XA0056534

IAEA-SM-358/30

ADVANCED MIXED FUEL ASSEMBLIES WITH HIGHER CONTENT FOR LIGHT REACTORS

W. STACH Siemens AG, Untemehmensbereich KWU, Erlangen, Germany

Abstract

The MOX introduction in LWRs (PWR and BWR) was started in Germany with initial steps of design in the early 70ies. The process of commercial utilisation of Pu recycling was based on these designs and initial MOX insertion at the Obrigheim-plant KWO (PWR) and the Gundremmingen-plant Unit A KRB-A (BWR). The needs of testing and validation of used methods could mainly be fulfilled by the insertion of initial MOX- FA reloads. Experiments with this early MOX-FAs have been conducted under fully realistic power reactor conditions for neutronic/nuclear and fuel/technological aspects. Optimising fuel cycle costs by increasing the final burnup to reduced generation of plutonium. Under properly defined boundary conditions thermal recycling in MOX-FAs reduces further the amount of Pu which has to be disposed of to final storage. Increasing the final burnup requires higher initial enrichments of U-fuel to be matched by an advanced design of MOX-FAs with higher Pu contents. The neutronic design of these MOX-FAs has to consider the licensing status of NPPs concerning the use of MOX fuel and the evolution of U fuel enrichment and burnup level. The Siemens Cycle Division, with more than 20 year's experience in the production of MOX fuel, has designed several advanced MOX FAs of different types for PWRs (14x14 to 18x18) as well as for BWRs (9x9 and 10x10) with averaged contents of fissile plutonium up to 5.85 w/o. Some reloads of this kind are at present under irradiation in different NPPs.

1. INTRODUCTION

For several years a shift towards higher burnup has taken place in all countries using light- water reactors (LWRs) in the effort to minimise amounts of , especially to reduce the number of fuel assemblies which have to be disposed of, to save resources of by reducing the numbers of new fuel assemblies, and to minimise fuel cycle costs to improve the economics of generation. This trend towards higher burnup is accompanied by the thermal recycling of plutonium from reprocessed fuel. Several countries including Belgium, , Germany, Japan and Switzerland are already using MOX (Mixed-Oxide) FAs (Fuel Assemblies) in LWRs or intend to do so [1]. For more than two decades, Germany has been gaining practical experience in the thermal recycling in both boiling water reactors (BWRs) as well as especially in pressurised water reactors (PWRs) [2], This was in accordance with the national Atomic Act. Until mid 1994, this was the only established method of nuclear waste disposal. The experience of SIEMENS (KWU) covers more than 20 years of MOX FA design ( physical and mechanical), production and insertion in BWRs as well as especially in PWRs, as can be seen in the Fig. 1. Thermal recycling of plutonium started under commercial conditions at Obrigheim (KWO) in 1972 (PWR) and at Gundremmingen Nuclear Power Plant Unit A (KRB-A) in 1974 (BWR). After the introduction of the improved production methods OCOM (Optimised Co- Milling) and A (U, Pu) C (Amrnonium-Uranyl-, Plutonyl-Carbonate) in 1980 MOX FAs have been used in other plants such as Neckarwestheim Unit 1 (GKN I) since 1982 and Unterweser (KKU) since 1984 [3], [4]. Up to 30 000 MOX fuel rods have been inserted per year in these reactors, some of them spending six irradiation periods in the reactor core.

389 As the national German fast breeder program had been cancelled, thermal recycling was the only way of using large quantities of plutonium for the energy production and helps to avoid accumulating separated Pu and the final disposal of plutonium.

70000y

SIEMENS M O X i MOX M.SBR 65000- () MOX ex. OCOM and AUPuC MOX ex. MIMAS 6* Irradiation Period I (Hanau) 60000- (Do99el, Cadaroche) th 55000- 5 Irradiation Period i.

1 50 MO 4" Irradiation Period o o 45000- 3rd Irradiation Period

40000- 2nd Irradiation Period X 35000. a Reload 30000-

FRAGEMA MOX 25000- | j 4th Irradiation Period 30000- RS HI [__j 3rt Irradiation Period 15000- Q] 2-™ Irradiation Period 10000- | Reload 5000.

PWR: KWO GKN i, GKN II BZN-1.B2N-2 KKU KKG KKP-2 KWG KBR KKGg KKI2

KRB-A GUN-C GUN-B

FIG. 1. Core Management Experience with commercial MOX Insertion in B WR andP WR by Siemens KWU Group (Status End of 1998)

The basic idea of thermal recycling is to insert MOX FA into normal LWR cores. As it is common practice to use MOX FAs together with U FAs in the same core it has to be the principal aim of nuclear MOX FA design to make them as compatible as possible with U FAs, so that they can be used instead of U FAs without any restrictions. Changes in cycle length are to be avoided. In order to achieve this, the fissile material content has to be dimensioned to obtain MOX FAs, which have the same burnup potential as the U FAs (burnup equivalence). Fuel management studies must be carried out to confirm that equilibrium cycle lengths of cores with MOX FAs are the same as those of cores without MOX FAs.

2. PLUTONIUM BALANCE

The debate and the public discussions on some aspects of the thermal plutonium recycling are often confused, above all with respect to the plutonium balance. The essential issue is the question, whether thermal recycling actually increases or decreases the amount of plutonium, which has to be handled. Especially uncertain definitions and general conditions within the complex system of the fuel cycle mostly cause this confusion. Quantitative statements can only be made with scenarios having properly defined boundary conditions.

390 In the following scenario, a PWR with 1300 MW electrical energy generation (1300 MWe) operating 320 equivalent full power days (efpd) per year is considered and discussed. The annual production of electric power is set to 10 billion kWh, which is a typical value for German nuclear power plants of the 1300 MWe class. Nuclear calculations to analyse the plutonium generation and to obtain a plutonium balance are done using the zero-dimensional burnup code KORIGEN [5]. In these calculations the cross-section libraries are based on the JEF nuclear data generated by the FZ Karlsruhe (former: Kern-Forschungszentrum-Karlsruhe). Some input parameters such as burnup dependent cross sections of and spectral parameters were calculated with the Siemens (KWU) standard design program system SAV90 [6] including the spectral code FASER. In general, increased burnup can be achieved by increasing the initial enrichment. If the cycle length remains unchanged, the number of new (fresh) fuel assemblies which have to be reloaded every year has to be reduced. In cores without MOX FAs, for example, the reload batch of 64 FAs (initial enrichment of 3.1 w/o 235U) can be reduced to 48 FAs with an initial enrichment of 3.7 W/O235TJ without any influence on the cycle length. Due to longer irradiation in the core, the total amount of plutonium generated in any FA increases. However, in terms of the amount of plutonium unloaded every year in the spent fuel assemblies of one batch, the smaller batch size results in a reduced final quality and quantity of plutonium. A scenario without thermal recycling is listed in the upper part of Table I. Increasing the burnup from 35 MWd/kg to 45 MWd/kg reduces the amount of unloaded fissile plutonium from 205 kg to 165 kg (- 20%) per year; the total amount of plutonium is reduced from 312 kg to 264 kg (-15%) per year. Thermal recycling can further reduce the amount of plutonium generated per year, as is apparent from Table I. The plutonium generated in 54 U FAs is just enough to fabricate 10 MOX FAs with a burnup

equivalent content of fissile plutonium Pufjss = (239pu + 241pu) Natural Unat is used as carrier material for the MOX FAs. The isotopic composition of the plutonium in the MOX FA

corresponds to the plutonium produced in an U FA with an initial enrichment of 3.1 W/O235TJ ancj a burnup of 35 MWd/kgHM considering a decay of241Pu of seven years between discharge of U FA and MOX FA reloading. The quality of the used plutonium, defined as the sum of the mass fractions

of the fissile plutonium 239Pu and ^41pu relative to the sum over all Pu isotopes, is 66 %. It is assumed that only spent U FAs are reprocessed and that burnt MOX FAs are consigned to the final storage. As can be seen from the Table I, the amount of plutonium, which has to be disposed of in the

case of recycling, is reduced to 107 kg Pufiss and 195 kg Putot> respectively.

TABLE I. PLUTONIUM BALANCES FOR DIFFERENT BURNUP AND WITH/WITHOUT RECYCLING (PRODUCED BY A PWR OF 1300 MW ELECTRICAL POWER)

Pu to be disposed of to final storage Without recycling with recycling (once through cycle) (one MOX generation) Reload kg Pufiss kg Putot kg Pufiss kg Putot

64 U FAs 205 312 48 U FAs 165 264 54 U FAs 10 MOX FAs 107 195 42 U FAs 6 MOX FAs 94 172

If U FAs with 3.7 w/o 235U are used to achieve higher burnup, 42 U FAs and 6 MOX FAs have to be loaded. Again an equilibrium scenario is assumed. Owing to the higher burnup, the quality of plutonium resulting from the reprocessing of the corresponding U FAs has decreased to 62 %. As it is a basic aim of MOX FA design to produce compatible and burnup equivalent MOX FAs, the MOX FAs have to be designed for higher burnup in this case (containing the same burnup potential as the U FAs with an 235U content of 3.7 w/o) and to compensate for lower Pu quality. Also here the carrier

391 material Utails was chosen, which allows the plutonium content to be increased in the MOX FA. In this case of recycling, 94 kg Pufiss and 172 kg Putot, respectively, have to be placed in the final storage, assuming again that MOX fuel is not reprocessed. These trends of reducing amounts of Pu to be disposed of continue for higher reload enrichments and higher batch burnup values. Although only the once-trough MOX fuel cycle is discussed here, reprocessing of spent MOX FAs has been demonstrated. The plutonium of the so-called second recycling generation can be used together with plutonium from reprocessed U FAs. The worsened Pu quality (- 55 % - 60 % depending on burnup and mixing ratio between plutonium of the first and the second recycling generation) can also be compensated by increasing the absolute content of fissile Pu isotopes in the MOX FA to about 7 w/o.

3. CURRENT STATUS OF LICENSING IN GERMANY

The present status of MOX licensing for German nuclear power plants has given rise to certain differences caused by the different procedures adopted by utilities and state authorities. An overall-view of the current licensing status of MOX licenses, which are in use or have been granted for German LWRs is given in Table II. The numbers of MOX FAs per reload or the total numbers in the core given in the table are restricted only by the licenses and not by technological limitations.

TABLE II. CURRENT STATUS OF MOX LICENSING FOR LWRS IN GERMANY

Reactor type Plant-Name Status of Pufls,-Content Number of MOX-FA-Content in License in w/o MOX-FAs the Core in % per Reload PWR: KWO in use 3.8 8 26 •GKNI in use 3.04 9 GKNII in use 3.8" 37 KKU in use 3.5 » 16 33 KKG in use *' 3.07 l: 16 33 KK12 in use equivalent to 24 50 4.0 w/o 235U

KWG in use 3.2 16 33 KBR in use equivalent to . 3) _ 3) 4.0 w/o 235U KKP-2 in use 4.65 24 37 "> KKE granted *' equivalent to 16 35 4.0 w/o 235U KWBA in preparation*) equivalent to 24 42 3.5 w/o23SU KWBB in preparation*' equivalent to 24 42 3.5 w/o asTJ KMK in preparation - BWR: GUNB/C in use "! 2.57/3.6 2x64 38 KKB in preparation - KKK in Drecaration -

1) changes in the carrier material and/or Pu-quality can be compensated 2) temporary restriction 3> according to the amount of Pu-generation in the plant (up to ~ 16 MOX FAs per reload) 41 max. nominal Putiss content of a fuel rod -) modification or extension in preparation

Based on the principles established for former MOX designs of FAs of the types 14x14, 15x15, and 16x16, a "standard" MOX FA was designed (Fig. 2) for use at five 1300 MWe plants (KKU, KKG, KWG; KKP-2, and KBR) and has been in service since the mid-80s with good operating results [7]. The average fissile plutonium content is 2.91 w/o in three FR (fuel rod)- types with different fissile plutonium contents. Four additional water rods (i. e. cladding tubes filled with water in connection with the water of the primary circuit) at the centre of the FA increase the moderation there in order to flatten the power distributions.

392 •XXXX XXXX XX X XXX >< >< XX • X Xo o o o >< r>< • 1.9w/oPufiss >< o o >< >< • 2.3w/oPufiss >< oX w wX o X X X Q] 3.3 w/o Pufiss >< X oXw wX o XXo X X ox>< LJ guide tube X 0 o X X >< W water rod XXo 0 o oXX x V X XX X ms<}\XX xXXX X X X22•

FIG. 2. Standard MOXFA (16x16), carrier material Una(,2.91 w/o average Puylss content

As it has been common practice in Germany in the past to license Unat as carrier material, planned changes to depleted uranium could not be compensated under some existing licenses by increasing the fissile plutonium content. In this case burnup equivalent MOX FA design could not be realised up to now. Here new licenses are necessary, which are in preparation, in licensing procedure or already granted. Lower Pu quality has to be treated in the same way, where burnup equivalence requires compensation for higher contents of the neutron-absorbing isotopes 240pu ancj 242pu by increasing the Puflss content.

4. MOX FA DESIGN WITH RAISED PuFjSS CONTENT

The above mentioned standard MOX FA had to be redesigned for several reasons: • The enrichment of U FAs loaded together with MOX FAs is increasing. In order to obtain equivalent MOX FAs, the content of fissionable plutonium nuclides has to be increased too. Usually, this requires a change in geometrical arrangement of the fuel rods with different Puflss contents within the FA.

• Instead on Unaf, depleted uranium (Utails) is proposed as carrier material. The lower content of the fissionable 23 5u -m t^e matrix material has to be compensated by increasing the Pufiss content. In this case the available plutonium can be concentrated in fewer FAs. • Changes in the Pu quality are caused by higher burnup of the reprocessed U FAs. This requires higher Pufiss contents to compensate for the effects of the neutron absorbing plutonium isotopes 240pu and 242pUj respectively.

4.1 MOX FA designs used to date

Designs for all boundary conditions have been established. Examples are given in the following chapters. In all cases of MOX FAs for PWRs the main features developed by Siemens • maximum three different fuel rod types (Pufiss contents) • water rods in the central region of the FA are established also international as a proven design.

393 4. I. IMOXFA designs compatible to enrichments ofUFAsupto 4.0 w/o

A first design with raised plutonium content was made for a NPP with 14 x 14 rod lattice in 1987 to match the increase in enrichment of the reload U FAs to 4.0 w/o 235U. Eight MOX FAs of that design were inserted in 1988. Based on the carrier material Unat, an average content of fissile material of 3.8 w/o Pufiss is used. In this case there is no need for the use of water rods. A further design was made for a NPP with a 16 x 16 rod lattice, triggered by the change in carrier material from Unat to Utails with compensation of the lower 23 5JJ content by a higher Puflss content at the same time. The enrichment of U FAs has been remained unchanged in this case. A FA design with an averaged Puflss content of 3.48 w/o (carrier material Utails with 0.25 w/o 235U) was realised. The mentioned 14 x 14 MOX FA and the 16 x 16 MOX FA are schematically shown in Fig. 3.

14x14 MOX FA 16x16 MOX FA 3.8 w/o Piifes in UK 3.48

A A A A A A A • • • • • • A X X A • O C) • A o xox ( ) * XX X • o o X • • xo o o o x w w X • • X X • • X X • o • • X o w w o X • o • xxo oxx • XX XX A • C) o ( ) ( ) • A A • . • • • . A xo A A A A A A xxxxxxxxxxxxxx x

w/'o Pu, 2.0 w/o Pu^ 3.3 w/o Piife • • guide tube E 1 water rod • 2.8 w/o Pu^ m r~j 4.3 w/o Pufts

4.1 w/oPufc l/~Si guide tube D

FIG. 3. Different MOXFA designs for higher burnup and use ofUfans as carrier material

Further MOX FA designs have been drawn up for other PWRs on the basis of U FA enrichments up to 4.3 w/o 235U and are shown in Fig. 4. For the German 1300 MWe NPPs with 16 x 16 rod lattice, a MOX FA compatible with U FAs with an enrichment of 4.0 w/o 235U has been designed. With depleted Uranium (with 0.25 W/O235TJ content) as carrier material the averaged Puflss content of the MOX FA is 4.2 w/o. For the 1300 MWe NPPs with an 18 x 18 rod lattice, MOX FAs compatible to enrichments of U FAs of 4.0 w/o 235U have been designed. The neutronic design calculations to a burnup equivalent MOX FA design with an averaged Pufiss content of 4.6 w/o with the carrier material Utails(0.25 w/o 23

4.1.2 MOXFA designs compatible to enrichments of UFAs exceeding 4.0 w/o 235JJ

For 14 x 14 rod lattices of Westinghouse type reactors, a newly designed MOX FA using Utails with 0.25 to 0.30 w/o 235jj as carrier material and an averaged Puf}ss content of 4.75 w/o fulfils the required burnup equivalence to U FAs with 4.25 w/o235u_ Because of the instrumentation tube at the centre of the FA it is not necessary to increase the moderation by adding water rods.

394 16x16 MOX FA -18x18 MOX FA

4.2 w/o Puflss in Utai}s 4.6 w/o Puflss lr ta £> • • xEXxxxX xXXxxXXlHH I xX K• • • x IXP X X o o 1 ix • o o o oXXP x Q o X x X X o u X X o o x x o o X X o o x X u o X X o w w o x x o o x X x x ww X X x X w w X x w w x x o o X o o X o o X x o b X X o x X O o X x o o x X M X, O o X • X o o o oxP X 0 0 x • XX x'xpi X JXIBI • •• XXX X X XX X X X•!•!• • • XKX xXxxxXxxXXXUBI

[•] 2.3 w/oPuflss LJ guide tube Q 2.3 w/o Puflss Q guide tube 13 3.4 w/o Pu fwi water rod flss [J 3.3 w/o Tufas • water rod • 4.7 w/o Puflss [H 5.0 w/o Pufiss

FIG. 4. Different MOX FA designs for higher burnupfor 16x16 and 18x18 FAs

14x14 MOX FA 15x15 MOX FA

4.75 w/o Pi^ in UaiIs 4.8 w/o Piife in Uafc X xxxx • • X XXXXXXXxxx • • X • x;X Xx• o o xx o o o oxx1 X x X X o o x o o o X xxo ox X o o X X x X X o w o X ox x X X X o o X x o O o X o o o X. X X X xx Q o o 0 x X X X X X X X X X X • x x x x • • • xXXXXXxxXxx • • g i8^>"- O guide tube • 2.6 w/o LJ guide tube PI ' ^S LJ instrumentation tube [X] 3.5 w/o fwi water rod • S.

FIG. 5. MOX FA designs for higher burnupfor 14x14 and 15x15 MOX FAs

For the 15 x 15 rod lattice type a MOX FA with the up to now highest content of Puflss has been designed. The requirement of burnup equivalence to U FAs with 4.3 w/o235]j can t,e fulfilled with an average content of 4.8 w/o Puf}ss, using Utails with 0.25 w/o 235U as carrier material. In the 15 x 15 rod lattice FA one central water rod is sufficient to flatten the radial power distribution. The MOX FA designs mentioned above are shown in the Fig. 5.

395 4.1.3 MOX FA designs for B WRs

For thermal recycling of plutonium in BWRs MOX FAs of the 9x9 and 10 x 10 rod lattice type have been designed. The MOX FA for the insertion in a BWR is in general much more complicated because of the much higher heterogeneity in comparison to PWR. The 9x9-1 BWR MOX FA contains 6 different MOX fuel rod types and 1 additional Gd poisoned U fuel rod type to avoid power peaks around the water channel and to reduce the initial reactivity. The averaged Pufiss content is about 3 w/o with carrier material Utails (0.25 w/o 235U). As an example for the progress in design of BWR MOX FAs a design for a 10 x 10 rod lattice type (ATRIUM 10TM) has been performed. The BWR 9x9 BWR MOX FA and the ATRIUM 10™ MOX FA are shown in the Fig. 6.

9x9-1 MOX FA ATRIUM 10™ MOX 3.Ow/oPufissinUlails 3.7w/oPufasinUails position of , position of control rod X X • • • • • A A A G A G A • 7 • • A G • A A • • G A • G G G • • • G / A G W G A i water • • • channel A 3x3 FR / G G G • • • A A a • • • • A A A G A G A G A • • • • A G X X G / • • X • • • • • • • i§ A i s 1.15 fc A 4.30 2.5w/o" U A 4.70 w/o Pufe c 1 c 1 4.70 w/o Pu, 1.85 w/oPiifo 1.70 w/o Pufe 5.52 w/oPife JL partial length X 5 3.95 w/o " U 2.55 w/o Puflss 5.52 w/o Puflss

2.30 w/o Pufo + 1.25 w/oGdjO, 5.52 w/o Pufe 3.40 w/oPuriss partial length •v/p. 235 2.60 w/o Pufe water channel 3.95 w/o U c 1 + 1.25 wioGdj03

FIG. 6. Different MOX FA designs for B WRs (9x9 and 10x10 rod lattice type)

5. CORE PERFORMANCE WITH MOX FA IN PWR

An equilibrium core based on full low leakage loading with 88 MOX FAs (45 % of the core) of a 1300 MWe NPP with 16 x 16 rod lattice type has been investigated in a licensing study. The equilibrium core has a reload batch of 20 MOX FAs and 24 U FAs (majority is Gd poisoned). Based on this study a license was granted for reload of MOX FAs with 4.2 w/o Puflss in Utails of Fig. 4. The important cycle characteristics are listed in Table III. The coolant or moderator temperature coefficient (MTC) tends to more negative vaiues for increasing piutonium content in the core. The Doppler coefficient is hardly influenced by plutonium. This is of importance with respect to the shutdown margin.

The boron worth decreases with increasing number of MOX FAs in the core. The boron control system must therefore handle larger concentration differences during reactor operation. The critical boron concentration must be raised during loading operations to keep the reactor at a required level of subcriticality. The use of boric acid with enriched ^B can increase the effective boron capacity.

396 TABLE in. U-MOX EQUILIBRIUM CYCLE (PWR WITH 1300 MW ELECTRICAL POWER)

MOX FA loading number / % of the reload batch (core) 20/45 Effect compared with number of reload MOX and U FAs 20 / 24 Uranium core

MOX FA type 16x 16 Pufiss content in w/o 4.2 235U content of carrier material 0.25

H5U content of U FAs in w/o 4.0

cycle length in efpd 315 same

MOX FA bumup in MWd/kgHM averaged discharge burnup 54 about same maximum FA burnup 60

initial boron concentration 1775 lower (without Xenon) in ppm

boron worth at BOC -5.5 lower in pcmlppm (U core ~ 8.3)

-73 higher MTC at EOC in pcm/K (U core ~ 55)

5.6 lower or same Net control rod worth at EOC in % Ap

As regards the net control rod worth for stuck-rod configurations, the data depend more on the loading scheme than on the MOX FA fraction in the core. Thus MOX fractions of up to 50% without need for increasing the number and the effectivity of control rod system were found possible. The primary operating results include information on cycle length, power distribution, reactivity coefficients, and control rod worth of cores containing MOX FAs. The reliability of the design methods is validated by measurements of these quantities. The neutron physics experience [7] is based on start-up measurements, in-service cycle monitoring and specific measurements required under licensing commitments.

6. CONCLUSIONS AND OUTLOOK

It is shown that, under properly defined circumstances, increasing of initial enrichment and final burnup combined with thermal recycling lead to decreasing amounts of plutonium to be disposed of. MOX FAs can be adapted without problems to be compatible with the design and enrichment of new U FAs, actual plutonium qualities and different carrier materials. MOX FAs inserted together with U FAs of an initial uranium enrichment exceeding 4.3 w/o 235U can be designed, as was demonstrated in design studies for a FA with 14 x 14 rod lattice type. With averaged Pufiss content of 5.84 w/o with the carrier material U^iis, the burnup equivalence (in the manner mentioned above) to U FAs with about 4.6 w/o 2j5U is achieved. First drafts of a 17 x 17 MOX FA for the European Pressurised Water Reactor (EPR) result in a comparable content of fissile plutonium of « 6.5 w/o.

397 REFERENCES "

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398