8 57.7
188.3 187.3
56.6% 56.4%
106.6 105.6
It can be remarked that the twice burning approach is more effective in making the discharge fuel more proliferation resistant.
37
Conclusion
9 Conclusion In this study different possibilities to extract more energy and to incinerate the fissile plutonium content in the discharge fuel of the B&B reactor are examined. At first, it can be remarked that the dual tier system of TWR and PWR is effective in incinerating the plutonium. As the total power produced from the initial heavy metal increases, the Pu discharged per unit electricity generated decreases. Three reprocessing processes with high proliferation resistance are studied in order to process the discharge fuel of B&B reactor namely, AIROX, Melt Refining (MR) process and FLURO-OX process. Using AIROX process it is not possible to achieve high burnup value without increase the M/F ratio significantly. When PWR is fuelled with MR processed fuel an additional burnup of 49.3 GWd/MTIHM can be achieved. Using inert matrix and annular MR fuels proved to be ineffective, even though the burnup increases the cycle length decreases. However the burnup can be increased further when the radius of the fuel is reduced from 4.11 mm to 3.46 mm, the achievable burnup value increases from 49.3 to 87.2 GWd/MTIHM. It should be noted that the standard FLURO-OX process removes uranium, with slight modification in the process uranium can be retained in the fuel; this modified FLURO-OX (MFO) process proves to be more efficient than the MR process and AIROX process, an additional burnup of 98.8 GWd/MTIHM can be achieved using this process. Instead of representing the PWR core with a pin cell an assembly model representation is used. In assembly the M/F ratio is 1.95 compared 1.65 of pin cell due to the water present in the control rod guide tubes and instrumentation thimble. The impact of this very significant when MR processed B&B discharge fuel is used, the average discharge burnup increases from 49.3 to 66 GWd/MTIHM this implies that using pin cell model the burnup value is underestimated. Using assembly model to represent the PWR core and with smaller diameter pins fuelled with MR processed B&B discharge fuel the average discharge burnup increases to about 95.2 GWd/MTIHM. Whereas using MFO process the achievable average discharge burnup is about 108.2 GWd/MTIHM. Using MR processed B&B discharge fuel in PWR core, the amount of plutonium discharged per unit electricity produced is 296.1 kg versus 537.7 kg/ GWeY for B&B reactor. The discharge plutonium content can be decreased further when the MR fuel is used in a PWR core with smaller diameter pins to about 203.9 kg/ GWeY. When MFO fuel is used PWR core the amount of discharge plutonium is 206.9 versus 204.3 kg/ GWeY when 5% enriched UO2 fuel is used. The natural uranium utilization of the dual tier TWR-PWR system is about 27% when MR processed B&B discharge fuel is used compared to 0.6% of conventional PWR.
38
Conclusion Instead of using once through cycle in the PWR, a twice burning cycle is suggested where the fuel is discharged from the PWR at 60 GWd/MT IHM and recycled into the PWR after reprocessing using FLURO-OX process. In this case an additional burnup of 61.8 GWd/MTIHM can be achieved in the second cycle and the amount of discharge plutonium can be decreased further to about 188.3 kg/ GWeY. Finally it can be concluded that MFO process is the best way to reprocess the B&B discharge fuel and moreover twice burning cycle in PWR is most effective than the once through cycle. The concept of multiple recycling has to be analysed further in future to maximize the reduction of plutonium and to increase the uranium utilization.
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Bibliography
10 Bibliography [1] S.M.Feinberg, βDiscussion comment,β in Rec. of Proc. Session B-10, ICPUAE, Geneva, Switzerland, 1958. [2] E. Greenspan, βA Phased Development of Breed-and-Burn Reactors for Enhanced Nuclear Energy Sustainability,β Sustainability, no. 4, pp. 2745-2764, 2012. [3] C. D. Sanzo, J. Vujic and E. Greenspan, βFeasibility of Fueling PWRs with Fuel Discharged from Breed and Burn Reactors,β Energy Procedia, 2014. [4] C. Ahlfeld, T. Burke, T. Ellis, P. Hejzlar, K. Weaver and C. Whitmer, βConceptual Design of 500 MWe Traveling wave Demonstartion Reactor Plant,β Nice, France, 2011. [5] C.Humrickhouse and J.Walter, βMelt refining of metallic fast reactor fuel,β in ICHEME Nuclear Fuel Conference, Mancheste, UK, 2012. [6] E.Greenspan and F.Ganda, βNeutronic Analysic of Hybride Fueled PWR Cores,β Nuclear Engineering and Design,239, pp. 1425-1441, 2009. [7] C. Shuffler, J. Trant, J. Malen and N. Todreas, βThermal hydraulic analysis for grid supported pressurized water reactor cores,β Nuclear Engineering and Design,239, pp. 1442-1460, 2009. [8] D. Majumdar, S. N. Jahshan, C.M.Allison and P. Kuan, Recycling of Nuclear Spent Fuel with AIROX Processing, U.S. Deparment of Energy, DOE Idaho Field Office, 1992. [9] N. Todreas and M. Kazimi, Nuclear Systems I, New York: HPC, 1990. [10] A. V. Nero, A Guidebook to Nuclear Reactors, Berkeley: University of California Press, 1979. [11] Y. Hirose, K. Mitachi and Y. Shimazu, βFeasibiity of Molten Salt Fast Reactors for emerging national tasks,β in ANS International High-Level Radioactive Waste Management, Charleston, U.S.A, 2015. [12] E. Greenspan, Maximum Fuel Utilization in Fast Reactors Without Chemical Reprocessing, Berkeley: Nuclear Energy Univeristy Programs , U.S Department of Energy, 2012. [13] βVery High Burn-ups in Light Water Reactors,β NEA, Paris, 2006. [14] βImpact of High Burnup Uranium Oxide and Mixed Uraniumβ Plutonium Oxide Water Reactor Fuel on Spent Fuel Management,β IAEA Nuclear Energy Series.
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Bibliography [15] βTechnical and economic limits to fuel burnup extension,β IAEA-TECDOC-1299 , San Carlos de Bariloche, Argentina, 1999. [16] K. Ikeda and H. Sekimoto, βTRU burning by dual tier system of LWR-SFR,β Progress in Nuclear Energy, vol. 53, pp. 902-908, 2011. [17] N. Takaki and H. Sekimoto, βPotential of CANDLE Reactor on Sustainble and Strengthened Proliferation Resistance,β Progress in Nuclear Energy, no. 50, pp. 114118, 2008. [18] N. Todreas and M. Kazimi, Nuclear Systems I, hpc, 1990.
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