Utilization of light water reactors for plutonium incineration

~

Pergamon

Ann. Nucl. Energy gol. 22, No. 8, pp. 507 511, 1995

0306-4549(95)0086-7

Copyright © 1995 Elsevier Science Lid Printed in Great Britain. All rights reserved 0306-4549/95 $9.5(1+0.(1(t

U T I L I Z A T I O N OF L I G H T WATER R E A C T O R S F O R PLUTONIUM INCINERATION ALEX GALPERIN Department of Nuclear Engineering, Ben-Gurion University of the Negev, Israel (Received 5 November 1994)

Abstract---In this work a potential of incineration of excess Pu in LWR's is investigated. In order to maintain the economic viability of the Pu incineration option it should be carried out by the existing power plants without additional investment for plant modifications. Design variations are reduced to the fuel cycle optimization, i.e. fuel composition may be varied to achieve optimal Pu destruction. Fuel mixtures considered in this work were based either on uranium or thorium fertile materials and Pu as a fissile component. The slightly enriched U fuel cycle for a typical pressurized water reactor was considered as a reference case. The Pu content of all fuels was adjusted to assure the identical cycle length and discharged burnup values. An equilibrium cycle was simulated by performing cluster burnup calculations. The material composition data for the whole core was estimated based on the core, fuel and cycle parameters. The annual production of Pu of a standard PWR with 1100 MWe output is about 298 kg. The same core completely loaded with the MOX fuel is estimated to consume 474 kg of Pu, mainly fissile isotopes. The MOX-239 fuel type (pure Pu-239) shows a potential to reduce the initial total Pu inventory by 220 kg/year and fissile Pu inventory by 420 kg/year. The following two fuel types: TMOX and TMOX239 are based on Th-232 as a fertile component of the fuel, instead of U-238. The amount of Pu destroyed per year for both cases is significantly higher than that of U-based fuels. Especially impressive is the reduction in fissile Pu inventory: more than 900 kg/year. The safety related reactivity coefficients were found negative, which indicates that the basic behaviour of a reactor core utilizing Pu and Pu Th based fuel types will be quite similar to that of a standard PWR core utilizing slightly enriched U fuel. It was also found that the reactivity control of a core based on Pu fuel as a fissile component will be more difficult due to a reduced reactivity worth of the soluble boron and control rod control mechanisms.

INTRODUCTION Utilization of p l u t o n i u m (Pu) fuel in Light Water Reactors (LWR's) of current technology m a y be considered as a partial solution for disposing o f the spent fuel p r o d u c e d by existing power plants as well as for disposing o f excess weapons-grade material. Utilization of Pu in L W R ' s at present is limited to loading of mixed oxide fuel ( M O X ) assemblies as a p a r t of a reload batch, usually n o t exceeding 3 0 4 0 % o f the fuel assemblies. The M O X fuel is fabricated by blending reactor grade p l u t o n i u m extracted from the discharged fuel with n a t u r a l or depleted uranium. It m a y be stated t h a t current utilization of the M O X fuel in L W R ' s is m o t i v a t e d mainly by fuel utilization a n d fuel cycle cost considerations. In this work a potential of incineration o f excess Pu in LWR's is investigated. Efficient destruction of Pu is a d o p t e d as a design objective. This m e t h o d of Pu incineration m a y be considered as an alternative

to the waste disposal option, i.e. Pu vitrification with high level radioactive waste followed by disposal of glass canisters, An economic analysis of the waste disposal o p t i o n was carried out, the results of the analysis were published a n d c o m p a r e d with the M O X option ( G o l d s c h m i d t a n d Verbeek, 1994). The analysis, based on current m a r k e t as well as long term prices, showed that Pu recycling in LWR's is economically justified. It should be noted t h a t this conclusion refers to LWR cores of a s t a n d a r d design, i.e. with M O X fuel assemblies constituting a b o u t onethird of the whole fuel load. Utilization of existing LWR plants for an effective incineration of excess Pu m e a n s t h a t one of the main design objectives of LWR cores, namely the maximizing o f the fuel utilization parameter, is replaced by an objective to maximize the destruction of fissile Pu per unit energy produced. In order to m a i n t a i n the economic viability of the Pu incineration o p t i o n (Barbrault, 1994) it should be carried out by the existing power plants without 507

508

Alex Galperin Table 1. Basic core and assembly parameters

Design heat output, MW (t) Average power density, MW/mt (HM) Total U (HM), metric ton (mt) Core active height, cm Total number of assemblies in core No. of fuel rods per assembly Fuel rod outside diameter, cm Pellet diameter, cm Clad thickness, cm Fuel rod pitch, cm Fuel assembly pitch spacing, cm In-core fuel management scheme Cycle length, full power days U density in UO2,g/cm~ Pu density in PuO2,g/cm3 Th density in ThO2,g/cm3

3360 38.4 88.0 365.00 193 264 0.950 0.820 0.057 1.260 21.5 3-batch 330 9.0 9.7 8.26

additional investment for plant modifications. Design variations are reduced to fuel cycle optimization, i.e. fuel composition may be varied to achieve optimal Pu destruction. In this work different fuel compositions were considered and corresponding fuel cycles were analysed. Material balance for each of the fuel mixtures was calculated in order to estimate the potential of Pu destruction. Basic parameters of the lattice, such as Doppler and void coefficients, soluble boron and Xe reactivity worth, were estimated in order to evaluate the performance of the core under consideration.

Table 2. Fuel composition summary

Fuel type U MOX MOX-239 TMOX TMOX-239

U-238 w/o

U-235 w/o

96.80 93.34 93.32

3.20 0.66t 0.68t --

Total Pu w/o

Th-232 w/o

6.0 2.5 7.0 3.7

-93.0 96.3

t U-235 part of natural uranium.

values. The considered fuel mixtures are presented in Table 2. Two different Pu compositions were considered: one of the Pu components of a fuel discharged from a PWR core and another representing weapon grade Pu. The PWR discharge isotopic composition of Pu was adopted for defining fuel composition in U-Pu and Th-Pu cases. The relative isotope content of the reload fuel was 0.55/0.25/0.14/0.06 of Pu-239, Pu-240, Pu-241 and Pu-242 respectively, representing the Pu isotopic composition of the fuel discharged from a typical PWR (about 40 GWd/t) and following 3 5 years of "cooling" period. This Pu vector was used for MOX and TMOX fuel types, with different fertile components: natural uranium for the MOX fuel and thorium for the TMOX fuel. Two additional fuel types MOX-239 and TMOX-239 were analysed to investigate the potential to incinerate weapon grade Pu, in this case approximated by pure Pu-239.

Pu BASED FUEL CYCLES

Fuel mixtures considered in this work were based either on uranium (U) or thorium (Th) fertile materials and Pu as a fissile component. The slightly enriched U fuel cycle for a typical pressurized water reactor (PWR) was considered as a reference case. At this stage the depth of a reactor core analysis was limited to lattice level only, which is assumed adequate for the comparison of the fuel cycle basic characteristics and an estimate of the material balances. A lattice of a typical 4-loop PWR plant was chosen as a platform for comparing performance of different fuel types. Main core and lattice parameters are summarized in Table 1. Five fuel types were considered, analysed and discussed in this work. The reference 3-batch cycle with 330 full power days per cycle was adopted for all cases, simulating an annual refuelling and resulting in an accumulated burnup value of the discharged fuel of about 38 GWd/t. The Pu content of all fuels was adjusted to assure the identical cycle length and discharged burnup

CALCULAT1ONAL M E T H O D S

At this stage the calculational methods were limited to lattice calculations only. The fuel assembly was modelled in I-D geometry by the cluster option of the WIMS-D code (Askew et al., 1966). The basic assumption adopted in this work is that detailed lattice calculations are adequate to predict the fuel composition evolution with fuel burnup, as well as main lattice parameters, such as reactivity coefficients and reactivity worth of control absorbers. Core parameters and fuel cycle data were estimated from the results of the WlMS burnup calculations by using the linear reactivity model and geometrical buckling. Assuming a 3-batch fuel management scheme and the linear reactivity model, end-ofequilibrium-cycle criticality values for each of the five fuel types were adjusted to produce identical cycle length. This was achieved by varying the fissile content of the fuel.

Pu incineration in LWR's

509

1.4 U 1.3 1.2

, ~ . _ . . . . ~ :. . . . .~. . . . . . ~ ~

t=

~

~

....................

MOX

..........

TMOX

........

MOX-239

- ....

TMOX-239

1.1 °~

-"."'-":"::"-':.:......:: .............

,....=.,.. ,.,.....,.~...........

1.0

0.9 0

10000

20000

30000

50000

40000

(MWd/t) Fig. 1. Criticality rundown curves. Accumulated

Burnup

Table 3. BOL fuelcompositionsummarytotal weightof an isotopein core (kg) Fuel type

U-235

U MOX MOX-239 TMOX TMOX-239

2830.5 590.6 609.8

Pu-239

Pu-240

Pu-241

Pu-242

3174.8 2378.6 3705.6 3528.1

1346.6

688.0

511.I

1562.8

796.2

599.6

RESULTSOF CALCULATIONSAND DISCUSSION An equilibrium cycle was simulated by performing cluster burnup calculations. The lattice K-infinity values are presented in Fig. 1, showing criticality rundown curves for all fuel types. A pronounced difference in the criticality behaviour of the fuel types with a single fissile isotope (U-235 in U fuel and Pu239 in MOX-239 and TMOX-239 fuels) is clearly demonstrated. The criticality swing of the MOX and TMOX fuels is significantly smaller than that of the U, MOX-239 and TMOX-239 fuels. This effect may be attributed to an additional fertile-to-fissile conversion process in a fuel containing Pu-240 isotope. The enhanced conversion results in a reduced burnup related criticality swing. Lower value of K-infinity at beginning-of-fuellife (BOL) for the MOX and TMOX fuels may be also attributed to the presence of the Pu240 isotope with very large absorption cross-section. The reduced burnup related criticality swing of the fuels based on reactor grade Pu demonstrated in Fig. 1, will reduce significantly the reactivity control requirements of the core. The material composition data for the whole core was estimated based on the core, fuel and cycle parameters presented in Tables 1 and 2. A summary of the total weight for all main actinide isotopes at BOL is shown in Table 3. A

U-233

corresponding fuel composition summary for the endof-life (EOL) at 38 GWd/t is derived from the WIMS burnup calculations and is shown in Table 4. The data given in Tables 3 and 4 may serve as a basis for a material balances table, which summarises production and destruction of the main fissile isotopes in each of the fuel types considered here. These balances are presented on an annual basis by assuming that each annual fuel reload constitutes one-third of the total core inventory. The summary of material balances is presented in Table 5, including U-233 fissile isotope for the Th based fuels. The potential of a given fuel type to incinerate excess Pu may be deduced directly from Table 5. The annual production of Pu of a standard PWR with 1100 MWe output is about 298 kg. The same core completely loaded with the MOX fuel is estimated to consume 474 kg of Pu, mainly fissile isotopes. The MOX-239 fuel type shows a potential to reduce the initial total Pu inventory by 220kg/year and fissile Pu inventory by 420 kg/year. The following two fuel types: TMOX and TMOX239 are based on Th-232 as a fertile component of the fuel. The amount of Pu destroyed per year for both cases is significantly higher than that of U-based fuels. Especially impressive is the reduction in

510

Alex Galperin Table 4. EOL (38 GWd/t) fuel composition summary total weight of an isotope in core (kg) Fuel type

U-235

Pu-239

Pu-240

Pu-241

Pu-242

U-233

U MOX MOX-239 TMOX TMOX-239

690.4 341.3 241.6

485.6 1794.6 825.6 963.2 345.0

213.3 1209.0 49 I.5 1238.5 392.2

130.7 786.3 249.9 806.0 272.3

63.9 51 I.l 108.1 589.7 97.9

1003.6 1022.2

--

Table 5. Pu and U-233 annual balance summary weight of an isotope in a single batch (kg) Total Pu

Fissile Pu U-233

Fuel type

Load

U MOX MOX-239 TMOX TMOX-239

0 1908 793 2221 1176

Discharge 298 1434 573 1199 369

Balance

Load

+ 298 -- 474 - 220 - 1022 - 807

0 1288 793 1501 1176

fissile Pu inventory: more t h a n 900kg/year. This p h e n o m e n o n m a y be attributed to the fact t h a t thermal a b s o r p t i o n cross-section of T h (7.4b) is m u c h higher t h a n that of U-238 (2.7b). This difference results in a higher initial fissile inventory of Th-based fuels, lower conversion ration a n d consequently in a higher fissile material c o n s u m p t i o n per unit energy produced. This conclusion holds even if U-233 buildup is taken into account in the overall fissile material balance. The U-233 buildup rate in T h is lower t h a n t h a t of Pu in natural U, which results in a m o u n t s of U-233 produced annually m u c h smaller t h a n a m o u n t s of Pu produced in a s t a n d a r d P W R core. Different fuel compositions considered in this work result in a different n e u t r o n spectrum of the reactor core a n d this in turn influences reactivity coefficients. Controllability o f a core based on Pu fuel is more difficult t h a n that o f a core based on U fuel (Paratte a n d Chawla, 1994). A preliminary evaluation of several reactivity related core p a r a m e t e r s were carried out in infinite lattice approximation. The results are given a n d discussed below. EQUILIBRIUM Xe REACTIVITY SWING AND SOLUBLE BORON WORTH

The reactivity swing due to Xe buildup ( A K ~ ) is an i m p o r t a n t c o n t r i b u t i o n to the overall reactivity control requirements a n d is strongly d e p e n d e n t on

Discharge 205 860 373 590 206

Balance

Balance

+ 205 428 -420 - 91 l - 970

0 0 0 + 112 + 114

n e u t r o n spectrum. The AKoo was evaluated for all fuel types considered a n d is shown in Table 6. It is shown t h a t the reactivity w o r t h o f equilibrium Xe is reduced significantly due to the presence o f Pu as well as T h isotopes. This effect m a y be attributed to higher thermal cross-sections o f T h as c o m p a r e d with U a n d o f Pu (especially Pu-240) as c o m p a r e d with U-235. Thus, the Xe c o m p o n e n t of reactivity control requirements is reduced by 1.5 2.0%. Similar effect m a y be n o t e d for the reactivity worth o f soluble b o r o n . The higher thermal cross-sections o f Pu isotopes cause partial shielding o f b o r o n a n d thus reduction o f its a b s o r p t i o n rate a n d subsequently reactivity worth. It should be n o t e d t h a t at this stage the control rod reactivity w o r t h was n o t calculated, but it is reasonable to assume t h a t a similar reduction of the reactivity w o r t h m a y be expected (see for example Paratte a n d Chawla, 1994).

REACTIVITY COEFFICIENTS Two main reactivity coefficients were calculated for all fuel types: D o p p l e r coefficient a n d void coefficient. The D o p p l e r coefficient was derived from composition o f two lattice calculations at BOL with fuel temperatures of 726 a n d 1000°C a n d void coefficient was o b t a i n e d by c o m p a r i n g the lattice K-infinity values with 0 a n d 10% void values, neglecting 3-D leakage a n d spectrum spatial variation effects. The

511

Pu incineration in LWR's Table 6. Reactivitycoefficientsand absorber worth

Fuel type u MOX MOX-239 TMOX TMOX-239

Equilibrium Xe

Soluble boron (I 000 pm) (AK)IK, %

Void coefficient ((~ 10 void) (AK)IK 10 2/% void

(AK)IK, %

Doppler coefficient (AK)/K I0 ~/°C

3.5 1.8 2.7 1.6 2.2

0.99 0.28 0.53 0.31 0.44

-0.16 - 0.22 -0.19 0.20 - 0.16

-2.8 - 3.5 2.9 3.8 - 3.2

estimates of the reactivity coefficients are summarised in Table 6. The Doppler coefficients of the M O X and T M O X fuel types are somewhat higher, i.e. more negative, than those of other fuel types, which may be attributed mainly to the presence of Pu-240 isotope. Similar effects may be noted for the void coefficient values: more negative values for the Pu based fuels in general and M O X and T M O X types in particular. This effect may be explained by a more pronounced effect of the spectrum hardening of the Pu based fuels in comparison with U based fuel, due to higher thermal cross-sections of Pu. CONCLUSIONS The main finding of this work was an increased potential of thorium-based fuel types to reduce initial amount of Pu. This conclusion is derived directly by considering material balances of the core (see Table 5). It was demonstrated that Th based core requires higher initial fissile loading, which leads to lower conversion ration and finally to a higher fissile isotopes (Pu) destruction per unit energy produce. Two additional conclusions may be derived from the results of the reactivity related parameters of the reactor lattices representing all considered fuel types. First: the safety related reactivity coefficients are

negative, which indicates that the basic behaviour of a reactor core utilizing Pu and P u - T h based fuel types will be quite similar to that of a standard PWR core utilizing slightly enriched U fuel. Second: the reactivity control of a core based on Pu fuel as a fissile component will be more difficult due to a reduced reactivity worth of the soluble boron and control rod control mechanisms. A potential solution of this problem may be found in utilization of enriched (in B-10 content) boron. The preliminary results and discussion presented above indicate a potential of utilizing P W R power plants of current technology for an effective incineration of Pu, weapon grade as well as discharged from reactor. It should be noted, however, that the final conclusion on a feasibility of the proposed fuel types may be obtained following detailed 3-D full core analysis, selected transient analysis and fuel cycle cost estimates. REFERENCES

Askew J. R., Fayers F. J. and Kemshell P. B. (1966) J. Bn Nucl. Energy Soc. 5, 564. Barbrault E (1994) Nucl. Europe Worldscan 3--4, 70. Goldschmidt P. and Verbeek E (1994) Nucl. Europe Worldscan 5-.6, 49. Paratte J. M. and Chawla R. (1994) "On the Physics Feasibility of LWR Plutonium Fuels Without Uranium" Ito be published).