INVESTIGATION OF AN INERTIAL CONFINEMENT FUSION-FISSION HYBRID REACTOR

SYSTEMS AND ENGINEERING TECHNOLOGY

INVESTIGATION OF AN INERTIAL CONFINEMENT FUSION-FISSION HYBRID REACTOR

INVESTIGATION OF AN INERTIAL CONFINEMENT FUSION-FISSION HYBRID

REACTOR

Kiranjit MejerPTNR Research Project 2009

Frazer-Nash ConsultancyUniversity of Birmingham

© Frazer-Nash Consultancy Ltd 2010. All rights reserved. Confidential and proprietary document. SYSTEMS AND ENGINEERING TECHNOLOGY

The Basic Concept

Fusion neutron sourceD + T → α + n + 17.6 MeV(n energy 14.1 MeV)

Sub critical fission blanket Neutron multiplier blanket Reflector

Benefits of a Hybrid

Waste transmutation – reducing inventory of HLW

Production of energy Development of fusion

technology Inherent safety

The Fusion-Fission Hybrid Reactor


Laser Inertial confinement Fusion-Fission Energy Engine

Inertial confinement fusion source

Surrounded by Beryllium blanket Spherical blanket of sub-critical

fission fuel Graphite blanket Pb-Li first wall coolant FLiBe (2LiF+BeF2) coolant Power conversion system

Image from ”Thermal and Mechanical Design Aspects of the LIFE Engine” R P Abbot et al, 2009


Multiplication Factor of Be Blanket

Pure 9Be – 1.85 gcm-3

Peak at 17 cm blanket thickness Factor ~ 2.06

Pebble packing fraction 60 % - 1.11 gcm-3

Factor ~ 1.81 at 16 cm Supported by “A Sustainable Nuclear

Fuel Cycle Based on Laser Inertial Fusion Energy” Moses et al, 2009

Neutron Multiplication Factor as a Function of Beryllium Blanket Thickness

0.5

0.7

0.9

1.1

1.3

1.5

1.7

1.9

2.1

2.3

0 5 10 15 20 25 30

Thickness/cm

Mul

tipl

icat

ion

Fac

tor

Neutron Multiplication Factor as a Function of Beryllium Blanket Thickness - 60 % packing

0.5

0.7

0.9

1.1

1.3

1.5

1.7

1.9

2.1

2.3

0 5 10 15 20 25 30 35 40

Thickness/cm

Mul

tipl

icat

ion

Fac

tor


Fuel Blanket Investigation

Below - Energy gain from fission blanket of natural Uranium 19.1 gcm-3

surrounding a Beryllium blanket

Energy Gain vs Fuel Blanket Thickness - Natural U

0

5

10

15

20

25

0 10 20 30 40 50 60 70 80 90 100

Fuel Blanket thickness/cm

Ene

rgy

Gai

n

Energy Gain vs Fuel Blanket Thickness - U-238

0

0.5

1

1.5

2

2.5

3

0 10 20 30 40 50 60 70 80 90 100

Fuel Blanket thickness/cm

En

erg

y G

ain

Above - Energy gain from fission blanket of pure 238U


Energy Spectrum of Neutrons

Neutron energy entering the fission blanket

~ 0.05 at 14 MeV Large proportion at thermal

energies

Maxwell-Boltzmann distribution peaks at 0.025 eV

Spectrum of neutrons returning from reflector shows same form

Thermal Neutron Energy Spectrum

0.00E+00

5.00E-03

1.00E-02

1.50E-02

2.00E-02

2.50E-02

3.00E-02

0.00E+00 5.00E-08 1.00E-07 1.50E-07 2.00E-07 2.50E-07 3.00E-07

Energy/MeV

Num

ber

(nor

mal

ised

per

sta

rtin

g n)

Fast Neutron Energy Spectrum

0.00

0.01

0.02

0.03

0.04

0.05

0.06

10.00 10.50 11.00 11.50 12.00 12.50 13.00 13.50 14.00 14.50

Energy/MeV

Num

ber

(nor

mal

ised

per

sou

rce

part

icle

)


Other Fuel Options

radius 1 cm

Outer radius 0.5 mm Kernel radius 0.3 mm

Buffer layer (C)

High-density Pyc

SiC

coated particles embedded in graphite matrix

30% TRISO

70% Carbon Fuel composition based on graphite pebbles containing TRISO particles

Image adapted from http://blogs.princeton.edu/chm333/f2006/nuclear/trisoball.jpg


Fission blanket energy gain and criticality

Fuel option Fission Energy Gain keff

Th-232 0.52 0.033

DU (0.26% 235U) 7.70 0.396

Natural U 16.08 0.576

LWR Spent Nuclear Fuel

27.60 0.720

Separated Transuranic Elements

183.14 0.966

Weapons grade plutonium

2.342


Coolant Effects

First wall coolant Pb83Li17

Primary coolant FLiBe (2LiF + BeF2)

6Li + n → 4He + T + Q7Li + n → 4He + T + n’ – Q

Tritium Breeding Ratio (TBR) – ratio of T produced to consumed For self sufficiency TBR > 1.05 Requires 6Li enrichment of 50% or more


Project Extensions

Improvements to the Model

Geometry – structural materials etc Fuel blanket compositions Temperatures Number of neutron histories Other fuel fabrication options Time dependent nature of the reactor - evolution of fuel with

breeding from fertile isotopes - flattening power output with 6Li content


Summary

Demand for clean, abundant energy and concerns over HLW management have led to renewed interest in the hybrid concept

MCNP modelling has demonstrated the viability of a number of fuel options particularly SNF

Enrichment of 6Li content in coolants can provide tritium self sufficiency for the reactor

Timescale for LIFE machine large

SYSTEMS AND ENGINEERING TECHNOLOGY

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MCNP Model

Isotropic, monoenergetic neutron point source

Pb-Li first wall coolant Beryllium multiplier blanket Fission Blanket Graphite reflector

Stochastic approach - uses random number generation and reaction cross section data to determine the ‘history’ of a particle

Many histories followed to give a representation of a real world situation


Documents

INVESTIGATION OF AN INERTIAL CONFINEMENT FUSION-FISSION HYBRID REACTOR