Separations for the Nuclear Fuel Cycle in the 21st Century - American

Chapter 6

Advanced PUREX Flowsheets for Future Np and Pu Fuel Cycle Demands

Downloaded by UNIV OF ARIZONA on August 9, 2012 | http://pubs.acs.org Publication Date: June 9, 2006 | doi: 10.1021/bk-2006-0933.ch006

O. D. Fox, C. J. Jones, J. E. Birkett, M. J. Carrott, G. Crooks, C. J. Maher, C. V. Roube, and R. J. Taylor Nuclear Sciences and Technology Services, B229 and BNFL Technology Centre, Sellafield CA20 1PG, United Kingdom

The reprocessing of nuclear fuel using PUREX-based solvent extraction has been undertaken in the U K on an industrial scale for over forty years. Current research includes the development of modified PUREX-style flowsheets for processing ever-increasing plutonium content fuels including M O X , fast reactor and even experimental and legacy fuels. Herein, results are presented on Np recovery in single cycle P U R E X flowsheets using hydroxamic acids. Flowsheets have been demonstrated on a centrifugal contactor rig. Aspects of Np chemistry relevant to the use of hydroxamic acids in process flowsheets for U/Pu(&Np) separation are reported.

Introduction The U K has undertaken Purex nuclear fuel reprocessing both in industrial deployment and research and development for over forty years. Spent fuel from the U K ' s first generation of reactors, Magnox, are reprocessed at the Magnox reprocessing plant and those from the more modem oxide fuel reactors (e.g. AGRs, PWRs and BWRs) at the Thermal Oxide Reprocessing Plant (Thorp); both plants are located at the Sellafield site in Cumbria, U K . 1

© 2006 American Chemical Society

In Separations for the Nuclear Fuel Cycle in the 21st Century; Lumetta, G., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 2006.

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Advances in Purex technology have reflected fuel cycle evolution: as evidenced through the development of flowsheets and equipment to reprocess both metal and oxide thermal reactor fuels. Research and development is continuing at B N F L in the area of aqueous (or solvent extraction) reprocessing to meet the anticipated future demands of the nuclear fuel cycle. This paper discusses how nuclear fuel cycles might develop in the future, how Purex technology may develop to respond to the challenges of advanced fuel cycle strategies and presents some current R & D work being undertaken within B N F L towards these ends. Developments in Purex technology are subject to a number of drivers to which different countries may ascribe different priorities depending on their nuclear energy strategy. At the more focussed level of developing future reprocessing plants three primary drivers of any development programme may be seen as> 1. Reductions in reprocessing costs. 2. Reduction in effluent volumes. 3. Minimisation of waste production. 1

However, at a more strategic level other issues become important:• Reducing the radiotoxicity of high-level waste. • Reducing or eliminating uncertainties about the long-term performance of waste repositories. • Utilizing the energy potential in the fissionable isotopes present in spent fuel. • Increasing the proliferation resistance of wastes and products. Drivers 1 to 3 are not necessarily independent since reductions in effluent and waste production may, of themselves, reduce costs, but this may be at the expense of process changes in fuel preparation and solvent extraction which may increase costs elsewhere in the cycle. A reduction in cost may be achievable by the use of intensified solvent extraction equipment. For example, there is current worldwide interest in the nuclear industry in the use of centrifugal contactors for liquid-liquid extraction in P U R E X and waste partitioning processes. Besides B N F L , flowsheets utilizing centrifugal contactor equipment for solvent extraction are being developed by, amongst others, American, European and Japanese groups. Centrifugal contactors offer a number of advantages over the pulsed-columns and mixersettlers employed on current reprocessing plants. These include; reduced risk of criticality due to design geometry; variable flowrates allowing high through-put 2

3

4


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of liquids and intensification of the process; compact equipment which markedly reduces capital costs; easy and rapid start-up and shut down procedures; short contact times with active feed reducing both radiolytic damage and hydrolysis of the TBP/hydrocarbon-diluent solvent. Set against the process cost drivers, there may be strategic benefits from improved partitioning of actinides and Fission Products (FP) which may justify increases in reprocessing costs, i f necessary. Elimination of 99.9% of the actinides can reduce the radiotoxicity of High Level Waste (HLW) from reprocessing by a factor of up to 100,000 compared to the direct disposal of spent fuel. Furthermore, the radiotoxicity of H L W is dominated by Fission Products (FP) for the first few hundred years but thereafter it is the presence of the TRans Uranium elements (TRU) which becomes the dominant contributor. Thus removing T R U from H L W can substantially reduce the time taken for it to decay to the radiotoxicity level of the natural uranium ore from which it originated. A target of less than 10 years, compared to ca. 10 years without T R U removal, seems possible and would simplify waste repository design. Retaining neptunium, americium and curium in recycled plutonium makes it substantially more radioactive and more proliferation resistant; but places much higher demands on reprocessing and fuel fabrication technology. 3

6

In order to derive significant benefits from segregating the T R U from H L W it is necessary to recycle and ultimately destroy the T R U in the nuclear fuel cycle through Partitioning and Transmutation (P&T). P & T offers the prospect of recovering energy from fissionable nuclides and consuming fissile material which might potentially be used in the production of fission weapons. P & T involves separating the FP from the actinides (Partition) so that the latter can be returned to a reactor for further irradiation (Transmutation), The recycled actinides may be removed by 'nuclear incineration' through fission or they may be converted to other fissile or short lived isotopes, these processes collectively being known as transmutation. The transmutation of actinides is more efficient i f a fast neutron energy spectrum is used and so requires the use of Fast Reactors (FR) or Accelerator Driven Systems (ADS). To fully close the fuel cycle with respect to T R U will require substantial investments in research and development to become viable. However, transmutation accompanied by multiple fuel recycle has the potential to reduce, by a factor exceeding 10 , the time taken for the H L W to decay to the activity level of the natural uranium from which it originated. Such a strategy helps to deal with objections about the long-term problem of nuclear wastes and about nuclear proliferation. If a transmutation strategy were adopted fuel reprocessing 2

2


92 would need to tolerate much higher radiation and heat outputs from spent fuel placing much greater demands on Purex technology and in the longterm requiring pyrochemical or other non-aqueous reprocessing techniques for multiply recycled fuel. In the nearer term, work on aqueous reprocessing flowsheets needs to address the problem of the routing of transuranium elements in the Purex process, in particular removing them from the highly active waste. 4


Control of Np in Advanced Fuel Cycles The first contactor in the P U R E X process extracts and separates the U and Pu from the majority of highly active fission products and dissolved fuel cladding. In acidic solutions, Np exists largely as Np(IV), [ N p 0 ] and [ N p 0 ] . Pentavalent Np is not significantly extracted into the TBP/alkane solvent; however, the tetra and hexa-valent Np ions are complexed by T B P and nitrate ions (see [ 1 ] and [ 2 ]) and extract into the U/Pu-loaded organic phase. (v)

+

2

(VI)

2+

2

Np(IV) + 4 N 0 " + 2TBP o N p ( N 0 ) ( T B P ) 0 + 2 N 0 " + 2TBP o N p 0 ( N 0 ) ( T B P ) 3

Np

( V I )

2 +

2

3

4

2

(VI)

3

2

3

2

2

[1] [2]

In the next contactor the U/Pu split is achieved by introducing a solution of reducing agent, commonly U(IV), to convert the Pu present to inextractable Pu(III), which is then rejected to the aqueous phase. However, Np is reduced by U(IV) to Np(IV) and any Np in the solvent will, therefore, accompany the uranium stream. A consequence of this is the requirement to operate a Uranium Purification (UP) cycle to remove Np. Therefore if Np can be diverted from the U before the U P contactor - and sufficient DFs for Pu and other TRUs and FPs are obtained in prior contactors - then the U P cycle and all its plant footprint, operation, wastes and costs can be eliminated. A number of options for Np routing have been considered, see Figure 1: • Route Np with FPs and M A s in first extraction contactor; and thereby send Np with the H L W and eventual vitrification • Remove Np immediately after U/Pu split contactor; that is, include a specific Np rejection contactor • Remove Np in U/Pu split contactor i.e. co-recover Np/Pu; this removes the need for a separate Np rejection contactor. For the first option Np would be directed with FPs and M A s and depending on fuel cycle scenario - either be vitrified (most likely) or recovered with the M A s in a second Highly Active (HA) extraction cycle.



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Figure 1. Options for Np routing in a PUREX-style flowsheet including 1) directing with FPs and MAs in HA contactor, 2) production of a Np-only product by use of a specific Np rejection contactor and 3) co-recovery with Pu in the Pu rejection contactor

For the second option a specific Np rejection contactor is required, which is absent for options one and three. Separation of Np after the U/Pu split would keep the Pu product free of Np (which is a burnable neutron poison) and this would be the preferred option for Pu-only recycle fuel cycles. However, this generates a pure neptunium stream and there appears to be little current or anticipated demand for such a product. The third option, which removes the need for a Np rejection contactor by producing a co-recovered Np/Pu product, is potentially the most convenient and justifiable option. Neptunium is a burnable poison (in thermal reactors) and could theoretically be included in a plutonium-based oxide fuel with adjustment to reactor operating conditions. In addition, recent concerns over nuclear weapons proliferation associated with the advancement of Pu-based fuels may be tempered by the manufacture of a Np-contaminated Pu product. Finally, Np is fissionable in fast reactors and therefore the fuel cycle can be closed with respect to Np in a P & T scenario. This Minor Actinide burning is also consistent with reducing the long-term radiotoxicity of the T R U elements in waste.


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Use of Hydroxamic Acids in Advanced Fuel Cycles for Np and Pu control

Np rejection - routing option 2, a Np-only product


Routing option 2 requires a hydrophilic ligand that selectively complexes Np(IV) over U(VI) at very low p H values and is reasonably stable to both acid and radiolytic degradation for the time scale of the separation process. B N F L identified hydroxamic acids (1) as suitable candidates for stripping tetravalent actinide cations into the aqueous phase while leaving hexavalent actinides in the T B P / O K phase. Two hydrophilic hydroxamic acids, formo- and aceto-hydroxamic acid, (FHA & A H A ) are considered suitable for this process as they form hydrophilic actinide complexes. 5

OH

H

R

/*=< H f

R = H, FHA R = Me, AHA

Work on the complexation of Np(IV) by F H A and A H A confirmed the suitability of this ligand for Np(IV) stripping and a flowsheet test of a N p rejection trial using A H A to selectively strip Np from a U-loaded solvent stream was undertaken at Sellafied. From the final Np concentrations in the exiting organic and aqueous solutions a decontamination factor (Np D F to uranium solvent product) of 270 was obtained. 5

Although successful, a number of problems were identified with the operation of the experimental rig and the analytical methods employed for data collection. The final value of the D F was limited either due to the analytical limit of neptunium detection and/or rig contamination. This problem arises from the difficulty of determining the concentration of low levels of Np in U-loaded solvent. In addition, early computational models over-predicted neptunium backextraction due to over-optimistic predictions on the stage efficiencies for the 1cm miniature annular centrifugal contactors employed. A significant volume of work was undertaken to understand and address the rates of actinide mass transfer in miniature centrifugal contactors to improve process modelling of the experimental flowsheets. 6


95 Np Flowsheet Trial


A second Np rejection trial (Np02) was undertaken to optimise the flowsheet with respect to Np stripping and improve on analytical limits of detection, see Figure 2. Alterations were made to the flowsheet compared to the first Np rejection trial at Sellafield. Briefly, in order to improve the Np DF a decrease in the solvent:aqueous flowrate ratio in the Np stripping section was made, dropping from 6:1 in NpOl to 4.2:1 in Np02. Clearly, this would lead to higher U levels in the Np product, but there appears little justification for producing a Np product free of U if the Np is to be returned to fuel manufacture. A glovebox-housed centrifugal contactor rig comprising three blocks of four-stage 1 cm centrifugal contactors was used to simulate the Np rejection contactor. The U - and Np-loaded solvent feed (SP1) entered at stage 17.

Figure 2. Flowsheet testedfor Np rejection trial at Sellafield

Product streams and profile stage samples were analysed for metals (U, Np, Tc) and acidity. A range of methods were employed depending on the concentration of analyte, the concentrations of interfering species and the sample matrices present. The methods employed included: off-line spectrophotometry, titrations, inductively-coupled plasma mass spectrometry (ICP-MS), X-ray fluorescence (XRF), and both gamma and low energy photon spectroscopy (LEPS). The concentration of Np in P2 aqueous product stream was monitored spectrophotometrically and showed that the Np concentration in the P2 aqueous product reached steady state after about 50 min. The absorption spectrum of the P2 product was characteristic of a Np(IV)AHA complex (A, = 732 nm). 8

max



96 During the course of the trial four product samples were taken and analysed for U , Np and acidity. Mass balances for U , Np and acid were determined across the flowsheet, see Table I. Uranium mass balances were excellent, averaging 99% for all four product samples taken. Interestingly, although in-line spectrophotometry indicated that the Np in the aqueous product had reached steady state after less than an hour, the level of Np in the aqueous product measured by LEPS did not plateau until the final two product samples, after 2h 41 min and 3h 26min. Similarly, good Np mass balances (96%) were not achieved until the final two samples; this is attributed to the gradual build-up of Np in the flowsheet during run up to steady state, a consequence of recycle in the U re-extraction cascade. Overall acid mass balances across the flowsheet were reasonably good.

Table I. Mass balances for Np rejection trial across four product samples Time (Hr/mins) 0.50 1.40 2.41 3.26

Uranium (%) 99 99 99 101

Neptunium (%) 75 86 95 96

Acidity (%) 83 87 90 90

Aqueous and organic product streams for the Np rejection trial are given in Table II. From the four product samples taken during the course of the trial, decontamination factors (DFs) were calculated based on a feed (II) of 31.4 mg/L Np, for the aqueous (P2) and analyses of the organic product ( I 2 P ) streams (Table 3). These results showed that the Np was very effectively stripped from the U-loaded solvent. Long count times were employed to provide better confidence in the magnitude of the DFs obtained. As mentioned previously, the mass balance for Np improves over the duration of the trial. In light of this, it is more reasonable to calculate the Np D F based on the final two product samples (at 2 h 40 mins and 3 h 30 mins) where the flowsheet is operating under steady-state conditions. Based on this, an average DF of 2150 for Np decontamination of the U product was obtained. In terms of Np recovery this corresponds to approximately 99.95%. This exceeds the proposed requirement for 99.9% M A recovery in Advanced Fuel Cycles in order to minimise losses of long-lived TRUs. The aqueous flowrate was increased in this trial in order to maximise Np DFs and ensure the backwash of a modest mass of uranium. The levels of U (

Separations for the Nuclear Fuel Cycle in the 21st Century - American

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