Nuclear Fuel Cycle Simulation System (NFCSS)

User Manual 2009-06-05

FOREWORD

Nuclear Fuel Cycle Simulation System (NFCSS, formerly known as VISTA) was developed by the International Atomic Energy Agency (IAEA) to estimate long term nuclear fuel cycle material and service requirements as well as to estimate the material arisings for a given nuclear fuel cycle option. The IAEA decided to disseminate NFCSS tool by creating a web based version of the tool and developed it. A Technical Document “Nuclear Fuel Cycle Simulation System (VISTA)” was published in 2007 to introduce the tool to the readers.

This user manual briefly provides the basic characteristics of the tool and then explains usage of the tool through the web based application. The manual then provides some sample scenarios and their outputs to demonstrate the capabilities of the tool to the readers.

CONTENTS

1. INTRODUCTION ............................................................................................................ 1

1.1. Background......................................................................................................... 1 1.2. Features............................................................................................................... 1 1.3. Scope .................................................................................................................. 2

2. DESCRIPTION OF NFCSS ............................................................................................. 2

2.1. NFCSS Information Flow................................................................................... 2 2.2. Nuclear Fuel Cycle ............................................................................................. 3 2.3. NFCSS Nuclear Material Flow Model ............................................................... 6 2.4. NFCSS Material Flow Calculation Model ......................................................... 8

2.4.1. Input Parameters: .................................................................................... 8 2.4.2. Output Parameters: ................................................................................. 9 2.4.3. Main Assumptions .................................................................................. 9 2.4.4. NFCSS Calculation Process.................................................................. 10

2.5. Fuel Depletion Model ....................................................................................... 12 2.5.1. Need to Track Actinides ....................................................................... 12 2.5.2. Selection of Actinides........................................................................... 12 2.5.3. NFCSS Reactor Model (CAIN)............................................................ 13 2.5.4. Cross-Sections Used ............................................................................. 14

3. DESCRIPTION OF NFCSS WEB SITE........................................................................ 15

4. NFCSS FULL VERSION............................................................................................... 21

4.1. Introduction ...................................................................................................... 21 4.2. Program Operation ........................................................................................... 22 4.3. Access to the NFCSS full version .................................................................... 23

4.3.1. Structure of NFCSS Pages.................................................................... 23 4.4. Scenario Preparation......................................................................................... 24

4.4.1. Creating a New Scenario or Selecting Existing Scenario (Scenario List Page) .............................................................................................. 24

4.4.2. Editing Scenario Description (Scenario Page)...................................... 26 4.4.3. Selecting Reactor Types and Associated Fuel Types (Scenario

Page) ..................................................................................................... 27 4.4.4. Editing Scenario Parameters Data (Scenario Data Page) ..................... 30

4.5. Calculations and Results................................................................................... 34 4.5.1. Isotopic Composition Calculations (Scenario Page) ............................ 35 4.5.2. Material Flow Calculations for Each Reactor Type (Scenario

Page) ..................................................................................................... 36 4.5.3. Overall Material Flow Calculations...................................................... 40

4.6. Working with Fuel Types ................................................................................. 41 4.7. Working with Reactor Types............................................................................ 46

5. CONCLUSION............................................................................................................... 50

REFERENCES......................................................................................................................... 50

APPENDIX I. INITIAL CORE LOADING AND FINAL CORE DISCHARGE ................ 51

APPENDIX II. THE LIST OF COLUMNS IN MATERIAL FLOW RESULT TABLES..... 54

1. INTRODUCTION

Nuclear Fuel Cycle Simulation System (NFCSS) is a scenario-based simulation system to estimate long-term nuclear fuel cycle material and service requirements as well as material arisings. The code uses simplified approaches to make estimations.

This user manual is intended to explain the users how to use NFCSS to model a nuclear fuel cycle option. The IAEA published a Technical Document on the subject in 2007. More detailed information can be found in [1].

1.1. Background

The International symposium on "Nuclear fuel cycle and reactors strategies: adjusting to new realities" was held from 3 to 6 June 1997 in Vienna. For the preparation of this symposium, an International Steering Group with representatives of 12 countries and 3 international organisations (IAEA, OECD/NEA and WNA) co-ordinated, from 1995, the work of six Working Groups. They studied different scenarios of energy consumption including nuclear electricity production. Nuclear electricity production and the associated nuclear fuel cycle requirements should then to be estimated.

Working Group 2 did a comparison of the existing calculation tools for that purpose. At that time, none of them were able to give simple and accurate estimations. Working Group 2, then, asked the Agency to adapt its old CYBA calculation model to take into account the two options for nuclear fuel back end strategy: direct disposal and reprocessing & recycling.

A new simulation system, named NFCSS (formerly known as VISTA), was developed in 1996. It was then intensively used to quantify the different scenarios fixed by the Working Groups of the Symposium.

The IAEA decided to further develop and disseminate NFCSS to the interested parties in Member States in 2003. The code was redesigned to include new reactor and fuel types and a web based software was developed. Currently NFCSS is available online at http://www-nfcis.iaea.org through an authorization procedure. It can be used by authorized users through the internet without installation of any software to the users’ computers. The scenario data and the results are stored centrally in IAEA servers, so the users can access to their scenario virtually from anywhere in the world.

1.2. Features

The NFCSS is able to calculate, by year over a very long period, nuclear fuel cycle requirements for all types of reactors. Calculations can be performed for a reactor, reactor park in a country or worldwide nuclear power plant park. Natural uranium, conversion, enrichment and fuel fabrication quantities should be estimated. Furthermore, the quantities and qualities (isotopic composition) of unloaded fuels can be evaluated to let the user apply a recycling strategy if desired.

Data inputs are reduced to a few basic data in order to let non-nuclear fuel specialists to develop different energy scenarios. The calculation speed of the system is quick enough to enable making comparisons of different options in a considerably short time. The NFCSS is designed to be an optimum mixture of accuracy, simplicity and speed.

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1.3. Scope

NFCSS calculations can cover the period ranging from the beginning of nuclear energy production to 2050 or 2100. In order to support estimations for the future term, NFCSS stores historical data in its database.

Historical data mainly come from the existing Agency database (PRIS). Other authoritative publications and consultant reports are also the sources for historical data. Future projection data can be calculated by using publications from different institutions. The IAEA’s Energy, Electricity and Nuclear Power Estimates up to 2030 (2007 edition is the latest edition) is one of the authoritative publications which was used to calculate future nuclear power projection data in NFCSS.

Fresh fuel requirements and spent fuel isotopic composition are then automatically calculated from a set of internal parameters that have been selected by experts and introduced in the program. The user may then choose to use spent fuel stockpiles to develop a recycling strategy. The estimation of accumulation of actinides including minor actinides is one of the capabilities of the simulation. Those accumulation estimations might be used to compare any future fuel cycle options for transmutation of minor actinides.

2. DESCRIPTION OF NFCSS

2.1. NFCSS Information Flow

The model uses simplified approaches to calculate the fuel cycle requirements and arisings. These simplified approaches make the code possible to estimate the long term fuel cycle service requirements for both open and closed fuel cycle strategies.

The main assumption in the model is that it is possible to simulate the nuclear fuel cycle by taking into account the evolution using different types of reactors during the years, without the precision of using a reactor by reactor database. In order to do this, commercially existing nuclear power plants are grouped into 7 types which are PWR, BWR, PHWR, AGR, GCR, RBMK and WWER. There are actually a few power plants which are not included in this grouping due to their insignificant share in the total nuclear power capacity. The different reactor types could be introduced to the system by the users if desired.

The NFCSS can be described as a simulation tool which makes calculations using a set of input parameters to produce a set of output parameters. The input parameters used in the model may be divided into three groups:

Strategy parameters - nuclear capacity and reprocessing-recycling strategies, reactor-type mixture and load factors, etc., on an annual basis;

Fuel parameters - average discharge burnup, average initial enrichment and average tails assay, etc., on an annual basis;

Control parameters - share of mixed-oxide fuel in the core of reactors using this type of fuel, lead and lag times for different processes, process loss coefficients, use of depleted or enriched uranium and the number of reprocessing cycles, etc., on annual basis.

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Fig. 1. NFCSS Information Flow.

The results are divided into the following groups:

Front End: Natural uranium requirements, conversion requirements, enrichment service requirements and fresh fuel requirements

Back End: Spent fuel arisings, total individual nuclides arisings including Uranium, Plutonium and Minor Actinides, reprocessing requirements.

Fig. 1 illustrates the input and output parameters of the NFCSS. The left side of the sketch gives the input parameters whereas the right side shows the list of output parameters.

2.2. Nuclear Fuel Cycle

Nuclear Fuel Cycle can be defined as the set of processes to make use of nuclear materials and to return it to normal state. It starts with the mining of unused nuclear materials from the nature and ends with the safe disposal of used nuclear material in the nature. A typical nuclear fuel cycle is given in Fig. 2.

To produce energy from Uranium in a nuclear reactor, it must be passed through in a series of different processes. The complete set of processes to make nuclear fuel from uranium ore is known as front end of the nuclear fuel cycle. The processes in the front end of the nuclear cycle are mining and milling, conversion, enrichment and fuel fabrication.

After producing energy in the reactor, nuclear fuel becomes spent fuel. Spent fuel has also to be processed in a storage facility or in a reprocessing facility if it is being recycled.

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Temporary storage, reprocessing, long-term storage, or final storage of spent fuel are together called back end of the nuclear fuel cycle

Fig. 2. Simplified diagram of the nuclear fuel cycle in recycle mode.

Mining & Milling: Uranium is an element that is widely distributed within the earth’s crust as ores. Its principal use is as the primary fuel for nuclear power plants. The uranium ore needs to be mined and then processed (milled) before being usable. Uranium ore is mined by open-pit or underground mining methods and the uranium is extracted from the crushed ore in processing plants or mills using chemical methods. Sometimes it is possible to pass chemical solutions to the ore beds and dissolve the uranium from the ore directly. This process is known as in-situ leaching. This is the first step in a nuclear fuel cycle. The feed for mining & milling process is uranium ore and the product is U3O8 concentrate, which is mostly called yellowcake due to its color and shape.

Conversion: The term conversion refers to the process of purifying the uranium concentrate and converting it to the chemical form required for the next stage of the nuclear fuel cycle. There are three such forms in common usage: metal, oxide (UO2 or UO3) and uranium

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hexafluoride (UF6). UF6 is the predominant product at this stage of the nuclear fuel cycle since it is easily converted to a gas for the enrichment stage, as employed in world’s most common reactor type (LWRs). For the PHWR fuel cycle, which generally uses natural uranium oxide as the fuel, conversion to the UF6 is not necessary. Uranium is purified and converted to UO2 or UO3 in this case. The Magnox fuel cycle uses natural uranium in metal form. The feed for this stage is U3O8 concentrate, and the products are UF6, oxide (UO2 or UO3) or metal, in applicability order.

Enrichment: Uranium naturally consists of about 0.7% of 235U isotope which is the main energy source in thermal reactors which are only commercially existing nuclear power plants in significant amounts. 99.3% of the natural uranium is 238U isotope. For LWR technology which is the most common reactor type in the world, it is impossible to build a LWR with the natural occurrence of 235U, so the 235U content should be increased with a special process. This process is called enrichment. There are two commercially available enrichment technologies: gaseous diffusion and centrifuge. Both techniques are based on the slight mass difference between 235U and 238U. So the enrichment is defined as the process of increasing the amount of 235U contained in a unit quantity of uranium. The feed for this stage is natural UF6 and the product is enriched UF6. The other output of the process is the uranium which has lower 235U content than the natural uranium. It is known as enrichment tail or depleted uranium.

Fuel Fabrication: Enriched uranium in UF6 form is converted to UO2 powder to make fuel for LWR technology. This powder then is formed into pellets, sintered to achieve the desired density and ground to the required dimensions. Fuel pellets are loaded into tubes of zircaloy or stainless steel, which are sealed at both ends. These fuel rods are spaced in fixed parallel arrays to form the reactor fuel assemblies. The whole process is referred as fuel fabrication. The similar procedure is adopted for natural uranium oxide fuel for some reactor types. The feed of this process is enriched or natural uranium oxide powder and the product is fuel assembly. The feed for Magnox fuel is uranium metal.

Reactor: The reactor itself is irradiator for nuclear fuel. It burns the fuel, produces energy and spent fuel. There are currently 7 commercially available reactor types in the world (classification is based on NFCSS assumptions): PWR, BWR, PHWR, RBMK, GCR, AGR and WWER. The feed for reactor is fresh fuel containing uranium or uranium/plutonium, in case of Mixed Oxide (MOX) fuel, for existing nuclear fuel cycle options. The product is the spent fuel consisting of newly generated nuclides such as fission products (Cs, I, …), minor actinides (Np, Am, Cm) and Plutonium as well as the uranium. The biggest part of the spent fuel is still uranium (more than 95% for the most reactor types).

Reprocessing: The spent nuclear fuel still consists of significant amount of fissile material that can be used to produce energy. The considerable amount of 235U is still contained in the spent fuel and there are new fissile nuclides that were produced during normal operation of nuclear reactor such as 239Pu. Some nuclear fuel cycle options consider taking out those fissile material from the spent fuel, re-fabricating it as fuel and burning in the reactor. MOX fuel is the most common fuel that uses reprocessed material. Reprocessing process is based on chemical and physical processes to separate the required material from spent nuclear fuel. The feed of this process is spent fuel and the products are reusable material and High Level Wastes (HLW).

Spent Fuel Storage: The spent fuel could be stored temporarily for future use or stored indefinitely. Spent fuel could be stored in pools (wet type, temporarily) or in silos (dry type).

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HLW Storage: The waste from fuel fabrication and reprocessing facilities are classified as HLW and requires careful treating. HLW is stored in special storage facilities after proper treatment.

2.3. NFCSS Nuclear Material Flow Model

NFCSS simulation tool is a two-layer computer model to calculate the overall nuclear material flow as well as the individual nuclide discharge and accumulation in a nuclear fuel cycle option.

As shown in Fig. 1, overall nuclear material flow is calculated by NFCSS main module and individual nuclide content is calculated by CAIN (Calculation of Actinide INventory) module which is a simplified fuel depletion code.

Overall material flow for a nuclear fuel cycle can be sketched by tracking the nuclear materials in each of the processes in the nuclear fuel cycle. NFCSS is capable of simulating different nuclear fuel cycle models with different reactor types and fuel types including non-existing fuel types (fuels with MA content) with the introduction of accurate libraries and data. Since commercially existing nuclear fuel cycle options in bulk amount are once through fuel cycle and U and Pu recycling in some reactor types, these two options are illustrated in Fig. 3.

The typical is a fuel cycle model with two fuel types. The first fuel type in the model is Uranium fuel from natural material whereas the second fuel type is the fuel using reprocessed material. The second fuel type in the system is mostly Mixed Oxide (U+Pu) fuel type since it is the only commercially available fuel from reprocessed material. Other fuel types such as fuels containing thorium or fuels with Minor Actinide content could also be used to investigate the future options.

The heavy metal starts flow from its natural location to its end location which is spent fuel storage or HLW storage in NFCSS modeling. NFCSS does not include final disposal of spent fuel in its model. In this flow the heavy metal spends some time in each of the processes or in between the processes. This is due to the technical and physical requirements of the handling of heavy metals.

Table 1 gives the list of typical waiting and process times for each step of a typical thermal reactor fuel cycle. Although there is always waiting or process times in each step, NFCSS assumes zero waiting and process times for conversion, enrichment, fuel fabrication and reprocessing in calculating nuclear material flow. Only cooling times in pools are considered in material flow calculations. Process time during reprocessing and cooling time before reprocessing is also taken into consideration in calculation of isotopic composition of reprocessed spent fuel.

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Fig. 3. Illustration of material flow simulated in NFCSS (UOX + MOX Fuel use case).

Table 1. Typical waiting and process times for a thermal reactor cycle.

Products or Services Lead Times

Nat-Uranium Procurement 2 years before loading

Conversion to UF6 1.5 years before loading

Enrichment 1 year before loading

UO2 fabrication 0.5 year before loading

Spent fuel storage in the reactor pool before transport 2 years after unloading

Spent UO2 or URE(* *) storage before reprocessing 5 years after unloading

Spent MOX storage before reprocessing 5 years after unloading

Reprocessing (Pu and RepU availability) 1 year

MOX fuel fabrication 1 year

URE(*) fuel fabrication 0.5 year * URE is UO2 fuel using reprocessed Uranium.

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2.4. NFCSS Material Flow Calculation Model

NFCSS material flow calculation model is based on the physical material flow model. NFCSS calculates annual average fuel cycle material and service requirements and material arisings, and whenever relevant, cumulative amounts by summing annual amounts. The mathematical model of the fuel cycle calculations has been developed by tracking the nuclear material from the beginning of the cycle to the end of the cycle based on material balance principle.

NFCSS is a scenario based simulation tool. Each problem is defined a scenario. There are input and output parameters for each scenario. The following list gives the definition of main input and output parameters:

2.4.1. Input Parameters:

Reactor Type: The type of the reactor which will be used in the scenario. There might be any number of reactor types in a scenario. It is possible to use reactor types which are included in the library of the NFCSS or to define new reactor types by defining the fuel type and fuel parameters for the reactor.

Fuel Types: The type of the fuel which. There are three types of fuels. One is Uranium fuel which is made by using natural uranium or enriched uranium. Second fuel type is the Uranium-Plutonium fuel type which is mainly used to utilize MOX fuel type in NFCSS code. The last fuel type is the free content fuel type and any isotopic composition can be used.

Nuclear Power: Net electric power capacity in MWe.

Load Factor: The ratio of the net electricity produced in the plant to the maximum producible electricity in the plant for the whole year.

Thermal Efficiency: The ratio of net electric capacity to the total thermal power capacity of the reactor.

Discharge Burnup: The average burnup level of the spent fuel at the discharge point.

Enrichment: Initial U235 Enrichment in Uranium fuel or Initial Total Pu content in MOX fuel.

Tails Assay: The amount of U235 in depleted uranium from the enrichment process.

Fuel Residence Time: The time duration that each of the equilibrium loading fuels must stay in the reactor.

Fuel Type 2 Share: The average ratio of the amount of second fuel type to the total fuel in the reactor. Most of the time fuel type 2 is MOX fuel.

Reprocessing Ratio: The ratio of amount of spent fuel to be reprocessed to the amount of total discharged spent fuel.

Uranium Source for MOX Fuel: The source of uranium in MOX fuel (depleted uranium or reprocessed uranium).

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2.4.2. Output Parameters:

Natural Uranium Requirement: The amount of natural uranium required to manufacture necessary fresh fuel.

Conversion Requirement: The amount of conversion service requirement.

Enrichment Service Requirement: The amount of enrichment service requirements (SWU).

Depleted Uranium Arising: The amount of depleted uranium arising from enrichment process.

Depleted Uranium Use for MOX fuel: The amount of depleted uranium used in MOX fuel.

Fuel Fabrication Requirements: The amount of fresh fuel requirement for each fuel type.

Electricity Production: The amount of electricity produced in the power plant.

Spent Fuel Discharge: The amount of spent fuel discharged from the reactor for each fuel type.

Spent Fuel Storage: The amount of spent fuel stored in interim storage facilities for each fuel type.

Spent Fuel reprocessing: The amount of spent fuel reprocessed for each fuel type.

Separated Pu amount: The amount of separated Plutonium from reprocessed spent fuel.

Amount of used Pu in MOX fuel: The amount of Plutonium used in MOX fuel.

Reprocessed U amount: The amount of separated Uranium from reprocessed spent fuel.

MA amount in Reprocessing Waste: The amount of Minor Actinides (Np, Am, Cm) in HWL from reprocessing process.

FP amount in Reprocessing Waste: The amount of Fission Products in HWL from reprocessing process.

2.4.3. Main Assumptions

Several assumptions are made in NFCSS material flow calculation in order to simulate the real nuclear fuel cycle. Below is the list of main assumptions:

• All calculations are performed annually. It is assumed that every reactor is loaded at the beginning of the year and discharged at the end of the year. All inputs must be given in yearly base and all the results will be calculated in yearly base.

• It is assumed that all mass reduction in heavy metal (U, Pu, MA, etc) amount is considered as fission product accumulation, although there is some mass loss during the operation of the reactor. This is an acceptable assumption for the purpose of the NFCSS.

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• Although it is possible to define every single reactor separately in a given scenario, it is feasible to group the reactors based on their neutronic characteristics. In that case, all inputs must be given in annual average mode and all the results will be calculated in annual average mode.

• Although NFCSS has a module to calculate isotopic composition of spent fuel after irradiation and after cooling, the NFCSS is also able to use isotopic composition table provided by the user. User can use more sophisticated computer model to calculate the isotopic composition to obtain more precise results.

• There are two modes of operation. The first mode is called Requirement Driven Mode. In this mode, all facilities work on requirement basis. For example, conversion requirement is calculated based on the fresh fuel requirement and then conversion facility is assumed to produce that amount of conversion services. There is no capacity limitation in this mode. In this mode, it is assumed that the fuel cycle facilities work in online mode. That means any material coming into the fuel cycle facility is assumed to be converted into product immediately. The second mode is Capacity Driven Mode. In this mode, output of each facility is driven by the capacity of the facility and the production over the requirement is stocked. The second mode of operation is not currently available in web version of NFCSS.

• Multiple recycling of material is possible only using the same fuel composition in the second fuel type.

• All material flow calculations are based on heavy metal masses. The unit is tHM (tonne Heavy Metal) unless otherwise noted.

2.4.4. NFCSS Calculation Process

To explain the calculation process, NFCSS is assumed to be running in “Requirement Driven Mode” of operation. In this case, calculation starts from the reactor. First of all, fresh fuel requirement is calculated for given reactor capacity and other parameters. Then two parallel calculation lines are performed. In the first line, simulation runs for the front end of the nuclear fuel cycle. The second line performs the calculation for the back end of the nuclear fuel cycle. The calculation process is shown in Fig. 4.

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Fig. 4. NFCSS Calculation Process.

In the front end of the cycle, annual fresh fuel requirements, annual enrichment service requirements, annual depleted uranium amount, annual conversion requirements and annual natural uranium requirements are calculated. The only interface between the first and second series of the calculations is the introduction of reprocessed material from the back end to the front end.

In the back end of the cycle, annual spent fuel discharge amount, annual spent fuel reprocessing amount, annual spent fuel storage amount, annual reprocessed material amount (U or Pu), and annual other material arisings such as MAs, annual Fission Products arisings and annual HLW arisings are calculated.

Material flow calculation algorithm is given in below list (calculation is performed for each reactor type and each year separately):

1. Fresh fuel requirement is calculated for each fuel type based on the given reactor capacity and other parameters. In NFCSS the total reactor capacity for any given year is calculated by: Total Power = Added Capacity + Continuous Capacity – Removed Capacity + Initial Capacity. where Added Capacity is the capacity which added to the system in given year; Continuous Capacity is the capacity coming from previous year; Removed Capacity is the capacity which is removed from the system; and finally Initial Capacity is the capacity which already exists in the system at the beginning of the scenario period.

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2. Based on the removed capacity and the fuel residence time in core, discharged spent fuel amount is calculated.

3. Using fresh fuel loading and spent fuel discharge, total fuel in core is calculated.

4. Amount of spent fuel which will be reprocessed after cooling is calculated and this fuel is routed to spent fuel cooling pool.

5. Amount of spent fuel which will be stored in interim storage facility is calculated.

6. Spent fuel amount in reprocessing pool is calculated.

7. Fuel fabrication requirement is calculated based on the fresh fuel requirement.

8. Enrichment requirement is calculated based on the fuel fabrication.

9. Conversion requirement is calculated based on enrichment and fuel fabrication.

10. Natural uranium requirement is calculated based on conversion.

11. Individual nuclides are calculated based on overall material flow and the isotopic composition in each step.

12. Cumulative stock values are calculated based on the annual accumulation.

2.5. Fuel Depletion Model

2.5.1. Need to Track Actinides

The actinide group includes elements from Th (Z=90) to Lr (Z=103), however major interest is given to the different isotopes of U and Pu (major components of fresh and spent nuclear fuel) and so-called Minor Actinides (MA - Np, Am and Cm) with extremely long life, high alpha (with energy of alpha-particles of 4-6 MeV, major contributors to the residual heat of spent fuel) and gamma radioactivity.

Assessment of the worldwide inventories of these elements/isotopes in spent fuel is important due to non-proliferation issues (for fissile isotopes of U and Pu) and radiotoxicity of long-lived MAs. The latter relates to open fuel cycle, when the safety of spent fuel storage and further final disposal should be justified and guaranteed, and to closed fuel cycle, when the same is applicable to immobilized high-level waste (HLW) containing MAs.

2.5.2. Selection of Actinides

The following criteria were applied to select actinides to be calculated by the NFCSS code:

• Short lived actinides or actinides of very small concentrations were excluded as unimportant for spent fuel/HLW in a long run term;

• Long lived actinides (half-life > 400 years) are assumed as stable.

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• Only 235U and 238U are considered as the initial nuclides in fresh uranium fuel. 234U is ignored because its transmutation into 235U is small.

Thus the following 14 nuclides have been selected and a simplified transmutation chain has been obtained:

235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 237Np, 241Am, 242mAm, 243Am, 242Cm and 244Cm.

Current CAIN module is limited to above nuclides, although NFCSS can be used to calculate material flow for fuel cycle options using other nuclides (i.e., thorium fuels). However, the user has to input the isotopic composition of spent fuel in thorium fuel case (note that the implementation of using other nuclides is not complete in NFCSS currently).

2.5.3. NFCSS Reactor Model (CAIN)

CAIN (Calculation of Actinide INventory of spent fuel) was developed by the IAEA for the needs of NFCSS simulation system to calculate isotopic composition of the spent fuel after irradiation and after cooling.

It solves Bateman's Equations for a point assembly like in ORIGEN code which is similar but commonly used tool with more details. CAIN code uses one group neutron cross sections. The cross sections could be selected from any standard library like ORIGEN library, or could be generated by using more complex lattice codes. Accuracy in the cross section library affects actually the accuracy of the NFCSS results. Hence, in a high accuracy need, a more complex lattice code might be used with a complex assembly model to generate average one group cross sections for the assembly. Then this cross section set might be used in CAIN.

There are currently 28 reaction chains and 14 decay chains are selected to be suitable for fresh fuels containing any of the 14 nuclides of the CAIN library. Other chains and other nuclides are ignored currently in CAIN. The transmutation chain is given in Fig. 5. More detailed information about the isotope and chain selection criteria can be found in [1].

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Fig. 5. Simplified transmutation chain.

Currently there are 24 fuel types for 13 different reactor types in CAIN library. The user can use any fuel type provided that the one-group neutron cross-sections and other fuel parameters are given by the user.

For further needs such as modeling thorium fuel cycle, new nuclides could be added to the cross section library. In that case, the reaction and decay chains have also to be expanded.

2.5.4. Cross-Sections Used

The CAIN code needs various inputs to evaluate the actinide inventory in the discharged fuel. They are cross sections and reactor constants such as specific power or neutron flux for 13 reactor types which are included in NFCSS fuel library.

The origins of the cross sections are described below:

• PWR, BWR and PHWR Uranium fuel cross section data are identical to the ORIGEN-2 library.

• RBMK uranium fuel cross sections were calculated by a consultant using a similar code to WIMS. 242mAm cross sections are extrapolated from PHWR library.

• AGR uranium fuel cross sections were calculated by the Canadian consultant using the WIMS code. Am and Cm cross sections were copied from RBMK data, because their neutron spectrums are similar. 242mAm cross sections are extrapolated from PHWR library.

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• GCR uranium fuel (0.71% enrichment) cross sections were linearly extrapolated by the above AGR cross sections of 1.6% and 2.6% enrichment at 0 to 4GWd/t range. Am and Cm cross sections were copied from PHWR data, because both reactors are using natural uranium fuel. 242mAm cross sections are identical from PHWR library.

• WWER cross sections (both uranium fuel and MOX fuel) were calculated by the a consultant.

• PWR and BWR MOX fuel cross sections are identical to ORIGEN-2 library.

• PBMR, GTMHR, PBGCFR and EFR cross-sections were calculated by a consultant.

• LMFR cross-sections are identical to ORIGEN-2 library.

More detailed information on CAIN module, existing cross-section library, its validation and verification against other codes and experimental results are given in [1].

3. DESCRIPTION OF NFCSS WEB SITE

NFCSS is a web based computer simulation system which is available in the INFCIS (The Integrated Nuclear Fuel Cycle Information System) web site (http://www‐nfcis.iaea.org/). The iNFCIS web site includes four databases and NFCSS published by the IAEA's Nuclear Fuel Cycle and Materials Section in the Division of Nuclear Fuel Cycle and Waste Technology.

The INFCIS web site is designed as a "one stop" resource for technical and statistical information about nuclear fuel cycle activities worldwide, as reported to the IAEA. The web site is available to everybody through a simple registration and login system. The entry page to the web site is given in Fig. 6.

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Fig. 6. Entry page to the iNFCIS web site.

Four other databases are published in the same web site in addition to NFCSS. They are:

• Nuclear Fuel Cycle Information System (NFCIS) covers civilian nuclear fuel cycle facilities around the world. It contains information on operational and non-operational, planned, and cancelled facilities. All stages of nuclear fuel cycle activities are covered, starting from uranium ore production to spent fuel storage facilities.

• World Distribution of Uranium Deposits Database (UDEPO) covers uranium deposits around the world, drawing on reports to IAEA technical meetings and other sources. It includes classification of deposits, technical information about the deposits, detailed geological information about regions, districts and deposits.

• Post Irradiation Examination Facilities Database (PIE) is derived from a catalogue of such facilities worldwide that the IAEA issued in the 1990s. It includes a complete survey of the main characteristics of hot cells and their PIE capabilities.

• Minor Actinide Property Database (MADB) is a bibliographic database on physico-chemical properties of selected Minor Actinide compounds and alloys. The materials and properties are selected based on their importance in the advanced nuclear fuel cycle options.

The NFCSS web site is accessible to all registered users of iNFCIS web site, however it is necessary to get authorized in order to use NFCSS full web version. The NFCSS web site consists of 7 tab pages as shown in Fig. 7.

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Fig. 7. Background page of NFCSS web site.

Background: This tab page is devoted to explain why NFCSS was needed. The objective of the system and its scope are also introduced under this section (Fig. 7).

Description: The NFCSS is briefly described in this tab page (Fig. 8).

Fig. 8. Description page of NFCSS web site.

Modeling: The modeling page (Fig. 9) gives little more information about material flow and reactor model of NFCSS. This section is made of three components. These are Nuclear Fuel Cycle, Material Flow and Reactor Model (Fuel Depletion and Burn-up Model).

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Fig. 9. Modeling page of NFCSS web site.

Example: This section is introduced to demonstrate the capacity of the NFCSS. The assumptions, scenario parameters and results were presented (Fig. 10).

Fig. 10.Example page of NFCSS web site.

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Calculation: Calculation page is made of two parts. The first one is used to introduce the model and input parameters and described in Modeling section of the NFCSS web site and the second part is basic calculation module. The basic calculation module has once through (one fuel type) and LWR Recycle Case (two for fuel type). The visitors of the web site can use this module to make simple nuclear fuel cycle material flow calculations by entering several fuel cycle parameters (Fig. 11).

Fig. 11. Basic calculation tool in NFCSS web site.

Full Version: NFCSS full version section is introduced to define and evaluate scenario parameters and to make calculations for front-end and back-end requirements. This is the full version of NFCSS. The users after authorization can define their own scenarios and then run the simulation. The details are given in this user manual (Fig. 12).

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Fig. 12. Full version page of NFCSS web site.

Help: The help section is devoted to introduce basic information about NFCSS and its web site (Fig. 13).

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Fig. 13. Help page of NFCSS web site.

4. NFCSS FULL VERSION

The NFCSS full version is a web based computer application which is designed to study back-end and front-end requirements of existing or planned nuclear fuel cycle scenarios for a reactor, reactor park in a country or in the world. The NFCSS full version is a scenario based tool so that the users are able to create scenarios and then work on them. The NFCSS is developed in such a way that a user could use existing fuel types and data files for reactor types to generate a new scenario or a user could define new fuel types and data sets for a reactor or a group of reactors to use in new scenarios to perform calculations for back-end and front-end requirements.

4.1. Introduction

The NFCSS full version is a web based computer application. The system is completely stored in the IAEA servers. The executable module is served by a web server in the IAEA Headquarters in Vienna. The scenario data and the results are stored in the IAEA database servers in Vienna. This approach has a lot of advantages:

1. There is no need to install any type of software in users’ client PCs.

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2. The users can access the NFCSS and their scenario data and results virtually using any PC in the world which has internet connection.

3. The NFCSS can be updated virtually any time. This is specifically important when there is a need to apply a patch. Applying patches and fixes is very easy because there is only one executable running for the whole users.

4. The users can use reactor and fuel types existing in the NFCSS library or can create their own reactor and fuel types.

5. The users are able to share their scenarios and results (currently not available).

6. More than one user can work on the same scenario (currently not available).

Below are some technical features:

1. The program has been fully tested and operational on MS Internet Explorer Version 6 and Version 7. The NFCSS is fully functional on FireFox Version 2 however, there are some display issues which needs to be solved.

2. The NFCSS application is an ASP (Active Server Page) application which runs on a web server in IAEA Headquarters in Vienna. The programming language of the system is Visual Basic.

3. The scenario data and the results are stored centrally in IAEA database servers (MS SQL Server Version 2005) in Vienna.

4. There is no need to install any third party component or software to users’ PC.

5. Exporting to and importing from MS Excel is available for some of the functions in NFCSS.

4.2. Program Operation

Since NFCSS is a scenario based simulation tool, the user needs to define the scenario to run the system. Below is the list of ordered steps to successfully run the NFCSS simulation tool.

1. Access to the NFCSS full version

2. Scenario Preparation

a. The user must Create New Scenario or Select Scenario to work

b. The user must edit scenario description (the scenario period is important)

c. The user must select the reactor and fuel types for the selected scenario

d. The user must enter scenario data for each of the reactor and fuel types in the selected scenario

3. Calculations and Results

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a. The user must make isotopic composition calculations for each fuel type in each reactor type in the selected scenario or must provide isotopic composition table

b. The user must make material flow calculation for each reactor type in the selected scenario

c. The user must make overall material flow for all reactor types in the selected scenario

The following sections explain each of the above steps in details and provide explanation for each of the input parameters.

4.3. Access to the NFCSS full version

Authorized users of the NFCSS can access to the NFCSS after login to http://www-nfcis.iaea.org by entering NFCSS web site. The NFCSS web site has 7 tabpages and one of the tabpage is Full Version. The user must select Full Version tab (Fig. 14)

Fig. 14. Accessing to NFCSS Full Version.

Unauthorized users will be able to see only Demo Scenarios in read only mode.

4.3.1. Structure of NFCSS Pages

After accessing to NFCSS full version, Full Version Home Page is displayed (Fig. 15). There is a breadcrumb in this page (red box in Fig. 15) to quickly access to the different parts of the system. First shortcut in the breadcrumb is the link to the entry page. The second one is the link to the main scenario list page. The user can create a new scenario, delete an existing scenario or select one of the existing scenarios to work with in this Scenario List Page. The third shortcut in breadcrumb is the link to the File List page. Finally the last shortcut in the breadcrumb is the link to the Scenario Page in which the users can edit scenario data, and run the scenario. The breadcrumb is always available in NFCSS in all pages.

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The below part of the page is the main part of the page (blue box in Fig. 15). This part is different for each part of the NFCSS.

The upper part is the heading part for NFCSS web site (pink box in Fig. 15. The heading is the same for all of the pages in NFCSS web site.

Another feature of the NFCSS application is that there is always confirmation dialogs when the user wants to add or delete something. For example when the user wants to create a new scenario a dialog like Fig. 17 is displayed.

Fig. 15. NFCSS Full Version Home Page.

4.4. Scenario Preparation

4.4.1. Creating a New Scenario or Selecting Existing Scenario (Scenario List Page)

The first active page in NFCSS simulation system is the Scenario List page. It is available through the link in the breadcrumb. When the user clicks on Scenario List link in breadcrumb, the main scenario list page is displayed (Fig. 16). The user can create a new scenario in this page, delete an existing scenario or select an existing scenario to work with.

This page gives the list of existing scenarios for the connected user and provides commands to do further.

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Fig. 16. Scenario List Page.

4.4.1.1. Creating a New Scenario

The user can use the “Create New Scenario” button (upper red box in Fig. 16) to create a new scenario. After confirmation (Fig. 17), the NFCSS will create a new scenario and displays the Scenario Page for the newly created scenario. The default code for the new scenario will be “New Code”.

Fig. 17. Confirmation dialog for creating a new scenario.

4.4.1.2. Copying an Existing Scenario

The user can copy one of his/her existing scenario in order to start a new study with similar data. The copying scenario is done by clicking button in the Scenario List page. After a confirmation dialog, the NFCSS copies the selected scenario with all the data. The isotopic composition and material flow results are not copied. Calculations must be done for the newly created scenario after changing the data if necessary.

4.4.1.3. Deleting an Existing Scenario

The user can use the button to delete an existing scenario (see blue box in Fig. 16). After confirmation, the NFCSS will delete the selected scenario and all related data and result records from the database.

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4.4.1.4. Selecting Existing Scenario to Work With

The user can select one of the existing scenarios to work with from the Scenario List page. There are links to the Scenario Page for each scenario in the list (below red box in Fig. 16). The user needs to click on the Scenario Code for the scenario to work with.

4.4.2. Editing Scenario Description (Scenario Page)

Following the selection of working scenario, the NFCSS displays the Scenario Page. Scenario Page is the center of defining, running and postprocessing a scenario. Fig18 shows the NFCSS Scenario Page (headings for the web site is not displayed). There are three tabpages in NFCSS Scenario Page (see red boxes in Fig18). The below part of the Scenario Page is the main part of the page and displays different options depending on the selected tabpage.

After creating a new scenario, the first thing to do is to enter the general scenario description

data. The tabpage in Scenario Page provides tools to do this (Fig18).

The explanation of each input parameter and the possible options are:

Scenario Code: The identification code of the scenario. The text can be entered freely in this field up to 50 characters.

Scenario Explanation: The explanatory information about the scenario. This field is a free text field of any length.

Scenario Reference: The source(s) of information which is used to define the scenario, if any. For example, the user might have used IAEA Nuclear Power Estimates to analyze the worldwide situation in the future. This is a free text field with any length.

Start Year: Beginning of the scenario period. The data must be entered in integer format.

End Year: End of scenario period. The data must be entered in integer format.

It is important to note that the start and end year difference defined as scenario period should be less than 200.

FabricationLoss, EnrichmentLoss, ConversionLoss, ReprocessingLoss: Loss coefficients for each of the steps in fuel cycle. The loss is given in percent. The material loss is calculated at the beginning of the step and then the rest of the calculation is performed assuming the remaining material without any loss. ReprocessingLoss is currently not implemented.

Volume of HLW Generated in Reprocessing: The volume of the High Level Waste generated during the reprocessing of spent fuel per tonne Heavy Metal of reprocessed spent fuel.

Volume of ILW Generated in Reprocessing: The volume of the Low and Intermediate Level Waste generated during the reprocessing of spent fuel per tonne Heavy Metal of reprocessed spent fuel.

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Fig18. Scenario Page in NFCSS.

Following the entry of each of the data, the data must be saved before continuing the work. In order to save the scenario description, the user must click

button in Scenario Page.

The user can export whole scenario data to MS Excel to locally store the scenario data and results, or to work on it in MS Excel. In order to export scenario description including all the data to MS Excel, button is used.

4.4.3. Selecting Reactor Types and Associated Fuel Types (Scenario Page)

After scenario description, tabpage in Scenario Page (Fig18) is used to specify reactor and fuel types and data for the scenario parameters for each of the reactor and fuel type. Scenario Parameters page (Fig 19) includes the list of reactor types which is included in the selected scenario. If there is not reactor type selected, the list is empty. There are four parts in this page: Reactor Type list (blue box in Fig 19), Fuel Types (1 and 2) (green box in Fig 19), Command bar (brown bar in Fig 19) and a button to add new reactor type to the scenario. Scenario period is also displayed in this page to remind the user.

button is used to add a new reactor type, in command bar is used to delete the selected reactor type and its related data, in command bar is used to edit reactor and fuel type characteristics, by clicking button new reactor type is added using the

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characteristics of existing one and finally in command bar is used to edit scenario parameters (nuclear capacity, enrichment level, etc.) for the selected reactor type.

Fig 19. Scenario Parameters tabpage in Scenario Page.

command from Fig 19 is used to edit the reactor type and associated fuel types in the selected reactor type. Fig. 20 is displayed when the user selects editing the reactor type information. Explanations of each field in this page:

Reactor Type: Free text information for the reactor type up to 50 characters.

Reprocessed Uranium Use: When user wants to use reprocessed uranium in the fuel type 1, RepUUse parameter should be entered “Yes”. If it is selected, the ratio to convert reprocessed Uranium to natural uranium is needed. For simplicity it could be 1.

Fuel Type 1: The type of the fuel for the Fuel Type 1 (fuel from natural material, usually Uranium fuel). The fuel type must be selected from a predefined fuel list. The list includes the fuel types from the NFCIS built-in fuel library and the fuel types which have been added by the user. The NFCSS has a built-in library for fuel types which could be used by any user. The data for those built-in fuel types can not be changed by the users. However, the users can create their own fuel types to be used in their scenario. After addition of user fuel to the system, only the user who owns the fuel type can use that specific fuel type in his/her scenario. The Fuel Types are explained in Section 4.5.1 4.6.

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Fig. 20. Reactor type and associated fuel types.

Fuel Type 2: The type of the fuel for the Fuel Type 2 (fuel from reprocessed material or depleted uranium, usually MOX fuel). The fuel type must be selected from a predefined fuel list. The list includes the fuel types from the NFCIS built-in fuel library and the fuel types which have been added by the user. If the reactor type does not use second fuel type, No Fuel options must be selected from the list.

Initial Core Fissile Content: Initial core fissile content is defined as a percentage of the fissile content of equilibrium cycle fuel loading. The initial loading enrichment for fresh core is not same as equilibrium cycle loading enrichment since not entire initial core has to stay in the reactor for the whole residence time. The initial core loading enrichment is less than the equilibrium cycle loading enrichment. There is a box to enter the core fissile content ratio to determine initial core requirements in NFCSS. The calculation procedure is also different for final core discharge and the methodology used in the NFCSS is discussed in APPENDIX I.

Initial Content: The flag which indicates how initial content of fuel will be entered. The NFCSS can calculate initial fuel content from the given enrichment or fissile content or the user can provide the table for initial content for each year. In later case, the initial content table must be provided in Isotopic Composition page (Section 4.5.1).

Depletion: The flag which indicates how isotopic composition of spent fuel during discharge time will be entered. The NFCSS can calculate it if it has cross section library for the fuel or the user can provide the table for isotopic composition for each year. In later case, the initial content table must be provided in Isotopic Composition page (Section 4.5.1).

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Decay: The flag which indicates how isotopic composition of spent fuel after given cooling period will be entered. The NFCSS can calculate it if it has half lives for the fuel or the user can provide the table for isotopic composition for each year. In later case, the initial content table must be provided in Isotopic Composition page (Section 4.5.1).

The command bar has more buttons in Reactor Type edit mode (see brown box in Fig. 20). button is used to delete reactor type, button is used to cancel all last edit and reset the form,

button is used to cancel editing and close reactor type edit mode; and button is used to save the last edits.

4.4.4. Editing Scenario Parameters Data (Scenario Data Page)

button from Fig 19 is used to edit the scenario parameters data for each the reactor type and associated fuel types. Fig. 21 is displayed when the user selects editing scenario parameters data.

Fig. 21. Scenario Data Page.

The scenario data page consists of several parts: There are five different tabpages for general characteristics, parameters for fuel type 1, parameters for fuel type 2, capacities of fuel cycle facilities and initial values (red box in Fig. 21); command buttons (brown box in Fig. 21); data part (blue box in Fig. 21); the reactor type and scenario period information (yellow box in Fig. 21); and the buttons for data charting (green boxes in Fig. 21).

When a new reactor type added to the scenario, records for scenario parameter data are also created with “0.0” values in each of the scenario parameter. The user, later, can edit the data records to complete the scenario.

The easiest data entry is done through MS Excel since MS Excel is known to many computer users already and it provides easy data editing features in tabular form. In this case, the data can be prepared in MS Excel and then imported into NFCSS using

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button. In order to do that, the data has to be prepared in the same column structure in MS Excel. The exporting data was explained in Section 4.4.2. The exported data may be used to prepare new data set or to modify existing ones and can be imported back to NFCSS.

The other way to enter data is to use NFCSS for whole. There is a practical way in NFCSS to enter data. In most cases, the data for several parameters is constant for the scenario period. The NFCSS can utilize this option by providing easy data preparation tool.

First of all, for any year (each row in the table represents a year in the NFCSS scenario since NFCSS makes calculations annually) the data can be entered by selecting a predefined reactor type from the reactor list (Working with reactor types is explained in Section 4.7.). In order to do that, button from DataYear column (leftmost column in the table) can be used. After the confirmation, Reactor Type Selection Page will be displayed (Fig. 22). The left part of the page shows the list of the reactor types which are available to the user and the right part of the page shows the details of the active reactor type. The active reactor type can be changed by clicking on the reactor type name (blue box in the Fig. 22). The active page will always be displayed in different background color in the list. If the active reactor type will be selected as base data, button in the Command column of the list must be used (red box in the Fig. 22). After clicking on the button, the data from selected reactor type will be copied to the given year.

Fig. 22. Reactor type selection page.

If the user wants to enter the data manually for any year, it can be done in the table directly. After entering the data manually it should be saved using button each time before doing any other command. There is another easy tool for the users to enter data to the NFCSS. If the scenario data is not changed for whole scenario period or part of it, there is a command to copy the data from any year to the rest of the scenario period. To do that, the user

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must click button in the leftmost column of the data page. Using that button, the data is automatically generated for every year of the scenario period after the selected year.

There is only one thing to do later. Since the Added Capacity and Removed Capacity are used to calculate Net Total Capacity, the data for Added Capacity and Removed Capacity should be entered individually for each year of the scenario period later.

Saving data will re-calculate Net Total Capacity each time.

Explanations of each field in this page (All parameters must be entered for the selected reactor type and the data must be calculated in annual average for given year):

Added Capacity (MWe): Net electric capacity added to the system.

Removed Capacity (MWe): Net electric capacity removed from the system.

Net Total Capacity (MWe): Total net electric capacity. This column is automatically calculated by the system using the data given for initial power, added and removed capacity.

Thermal Efficiency (%): The ratio of net electric capacity to the total thermal energy produced in the reactor.

Load Factor (%): The ratio of the net electricity produced in the plant to the maximum producible net electricity in the plant for the whole year.

Enrichment Tails Assay (%): The amount of U235 in depleted uranium from the enrichment process.

Fuel Type 2 Share (%): The average ratio of the amount of second fuel type to the total fuel for the reactor type. Fuel type 2 is MOX fuel for the most of the existing reactor types, if any.

Residence Time (yr): The time duration that each of the equilibrium loading fuels must stay in the reactor. In reality, multi-batch fuel management schemes are used in reactors and a fraction of core refueled. If n is the number of fuel batches, the fraction of core refueled is 1/n. If the operating cycle length is Tc then the residence time to be used in NFCSS is equal to (n Tc) (closest integer must be given to the NFCSS). The NFCSS will calculate annual average loading or discharge instead of actual loading every operating cycle.

Discharge Burnup (GWd/tHM): The average discharge burnup level of the spent fuel at the discharge point for each of the fuel type.

Enrichment (%): Initial fissile U235 in Uranium fuel or Initial Total Pu content in MOX fuel for each of the fuel type.

Reprocessing Ratio (%): The ratio of amount of spent fuel to be reprocessed to the amount of total discharged spent fuel. It is assumed in NFCSS that some portion of the spent fuel is sent to cooling pool for reprocessing and the rest is sent to interim storage facilities. The reprocessing ratio defines the amount of spent fuel which is sent to the cooling pool for reprocessing for given year and not the amount of reprocessed spent fuel for the given year.

Capacities of fuel cycle facilities (tHM or SWU): The alternative running mode of NFCSS is “Capacity Driven Mode” which will be implemented in the future requires the capacities of each fuel cycle facility. Only reprocessing facility capacities have been implemented so far.

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Reprocessing Capacity (tHM): The maximum amount of reprocessing which can be performed in given year.

Initial Reactor Power (MWe): Total net electric capacity of selected reactor type at the beginning of the scenario period. The reactor is assumed to be running in equilibrium mode at the beginning of the scenario period. This data should be supplied in the initial data section and it is not part of data import from MS Excel.

Initial Reprocessed Pu Stock (tHM): The total amount of Reprocessed Pu which is available for recycling at the beginning of scenario period.

Initial SF Stock (tHM): The total amount of spent fuel (for each fuel type) which is ready for reprocessing at the beginning of the scenario period.

4.4.4.1. Considerations for Scenario Parameters Data

While specifying parameters for different reactor types, the consistent data set needs to be entered. In order to do that below considerations needs to be taken:

1. Enrichment and Burnup relations may be derived depending on reactor types. For PWR, the discharge burnup and enrichment relation for 3-batch core is

0.109.12 −= εxBd

And for, 4 batch core

02.107.13 −= εxBd

ε is defined as enrichment and Bd is defined as discharge burnup.

For PHWR

830.626.22 −= εxBd

where ε is enrichment level of fuel in percent, BBd is the discharge burnup (GWd/tHM)

2. Enrichment of Fuel Type 2 to Enrichment of Fuel Type 1 relation may be derived by keeping equal fissile content in each fuel type. For example, if the enrichment of fuel type is 4 %, that means the fissile content is 4 %. The fissile content of the second fuel type will be the same if the same discharge burnup level is used. So, if the fissile content in Pu in MOX fuel is about 60%, then the total Pu content in the MOX fuel should be at least 6.66 %.

3. Specific power of fuel and discharge burnup relation: Specific power is entered in the fuel type information. The specific power is defined as the power produced in unit amount of fuel. Specific power is only used to calculate the irradiation period of the fuel and that is used to calculate isotopic composition of the spent fuel after irradiation. However, consistent discharge burnup, residence time and specific power

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need to be entered in order to keep the accuracy higher. The relation between those three is defined as ./ RTBSP d=

4. Fuel type must be consistent with the burnup. See Table 2 for the recommended fuel types for different burnup levels for different fuel types.

5. First Core Loading Requirements:

To define initial fuel loading requirements, initial core enrichment ratio needs to be specified depending on fuel cycle scenarios and target burnup for initial core. In real practice, different reactors apply different enrichment levels for initial core loading. For instance,

• TURKEY POINT-PWR-2200 MWt 3.1, 2.55, 1.85 w/o enriched fuels

• INDIAN POINT #2-PWR-2758 MWt 3.2, 2.7, 2.2 w/o enriched fuels

• GINNA PWR 1300-MWt 3.48, 2.78, 2.44 w/o enriched fuels

• AP-1000 (PWR) 3415 MWt 2.35-4.45 w/o eq. 4.8 w/o

To define initial core enrichment ratio, the following procedure is suggested:

Assume number of batch is n (in NFCSS this is residence time since each batch is assumed one year), enrichment of reload fuel at equilibrium core is εeq, enrichment of discharged fuel at the end of cycle is εd.

Since enrichment (fissile material weight percent) decreases linearly as a function of burnup, the following model might help to define equilibrium composition of fuel batches at the end of each cycle and use the equilibrium cycle results to define initial core requirements:

Hence the core average enrichment in initial core is

( )( 1) ( 1)2

eq dinitial core ave

n nn

ε εε − −

+ + −=

For PWR’s, let us assume that n = 3, εd =1.5 w/o and εeq =3.2, w/o,

owavecoreinitial /50.2=−−ε

4.5. Calculations and Results

The last tabpage in Scenario Page is the Calculations and Results page. This tabpage is used to make all calculations in the NFCSS and then see the results (Fig. 23).

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Calculations and Results tabpage displays the list of each reactor type and associated fuel types. Blue box in Fig. 23 is the list of reactor types and associated fuel types. Every row in the list corresponds to a reactor type.

The calculations must be performed in the correct order in order to get the appropriate results. The correct order of the calculations is:

1. Isotopic composition calculations for each fuel type in each reactor type

2. Material flow calculations for each reactor type

3. Overall material flow calculations

Fig. 23. Calculation and Results tabpage in Scenario Page.

4.5.1. Isotopic Composition Calculations (Scenario Page)

In order to start the NFCSS calculations, the user first needs to make sure that isotopic compositions of initial fresh fuel, spent fuel at discharge time and spent fuel after a cooling period. The compositions could be provided as tables, or could be calculated by the NFCSS depending on the selected options in definition of fuel types (see section …..).

In order to perform isotopic composition calculations, initial composition of fresh fuel and other fuel parameters have to be provided. If all data is correctly entered for a fuel type, button is used to perform isotopic composition of spent fuel. The calculation might take several seconds depending on the length or scenario period. The button is available for each of the fuel types for each of the reactor types. button is used to see the result of isotopic composition calculation for each fuel type.

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If the isotopic composition is to be provided by the user, button is again used. When the user clicks on button, isotopic composition page is displayed (Fig. 24). This page has three tab pages (red boxes) for each of the cycle positions: Beginning of the cycle, end of the cycle and after cooling. The page has also two buttons (blue boxes) to import isotopic composition from MS Excel of to export isotopic compositions to MS Excel. In order to import the table from MS Excel, the same column structure (order of the columns and number of columns are important) has to be prepared in MS Excel. Then the data must be selected together with the header row and copied into clipboard (Ctrl-C makes it in MS Excel). After having the data in Clipboard, button is used to import data from MS Excel to the NFCSS. Similarly button can be used to export data from NFCSS to MS Excel.

Below part of isotopic composition page displays the data for each year and each isotope (and some isotope groups).

Fig. 24. Isotopic Composition Page.

4.5.2. Material Flow Calculations for Each Reactor Type (Scenario Page)

After completing isotopic composition calculations or providing the isotopic compositions, the user then can make material flow calculations for each of the reactor types. Fig. 25 shows

the command bar in red box ( ) for the material flow calculations. In this command bar, button is used to perform material flow calculations for the selected reactor type, button is used to see the material flow result in table form and button is used

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to see the material flow result in diagram form. If the isotopic compositions are not ready for any reactor type, the command bar will display a warning message.

Fig. 25. Calculations and Result tabpage showing material flow buttons..

When the user clicks on the NFCSS start material flow calculations. The calculation is completed usually in several seconds depending on the length of scenario period. During the calculations the NFCSS displays a Wait message on the screen. After completing the calculations the NFCSS returns to the Calculations and Results tabpage.

Following the calculation, the user might want to see the results. The results can be displayed in two different formats: tabular and diagrammatic. When the user clicks button, material flow result page is displayed for the selected reactor type (Fig. 26). This page consists of four tabpages: General, Front End, Back End and Nuclide Group tabpages provides the different results from the material flow calculation. The table displays the pre-selected result columns from the NFCSS full result table. The selection of the columns by the users will be implemented in the future.

The user can see the graph for each of the result column by clicking on button in each column heading. The chart form is shown in (Fig. 27).

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Fig. 26. Material flow result page in table form.

Fig. 27. Example result in chart.

The NFCSS calculates and displays below outputs (and more) in addition to the summary of some input parameters for selected reactor type for each year:

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Nuclear Capacity: The total net electric capacity.

Electricity: The total electricity amount generated in the power plant.

Natural Uranium Requirement: The amount of natural uranium required to manufacture necessary fresh fuel.

Conversion Requirement: The amount of conversion service requirement. Conversion service requirement is calculated for the end product from the conversion facilities. For example for conversion to UF6 the only requirement for conversion to UF6 is calculated and not conversion to UF4 if it is done in a different facility.

Enrichment Service Requirement: The amount of enrichment service requirements (SWU).

Depleted Uranium Arising: The amount of depleted uranium arising from enrichment process.

Fuel Fabrication Requirements (for two fuel types): The amount of fresh fuel requirement.

Total Fuel in Reactor Core: The total amount of fuel in the reactor core.

Natural Uranium Saving due to Use of MOX: Saved amount of natural uranium as a result of MOX fuel use.

Natural Uranium Saving due to Use of reprocessed Uranium: Saved amount of natural uranium as a result of reprocessed uranium use.

Spent Fuel Discharge (for two fuel types): The amount of spent fuel discharged from the reactor.

Spent Fuel Storage (for two fuel types): The amount of spent fuel stored in interim storage facilities.

Spent Fuel reprocessing (for two fuel types): The amount of spent fuel reprocessed.

Separated Pu amount: The amount of Plutonium separated from reprocessed spent fuel.

Used Pu in MOX fuel: The amount of separated Plutonium used in MOX fuel.

Depleted Uranium Used in MOX fuel: The amount of depleted uranium used in MOX fuel.

Separated U amount: The amount of separated Uranium from reprocessed spent fuel.

MA amount in Reprocessing Waste: The amount of Minor Actinides (Np, Am, Cm) in HLW from reprocessing process.

FP amount in Reprocessing Waste: The amount of Fission Products in HLW from reprocessing process.

In order to see the results in diagrammatic form, button needs to be used. The result page in diagrammatic form is shown in Fig. 28. The user can select the output year from the list (blue box in Fig. 28). Immediately after selecting the year, the diagram will be updated automatically to show the results for the selected year.

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The user can get the chart for results of each of the fuel cycle steps by clicking on the box of steps. The mouse cursor will automatically change when the mouse comes over a chartable parameter in the diagram.

The complete list of result parameters with their column headings and the short explanation can be found in APPENDIX II.

Fig. 28. Material flow result in diagrammatic form.

4.5.3. Overall Material Flow Calculations

The material flow results of each of the reactor types can be combined using overall material

flow calculations. is used to make overall material flow calculations. The result of the combination is displayed in table or diagram form similar to the result for each reactor type.

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The all calculations can be done by clicking only however, depending on the performance of the server at the time of request, the program operation might not be completed. In that case, performing all calculations separately is needed.

4.6. Working with Fuel Types

One of the important features of the NFCSS simulation system is its module to calculate isotopic composition of the spent fuel at the discharge time or after a decay period. The module is called CAIN and the details of CAIN module is given in section 2.5. In order to make the isotopic composition calculations, the CAIN needs to know some parameters of the fuel. These parameters are mainly initial content of the fuel, the specific power of the fuel, burnup average one group-neutron cross sections and burnup average neutron flux. All these parameters form the fuel type information. The NFCSS has a built-in fuel type library which includes a number of typical fuel types for the existing and planned reactor types. The list of existing fuel types from the NFCSS library and their reference information is given in Table 2.

4.6.1.1. Fuel Type List

The users can access the available fuel type list by selecting Fuel List link from the NFCSS breadcrumb (Fig. 29).

Fig. 29. Fuel List in NFCSS breadcrumb.

This gives the list of fuel types which are available to the user. The list will show the fuel types from the NFCSS built-in library and the user’s own fuel types. Other users’ fuel types will not be displayed in this list.

The fuel list page is shown in Fig. 30. The list shows the names of each fuel type and the owner. Clicking on the fuel type name displays the fuel data page for the selected fuel type.

Fig. 30. The fuel list page.

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4.6.1.2. Adding a New Fuel Type

In addition to the selection of the fuel type from the built in library of the NFCSS, the users can add their own fuel types to the fuel list. The user added fuel types will only be available to the creator of the fuel type.

button is used to create a new fuel type and to add it to the library. The newly created fuel type will be displayed at the end of the list. The user then goes to the fuel data page to define the details of the new fuel type by clicking on the name of the new fuel type.

4.6.1.3. Copying Existing Fuel Type

The users can copy one of the existing fuel types from the list to create a new fuel type. They can later revise the data for the newly created fuel type and use in their scenario. In order to copy the fuel type, button from the command column of the fuel list table must be used. After that a new fuel type will be added to the end of the list with “New Fuel” name and with the same data as copied fuel type.

4.6.1.4. Editing Fuel Type

Clicking on the name of the fuel type opens the fuel data page for the selected fuel type. The fuel data page is shown in Fig. 31. The page has two parts: First part is the command part and the other part is the data part. The data part consists of two tabpages: Fuel Characteristics and Cross Sections. In the fuel characteristics tabpage, fuel parameters such as specific power, reference enrichment and etc are entered. In the cross section tabpage, fission, capture and (n, 2n) cross sections are entered. After entering the data the data needs to be saved by clicking on the button.

If the user wants to delete one of owned fuel types, button needs to be used. The user can cancel the last edit any time and reset the data to the status on last save by clicking

42

Table 2. Fuel Type library of NFCSS and recommended burnup levels for selection.

Reactor Type Reference Discharge Burnup (GWd/MWe)

Fuel Type from NFCSS Library

Reference for Fuel Type

33 PWR UOX 1 ORIGEN PWRU

50 PWR UOX 2 ORIGEN PWRUE

50, with 3 batch loading PWR UOX 3 ORIGEN PWR50

PWR

33 PWR MOX 1 ORIGEN PWRPUPU

33 BWR UOX 1 ORIGEN BWRU

43 BWR UOX 2 ORIGEN BWRUE

BWR

33 BWR MOX 1 ORIGEN BWRPUPU

7 PHWR U 1 ORIGEN CANDUNAU PHWR >7 PHWR U 2 ORIGEN CANDUSEU

GCR Consultant

AGR Consultant

RBMK Consultant

40 WWER U 1 Consultant

40, civilian grade Pu WWER MOX 1 Consultant

WWER

40, weapon grade Pu WWER MOX 2 Consultant

PBMR Consultant

GTMHR Consultant

PBGCFR Consultant

EFR Consultant

LMFR LMFR CORE ORIGEN AMOPUUUC

LMFR RADIAL ORIGEN AMOPUUUR

LMFR AXIAL ORIGEN AMOPUUUA

43

Fig. 31. Fuel Data Page.

The description of the fuel parameters and possible options are:

Reference Enrichment (%): Enrichment level of reference fuel from which the cross sections are calculated. This parameter has no effect on the NFCSS calculations.

Power Density (kW/kg): Power density (or specific power) of the fuel. Power density is used to determine the irradiation period which is used in CAIN module to calculate isotopic composition of spent fuel during the discharge time. It is also used to calculate the initial core loading requirements. This parameter must be consistent with Residence Time and Discharge Burnup data in the scenario data part.

Fuel Type Code: A short code to represent the fuel type.

Fuel Type: The type of the fuel. NFCSS has currently three types of fuel in its library. “U Fuel”, “U-Pu Fuel” and “Free Content”. The calculation of initial content is different for each fuel type. The details are given in Initial Content parameter below.

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Conversion Type: Type of conversion which is necessary to manufacture the fuel. There are three types of conversion. “UF6” is used for enriched fuel. “UO2” is typically used for PHWR fuels and “UMetal” is used for GCR fuels. For enriched uranium fuels, reconversion to UO2 after enrichment is not considered as a conversion step. It is considered as internal part of fuel manufacturing.

Cooling Time (yr): Minimum time required for cooling before reprocessing the fuel. T

Manufacturing Time (yr): Minimum time required for manufacturing the fuel.

Reprocessing Time(yr): Minimum time required for reprocessing the fuel.

Fuel Type Description: Free text for description of the fuel type.

Information Source: The reference where the fuel type information is received from.

Burnup Average Neutron Flux (n/(cm2.sn)): Burnup average one group neutron flux.

Fuel Initial Contents: The table which is used to calculated the initial compositions of the fresh fuels.

If the fuel is "U Fuel" type, initial content has no meaning since the content will be calculated from the given enrichment level.

If the fuel is "U-Pu Fuel" the initial content must give the Pu vector which gives the contents of Pu isotopes in total Pu amount and the enrichment of uranium part of the fuel. In this case the initial content will be calculated from given total Pu content.

If the fuel is "Free Content" the initial content must be entered as complete since the initial content of the fuel

Cross Sections (barn): Burnup average one group fission, capture and n2n cross sections for each isotope in the fuel (Fig. 32).

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Fig. 32. Cross Sections part of fuel data page.

4.7. Working with Reactor Types

The NFCSS has a built-in reactor type library which includes a number of typical reactor types for the existing and planned reactor types. The list of existing reactor types from the NFCSS library and their reference information is given in Table 3.

4.7.1.1. Reactor Type List

The users can access the available reactor type list by selecting Reactor List link from the NFCSS breadcrumb (Fig. 33).

Fig. 33. Reactor List in NFCSS breadcrumb.

This gives the list of reactor types which are available to the user. The list will show the reactor types from the NFCSS built-in library and the user’s own reactor types. Other users’ reactor types will not be displayed in this list.

The fuel list page is shown in Fig. 34. The list shows the names of each reactor type and the owner. Clicking on the reactor type name displays the reactor data page for the selected reactor type.

46

Fig. 34. The reactor list page.

4.7.1.2. Adding a New Reactor Type

In addition to the selection of the reactor type from the built in library of the NFCSS, the users can add their own reactor types to the reactor list. The user added reactor types will only be available to the creator of the reactor type.

button is used to create a new reactor type and to add it to the library. The newly created reactor type will be displayed at the end of the list. The user then goes to the reactor data page to define the details of the new reactor type by clicking on the name of the new reactor type.

4.7.1.3. Copying Existing Reactor Type

The users can copy one of the existing reactor types from the list to create a new reactor type. They can later revise the data for the newly created reactor type and use in their scenario. In order to copy the reactor type, button from the command column of the reactor list table must be used. After that a new reactor type will be added to the end of the list with “New Reactor” name and with the same data as copied reactor type.

4.7.1.4. Editing Reactor Type

Clicking on the name of the reactor type opens the reactor data page for the selected reactor type. The reactor data page is shown in Fig. 35. The page has two parts: First part is the command part and the other part is the data part. After entering the data, the data needs to be saved by clicking on the button.

If the user wants to delete one of owned reactor types, button needs to be used. The user can cancel the last edit any time and reset the data to the status on last save by clicking

47

Table 3. Reactor Type library of NFCSS.

Reactor Type

Net Capacity (MWe)

Thermal Efficiency (%)

Load Factor (%)

Fuel Residence Time (yr)

Tails Assay(%)

Initial Enrichment(%)

Discharge Burnup (GWd/t)

Reference

PWR 1000 33 80 4 0.3 4 45

BWR 1000 33 80 4 0.3 3.8 42

PHWR 600 30 80 1 0.3 0.711 7

RBMK 925 28 80 3 0.3 2.05 15

AGR 600 40 80 3 0.3 3 30

GCR 220 26 80 1 0.3 0.711 5

WWER 950 30 80 4 0.3 4 45

PBMR

GTMHR

PBGCFR

EFR

LMFR

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Fig. 35. Reactor Data Page.

The description of the fuel parameters are given in the input parameters description (Section 4.4.4).

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5. CONCLUSION

The Nuclear Fuel Cycle Simulation System NFCSS has following features;

• A simplification to nuclear fuel cycle requirements modeling, by not reflecting the actual fuelling pattern reactor by reactor.

• A method to calculate reactor type average requirements

• A system capable of simulating the evolution of key fuel cycle features for given parameters, e.g., load factors, fuel burnup and enrichment, tails assay, reprocessing fractions, and recycling related parameters. All parameters can be changed from reactor type to reactor type, and from time period to time period.

• A tool to track the individual actinide content in any stage of the nuclear fuel cycle including in the discharged and stored spent fuel even after a period of time. So it can be used to evaluate different fuel cycle options in terms of their effectiveness in reducing the radiotoxicity of the spent fuel or reducing the natural nuclear material requirements.

• A tool designed for being an optimum mixture of accuracy, simplicity and speed.

The development of the NFCSS is an ongoing process. New features will be added to the system continuously. Web based centrally located structure of the NFCSS allows continuous improvement and upgrading of the system. Although this user manual will be updated each time a modification is performed in the NFCSS system, the most up-to-date information can be seen at the NFCSS application itself on the internet.

REFERENCES

[1] IAEA TECDOC-1535, Nuclear Fuel Cycle Simulation System (VISTA), 2007, Vienna.

[2] IAEA Power Reactor Information System (PRIS) [3] IAEA iNFCIS Web Site, http://www-nfcis.iaea.org.

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APPENDIX I.

INITIAL CORE LOADING AND FINAL CORE DISCHARGE

NFCSS is a simulation system which tries to represent the travel of nuclear material in the real world. In order to do that, like in every other simulation systems, there are assumptions to enable simulation models to run. One of the assumptions of the NFCSS model was about initial core loading and the final core discharge.

In NFCSS’s previous status, initial core loading for a new power added to the reactor type (basically adding a new reactor to the electricity grid) had been calculated using the Residence Time parameter which is given in scenario input parameters in whole years (etc. 4 years, 5 years). This has an effect in the resulting initial core loading depending on how the actual residence time differs from the whole integer. For example if the real residence time is 4.65 years then we should enter Residence Time as 5 (nearest integer).

In real operation of a nuclear power plant, loading and unloading of the fuels performed according to scheme which is given in Figure 1. Upper part of the figure shows the fuel types in the reactor core (initial core fuel or equilibrium core fuel) for whole life of the reactor. Cycle (1) is initial loading and Cycle (n) is the final discharge. Middle part shows the loadings to the reactor at every cycle. In the first cycle, only initial core fuel type is loaded into the reactor. In the following cycles, equilibrium fuel type is loaded. The lower part of the figure shows the discharges from the reactor. In the first cycle there is no discharge from the reactor. In the following cycles, first initial core fuel type is discharged and then equilibrium fuel type is discharged. In the final cycle full core containing equilibrium fuel is discharged from the reactor.

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Figure 36 Schematic illustration of loadings in and discharges from the reactor.

NFCSS model has been modified to reflect the real situation of refuelling scheme of the reactors. NFCSS first calculates the material flow and then calculates the composition in each stage (beginning of cycle and end of cycle stages).

NFCSS calculations are based on annual estimations, not cycle-wise estimations. So that the initial core loading and final core discharge must be estimated different than the equilibrium loading and discharge. Initial core loading is calculated using the Irradiation Time in which is defined as the time the fuel is irradiated at full power in the reactor. The equations used for calculating loadings are:

365*)(1000*)()(

yrtyPowerDensiyrBurnupyrnTimeIrradiatio =

1000*)(*)(

365*)(*)()(yrBurnupyriciencyThermalEff

yrityAddedCapacyrnTimeIrradiatioyreLoadingInitialCor =

1000*)(*)(365*)(*)()(

yrBurnupyriciencyThermalEffyrpacityExistingCayrLoadFactoryrngmCoreLoadiEquilibriu =

The discharges are calculated in below formula:

1000*))(Re(*))(Re(365*))(Re(*)(

)(argyresidenceTimyrBurnupyresidenceTimyriciencyThermalEff

yresidenceTimyrLoadFactoryrpacityExistingCatemCoreDischEquilibriu

−−−

=

)1()1(*)(Re)(arg

−−

=yrityTotalCapac

yrnCoreTotalFuelIyritymovedCapacteischFinalCoreD

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Units:

Burnup MWd/kg

PowerDensity kW/kg

IrradiationTime year

Capacities MWe

Loading-Discharge tonnes

ResidenceTime year

One assumption is made in this calculation. Although we load the fuel into the reactor in its every operation cycle in real life, we assume that the fuel is loaded and discharged every year. Since we are using IrradiationTime parameter in initial core loading and final core discharge calculations, annual loading is not exactly the same as full core amount divided by Residence Time defined by integer number mentioned previously.

Another important assumption is made in the calculation of isotopic composition of discharged fuel. Although the composition of initial core fuel is different than the equilibrium fuel, we assume that the composition of the discharged fuel is the same in the initial core and the equilibrium core. Similarly, although the average burnup level of the final core discharge is usually lower than the equilibrium fuel, we assume the composition of the final core discharge is the same as the one of the discharged fuel in equilibrium core.

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APPENDIX II.

THE LIST OF COLUMNS IN MATERIAL FLOW RESULT TABLES

Column Heading Explanation

TotalCapacity Total Nuclear Capacity

TypeCapacity Nuclear Capacity

Electricity Annual Electricity Generation

F2Share Share of Second Fuel Type

Enri_1 Enrichment of Fuel 1

Bu_1 Discharge Burnup of Fuel 1

Rep_1 Reprocessing Ratio for Fuel 1

Enri_2 Enrichment (Total Pu Content) of Fuel 2

Bu_2 Discharge Burnup of Fuel 2

Rep_2 Reprocessing Ratio for Fuel 2

NatU Natural Uranium

ConversionUF6 Conversion to UF6

ConversionUO2 Conversion to UO2

ConversionUMet Conversion to UMetal

ConversionTot Total Conversion

ENR_In Enriched Input

SWU Enrichment SWU

Enr_DepU Depleted Uranium

ENR_U Enriched U

FF_1 Annual Fresh Fuel for Fuel 1

FF_2 Annual Fresh Fuel for Fuel 2

TotalFF Annual Fresh Fuel

FuelInCore Total Fuel In Core

NatUSaveMOX NatU Saving From MOX use

NatUSaveRepU NatU Saving From RepU use

SFD1 SF Discharge for Fuel 1

SFD2 SF Discharge for Fuel 2

SFD SF Discharge Total

SFS1 SF Storage for Fuel 1

SFS2 SF Storage for Fuel 2

SFS SF Storage Total

SFSStock1 SF Storage Stock for Fuel 1

SFSStock2 SF Storage Stock for Fuel 2

SFSStock SF Storage Stock total

SFRS1 SF Storage for Reprocessing for Fuel 1

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Column Heading Explanation

SFRS2 SF Storage for Reprocessing for Fuel 2

SFRS SF Storage for Reprocessing Total

SFRSCP1 SF Cooling Pool for Fuel 1

SFRSCP2 SF Cooling Pool for Fuel 2

SFRSCP SF Cooling Pool total

SFRSCPSto1 SF Cooling Pool Inventory for Fuel 1

SFRSCPSto2 SF Cooling Pool Inventory for Fuel 2

SFRSCPStock SF Cooling Pool Inventory total

SFRSRP1 SF Ready for Reprocessing Pool for Fuel 1

SFRSRP2 SF Ready for Reprocessing Pool for Fuel 2

SFRSRP SF Ready for Reprocessing Pool total

SFRSRPSto1 SF Ready for Reprocessing Pool Inventory for Fuel 1

SFRSRPSto2 SF Ready for Reprocessing Pool Inventory for Fuel 2

SFRSRPStock SF Ready for Reprocessing Pool Inventory total

SFR1 SF Reprocessing for Fuel 1

SFR2 SF Reprocessing for Fuel

SFR SF Reprocessing total

DepU Depleted Uranium

DepUUsed Depleted Uranium Use

DepUAccu Annual Depleted Uranium Accumulation

DepUStock Depleted Uranium Stock

RepU Reprocessed Uranium

RepUUsed Reprocessed Uranium Use

RepUAccu Annual Reprocessed Uranium Accumulation

RepUStock Reprocessed Uranium Stock

RepPu Reprocessed Plutonium

RepUUsedNatUEq Natural Uranium Equivalent of Used Reprocessed Uranium

RepPuUsed Reprocessed Plutonium Use

RepPuAccu Annual Reprocessed Plutonium Accumulation

RepPuStock Reprocessed Plutonium Stock

RepMA Reprocessed MA or MA in HLW

RepMAUsed Reprocessed MA Use

RepMAAccu Reprocessed MA Accumulation

RepMAStock Reprocessed MA or MA in HLW Stock

RepFP FP in HLW

RepFPStock FP in HLW Stock

ILWVolume ILW Volume

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Column Heading Explanation

ILWVolumeStock ILW Volume Stock

HLWVolume HLW Volume

HLWVolumeStock HLW Volume Stock

56