Approaches to Minimization of High Level Wastes in ... on AFC/Russia_… · Approaches to Minimization of High Level Wastes in Advanced Fuel Cycles 1 presented by Alexander Grol,

Approaches to Minimization of High Level Wastes in Advanced Fuel Cycles

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presented by Alexander Grol,

National Research Centre “Kurchatov Institute”, Moscow

[email protected]

IAEA Technical Meeting on Advanced Fuel Cycles to Improve the Sustainability of Nuclear Power Through Minimization Of High Level Waste

17–19 October 2017, Vienna, Austria

NATIONAL RESEARCH CENTRE

«KURCHATOV INSTITUTE»

Intro (1/2)

According to the Russian Federation federal target program "Nuclear Energy Technologies of New Generation for the Period of 2010-2015 and for the Future up to 2020” the State Atomic Energy Corporation ROSATOM has to direct its efforts on:

creation of conditions necessary for the maintenance and growth of the nuclear power complex;

creation of system that reliably ensures nuclear and radiation safety;

solution of adjourned environmental problems that arose at the first stages of development of the country's nuclear industry.

To solve existing problems it’s necessary to develop nuclear energy system with closed fuel cycle

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Intro (2/2)

The results achieved in the field of closing fuel cycle:

Industrial technology of aqueous chemical reprocessing of the spent nuclear fuel with uranium and plutonium extraction and separation of highly active waste (RT-1);

Monorecycling of uranium obtained from the spent nuclear fuel (LEU 2,8% for RBMK);

Experimental technology of production mixed oxide uranium-plutonium fuel (MOX). The MOX-fuel production facility for the power unit with the BN-800 reactor was built.

Now a lot of research are aimed on multirecycling of spent nuclear fuel

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Minimization of the Radiation Burden Using FA with Intra-Assembly Heterogeneity in FR

• One of the possible ways to reduce the radiation burden in the fuel cycle is the separate recycling of spent fuel from the breeding and active zones of the core.

• The objective of this research is estimation of the opportunity of using fuel assemblies with intra-assembly heterogeneity in terms of reduction of the required time for storage of the part of the SNF before reprocessing.

• The concept of this approach is closing of the fuel cycle on the first stage of the transition of nuclear energy system to fast reactors using hydrometallurgy reprocessing.

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• 1 – wrapper;

• 2 – fissile fuel element;

• 3 – fertile fuel element;

• 4 – coolant;

• 5 – distance element.



Theoretical Advantages/Disadvantages

Advantages

1. Reduction of storage time

2. Reduction of radiation burden on personal and equipment

3. Natural U economy

4. Involving of depleted U in fuel cycle

5. More cheap and practical method of storage of Pu in SNF form

6. Accumulated in fast reactors Pu has higher quality

7. Less HLW during reprocessing per unit of Pu in SNF

8. Burnup of the reprocessed fuel from fertile fuel pins is lower => less FP => less problems with precipitation

Disadvantages

1. Uncertainty of the scale of building nuclear power plants with fast reactors in Russia

2. Long-term storage of the spent MOX fuel => economic effect

3. Implementing plutonium separation during reprocessing => non-proliferation issues

4. Necessity of developing dense fuel production technologies in Russia

5. FR SNF contains a lot of fissile material => PUREX – water technology => nuclear safety issues

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Initial Data

• Object of the research – core of the sodium fast reactor of high power

• Fuel operation mode:

― Duration of the irradiation – 5 micro campaigns;

― Duration of storage in IRS – 2 years.

• Fuel for the fertile fuel elements:

1) Metal depleted uranium with a 6…10% zirconium addition. Density of the fuel tablet – 15,4 g/cm3;

2) Depleted uranium nitride. Effective density – 11,5 g/cm3.

• Preservation of the:

1) reactor micro campaign;

2) power;

3) design decisions.

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Choice of Fuel Composition for IAH FA

• Designing of the models of different level of complexity for neutron physics calculations

― 2D and 3D models of fuel cell and FA

― full reactor core models;

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Range of Options to Select Fuel Compositions for IAH FA

U(depl)Met+8%Zr U(depl)N(nat)

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25 25.5 26 26.5 27 27.5 28x_Pu, %

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12

13

14

15

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gam_UZr, g/cm^3

K_infdelta_Ro

26 26.5 27 27.5 28 28.5 29x_Pu, %

10.0

10.5

11.0

11.5

12.0

12.5

13.0

gam_UN, g/cm^3

K_infdelta_Ro

• Dependence of the multiplication properties on

• {effective density; amount of Pu}, mass fraction of Zr = сonst} ― K_inf – multiplication factor of FA in the end of the equilibrium micro campaign ― Delta_Ro – reactivity margin for one micro campaign, rel.un.



Choosing the Optimal Option (nitride/metal)

Estimated functionals: ― k_crit – average value of the multiplication factor k-inf of the FA obtained

for the ends of five micro campaigns; ― k_start - average value of the multiplication factor k-inf of the FA obtained

for the beginnings of five micro campaigns; ― dRO = 1/k_crit – 1/k_start.

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MOX 18,2% Pu

UN, 28%Pu Eff. dens=11,5 g/cm3

8%Zr, 26% Pu Eff. dens=14 g/cm3

k_start 1,2478 1,2400 1,2288

k_crit 1,2146 1,2126 1,2143

dRO 0,022 0,0182 0,0098



Evaluation of Radiation Characteristics

A set of characteristics sufficient for the initial assessment of the radiation situation and heat release in the fuel cycle :

• Activity;

• Heat release;

• Power and spectrum of neutron radiation;

• Power and spectrum of photon radiation;

• Equivalent dose rate;

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SNF Composition of MOX FA and IAH FA of Fast Reactor

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• Homogeneous MOX FA

• Fissile pins of IAH FA

• Fertile pins of IAH FA

Fuel composition Pu composition



Results of Evaluations of Radiation Characteristics (1/3)

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Activity

Activity of the fertile fuel pins about ten times lower than activity of both fissile and homogeneous fuel pins Heat release A similar situation is for heat release



Results of Calculations of Radiation Characteristics (2/3)

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Power of neutron radiation

The neutron radiation of MOX fuel is due to the spontaneous fission of actinides and (α,n) reaction on heavy isotopes of oxygen 17O and 18O presenting in natural oxygen.

The neutron radiation of IAH FA with metal fuel elements is mainly due to spontaneous fission of actinides. Power of photon radiation



Results of Calculations of Radiation Characteristics (3/3)

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The equivalent dose rate from unprotected fuel assemblies After holding FA in the IRS, the radiation characteristics of the fertile and fissile fuel pins differ greatly, which makes it reasonable to recycle them separately



Results

• It is shown that after storage in the IRS the radiation characteristics of the fertile and fissile fuel elements differ greatly, which makes it reasonable to recycle them separately

• Estimations show that level of heat release of fuel pins of homogeneous MOX FA is reduced to 7 W per liter (acceptable value for PUREX) after 12 years of storage

• Fertile fuel pins of the IAH FA achieve that level of heat release in 1,5-2 years after pulling out of inner reactor store

• Average burnup of IAH FA in this case is ~116 MWd/kgHM, but average burnup of fertile fuel pins is ~40 MWd/kgHM

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Useful R&D Directions

• Improvement of dense fuel production technology. Dense fuel based on the metal uranium is not a priority option as a reactor fuel in Russia. There are still some problems with achieving necessary burnups with dense nitride fuel

• Economic evaluations of a large number of aspects of this approach

• This technology is a temporary measure before switching to a hybrid PH-process and then to pyrochemical technology. That is why it is essential to continue intensive research for the development of these processes

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Assessments of Perspective of Uranium Recycling in LWR Fuel Cycle

The main purpose of this option is closing fuel cycle inside nuclear energy system with light water reactors. The principle consists in extracting uranium and plutonium from UOX spent nuclear fuel using hydrometallurgical technology (PUREX) and use them in new fresh fuel. Plutonium can be used one time in MOX fuel for LWRs. Reprocessed uranium mixed with natural (or natural and depleted) uranium can be recycled up to six times .

Multiple usage of recycled uranium within the closed fuel cycle of LWRs would make it possible to extend the resource basis of nuclear energy and to achieve certain advantages compared to once-through fuel cycle.

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Theoretical Advantages/Disadvantages

Advantages 1.Natural U economy

2. Reducing of waste amounts to be disposed of since SNF contains over 90% of uranium

3. Opportunity to close fuel cycle without fast reactors

4. Involving of depleted U in recycling

5. Option of using MOX fuel from reprocessed Pu is also not excluded

6. Described below processes can be applied on the existing facilities in Russia

7.The same principles can be applied to the unseparated U/Pu mixture for REMIX fuel production

Disadvantages

1. Closing fuel cycle in LWRs system is a temporary measure

2. Price of the natural uranium is still really low. This circumstance plays negative role in all discussions about closed fuel cycle

3. Plutonium separation and storage => non-proliferation issues

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Restrictions in Fabrication of Fuel from Recycled Uranium

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• 232U is a threat to radiation safety due to presence of the 208Tl in the decay

chain => high energy photons (~ 2,61 MeV) => necessity of the protection

for people and equipment. 232U content limited by ~ 2,0•10-7 U wt. % for

LWR fuel. At present time “MSZ” is licensed to RBMK fuel production

with 232U content ~ 5,0•10-7 U wt. %

• 236U is an additional neutron absorber => necessity of compensating its

presence by raising amount of 235U.

• 234U content is limited by 20000 μg/g 235U according to ASTM C996-96

specification due to its alpha activity. Strong internal alpha-particles source

causes dissociation of the UF6 to nonvolatile compounds (UF5, UF4, UF3,

possibly U2F9 и U4F17) which impairs quality of the enrichment product.



Scenarios of Multiple Recycling of Reprocessed Uranium (1/3)

The following sequence of uranium fuel irradiation scenarios was considered:

1) Fresh fuel with average enrichment of 4,6% is fabricated from natural uranium, irradiated in the core up to burnup 46,9 MWtdays/kg U, cooled for 5 years and reprocessed.

2) Recycled uranium from reprocessed SNF comes to the separation cascade for further enrichment. Natural uranium (or natural and depleted uranium) is supplied to this cascade too. Obtained fresh fuel containing reprocessed uranium is returned into the core, irradiated, cooled and reprocessed.

3) Then recycled uranium is sent to the separation cascade again, etc. The total of five consecutive recycled uranium cycles was considered.



Scenarios of Multiple Recycling of Reprocessed Uranium (2/3)

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Contribution of the Uranium Isotopes to Alpha Activity of LWR Spent Fuel (Bq/MT)

Storage, days 3 30 365 730 1095 1825 3650

232U 6,90E+08 7,23E+08 1,08E+09 1,38E+09 1,61E+09 1,91E+09 2,19E+09

233U 1,43E+06 1,44E+06 1,52E+06 1,59E+06 1,67E+06 1,82E+06 2,20E+06

234U 5,01E+10 5,02E+10 5,07E+10 5,12E+10 5,17E+10 5,28E+10 5,53E+10

235U 8,18E+08 8,18E+08 8,18E+08 8,18E+08 8,18E+08 8,18E+08 8,18E+08

236U 1,43E+10 1,43E+10 1,43E+10 1,43E+10 1,43E+10 1,43E+10 1,43E+10

238U 1,14E+10 1,14E+10 1,14E+10 1,14E+10 1,14E+10 1,14E+10 1,14E+10

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According to ASTM C996-96 specification concentration ratio of 234U to 235U should be below 0,02. In our case in reprocessed uranium this ratio is 0,01943 (before enrichment). That’s why we can use only cascade schemes that involve dilution of the reprocessed uranium.



Cascade Schemes of the Additional Enrichment of Reprocessed Uranium with 235U Using Two Supply Sources

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F, E, P, W, – flows of natural U supply, reprocessed U supply, product and waste respectively Cascade scheme - addition of reprocessed U into intermediate point of the cascade;



Cascade Scheme of the Additional Enrichment of Reprocessed Uranium Using Three Supply Sources

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F1, F2, F3 – flows of depleted, natural and reprocessed uranium respectively; - components concentrations in flows F1, F2, F3. P и W – flows of product and waste with concentrations



Scenarios for Cascade Schemes

• Scenarios for cascade scheme with two supply flows:

Licensed for fuel made of reprocessed uranium reactor is loaded by fuel

assemblies made of reprocessed U obtained from spent fuel of several

reactors => we have an assumption that resources of reprocessed U is

unlimited. In other words maximal load of reprocessed U into the core is

modeled.

Reprocessed U obtained after work of a reactor is used for fabrication of

fuel for the same reactor. We shall call this scenario “one to one” further.

• Scenario for cascade scheme with three supply flows:

In this case only scenario “one to one” is considered

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Consumption of Natural U, Reprocessed U and Separation Work

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Relative consumption of natural uranium

Relative separation work

Relative consumption of reprocessed uranium



Assessments of Dependence of Natural U and Separation Work Consumption on Natural U Cascade Parameters (“+” – saving; “-” –

overspending)

Recycle

Two supply flows Three supply flows

Maximum Consumption of Reprocessed U

“One to one” “One to one”

Natural U, %

Separation Work, %

Natural U, %

Separation Work, %

Natural U, %

Separation Work, %

1 41,96 9,81 19,15 4,47 50,90 -47,09

2 11,94 1,63 19,02 3,70 51,07 -48,11

3 13,33 2,16 13,11 2,10 50,90 -58,50

4 13,73 2,30 13,58 2,28 50,72 -58,22

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Results (1/2)

• Estimations show effectiveness of different cascade schemes designed for enrichment of reprocessed uranium.

• In case of scenario “one to one”, when we use reprocessed uranium from one spent fuel assembly to produce one fuel assembly, full usage of reprocessed fuel can be maintained only on first two recycles.

• Degradation of isotope composition and regulatory limitations doesn’t allow us to use all of the reprocessed uranium. We can load unspent product into another core but natural uranium saving would be less in this case.

• In case of scenario “Maximum Consumption of Reprocessed U” its full usage is possible only on first recycle.

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Results (2/2)

• Scenario of the “Additional Enrichment of Reprocessed Uranium Using Three Supply Sources” (natural U, reprocessed U, depleted U) showed itself as the most effective from the natural uranium saving point of view.

• On the other hand, cost of stable 50% of natural uranium saving is 50% overspending of separation work (60% after third and fourth recycle).

• Such kinds of schemes hypothetically can provide us with almost any degree of natural uranium saving (50% in this work is just an illustration). But in process of choosing ratio between spending of natural uranium and separation work we need to keep in mind that this ratio depends on parameters of depleted uranium (ex. density of 235U). That’s why in case of using these schemes (we can actually apply them on the existing enrichment facilities) we have to perform inventory of depleted uranium reserves

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Useful R&D Directions

• Economic evaluations of a large number of aspects of this approach. For example, analysis of depleted uranium reserves, their quality and the effect of it on the cost of the processes

• Continued research in the field of improving hydrometallurgical technology (increasing lifetime of the equipment, implementation of more resistant carriers for sorption and adsorption, etc.)

• This technology is a temporary measure before switching to a hybrid PH-process and then to pyrochemical technology. That is why it is essential to continue intensive research for the development of these processes

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Conclusion

• Nowadays more and more people in nuclear community talk about future perspectives and inadmissability of further postponement of nuclear waste and fuel supply problems.

• Closing of fuel cycle through monorecyling and on the next step through multirecycling of spent nuclear fuel is a solution that can help to cope with this problems.

• We believe that pooling of efforts on the international level to solve this problems will be useful for all participants of this process and future generations.

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Acknowledgements

Authors participated in the work:

P.N. Alexeev1, A.L. Balanin1, V.Yu. Blandinskiy1, A.A. Dudnikov1, P.А. Fomichenko1, V.А. Nevinitsa1, S.А. Subbotin1, A.Yu. Smirnov2, G.A. Sulaberidze2

1National Research Centre «Kurchatov Institute» 2National Research Nuclear University “MEPhI”

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Thank you for attention

33 NATIONAL RESEARCH CENTRE


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Approaches to Minimization of High Level Wastes in ... on AFC/Russia_… · Approaches to Minimization of High Level Wastes in Advanced Fuel Cycles 1 presented by Alexander Grol,