technology for radioactively contaminated wastewater...

Technology for radioactively contaminated wastewaterfrom the nuclear power plants

processing.

Danube WATER Project, MIS ETC 166

Technology for radioactively contaminated wastewater from the nuclearpower plants processing.Data for reporting: 01.07.2015

PROJECT MANAGER Dr. Mary-Jeanne ADLER

PROJECT RESPONSIBIL PP6Dr. eng.Aurelia PISCUREANU

PROJECT RESPONSIBIL PP7Dr. Arsene Carmen

PROJECT RESPONSIBIL PP8Prof.Dr.eng. Dima RomulusDr.eng. Dinculescu DanielDumitru

PROJECT RESPONSIBIL PP13Assoc. prof. Dimitar TONEV

EXPERTS CONTRIBUTING PP6 – ICECHIM

EXPERTS CONTRIBUTING PP7-RATEN-ICN

Aurelia PISCUREANUDana VARASTEANUIrina CHICANMircea RUSEGeorgeta NEGRAUMariana MATEESCUCarmen ANDREIASNicolae TIGANILA

Arsene CarmenDulama MirelaDeneanu NicoletaDianu MagdalenaOlteanu MirelaBujoreanu DănuţBujoreanu LilianaIchim CameliaBadea SilviuPătrulescu PetreIvan ConstanţaIoniţă AureliaBobaru Constantin

EXPERTS CONTRIBUTING PP8-UPB

EXPERTS CONTRIBUTING PP13-INRNE

Cârlanaru VasilePopescu Felicia

Dinculescu Daniel DumitruGijiu Cristiana LuminitaMares Alina Monica

Prof. Kiril Asenov KREZHOVAssoc. prof.Tsvetana PetrovaNONOVAAssoc. prof. Anna AndreevaDAMIANOVAAssoc. prof. Hristo NedialkovPROTOHRISTOV

Liuben Todorov DOBREVAssoc. prof. Lachezar StoyanovGEORGIEVPetko Borislavov KRASTEVTsvetana JordanovaSIMEONOVAZahari Rangelov ZAHARIEVSevka Velcheva IVANOVANeli Nikolaeva ZAHARIEVAElena Georgieva GELEVAAlexandar Dechkov MLADENOVBozhidar Slavchev SLAVCHEVEvgeni Petrov POPOV

National Research & DevelopmentInstitute for Chemistry & Petrochemistry– ICECHIM, Bucharest RO– PP6

Authonomus Company of Technologiesfor Nuclear Energy – Institute forNuclear Research, Piteşti RO – PP7

University POLITEHNICA of BucharestRO – PP8

The Institute for Nuclear Research andNuclear Energy, Sofia BG – PP 13

“Danube WATER Integrated Management”

Technology for radioactively contaminated wastewater from thenuclear power plants processing

Leader of working group for Activity 12: National Research & Development Institutefor Chemistry & Petrochemistry – ICECHIM – PP6

Legal representative: Project Coordinator: PhD Sanda VELEA PhD Aurelia PISCUREANU

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CONTENT

1. INTRODUCTION…………………………………………………………………….. ……...11.1 The purpose and importance of work performed ………………...............................……....11.2 Legal basis ……………………………………………………………………...……. …........31.3 The phases of research performed ……………………………………………..…... …........42. PERFORMED WORKS AND THEIR RESULTS…………………………….…… .….......52.1 Laboratory works.……………………………………………………………….…... ..……..5

2.1.1 The laboratory works performed at ICECHIM, in non-radioactive simulatedsystem……………………………………………………………..………..…….……….. …........5

2.1.2. The laboratory works performed at ICN Pitesti, with radioactive waste waterstype CANDU .……………..……………………………………………………….……... …........9

2.1.3. The laboratory works performed at IRNE Sofia, with radioactive wastewatertype Kozloduy technology …….………………………………………………….….…… …......12

2.1.4. Modeling of laboratory procedure …...…………………………………………...…......292.2 Pilot scale testing of the decontamination process of radioactive wastewater fromCernavoda NPP ………………………………………………………………………...….…......33

2.2.1. Ion exchange resins used within experiments …………………………………...…......332.2.2. Aqueous radioactive waste processed………………………………..…..……... …......332.2.3. Methods and apparatus ………………………………………...…………….…. …......332.2.4. Test description …………………………………………………………….….... …......342.2.5. Experimental results……………………………………………………….…… …......36

3. TECHNOLOGY FOR TREATMENT OF AQUEOUS RADIOACTIVE WASTEFROM CERNAVODA NPP……………………………………………….……………. ……..383.1 Flow chart of technology for treatment of liquid radioactive waste fromCernavoda NPP………………………………………………………………………….. …......383.2 Pilot installation design…………………………………………………………….... …......423.3 Operation procedure……………………………………………………………….... …......46

3.3.1 Preparing ion exchange resins for utilization…………………………………... …......463.3.2 Preparing pilot- installation for operation………………………………………. …......463.3.3 Loading the columns with ion exchange resins………………………………………......463.3.4 Settleable solids separation and organic phase separation......................................…......473.3.5 Aqueous radioactive waste clean-up......................................................................... ….....473.3.6 Effluent storage and discharge……………………………………………………..…......483.3.7 Discharging spent ion exchange media…………………………………………. …......48

3.4 Processing parameters for treatment by ion exchange of aqueous radioactivewaste from Cernavoda NPP…………………………………………………………….. …......48

3.4.1 Chemical and radiochemical characteristics of aqueous radioactive waste……....…......493.4.2 Selected ion exchange resins – characteristics………………………………….. …......503.4.3 Establishing parameters which control the ion exchange resins performances….…......503.4.4 Influent flow rate and flow velocity through ion exchange resin bed………….. …......51

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4. RECOMMENDATIONS ON THE TYPE, DIMENSIONING OF SPECIFICEQUIPMENT AND CONSTRUCTION MATERIALS................................................. ….....525. RECOMMENDATIONS ICN ON THE WORK SAFETY MEASURES……….... …......646. RECOMMENDATIONS ON THE WORK SAFETY MEASURES AT INRNE… .........667. SUMMARY OF THE RESULTS AND CONCLUSIONS……………….................... ……..697.1 Technology efficiency……………………………………………………………...…... .........697.2 Cost consideration…………………………………………………………………....... .….....71REFERENCES……………………………………………………………………..…....... ...…...73

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1. INTRODUCTION

1.1 The purpose and importance of work performed.

In Romania is in force the Order no. 156/2005 of CNCAN President for approval of regulation on

the classification of radioactive waste [1].

According to this regulation, radwaste is placed in following categories: excluded radioactive waste

(EW), transitional radioactive waste (TW), very low level radioactive waste (VLLW), low and interim

level short lived radioactive waste (LILW-SL), low and interim level long lived radioactive waste (LILW-

LL) and high level radioactive waste (HLW), based on the requirements for assuring the isolation from

biosphere during the radioactive waste disposal.

Our activity refers to low level waste, which contains radioactive and non-radioactive components

that may affect humans and the environment. Therefore the management of such wastes should take into

account both radioactive and non-radioactive components and their associated hazards.

Waste characterization is necessary to ensure the selection and application of appropriate treatment

and conditioning options to satisfy requirements for waste storage and final disposal to protect human

health and the environment.

In defining the required treatment for stabilization of each waste stream, it is necessary to identify

the radioactive and non-radioactive hazardous constituents. For both the radionuclides and the chemically

toxic contaminants, it is important to know their chemical forms and concentrations. It is also important

to have an understanding of their expected behavior in treatment processes, their toxicity, and the

potential release rates of the components from the final waste form.

Treatment of liquid radioactive waste involves the application of several steps such as filtration,

precipitation, sorption, ion exchange, evaporation, membrane separation to meet the requirements both

for the release of decontaminated effluents into the environment and the conditioning of waste

concentrates for disposal [2]

At Cernavoda NPP [3] radioactive wastes waters level 1 – from the laundry, underground drainage

pits of the spent fuel bay, the underground drainage pits of shower rooms, laboratories, floor drainage,

with radioactivity ranging between 3.7x102 Bq/l and 3.7x10-1 Bq/l and level 2 from the heavy water

upgrading system, the equipment decontamination system, rubber objects laundry with radioactivity

ranging between 3.7x104 Bq/l and 3.7x10 Bq/l.. Level 3 are higher radioactive wastes – resulting from the

reactor drainage systems, spent fuel storage bays, spent resin storage vaults, whose radioactivity ranging

between 3.7x106 Bq/l and 3.7x104 Bq/l

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The liquid wastes are collected in five epoxy concrete vaults (50 m3) located in the S/B basement.

Two of the vaults are used for the collection of the level 2 and 3 radioactive wastes; the other 3 vaults are

aimed to collect the level 1 low radioactive waste.

The capacity of each vault (50 m3) represents about the double capacity of the daily average

quantity of radioactive liquid wastes produced by the plant so that the system can accommodate possible

emergencies or abnormal situations of the plant operation.

Normally, the liquid waste radioactivity is sufficiently low so that the discharge may be allowed

without any previous decontamination. For special cases in which such decontamination is mandatory, an

ion-exchange filter, a feed pump and the necessary auxiliary equipment is provided.

The radioactivity of the discharged wastes is continuously measured by Liquid Effluent Monitor

which, when the limit value of 5x103 Bq/l is reached automatically closes the discharge pipe line circuit

and the water is to be directed back into the vault for analysis and decontamination.

Kozloduy NPP [4] have WWER–440 and WWER-1000 reactor units. These were constructed

using a “double-unit” concept — two reactors, one common auxiliary building. The following storage

facilities were foreseen:

·8300 m3 for evaporator concentrates

·2300 m3 for spent resins

·4500 m3for LILW dry solid waste.

Danube river, using its water for plant cooling needs as well as a recipient for liquid releases.

National environmental limits significantly restrict release of boron containing compounds.

The possible solutions for reduction of waste are:

·boric acid recovery system for primary circuit water (primary coolant) for WWER-1000 reactor

units

·evaporator- distillate can be re-used as make-up water.

· partly replaced of insulating wool by reusable material.

· improved maintenance of pump seals and drain valves to avoid leaks.

The technology for conditioning of both solid and liquid waste is based on cementation method,

using steel-concrete container as a package.

The technology for liquid waste conditioning includes following processes:

· Transportation of liquid waste from the storage tanks in the nuclear unit’s auxiliary

· buildings to the liquid waste processing facility

· Concentration of the liquid liquid waste (if necessary) through evaporation,

· Cementation and filling the mixture in a package (steel-concrete container).

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An additional system for decontamination of metal liquid wastes was placed in the facility’s

building.

References indicate that the ion exchange resin is used in the water purification system both in

nuclear research and power reactors. Combined with active carbon, the resin removes the dissolved

elements from water when the nuclear reactor is operating. After its consumption, it becomes a special

type of radioactive waste. The usual treatment of this type of waste is the immobilization with Portland

cement, which is simple and low cost. However, its low capacity of immobilization and the increase

volume of waste have been the challenges.

For the conditioning of ion-exchangers generated from operation of Cernavoda NPP Unit 1

techniques of direct immobilization in cement, bitumen and organic polymers have been experimented

within ICN Pitesti. The selected process for conditioning of spent resins is their bituminization. In view of

final disposal, the container that includes the waste bitumen block is cemented and confined in another

with a larger capacity [5].

According to those above presented, we focused in this project on developing an efficient

technology, based on ion exchange resin (nuclear use type), for decontamination of radwaste waters

resulting from Cernavoda.and Kozloduy NPP.

1.2. Legal basis

The works for elaboration technology for decontamination of aqueous radioactive waste from

Cernavoda and Kozloduy NPP were objective of the 12th activity “Elaboration of a new technology for

wastewater processing and conditioning the organic liquid wastes radioactively contaminated from

nuclear power plants” within cross-border Romania Bulgaria MIS-ETC WATER 161 project – Danube

water integrated management.

Partners who have implemented this activity were:

· National Research and Development Institute for Chemistry and Petrochemistry- ICECHIM

Bucharest Romania – partner 6 in the project;

· Institute for Nuclear Research (under Regia Autonoma Tehnologii pentru Energie Nucleara-

RATEN)Mioveni, Romania- partner 7 in the project;

· University POLITEHNICA of Bucharest, Romania- partner 8 in the project;

· Institute for Nuclear Research and Nuclear Energy-IRNE, Sofia Bulgaria, partner 13 in the

project.

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1.3. The phases of research performed

Development of the technology for radwaste waters decontamination was conducted in two phases,

in the project called subactivities:

· 12.1- Elaboration of a new technology for processing the waste waters from the nuclear plants

Cernavoda and Kozloduy. The results were presented in Technical report No.1- December 2013.

· 12.3Testing the technology for processing the waste waters from the nuclear plants Cernavoda and

Kozloduy final document may 2015.

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2. PERFORMED WORKS AND THEIR RESULTS

2.1. Laboratory works

The goal of activity 12.1, carried on by ICECHIM specialists together with those of Romanian

partners organizations in the project –RATEN-ICN, UPB and Bulgarian partner –PP13-INRNE is to

elaborate the technology for decontamination of radioactive waste waters, with low radioactivity, from

Cernavodă and Kozloduy NPP.

The experimental works carried by Romanian partners had the goal to study the removal of

radionuclides (Co, Cs, Ag,) and some surfactants (linear alkyl benzene sulfonic acid sodium salt –

LABSNa) from aqueous radioactive waste by applying the ion exchange resins- column technique. The

LABSNa anionic surfactant was selected in the experiments because it’s the main component of cleaning

products and in this case, the radioactive waste waters might contain such contaminated ions after the

equipments from the nuclear plants are washed.

Since the decontaminated solution contains a mixture of anionic surfactant (LABSNa) and cations,

both anionic and cationic resins are required for decontamination.

The experiments were performed first in a simulated non-radioactive mono-component or bi-

component system (LABSNa and/or a cation), and then are checked in radioactive multi-component

system.

Bulgarian partner-PP13 has conducted laboratory experimental works for removal radionuclides Sr,

Fe, Ni and Pu from simulated secondary waste from the decontamination processes, applied at Kozloduy.

2.1.1. The laboratory works performed at ICECHIM, in non-radioactive simulated system

The experimental works carried out in simulated non-radioactive mono-component or bi-component

systems have used solutions which contain cations: cobalt and/or sodium, and anions: chlorine and/or

LABS-.

Experiments in simulated non radioactive system were conducted in an installation consisting of:

· glass column containing ion exchange resin with inner diameter of 1.8 cm and height of 20 cm.

Height of the ion exchange resin was varied between 4.5 and 10 cm, and the flow passage of the solution

through the ion exchange resin was varied between 0.83 cm3/min (50 cm3/h) and 6.67 cm3/min (400

cm3/h);

· dropping funnel with a volume of 500 mL, for transfer the solution (simulated) containing cobalt

ions and/or LABSNa into the glass column;

· collection vessel for effluent [6].

In the experiments were tested the following ion exchange resins from Purolite Comp:

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NRW 160- Cationic exchange resin for nuclear use: Ionic form: H+; Strong acid cation; Total

capacity 2.1 eq/L; Moisture retention: 43-48 %. Application: Polishing steam generator blow down and

layering on polishing mixed beds. Primary purification cation beds for delithiation and outage clean

up.Selective for137Cs.

NRW 1600- Cationic exchange resin for nuclear use: Ionic form : H+ Strong acid cation; Total

capacity 2.1 eq/L; Moisture retention: 43-48 %; Mean diameter: 570 ± 50 μm; Application: Cation resin

vessels or layering for added cation capacity.

NRW 5050 Anionic exchange resin for nuclear use: Ionic form: OH- Macroporous Strong Base

Anion; Total capacity 0.9 eq/L; Moisture retention: 53-58 %; Application: Porous structure designed to

give greater resistance to surface fouling in a wide range of nuclear applications.

NRW 3550 Mixed bed products for nuclear use: Ionic form H+/ OH-; Cation capacity: 2.1 eq/L;

Anion capacity: 0.9 eq/L; Equivalent ratio: 1:1 Macroporous/Macroporous; Application: primary

polishing and cleanup systems, steam generator blow down demineralization for high organics water and

boron removal.

Selection of the optimal type of resin was performed by evaluating the following parameters:

·breaking point (selection criteria–highest value) and

·saturation point (selection criteria – the largest volume of waste).

Analytical methods in simulated systems

Determination of Anionic Surfactants (LABSNa) through two-phase titration

The analytical method is based on titration of anionic surfactant with a cationic surfactant (Hyamine

1622 –Benzethonium chloride) in the presence of an indicator which consists of a mixture of cationic dye

(Dimidium bromide) and an anionic dye (Disulphine Blue). Titration takes place in two phases, the

aqueous phase and chloroform phase, at the end point the organic phase turning from pink to blue-gray.

Content of LABSNa [mg/L] = V x f x d x 348 x 0.004

Where: V is the volume of 0.004 M Hyamine solution used at titration; f is the correction index for

0.004 M Hyamine solution; d is the dilution ratio; 348 is molecular weigh of LABSNa.

Determination of cobalt content

The method is based on Co complexation with ethylenediaminetetraacetic acid disodium salt

(dihydrate) in the presence of murexid as indicator.

Content of Co2+ [mg/L] = 5893.0´´´ dfV , where:

V is the volume of 0.01 M of Complexon III used for titration; f is the correction index for solution

0.01 M of Complexon III; d is the dilution ratio; 0.5893 the amount of Co2+ corresponding to 1 mL of

0.01 M of Complexon III.

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Synthetic dynamic laboratory data obtained are presented in Tables 1-3 and fig 1-2 [6].

Table 1. Breaking point of anionic resins in a simulated LABSNa systemNo.crt. Resin

LABSNa initialconcentration,

mg /L

Breaking pointBV/h Volume of the solution passed on

the resin at saturation, mLMinutes Bed volumes

1 Purolite NRW 5050Anionic resin 1,000 200 62.89 13.03 3800

2 Purolite NRW 5050Anionic resin 494.16 2475 157.23 4.44 10400

3 Purolite NRW 3550Mixed resin 490 1790 172.96 5.86 6800

Table 2. Breaking point of cationic resins in a simulated cobalt dichloride systemNo.crt. Resin

Cobalt initialconcentration,

mg /L

Breaking pointBV/h Volume of the solution passed

on the resin at saturation, mLMinutes Bed volumes

1 Purolite NRW 160Cationic resin 979.85 180 47.15 14.29 2200

2 Purolite NRW 160Cationic resin 447.36 940 62.89 4.66 4255

3 Purolite NRW 1600Cationic resin 447,82 720 47.16 4.88 3675

4 Purolite NRW 3550Mixed resin 447.36 150 17.46 5.86 1200

5 Purolite NRW 3550Mixed resin 241.61 735 47.16 5.50 1800

Table 3. Breaking point of mixed resins in a simulated CoCl2 + LABSNa system

No.crt. Resin

LABSNa initialconcentration,

mg /L

Co initialconcentration,

mg/L

Breaking pointBV/h

Volume of the solutionpassed on the resin at

saturation, mLMinutes Bed volumes

Co LABSNa Co LABSNa1 Purolite NRW 3550

Mixed resin 434.16 247.83 865 3530 91.68 288.14 5.73 10200

2 Purolite NRW 160 +Purolite 5050 250 223.68 1100 1270 78.61 212.26 5.33 14700

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Figure 1. The breaking curve of NRW3550 resin for Co2+ (affluent conc.247.83 mg/L) and LABSNa(affluent conc.434.16 mg/L) versus time

Figure 2. The breaking curve of NRW160 and NRW 5050 resins for Co2+ (affluent conc.245.36 mg/L)and LABSNa (affluent conc.244.18 mg/L) versus time

In simulated, non radioactive systems the best results are obtained for:

· Purolite NRW 5050 for anionic surfactants (LABSNa);

· Purolite NRW 160 for Co2+ retention, followed by Purolite NRW 1600 and Purolite NRW 3550;

· Couple Purolite NRW 160 and Purolite NRW 5050 and mixed resin Purolite NRW 3550 both for

anionic surfactant and Co2+ retention.

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2.1.2. The laboratory works performed at ICN Pitesti, with radioactive waste waters type CANDU

The purpose of the work was to improve the process for treatment of liquid radioactive waste

generated during the decontamination operations.

Sorption of radionuclides Cs-137 and Co-60 from real waste solution that are generated at

Cernavodă NPP is studied. The study includes:

- Dynamic experiments and construction of the breakthrough curves for different flow rates

- Determination of the decontamination factor variation with treated waste volume.

The study was performed with one type of spent decontamination solutions with two different

values for radioactive concentration. These solutions represent the category 2&3 radioactive wastes from

Cernavodă NPP, after the filtration of suspended solids. The categories 2 and 3 radioactive wastes from

Cernavoda NPP have an activity concentration ranging between 3.7x10 Bq/l and 3.7x104 Bq/l,

respectively 3.7x104 Bq/l and 3.7x106 Bq/l , in conformity with an operational classification. Usually, the

categories 2 and 3 wastes are collected together. The resulting mixture has an activity ranging between

3.7x101 Bq/l and 3.7x105 Bq/l.

The first radioactive solution was treated on Purolite NRW1600 ion exchanger, at 18 BV/h

(BD=bed volume) and the second solution was treated on the same ion exchanger, at 10 BV/h,

respectively 50 BV/h.

Measurement apparatus and instruments

Characterization of the solutions as concerns the radioactive composition was performed by

measurement concentrations of gamma emitting radionuclides with a high resolution spectrometer

composed of: a HPGe detector with 25% relative efficiency, a low background shielding and a Canberra

spectrometric analyzer, model InSpector. The analysis software was GENIE-2000 from Canberra. The

spectrometer was energy and efficiency calibrated in the range 60 keV – 1500 keV, by considering all

counting geometries used in the experiments. The nuclear properties values used for data reduction were

gathered from the original generic library of the software GENIE 2000 v. 1.2.

The pH, conductivity and TDS measurement was done by using a Mettler Toledo multiparameter

device.

A series of dynamic column experiments was performed with Purolite NRW-1600 resin. The

original solutions with non-adjusted pH were used. This solution represented a real spent decontamination

solution generated in the decontamination process at Cernavoda NPP. It did not contain anionic active

surface agents and the main radioactive contaminants were Ag-110, Co-60, Sb-125, Cs-134, Cs-137, Zr-

95, Nb-95 and H-3. After the filtration of suspended solids, Co-60, Cs-137, H-3 and Ag-110 (only for a

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solution) were detected into spent decontamination solutions. The H-3 can be used like a tracer for the

experimental tests because H-3 is not fixed by an ion exchange process.

The experimental conditions of tests performed were:

- dynamic column experiment

- batch operation

- descendent flow (gravitational)

- fixed bed

- ambient temperature (20±2°C).

The experimental data performed at 18 BV/h, 50 BV/h and 10 BV/h with purolite NRW 1600 are

shown in the table 4-6.

Table 4. Ce/C0 evolution (18 BV/h flow rate)Volume

[BV]Time[min]

Ce/C0 (1)

Ag-110m Co-60 Cs-13785 267 9.7E-02±0.02 1.3E-01±0.06 4.0E-04±6.2E-05169 538 1.2E-01±0.02 1.4E-01±0.06 5.7E-04±8.3E-05231 742 1.2E-01±0.01 1.7E-01±0.06 6.8E-04±1.0E-04358 1158 1.5E-01±0.03 2.5E-01±0.09 1.1E-03±8.7E-05444 1436 1.8E-01±0.03 4.1E-01±0.10 1.9E-03±1.3E-04548 1772 3.2E-01±0.07 7.9E-01±0.19 1.9E-02±1.0E-03656 2139 7.2E-01±0.06 9.6E-01±0.16 2.7E-01±1.1E-02677 2207 7.3E-01±0.10 9.9E-01±0.23 2.7E-01±9.8E-03696 2270 7.8E-01±0.07 1.2E+00±0.19 4.7E-01±1.6E-02717 2347 9.0E-01±0.10 NA 5.5E-01±1.9E-02

Table 5.Ce/C0 evolution (50 BV/h flow rate)Volume

[BV]Time[min]

Ce/C0(1)

Co-60 Cs-137100 123 5.9E-02±0.016 8.8E-04±0.00052183 225 5.7E-02±0.016 < 9.4E-04267 325 < 6.4E-02 < 1.4E-03350 420 1.3E-01±0.026 1.1E-03±0.00052433 520 1.8E-01±0.012 1.9E-03±0.00021517 620 1.9E-01±0.021 1.4E-02±0.00140600 720 2.6E-01±0.017 7.8E-02±0.00713642 770 3.1E-01±0.032 9.7E-02±0.00896667 800 3.3E-01±0.033 1.9E-01±0.01739750 900 3.3E-01±0.033 1.9E-01±0.01739833 1002 3.7E-01±0.033 5.6E-01±0.05123917 1102 5.2E-01±0.049 1.3E+00±0.11489

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Table 6 Ce/C0 evolution (10 BV/h flow rate)Volume

[BV]Time[min]

Ce/C0 (1)

Co-60 Cs-13797 680 5.4E-03±1.6E-03 5.2E-05±3.1E-05

177 1235 6.7E-03±1.6E-03 5.2E-05±2.1E-05258 1810 8.5E-03±1.1E-03 5.2E-05±1.0E-05339 2375 1.0E-02±1.1E-03 5.2E-05±2.1E-05419 2950 1.3E-02±1.6E-03 3.6E-05±3.1E-05500 3530 1.5E-02±2.1E-03 6.2E-05±3.1E-05581 4150 2.3E-02±2.2E-03 8.8E-05±3.1E-05661 4755 N/A 2.3E-03±2.2E-04742 5370 2.6E-02±2.2E-03 5.0E-02±4.5E-03798 5810 4.4E-02±3.9E-03 1.3E-01±1.2E-02823 6075 5.7E-02±5.4E-03 2.6E-01±2.3E-02887 6515 5.3E-01±3.5E-02 2.3E+00±7.8E-02

Decontamination factor (DF) is the ratio of the initial amount of a specified activity to the remaining

amount of residual activity after the treatment. The DF may be expressed relative to gross activity or for a

particular radionuclide.

The decontamination factor of the radionuclides for each fraction of the effluent was calculated and

then plotted vs. the volume of treated solution expressed in the units of bed volumes of the ion exchange

resin (fig. 3 and fig 4).

Figure 3. Decontamination factor for Cs-137

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Figure 4. Decontamination factor for Co-60

Experimental results and decontamination factors obtained show the applicability of Purolite NRW-

1600 ion-exchange resin to treatment of liquid radioactive waste generated during the decontamination

operations for the selected experimental conditions.

2.1.3. The laboratory works performed at IRNE Sofia, with radioactive wastewater type Kozloduytechnology

The experimental works were conducted with following resins: Purolite A-100, Purolite A-500,

Purolite NRW-5070, Purolite NRW-160, Purolite NRW-5050, Purolite NRW-3550, Purolite NRW-1600,

Amberlite IR 120, Amberlite IRA 410.

Preparing the experimental tasks:

· Experiments for optimization of exchange columns loaded with Purolite NRW-160, Purolite

NRW-1600, Purolite -3550, Purolite- 5050 and Amberlite IRA410 Cl resins in respect to the industrial

decontamination solution for steels.

· Isotopic and elemental experiments with various matrix solutions that correspond to those applied

in decontamination activities at NPP.

· Tests and experiments on procedures for waste water processing and conditioning of

nonradioactive solutions containing target elements Cs, Co, Mn, Fe, Ni, Sr and Sm with different pH.

Prepared Radioactive Samples (Test solutions):

· Test solutions containing known amount of standard cocktails including gamma rays emitting

radionuclides (gamma emitters) such as 137Cs, 57Co, 60Co, 241Am, 85Sr, 109Cd, 123mTe and 88Y;.

· Test solutions acting as “Simulated waste solutions” like secondary waste from the

decontamination processes, including pure beta emitter. (Отделни разтвори for 90Sr, 55Fe, 63Ni, 241Pu)

with different decontamination chemical agents. The prepared test solution was left for certain time to

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become fully homogeneous. When using a tracer of radioactive isotope with a known activity, with the

aim for obtaining measurable mass quantities, also a stable isotope of the element was added to the

solution as carrier. The test solution was allowed to stand for achieving also the isotopic exchange. The

homogenised solution was then let through the bed at a given service flow rate (bed volumes per hour).

Time and service flow rates are specified in the corresponding details of the particular experiment.

The prepared test solution was left for certain time to become fully homogeneous. When using a

tracer of radioactive isotope with a known activity, with the aim for obtaining measurable mass quantities,

also a stable isotope of the element was added to the solution as carrier. The test solution was allowed to

stand for achieving also the isotopic exchange. The homogenised solution was then let through the bed at

a given service flow rate (bed volumes per hour). Time and service flow rates are specified in the

corresponding details of the particular experiment

Instruments and methods:

Standardised and well-known methods in the practice are used, such as gamma spectrometry, low

background radiometry of total beta activity and isolated radiochemical radioactive strontium, liquid

scintillation spectrometry of beta emitters (3H, 14C, 90Sr, 63Ni, 55Fe, 241Pu) and alpha – spectrometry of

transuranium elements.

The measuring methods are:

· Universal device UVJ-01 for measurements of total beta activity of samples. The device uses a

Geiger detector with 0.23 cps/Bq efficiency for Sr90/Y90 (MK-30 Measuring Chamber) and background

≤ 2 cps.;

· Multi-Detector Systems PIC-MDS-8 for measurements of total a/b activity. The device features

low background 0.03 – 0.07 cpm for alpha and 0.4 – 0.7 cpm for beta activity. Efficiency varies between

35 % and 55 % depending on energy and geometry of sources.

· Two gamma radiation spectrometers for determination of specific activities of gamma-ray

emitting radionuclides- ORTEC system equipped with a high purity Ge (HPGe – GMX 50P4) detector in

low-background shielding. The system provides an energy resolution of 2.3 keV and relative efficiency of

54.9 % for the 1332.5 keV 60Co line and DSP spectrometer equipped with a coaxial high purity Ge

detector in low-background shielding. The system provides an relative efficiency of 20.0 % for the 1332.5

keV 60Co line.

· Packard Instruments low-level Tri-Carb® 2770 TR/SL system for liquid scintillation analysis. The

analyzer can detect small amounts of beta radioactivity. Reliable results are produced by this analyzer

even for quenched samples. This system, specially configured for extremely low level radioactivity

analysis, includes several features to reduce the background and to improve counting performance. A

14

built-in electrostatic controller minimizes the effects of static electricity. The automatic data reduction

feature includes averaging of repeat sample counts, percent Coefficient of Variation (CV), low count

rejection, and result normalization;

· Varian model Vista MPX with radial torch and CCD array detector for inductively coupled

plasma Optical Emission Spectroscopy (ICP-OES). The system is used for simultaneous quantitative

measurement of all elements in the sample analyzed. ICP-OES is analytical technique used for the

detection of trace elements and their mass concentrations in the samples.

· Varian 810-MS – floor-mounted inductively coupled plasma mass spectrometer (ICP-MS). Its

feature patented 90 degree reflecting ion optics system for gigahertz sensitivity and low background and

interferences. The Varian ICP-MS system includes a sample introduction system, solid state 27 MHz RF

generator, and patented Turner Interlaced Coils. Full PC control of plasma positioning, triple stage

vacuum system, all plasma gas flows, mass analyzer, and Discrete Dynode Electron Multiplier (DDEM)

detector is also included. Unique DDEM detector provides nine decades of dynamic range in an all-digital

pulse design. The Varian 810-MS system also features a Collision Reaction Interface (CRI) providing

interference-free analysis using simple collision and reaction gases.

The experimental setup for evaluation of the resin capacity is presented in fig 5

Figure 5. Simplified drawing of experimental setup for determinationof resin capacity in laboratory

Experimental works followed several phases:

· Decontamination capacity of resins tested with a simulated model solution of 0.5% HNO3, which

is used widely in NPP practice for decontamination of steel and stainless steel surfaces.(Fig.6-Fig 7).

Experimental results obtained show (Fig 6) the applicability of ion-exchange resins NRW-1600, NRW-

160 and NRW-3550 to decontamination of the above key nuclides in the selected experimental

conditions.

15

Figure 6. Histogram presentation of DF of NRW1600, NRW160, NRW3550for 137Cs, 60Co, and 241Am

Resin Amberlite IR 120 presents a lower activity than the above tested resins.

Figure7. Histogram presentation of DF of AMBERLITE IR 120 Na and AMBERLITE IRA 410 Cl forCo, Fe, Ni and Sr

· Decontamination capacity of cation resin Amberlite-IR120 for specific decontaminants used inNPP practice

The following decontaminants were selected:

- 0.5% HNO3 - nitric acid is used for decontamination of steel components and stainless steel

surfaces.

- 10% H2C2O4 – oxalic acid is widely used for decontamination of wood and concrete.

- 3% NaOH – is used for decontamination of porcelain and many polymer materials, such as PE,

HDPE, PVC etc.

- Surfactant solution – most commonly used for decontamination of textile materials.

The contaminants were labelled with known amount of 90Sr, which is a beta-emitting radionuclide

with great contribution to the radioactivity of low-level nuclear wastes.

16

Figure 8. Histogram presentation of DF of AMBERLITE IR 120 Na for selected decontaminants.

The resin Amberlite-IR120 is suitable for decontamination of 90Sr originating from decontaminant

matrices based on oxalic acid and nitric acid. Respectively, this resin is not appropriate for

decontamination of low-level wastes originating from decontaminants NaOH and Surfactant solution.

· Determination of 90Sr selectivity in a matrix of 10 % H2C2O4 passed through a column loaded

with Purolite-NRW 5070, Purolite-A-100, Purolite-A-500, Purolite-NRW 3550, Purolite-NRW 1600,

Purolite-NRW 160 (Table 7).

Table 7. DF and RF of resins Purolite-NRW 5070, Purolite-A-100, Purolite-A-500, Purolite-NRW 3550,Purolite-NRW 1600 and Purolite-NRW 160 for Sr

Parameter: Purolite-NRW5070

Purolite-А100

Purolite-А500

Purolite-NRW3550

Purolite-NRW1600

Purolite-NRW160

Sr (before column), [cps/g] 7.831* 7.831* 7.831* 7.831* 7.831* 7.831*Sr (after column), [cps/g] 6.135* 4.074* 5.513* 0.002* 0.004* 0.020*RF(Retention factor), [%] 21.66 47.98 29.60 99.97 99.95 99.74DF(decontamination factor) 1.3 1.9 1.4 3915 1958 392

*- Value calculated as the mean of 5 aliquots. The combined standard uncertainty of these aliquots is within 2.5%.

Resins NRW 1600 and NRW 160 are suitable for the effective retention of Sr in solution with a

matrix of 10 % oxalic acid under the given conditions (mass concentration of Sr of 1.5mg, ratio

resin/solution – 1/10 and a flow rate of 10 bed volumes/hour).

· Determination of the selectivity and capacity of Purolite-NRW 160, Purolite-NRW 1600 and

Purolite-NRW 3550 for 55Fe in matrix 0.5% HNO3 (Table 8)

Table 8. DF and RF of resins Purolite-NRW 3550, Purolite-NRW 1600 and Purolite-NRW 160for 55Fe after passing by 10 and 20 column volumes of model solution

Parameter: Purolite-NRW160

Purolite-NRW1600

Purolite-NRW3550

Fe (before column), [cps/g] 0.2347* 0.2347* 0.2347*Fe (after column 10 c.v.) [cps/g] 0.000523* 0.000458* 0.0051*Fe (after column 20 c. v.), [cps/g] 0.052901* 0.105835* 0.0108*RF(Retention factor after 10 c. v.), [%] 99.78% 99.80% 97.81%

17

RF(Retention factor after 20 c. v.), [%] 77.46% 54.90% 95.39%DF (decontamination factor after 10 c. v.) 448 513 46DF (decontamination factor after 20 c. v.) 4 2 22

· Determination of the capacity of Purolite-NRW 1600 with respect to 60Co, 137Cs and 241Am, from

a matrix of 0,5% HNO3, in the standby saturation of the resin.

· Determination of the selectivity of ion exchange resins Purolite NRW-1600 for radionuclides137Cs, 241Pu, 94Nb and 63Ni contained in model solutions. Test solutions containing known amount of

standard cocktails including the radionuclides 137Cs, 241Pu, 94Nb and 63Ni were prepared. The conclusions

are the following:

- Experimental results and retention and decontamination factors obtained show the applicability of

ion-exchange resins NRW-1600 to decontamination of 241Pu under the selected experimental conditions.

- When comparing the results for the retention and decontamination factors for the β-emitter 241Pu,

it is clear that the possibility of effective retention decreased with increasing column volumes omitted.

- The results for the retention and decontamination factors for the β-emitter 94Nb are similar to that

for 241Pu. The possibility of effective retention decreased with increasing column volumes omitted.

- The ion-exchange resins NRW-1600 and NRW-160 are suitable for decontamination of 63Ni

originating from decontaminant matrix based on nitric acid.

· Determination of the capacity of Purolite-NRW 1600 with respect to 60Co, 137Cs and 241Am, from

a matrix of 0,5% HNO3, in the standby saturation of the resin.

18

Figure 9. Capacity of Purolite-NRW 1600 with respect to 60Co, 137Cs and 241Am

· Determination of the selectivity of ion exchange resins Purolite NRW-1600 for radionuclides137Cs, 241Pu, 94Nb and 63Ni contained in model solutions. Test solutions containing known amount of

standard cocktails including the radionuclides 137Cs, 241Pu, 94Nb and 63Ni were prepared. The conclusions

are the following:





- The results for the retention and decontamination factors for the β-emitter 94Nb are similar to that


- The ion-exchange resins NRW-1600 and NRW-160 are suitable for decontamination of 63Ni

originating from decontaminant matrix based on nitric acid.

19

0

100

200

300

400

500

600

700

800

900

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53

Channel

Cou

nts

Figure 10. β – spectrum of 241Pu in a solution with matrix 0.5% HNO3

0

50

100

150

200

250

300

350

400

1 4 7 10 13 16 19 22 25 28 31 34 37 40 43 46 49 52

Channel

Cou

nts

Figure 11. β – spectrum of 241Pu in a solution with matrix 10% H2C2O4

· Gamma spectrometry of 137Cs:

The resin NRW-1600 was preliminary conditioned with 5 column volumes of pure solution of 0.5%

HNO3 (used for decontamination of surfaces).

The test solution contained 0.5% HNO3 to which a known quantity of radioactive 137Cs was added.

With the aim for obtaining measurable mass quantities also a stable isotope of Cesium was added as

carrier. The model solution prepared this way was left for 10 days to achieve isotopic exchange and to

become fully homogeneous. The homogenised solution was then let passing through the bed at a service

flow rate of 10 BV/h (bed volumes per hour) at room temperature.

Figures 12 and 13 show the results obtained by gamma spectrometry of the initial solution and the

eluate after the column.

20

Figure 12. Retention factor of NRW-1600 with respect to 137 Cs in 0.5% HNO3

Figure 13. Decontamination factor of NRW-1600 with respect to 137Cs in 0.5% HNO3

· Liquid-Scintillation Spectrometry of 241Pu:

The bed resin NRW-1600 was preliminary conditioned with 5 column volumes of pure solution of

0.5% HNO3.

The resin was tested with an initial solution containing 0.5% HNO3 to which a known quantity of

radioactive 241Pu was added. The model solution prepared this way was left for 3 days to achieve isotopic

exchange and to become fully homogeneous. The homogenized solution was then let through the bed at

service flow rate of 10 BV/h (bed volumes per hour) at room temperature. For the 241Pu activity

determination, from the solution were taken by three aliquots of 10 ml each mixed with 10 ml scintillation

cocktail ULTIMA GOLD™ LLT in the standard vessel for scintillation measurements. Analysis of

aliquots of material before the bed and eluted solution was carried out by means of the Packard

Instruments low-background Tri-Carb® 2770 TR/SL system (PACKARD TRI-CARB 2770 TR/SL) for

liquid scintillation analysis. The experimental data were treated using the software package Quanta Smart

and Microsoft Excel.

Figure 14 shows the results for the retention and decontamination factors for the β-emitter 241Pu.

21

Figure 14. Retention and decontamination factors with respect to 241 Pu in 0.5% HNO3





· Liquid-Scintillation Spectrometry of the β-emitter 94Nb

The bed resin NRW-1600 was conditioned with 5 c.v. of pure solution of 0.5% HNO3.

The test was carried out with an initial solution containing 0.5% HNO3 to which a known quantity

of radioactive 94Nb was added. With the aim for obtaining measurable mass quantities also a stable

isotope of Niobium was added as carrier. The experiment was performed by means of the pre-calibrated

as specified above Packard Instruments low-background Tri-Carb® 2770 TR/SL system at room

temperature using a model solution prepared in very similar way as for 241Pu.

Figure 15 shows the results for the retention and decontamination factors for the β-emitter 94Nb

22

Figure 15. Retention and decontamination factors with respect to 94Nb in 0.5% HNO3

The results for the retention and decontamination factors for the β-emitter 94Nb are similar to that


· Comparative measurements on the decontamination capacity of ion-exchange resins Purolite

NRW-1600 and Purolite NRW-160 for the beta-emitter 63Ni in a matrix of 0.5% HNO3

Experiments were conducted to compare the capacity of above said resins for 63Ni. To a known

volume of 0.5% nitric acid solution were added known quantities of the radioactive isotope 63Ni and

stable isotope of nickel as the carrier The solution was allowed to mature in order to achieve isotope

exchange and complete homogenization. The solution was then passed through the said resins,

conditioned preliminary with 5 column volumes of fresh solution of 0.5% nitric acid.

The samples were passed through the column at a speed of 25 column volumes per 1 hour. Fifty

column volumes were passed through in 5 portions of 10 c. v. each one.

The decontamination factors for 63Ni, under the particular conditions and matrix, were determined

by liquid scintillation spectrometry using the spectrometer PACKARD TRI-CARB 2770 TR/SL.

For determining the activity of each solution of 63Ni, by 3 aliquots of 2 ml each one were taken,

which were mixed with 18ml of scintillation cocktail ULTIMA GOLD XR in a plastic cuvette.

The laboratory tests show that NRW-1600 and NRW-160 have a high selectivity for 63Ni. The

decontamination factors for 63N of the two resins are presented in Figure 16.

23

0

30

60

90

120

150

180

210

240

10 20 30 40column volumes

DF

Purolite NRW-1600Purolite NRW-160

Figure 16. DF of Purolite NRW-1600 and Purolite NRW-160 for 63Ni in 0.5% HNO3 matrix

The results show that under the conditions of the experiment:

- The resin Purolite-NRW160 can be used for decontamination of the radionuclide 63Ni, however

the resin Purolite-NRW1600 is more effective for 63Ni sorption of matrix solution with 0.5% HNO3.

- The capacity of both resins decreases with increasing volume of the passed through them model

solution.

· Gamma spectroscopic determination of the decontamination capacity of Purolite NRW-1600for 241Am, 137Cs and 60Co (Part 2.)

First, the two gamma spectrometric systems with high purity Ge detectors of 50% and 20 % relative

efficiency for the 1332 keV line of 60Co were recalibrated with respect to energy and efficiency and

comparative experiments for optimization of accuracy and counting time were carried out.

For the experiments test solutions were taken after transmission of the initial solution through

columns loaded with preliminary conditioned resin volume of 2 ml: 200, 225 и 250 c. v. The samples

after transmission through the columns were with a volume of 46.580 ml (for 200 c.v.); 50.070 ml (for

225 c.v.) and 49.840 ml (for 250 c.v.). The results from the gamma spectrometry measurements (Bq.l-1,

SE %) are displayed below.

Table 9. Results from the gamma spectrometry measurements (Bq.l-1, SE %)Sample Sample type 241Am,

Bq.l-1, (SE %)137Cs,

Bq.l-1, (SE %)60Co,

Bq.l-1, (SE %)S – 0 Initial solution 216000 ± 11020

(5.1 %)112600 ± 5193

(4.6 %)57110 ± 2354

(4.1 %)S – 1 25 column

volumes344.9 ± 27.6

(8.0 %)< 5.82 73.21 ± 7.76

(10.6 %)S – 2 50 column

volumes41.20 ± 4.29

(10.4 %)10.25 ± 2.02

(19.7 %)113.30 ± 5.55

(4.9 %)S – 3 75 column

volumes33.61 ± 3.42

(10.2 %)16.82 ± 1.93

(11.5 %)174.40 ± 7.42

(4.1 %)

24

S – 4 100 columnvolumes

35.75 ± 4.50(12.6 %)

28.47 ± 1.31(4.6 %)

179.53 ± 7.18(4.0 %)


37.36 ± 3.47(9.3 %)

41.34 ± 2.86(6.9 %)

184.4 ± 7.7(4.2 %)


37.73 ± 4.60(12.2 %)

42.81 ± 4.49(10.5 %)

1243.0 ± 48.96(3.9 %)


38.13 ± 3.24(8.5 %)

43.14 ± 3.11(7.2 %)

1359.0 ± 53.00(3.9 %)


39.49 ± 3.39(8.6 %)

251.70 ± 14.13(5.6 %)

4000.00 ± 155.30(3.9 %)

Figure 17. Variation of 241Am, 137Cs and 60Co concentration after handling with the Purolite NRW-1600resin.

Figure 18. Histogram presentation of DF of Purolite NRW-1600 for 241Am, 137Cs and 60Co

25

Figure 19. Decontamination factors (DF) of ion-exchange resin Purolite NRW-1600 for radionuclides241Am, 137Cs and 60Co.

Figure 20. Histogram presentation of DF of Purolite NRW-1600 for 241Am, 137Cs and 60Co

The obtained removed activity data show high percentage of retention of analyzed radionuclides

(241Am, 137Cs, 60Co). This result is confirmed also by the for the decontamination factors (DF) data. For

decontamination of the radionuclide 63Ni the resin Purolite-NRW160 can also be used, however the resin

Purolite-NRW1600 is more effective for 63Ni sorption of matrix solution with 0.5% HNO3.

Therefore, it can be concluded that under the conditions of present experiments the resin Purolite

NRW-1600 is good cleaner for liquid radioactive wastes containing radionuclides such as 241Am, 137Cs,60Co and 63Ni.

26

· Determination of the amount of retained Co2+ in Purolite NRW 160 ion exchange resin

Experiments were conducted to determine the degree of adsorption of Co2 + ions in the resin Purolite

NRW 160 by measuring the time of mixing, in which the equilibrium is established. The amount of Co2 +

ions retained in the volume of the resin is determined.

The preparation of the experiment included a theoretical study and preparation of the initial solution

of known activity of 60Co;

Seven identical samples containing 100 mL of the first solution and 1 g of resin were prepared.

Description of the prepared samples:

· М1=100.03g (20,892 Bq.g-1)

· М2=100.03g (20,892 Bq.g-1)

· М3=100.03g (20,892 Bq.g-1)

· М4=100.02g (20,890 Bq.g-1)

· М5=100.04g (20,894 Bq.g-1)

· М6=100.00g (20,886 Bq.g-1)

· М7=100.00g (20,886 Bq.g-1)

Over different periods of time (5, 25, 45, 60, 70, 90 and 120 minutes) the samples were mixed and

gamma spectrometric determination of the specific activity of 60Co at the seven samples was performed.

Before the measurements a recalibration of gamma spectrometric instrumentation and gamma

spectrometric measurements on model samples was carried out. Treatment of the obtained spectra and

analysis of results was performed. Discussion on the results was carried out with a view of optimizing the

conditions of the experiments at next initial concentrations of solution.

Taking into account the conclusions from the above experiments the work continued with the next

step as follows:

Starting solution of known 60Co activity (20.13 Bq.g-1 ) has been prepared from which 7 identical

samples were made –each containing 100 mL from the initial solution and 1 g Purolite NRW 160. The

samples were homogenized using an electromagnetic stirrer for different time periods: 5, 25, 45, 60, 70,

90 and 120 minutes. After that by an aliquot of 20 mL from the homogenized solutions were taken. These

aliquot samples were diluted to a volume of 100 mL and the specific activity of 60Co has been measured.

The results are presented in Table 10.

27

Table 10. 60Co specific activity (Bq.g-1), DF and removed activity (%)

DF = Radiation level before decontamination/Radiation level after decontamination.** % Activity removed = (1 – (1/DF))100

The gamma spectrometry measurements were carried out using HPGe detector (Ortec) with 54 %

counting efficiency and energy resolution of 2.3 keV (at 1332 keV).

0 5 25 45 60 70 90 120Time [min]

0

2

4

6

8

10

12

14

16

18

20

22

Co-

60 [B

q.g-1

]

Figure 21. 60Co specific activity in solution after interaction with the Purolite NRW 160 resin

0 5 25 45 60 70 90 120Time [min]

0

10

20

30

40

50

60

70

80

90

100

Rem

oved

act

ivity

, [%

]

Figure 22. Effect of contact time on removal of 60Co by Purolite NRW 160 resin

Sample Co-60removal time,

[min]

Samplemass,

[g]

Specific activity ±Uncertainty,

[Bq.g-1]

DecontaminationFactor (DF*),

Removedactivity**,

[%]S-1 5 100.47 14.56 ± 0.39 1.38 27.54S-2 25 100.52 12.02 ± 0.40 1.67 40.12S-3 45 101.23 9.19 ± 0.33 2.19 54.34S-4 60 100.24 5.40 ± 0.14 3.73 73.19S-5 70 100.09 3.03 ± 0.08 6.64 84.94S-6 90 100.45 2.98 ± 0.08 6.75 85.18S-7 120 100.66 3.01 ± 0.08 6.69 85.05

28

· Determination of the amount of retained Co2+ in Purolite NRW 160 ion exchange resin by

means of spectrophotometry

A spectrophotometric technique has been applied for determination of cobalt (Co) as Co2+. A Hach

DR 2800 Portable Spectrophotometer has been used. For the experiment a starting solution with a

concentration of 1,928 mg/ml of Co has been prepared.

The samples were pretreated with a cation exchange resin Purolite NRW 160, and the

concentrations of Co2 + were measured at different period of times: 5, 25, 45, 60, 90 и 120 min.

For the spectrophotometric analyses to the samples were added additional reagents (EDTA, PAN,

phthalate-phosphate).

After the sample is buffered and pyrophosphate is added to mask any Fe3+, the cobalt reacts with 1-

(2-Pyridylazo)-2-Naphthol indicator (PAN) and the indicator forms complexes with most of the metals.

After color development, EDTA is added to destroy all metal-PAN complexes except cobalt. In the

experiments the concentrations of 1,88 and 1,82 of Co are used. The results are shown in Fig.23.

Figure 23. Time dependence for the initial concentration of 1,88 mg/L (I) and 1,82 mg/L (II) of Co aftertreatment with a cation exchange resin Purolite NRW 160

· Simultaneous determination of the amount of retained Cobalt and Nickel byspectrophotometry (continuation)

The 1-(2 Pyridiazo)-2-Naphtol (PAN) method has been successfully adapted and applied for

determination of the amount of Co2+ and Ni2+ ions after treating test solutions with ion exchange resins.

The method has been adapted on the basis of our previous experience in the analysis of presence of

transition elements in water and wastewater. The Hach DR 2800 Portable Spectrophotometer has been

used.

Short description: After pyrophosphate is added to buffer the sample and mask any Fe3+, the nickel

ions react with the 1-(2-Pyridylazo)-2-Naphthol indicator. The indicator forms complexes with most

metals present. After color development, EDTA is added to destroy all metal-PAN complexes except

nickel and cobalt. Spectrophotometers automatically adjust for cobalt interference by measuring the

29

absorbance of the sample at both 560 nm and 620 nm. This method is unique because both nickel and

cobalt can be determined on the same sample with a spectrophotometer. The results for nickel and cobalt

are shown in Figure 24.

Figure 24. Most of the cobalt and nickel are absorbed from the resin on the 5-th minute.

In conclusion the resins Purolite NRW 160 and Purolite NRW 1600 are suitable for

decontamination of radwaste water.

2.1.4. Modelling the radions separation

In order to model the mass transfer of anion in the resin, the following assumptions have been

adopted:

· The resin is a packed bed of spherical particles

· Particles diameter are uniform

· The flow across the fixed bed is symmetrical and uniform

· Anion retention occurs at the particles surface all along the fixed bed

· The anion adsorption process is isothermal

· The physical properties of the solution have been assumed to be equal to those of pure water at

room temperature

· Mass transfer only occurs in the axial direction of the resin bed

Mass transfer model

The evolution of the anion concentration is represented with the mass conservation equation as

following:2

ax 2

C q C Cu D 0t t z z

¶ ¶ ¶ ¶e × + + e × × - e × × =

¶ ¶ ¶ ¶(1.1)

30

In this equation, the first two terms represent accumulation in the pores and at the particles surface

(due to adsorption), whereas the two last terms take into account convective and diffusional mass

transport in the bed axial direction.

The variation of the anion concentration C in the radial direction of the bed has been neglected

because mass transfer in this direction is faster than in the path of the length of the bed Lp given by:

Lp » tax×L (1.2)

Where tax, is the tortuosity in the axial direction set equal to 1.4 for a bed of spherical particles [7]

and L is the bed height.

This has been validated by calculating the time to cross the column diameter (dc/kc) and comparing

it with the time to cross the bed length (Lp/u) [8].

p c

c

L du k

> (1.3)

The local mass transfer coefficient kc has been evaluated with the relationship:31

axg

axc ScRe03.164

dDk ú

û

ùêë

é××÷

øö

çèæ

e+×= (1.4)

with:

h×e×r×

=ud

Re g (1.5)

and

axax D

Sc×rh

= (1.6)

The velocity of flow in the pores is given by:

2c

4 Fud×

=p × × e

(1.7)

where F is the feed volumetric flow rate, dc column diameter and e the bed porosity.

The concentration of the ion in the adsorbed phase is given by Langmuir equation:

am

a

K Cq q1 K C

×= ×

+ ×(1.8)

where Ka is the equilibrium constant and qm is the resin capacity.

The initial and boundary conditions have been set as below:

At t = 0, for z ≥ 0 , C = 0

At z = 0 for t > 0, F axCu C u C Dz

¶× = × - ×

¶

31

At z = L for t > 0, C 0z

¶=

¶

Mathematical resolution

Using the experimental data provided by our partners, the equation (1.1) was used to estimate the

parameters of separation characteristics: maximum concentration in solid (qm), axial dispersion

coefficient (Dax), and the coefficient of adsorption (Ka).

The interstitial velocity was calculated from the volumetric flowrate:

2c

4 Fud×

=p × ×e

where F is the fluid volumetric flowrate, m3/s;

dc – column diameter, m;

e – bed porosity, m3/m3;

For the parameter estimation, in the Matlab program was implemented an optimization procedure,

by genetic algoritms. The calculation is done when are no signs of significant improvement of the

solution. This strategy involves algorithm monitoring and verification of the progress.

Some results are presented in the table 11

Table 11.Estimated parametersIonic

speciesFlowrate(BV/h)

Dax (m2/s) qm (g/L) Ka (L/g)

RATEN-ICNNRW 1600

Co-60 10 2.96·10-9 9.556 112.3218 2.87·10-9 3.724 91.0418 2.8527·10-9 4.3025 78.808550 1.047·10-9 7.6611 92.30

Cs-137 10 1.342·10-9 7.593 102.7018 2.696·10-9 5.629 109.7550 1.055·10-9 7.5 91.04

Ag-110m 18 2.94·10-9 6.17 76.08ICECHIMNRW 1600

Co-59 4.41 7.3641·10-7 189.0 1.106NRW 160

Co-59 0.24 4.7101·10-6 231.71 0.37NRW 3550

Co-59 5.736 6.7·10-9 506.23 1.56LabsNa 5.736 1.47·10-6 824.98 1.30

32

Verification of the scale up plant operation

The estimated parameters were used to calculate the concentrations variations in time and space for

the new ion exchange columns, after the scale up, and comparing with the results provided by the PP7

partner – RATEN –ICN.

It must be noted that the radions initial concentrations of the real wastewater were not been known,

only the order of magnitude.

For the Co-60 separation, for example, the calculated and the experimental effluent concentrations

Ce (relative to the feed concentration C0) are presented in figures 25 and 26.

Figure 25. Co-60 effluent concentration versus time

Figure 26. Co-60 concentration versus time and bed length

33

It can be concluded that for this specific radwaste sample the bed volume was overestimated, an

important length of the bed remaining unused after 1 day of operation and can be used for future

treatments of other radwaste quantities.

2.2 Pilot scale testing of the decontamination process of radioactive wastewater from Cernavoda

NPP

2.2.1 Ion exchange resins used within experiments

To determine the type of ion exchangers that can be used successfully to treat aqueous radioactive

waste from Cernavoda NPP were recommended the ion exchanger pairs Purolite NRW160 + Purolite

NRW5050 and PuroliteNRW1600 + Purolite NRW5050, and the mixed bed resin Purolite NRW3550.

For pilot scale tests has been selected pair Purolite NRW1600 + Purolite NRW5050, combination that led

to the best results both to the radioactive species retention, and to the retention of the surfactants present

in aqueous radioactive waste arising from the decontamination processes.

2.2.2.Aqueous radioactive waste processed

Pilot-scale experiments were performed using liquid radioactive waste from Cernavoda NPP.

The radioactivity of the liquid waste processed at pilot scale is governed mainly by 137Cs, 134Cs,60Co, 110mAg, 125Sb, 95Zr, 95Nb, 3H and to smaller extent by 54Mn, 144Ce, 113Sn, 106Ru, 46Sc and 51Cr. After

filtration of suspended solids, 60Co, 134,137Cs, 54Mn, 95Zr, 95Nb and 3H were detected into aqueous

radioactive waste and the global gamma activity level was ~103 Bq/l.

The H-3 is not removed by ion exchange.

2.2.3 Methods and apparatus

To evaluate the effectiveness of the treatment process were determined: radioactive composition,

conductivity, pH and total dissolved salts content (TDS). Analyses were performed for the initial waste

(influent) and for elute (effluent).

The radioactive composition of aqueous radioactive waste was performed by measurement of

gamma emitting radionuclides concentrations with a high resolution spectrometer composed of a HPGe

detector with 25% relative efficiency, a low background shielding and a Canberra spectrometric analyzer,

model InSpector. The analysis software was GENIE-2000 from Canberra. The spectrometer was energy

and efficiency calibrated in the range 60 keV – 1500 keV, by considering all counting geometries used in

the experiments. The nuclear properties values used for data reduction were gathered from the original

generic library of the software GENIE 2000 v. 1.2.

34

The pH, conductivity and TDS measurement was performed with a Mettler Toledo multiparameter

device.

Analyses were performed for the initial waste (influent) and for effluents which must be discharged.

2.2.4 Test description

Pilot-scale test has treated 1.2 m3 aqueous radioactive waste from Cernavoda NPP. After separation

of the organic phase, settling and filtering suspension, the wastewater is treated by ion exchange. The ion

exchange resins used were cationite Purolite NRW1600 and anionite Purolite NRW5050.

The effluent is collected in containers with a capacity of 130 liters. At intervals of 60 minutes they

were determined, on fractions of effluent, pH, conductivity, gamma emitting radionuclides concentration

and tritium concentration. Effluent monitoring results of chemically and radioactively will indicate when

the resin is spent.

At the end of the test, representative effluent samples were collected; their physico-chemical

parameters were analyzed and were determined concentrations of tritium and gamma emitting

radionuclides.

The experimental conditions of tests at pilot scale performed were:

- dynamic experiment (column)

- batch operation (8 h daily)

- descendent flow (gravitational)

- fixed bed

- ambient temperature (20±2°C).

The characteristics of cationit column (Purolite NRW1600) were:

- Di = 160 mm

- Hcoloana = 800 mm

- hlq.col. = 640 ± 10 mm

- hpat = 500 mm

- Vpat = 10 L.

The characteristics of anionit column (Purolite NRW5050) were:

- Di = 160 mm

- Hcoloana = 800 mm

- hlq.col. = 640 ± 10 mm

- hpat = 300 mm

- Vpat = 6 L.

35

The test was performed at 10 BV/h flow rate for anionit and 6 BV/h flow rate for cationit. The

influent characteristics were pH = 6.56; conductivity = 0.226 mS/cm; TDS = 0.108 g/l.

Effluent monitoring

In the table 12 are presented concentration values emitting radionuclides in samples of effluent

normalized range from gamma emitting radionuclides influent concentration.

Table 12. Ce/C0 evolution for pilot test

SampleTimemin

Vol.BV

Ce/C0(1)

54Mn 110mAg 134Cs 137Cs 60Co (2) 95Nb (2) 95Zr (2)

Effluent 1 80 5 <0.44 N/A N/A N/A < 0.01 N/A N/AEffluent 2 140 9 <0.47 N/A N/A N/A < 0.02 N/A N/AEffluent 3 200 13 <0.47 N/A N/A N/A < 0.01 N/A N/AEffluent 4 280 18 <0.52 <0.28 <0.10 <0.01 0.01±0.007 < 0.07 < 0.22Effluent 5 340 21 <0.52 <0.28 <0.10 <0.01 0.01±0.007 < 0.06 < 0.22Effluent 6 406 25 <0.54 <0.28 <0.10 <0.01 < 0.01 < 0.07 < 0.21Effluent 7 466 29 <0.52 <0.28 <0.10 <0.01 < 0.01 < 0.06 < 0.21Effluent 8 536 34 <0.54 <0.28 <0.10 <0.01 < 0.01 < 0.07 < 0.23Effluent 9 596 37 <0.55 <0.28 <0.10 <0.01 < 0.01 < 0.05 < 0.25Effluent 10 666 42 <0.51 <0.27 <0.09 <0.01 0.02±0.005 0.07±0.031 0.13±0.095Effluent 11 731 46 <0.51 <0.27 <0.09 <0.01 0.02±0.005 0.10±0.025 0.11±0.046Effluent 12 801 50 <0.52 <0.28 <0.09 <0.01 0.03±0.005 0.09±0.025 0.14±0.047Effluent 13 861 54 <0.52 <0.28 <0.10 <0.01 0.02±0.005 0.13±0.031 0.14±0.048Effluent 14 931 58 <0.52 N/A <0.10 <0.01 0.02±0.005 0.15±0.032 0.19±0.058Effluent 15 991 62 <0.54 <0.24 <0.10 <0.01 0.02±0.005 0.16±0.036 0.22±0.054Effluent 16 1061 66 <0.52 <0.23 <0.10 <0.01 0.02±0.006 0.18±0.037 0.19±0.057Effluent 17 1186 74 <0.51 <0.23 <0.10 <0.01 0.02±0.005 0.21±0.042 0.20±0.059

Note: The values preceded by „<” were calculated with DL (detection limits).(1) – Ce= effluent radionuclide concentration ; C0= influent radionuclide concentration(2)- The expanded uncertainties are indicated with a coverage factor K=2

To identify the moment in which ion exchange is inefficient it establishes a limit value of effluent

radioactive concentration corresponding either of the release criteria specific for each nuclear facility,

approved by the national regulatory body, or the operational limit 5x103 Bq/l , limit value at which LEM

automatically stops the release, at Cernavoda NPP.

For performed tests, the treatment capacity of the resin was calculated considering the amount of

radioactive effluent concentration limit 5x103 Bq/l, the value at which LEM automatically stops the

discharge of Cernavoda NPP.

In case that radioactive effluent concentration limit imposed is a different value set based on criteria

approved by CNCAN release into the environment, the proposed treatment capacity is recalculated based

resins breakthrough curves obtained.

36

2.2.5 Experimental results

The test carried out for treating by ion exchange of ~ 1.2 m3 waste originated from Cernavoda NPP,

using the pilot installation for testing of the resins Purolite NRW5050 and Purolite NRW1600, led to

approx. 1190 liter effluent.

The breakthrough curves of gamma emitting radionuclides through two columns of Purolite

NRW1600 cation exchanger and Purolite NRW5050 anion exchanger, respectively, are plotted in figure

27

0,0

0,2

0,4

0,6

0,8

1,0

0 20 40 60 80

C e/C

0

Volume, BV

CO-60 NB-95 ZR-95

Figure 27. Breakthrough curves through Purolite NRW 1600+NRW5050 columns (6-10 BV/h flow rates)

In the graph of the Ce/C0 variation, data for 137Cs, 134Cs, 54Mn and 110mAg are not represented

because after treating of 1.2 m3 waste, the concentration of these radionuclides in effluent was kept below

detection limits.

The variation of effluent conductivity and TDS (total dissolved salts) with treated wastewater

volume, expressed as BV (bed volume) is shown in figure 28

0

2

4

6

8

10

12

14

0

40

80

120

160

0 20 40 60 80

pH

Cond

ucti

vity

, uS/

cm

Volume, BV

cond pH

Figure 28. The variation of effluent conductivity and TDS (total dissolved salts) with treated wastewatervolume

37

The decontamination factor of the radionuclides for each fraction of the effluent was calculated and

then plotted vs. the volume of treated solution expressed in the units of bed volumes of the ion exchange

resin (fig. 29).

Figure 29. Decontamination factors for Purolite NRW 1600+NRW5050 ion-exchange resins

Experimental results and decontamination factors obtained prove the applicability of Purolite

NRW1600+NRW5050 ion-exchange resins to treatment of radioactive wastewater generated at

Cernavoda NPP for the selected experimental conditions.

38

3. TECHNOLOGY FOR TREATMENT OF AQUEOUS RADIOACTIVE WASTE FROM

CERNAVODA NPP

The objectives proposed for the aqueous radioactive waste treatment technology elaborated were:

- maintaining the radioactive content of aqueous waste from Cernavoda NPP below the LEM

control limit or below DEL.

- an increased volume reduction factor

- a simple, secure and easy to operate installation

- an acceptable treatment cost.

The ion exchange process was selected for the treatment of aqueous radioactive waste from

Cernavoda NPP, taking in consideration the following:

P The treated liquid radioactive wastes are diluted aqueous solutions (£ 0.1%), reason for which

the ion exchange resins are spent slowly, and the spent ion exchange media volume is much less than the

treated waste volume.

P The effluents radioactivity is low enough to be discharged.

P The spent resins radioactivity doesn’t influence their stability.

P The treatment process was tested at pilot-scale, using aqueous radioactive waste from Cernavoda

NPP.

P The components of installation for treatment by ion exchange are simple and easy to operate.

3.1 Flow chart of technology for treatment of liquid radioactive waste from Cernavoda NPP

The mostly synthetic organic ion exchange resins are based on a copolymer of styrene and

divinylbenzene which are fixed functional groups, like –SO3H, -(SO3Na) for strong acidic cation

exchangers, - COOH for weak acidic cation exchangers and –NH3+ or –N2

+ functional groups for anion

exchangers.

The selectivity of ion exchange processes depends on the following factors:

P The ion exchanger type

P The ions valences, the ionic species present in waste and their concentrations

P The type of counter ions and their concentration

P Ion exchange parameters (temperature, pressure, flow rate, ion exchange capacity).

Ion exchange processes can be operated in batch or continuous modes. We chose the continuous

operation, when the media is contained in a vessel and the liquid is pumped through the media under

pressure. Separate vessels can be used for cation and anion exchange, or the media can be combined into

39

a single vessel (termed a „mixed bed system”) Typical separate bed and mixed bed ion exchange units are

shown in figure 30.

a) b)Figure 30. Schematic diagrams of the (a) separate bed and (b) mixed bed ion exchange systems. Figure

first published in Radioactive Waste Technology (Moghissi, A., Godbee, H.W., Hobart, S.A., Eds), ASMEPress, New York (1986)

Several strong acidic cation exchangers and anion exchangers, respectively, manufactured at

Victoria – Brasov County (Romania) by PUROLITE Co., have been tested. The pilot-scale testing results

recommend the NRW1600 and NRW5050 ion exchange resins for decontamination of aqueous

radioactive waste from Cernavoda NPP.

NRW 1600 is a strongly cationic exchanger, R-H form, and NRW5050 is a strongly anionic

exchanger, R-OH, both manufactured by Purolite Co.

The flow chart of the aqueous radioactive waste treatment technology is presented in Figure 31.

40

Figure 31. Flow chart of the aqueous radioactive waste treatment technology

The ion exchange pilot installation design is shown in Figure 32.

The aqueous radioactive waste from Tank R1 is transferred into Tank R2, with pump P, until the

level gauge shows the 200 liter level. Before the transfer of aqueous waste, the concentration of gamma

emitting radionuclides, the H-3 concentration, the pH value, the conductivity and TDS have been

determined. The pump P is started and the waste is firstly filtered through the filters F1, F2 and MF,

respectively, then the waste flow into tank R2. The pump P stays on, until the tank is full filled. The valve

R6 of tank R2 is opened and the aqueous radioactive waste from the feeding tank enters into the column

Λ < CMA Discharge

Λ > CMA

organic radioactive waste

radioactive sludge

solid radioactive waste(filters)

spent cationic resin

spent anionic resin

EFLUENTRADIOACTIVITY

ANALASYS

ION EXCHANGE(NRW5050 anionic resin)

ION EXCHANGE(NRW1600 cationic resin)

FILTRATION

ORGANIC PHASESEPARATION

SETTLING

Aqueous radioactivewaste/wastewater

41

C1, on the upper side, and it flows through the cationic resin bed, with a constant fixed rate. In C1

column, the cations uptake on cationic resin occurs and the H+ ions pass into solution.

Figure 32. The ion exchange pilot installation designC1,C2 cationic resin and anionic resin columns, respectivelyR l waste tankR2 feeding tankCS separation column (coalescence)V1 organic waste containerV2,V3 sludge containerV4 effluent containerF1, F2 cartridge filters (suspended solids)MF microfiltration cartridgePA feeding pump (peristaltic)P filtration pump

After transiting the cationit column, the decationized waste runs through the anionit column (C2)

from top to bottom where the anions are retained, and the OH- ions are I into the solution. The H+ ions in

the effluent from the cationite column will be neutralized with HO- ions from ionic exchange on the

anionite to generate water.

R2

R1

V2

F1

PA P

F2 MF

V3

V4

V1

Aqueous radioactivewaste/wastewater

Organicphase

CS

C1

C2

sampling

sampling

sludge

sludge

effluent

R6R1

R2R3

R4

R5

R7R8

R9

R10

42

The conductivity of the final effluent is reduced due to ionic species retained on the resin and the

generation of the hard dissociable water. Because of this effluent conductivity measurement during the

process technology is needed to assess the degree of waste decontamination. Increasing the conductivity

of the effluent is produced by major ions leak from the waste, often caused by depletion of the resin, but

also by some operating incidents that can lead to leaks, to the creation of preferential channels in the resin

mass or involving the resin in the effluent.

Two cations columns and two anions columns successively linked are proposed to be used at the

industrial scale to extinguish these deficiencies.

After passing through the ions exchanger columns, the effluent is collected into a collecting

container (V4) and stored until it is determined the radioactivity concentration. If the radioactivity

concentration of the effluent is less than the maximum agreed level, the effluent is discharged, otherwise

the effluent is recirculated. The columns are gravity fed and establishing the flows and interrupting-

starting the process is done through valves mounted on the link pipes of the installation components. The

installation components are mounted on a rack and positioned in such a way that it ensures gravity

feeding of the columns.

3.2 Pilot installation design

The pilot installation has been manufactured at ICN Pitesti for testing the ion exchanger

materials performances used for aqueous radioactive waste decontamination. The pilot installation

consists of two series columns, each with 160 mm in diameter and 800 mm high. The resin beds were 500

mm and 300 mm deep, respectively. The apparent contact time was 10 min for cationit and 6 min for

anionit, i.e. a maximal flow rate of 20 bed volumes/h.

Lab-scale tests showed that the pilot installation could treat 700 bed volumes before a break-

through to occur. The highest decontamination factor achieved by the nuclear grade Purolite resins was

104.

The pilot installation used for ion exchange treatment of the aqueous radioactive waste from

Cernavoda NPP has the following main components:

- column for organic phase separation (figure 33a)

- two cartridge filters and one microfiltration cartridge for particulate material separation

- radioactive waste storage tank of 1 m3 capacity (figure 33b)

- cationite column feeding tank, with level gauge glass (figure 33c)

- two ions exchanger columns, each of them with level gauge glass (figure 33d)

- corrosion-proof pump for waste filtration/feeding of the pilot installation

- secondary waste collecting containers, effluents respectively.

43

a) b) c) d)Figure 33. Components of the ion exchange pilot installation

The pilot installation was used to set the optimal treatment parameters.

The technical specifications for the pilot installation are:

Radioactive waste storage tank is made from fiber glass reinforced polyester and has a volume of 1

m3. On the upper side, the tank has a central fitting (Dn=400mm) used for waste transfer pump and on the

bottom side there is a Dn=50mm fitting with a valve for sediment exhausting. On the lateral side, there is

a Dn=50mm fitting with a valve for waste filling to the feeding tank.

The radioactive waste storage tank is corrosive-proof, both to acid and alkaline agents.

The feeding tank has a 200L volume, it is a stainless steel cylinder having D=600mm, h=650mm

and a funnel-shaped bottom of h=160mm. A connecting pipe with valve for feeding the cationite column

is on the inferior side panel and a fitting for exhausting of the possible colloidal sludge sediments is

attached to the bottom of the feeding tank. The upper side of the tank is a detachable lid for cationite

feeding coupled with the exhausting hose from the microfilter.

The feeding tank has a level gauge glass on a height of 570mm. The tank is mounted on a frame to

ensure the height difference required for feeding.

Filtration/transfer pump of the waste from the storage tank to the feeding tank is a corrosion-proof

pump that ensures a pressure of 4 bars, maximum allowed for microfiltration.

Ions exchanger resin columns are made of stainless steel (D=160mm and h=800mm) with flanges

on the inferior side. The column lid has a connecting pipe to the waste feeding tank. The columns have a

tronconical shaped bottom, flange mounted and are equipped with glass level. The bottom flange of the

column secures a perforated plate, which supports the stainless steel sieve with mesh size up to 0.3 mm.

44

The sieve retains the resin pearls. On the bottom-side panel, the column has a closed connecting pipe with

a blind flange for exhausting of the spent resin. Also, on the bottom side, the column is provided with an

effluent exhausting fitting and a valve for sampling. The effluent exhausting fitting has a valve and it is

connected through the stainless steel pipe to the feeding fitting of the anionite column.

The two columns (cationite and anionite) are identical; the cationite column is positioned under the

feeding tank and the anionite column is placed in such a manner that there is a level difference fall

between the two columns. This position ensures the feeding of the columns through free fall. All the pipes

are detachable, stainless steel made.

Effluent container is a 130L tank, located near to the anionite column.

Cartridge filters

Two cartridge filters, Dual-Gradient Density (DGD) type, assembled into Big Blue® filter housings,

model #10 Big Blue® - ¾” (manufactured by Pentair Filtration Inc. – Pentek) were installed for the solid

suspension separation.

The installed filters use cartridges manufactured by Pentair Water Treatment – Pentek Filtration,

with the following performance data:

- filter media: polypropylene

- temperature range: 4.4 ÷ 62.8 °C

· filter model DGD-7525

-pre-filter micron rating (nominal): 75 µm

-post-filter micron rating (nominal): 25 µm

-flow rate = 38 l/min (initial ΔP<0.1 bar)

· filter model DGD-2501

-pre-filter micron rating (nominal): 25 µm

-post-filter micron rating (nominal): 1 µm

-flow rate = 38 l/min (initial ΔP<0.1 bar).

The physico-chemical properties of the liquid waste from Cernavoda NPP impose an additional

filtration step for the prevention of resin bed fouling. With this purpose, a microfiltration pleated cartridge

filter was installed within the pilot installation.

The microfiltration cartridge is used for the removal of particulate material from wastewater. The

separation is achieved by dead end filtration, when the particles accumulate on the filter surface.

The installed microfiltration cartridge uses Memtrex-NY filter, manufactured by GE –

Water&Process Technologies, with the following characteristics:

- filtration media: Hydrophilic Nylon66 Membrane

45

- support layers : Polyester Microfiber

- filtration area: 0.64 m2

- pore size rating: 0.45 Micron Absolute

- retention efficiency at 0.45 micron: 99.9 % .

Operational limits of microfiltration cartridge:

- Max. Oper. Temp.: 82º C at 0.69 bar, in water

- Max. Forw. Diff. Pressure: 4.1 bar at 21°C

- Max. Rev. Diff. Pressure: 2.1 bar at 21°C.

The dead end filtrations have the pressure gradient as the process driving force and, therefore a

pump is needed to assure the waste flow. Pressure gauges are installed for differential pressure

monitoring.

Figure 34 shows the ICN Pitesti pilot installation used to carry out the test.

Figure 34. The pilot ion exchange installation

46

3.3 Operation procedure

3.3.1 Preparing ion exchange resins for utilization

The NRW 1600 cationite does not require regeneration, because it is supplied by the manufacturer

in the form R-H, however, before use it must be swelled (hydrated) and expanded, then to activate the

functional groups and for the mechanical removal of impurities, insoluble substances, air bubbles or gas

bubbles it will be maintained in demineralized water for 24 hours.

The NRW 5050 anionite must be kept in demineralized water for 24 hours for pearls expanding and

functional groups activation, after which it is washed with demineralized water, repeteadly, until the

washing water is colourless with a neutral pH.

Unlike the A-600 anionite, used currently, the NRW 5050 anionite recommended does not require

conversion because it is delivered by the manufacturer in R-OH form.

After completion of these preliminary operations, the ions exchangers are ready to be filled up into

the columns.

3.3.2 Preparing pilot- installation for operation

To avoid contamination of the effluent before loading the columns with resin, it is necessary the

decontamination of the columns, of the connection pipes and of the containers for collecting effluents.

The decontamination is performed using a solution of up to 1 g/L oxalic acid, citric acid and EDTA. The

washing water is analyzed through gamma spectrometry in order to assess the unfixed contamination of

the installation. Before loading the resins it is mandatory to check the state of the stainless steel sieve, and

in case this is damaged, it should be replaced. A leakage test must be done to all valves and connection

fittings of the pilot installation.

3.3.3. Loading the columns with ion exchange resins

The resin must always be maintained in fluid slurry. The desired amount of resin for the column test

is measured, using a graduated onizi. About one bed volume of fluid (a volume of fluid equal to the

volume of resin used) is placed into the column, then with the outlet valve closed, pour the re-hydrated

and measured resin slurry into the column.

For optimum flow properties, it is a good practice to backwash-stratify the resin bed prior to the

introduction of feed. Backwash-stratification redistributes the resin in the column, packing the resin

so that the larger beads are toward the bottom of the bed and the smaller ones are closer to the top.

Backwash-stratification is accomplished by placing the column in an up-flow mode, using barren fluid

(usually deionized water) to backwash expand the resin bed. Normally, the unused space in the top of the

column (freeboard) is large enough to allow the complete expansion of the resin bed by approximately

47

80-100%. After allowing the bed to expand for 15 – 20 minutes, gradually reduce the flow to zero,

allowing the bed to gravity settle. Do not force the bed to pack by draining the column through the outlet.

This can result in extremely poor flow characteristics, possibly rendering your data useless. Also, make

sure that the column is not allowed to „run dry”. This will cause cracking and channeling within the resin

bed. If air does happen to enter the bed, re-fluidize and re-pack the bed before continuing. Once the bed is

packed it is usually a good idea to pass several bed volumes (BVs) of de-ionized water through the bed to

rinse out any trace of accumulated degradation products, common in resins that are stored for long

periods of time prior to use [http://www.dow.com/liquidseps].

3.3.4. Settleable solids separation and organic phase separation

The liquid radioactive waste transport container is placed near the pilot installation. A corrosion-

proof pump is used to feed the waste to the treatment installation. The pump sorb is sunk in the

radioactive waste container and the pump discharge path is coupled either to the reservoir of the

installation, if the organic phase is missing, or to the organic phase separation column.

The aqueous phase is transferred in the radioactive waste storage tank of the pilot installation and

the organic phase is stored for treatment.

After the solid particles sedimentation, the slurry decantated in the storage tank is discharged by

opening the valve mounted at the bottom of the tank and is transferred to the cementation unit for

conditioning according to approved technology. This operation is carried out after treating several waste

batches and only if sediments are detected.

3.3.5. Aqueous radioactive waste clean-up

The connections between the individual components are done by verifying the tightness of the

connections. By starting pump P, the waste is transferred from storage tank in the feeding tank, through

feeding tanks-columns are opened and the flow is adjusted to the 1L/min.

The optimal flow rate was determined considering the resin bed height, the column section, and the

results obtained at laboratory level and the recommendations from the specific literature.

First, the waste will pass through the filter cartridges, where particles with dimensions >0.45µm are

retained, then through the column with cationit, where the decationisation occurs and then through anionit

where the anionit species are uptaken.

The normal operating conditions are:

· the liquid that will be treated must be clear, the admissible suspended particulate content being

10mg/l;

48

· the influent must be distributed uniformly on the bed of resin, this being achieved by maintaining

the level of the aqueous waste above the upper limit of the resin bed.

3.3.6. Effluent storage and discharge

The effluent resulted from the process is collected in the storage tanks. In order to discharge the

effluent, a representative sample is taken for determining the concentrations of the gamma emitting

radionuclides, H-3 and pH.

In case the activity does not exceed admissible activity and pH is the range of 6-8.5, the effluent is

discharged. In case the activity exceeds the maximum admissible levels, the effluent will be reintroduced

in the treatment process. If the activities allow the discharge, but the pH is out of recommended range, the

pH will be adjusted with a solution of HCl 6N or NaOH 6M, depending of the type of the effluent: basic

or acid.

The neutralization is made by adding, while stirring, NaOH or HCl solution. The solution volume

shall be established by laboratory tests.

3.3.7. Discharging spent ion exchange media

The resin exhaustion is highlighted by the determination of conductivity, pH and radioactivity. The

pH change, the increased activity and increased conductivity of effluent indicates that the resin is

exhausted. In this case, the feeding valve of the waste columns is closed and the liquid from the column is

let to drain. The connection valves are closed and the flanges on the side of the column are switched off.

The discharge is done using a water jet and the suspension is collected in a 60 l drum. After discharge, it

will be measured the dose rate at the wall of the drum, the unfixed contamination and the activity of the

resin. The spent resin will be stored as solid waste and conditioned using the spent ion exchange

conditioning plant, from ICN Pitesti, by applying a licensed technology.

3.4. Processing parameters for treatment by ion exchange of aqueous radioactive waste from

Cernavoda NPP

The analysis of the effluents resulted from experiments carried out on the pilot installation led to

values of the activity that allow their discharged in accordance with environmental legislation However,

the inadequate choice a some parameters of the process may lead either to obtaining some effluents with

activity greater than accepted values, either to decreasing the efficiency of treatment process, by getting a

volume reduction factor lower than expected.

The main parameters that influence the treatment process are:

49

-the composition and physico-chemicals characteristic of waste;

-the type of ion exchange resins and the form in which are used;

-the flow rate and velocity through the columns;

-the depth of the resin bed crossed by waste.

From the results obtained for pilot scale tests, the optimal values of process parameters have been

established for the ion exchange treatment technology of aqueous radioactive waste from Cernavoda

NPP.

3.4.1. Chemical and radiochemical characteristics of aqueous radioactive waste

The process efficiency and effluent quality are influenced by characteristics of aqueous radioactive

waste. The concentration of total solids in the wastewater should be as low as possible and, after filtration

less than 10 mg/l. The concentration of total dissolved salts (TDS) in wastewater should be very low to

achieve a maximum efficiency of ion exchange process. It was established the ion exchange process at an

agreed content of maximum 2.5 mg/l total dissolved salts, but TDS should be less than 1.5 g/l in order to

reach the optimum efficiency.

The pH value of wastewater influences the ion exchange by elution/leaching ions or by increasing

concentration of the ionic dissolved species. The pH range recommended for influent is 1 – 8.

The waste radioactivity can affect the process efficiency as ion exchangers have limited stability to

radiation, a high dose of radiation causing the degradation of the resin which has the effect of lowering

the exchange capacity.

In the specialty literature it was reported that at a dose of 109 R the exchange capacity of strong acid

cation decreases with 10% and of strong basic anions with 40%. In the experiments carried out at pilot

scale, it was found that the waste radioactivity did not produce significant changes in process efficiency

and did not create operational problems.

Since aqueous radioactive waste from Cernavoda have low activity, the resin transfer activity does

not lead to major changes in the structure and its exchange capacity.

The required characteristics of the aqueous waste that can be effectively treated by ion exchange

are:

- total dissolved salts(TDS): max. 2.5g/l;

- pH:1-8;

- suspended particulate (after filtration): 10mg/l;

- radioactivity conc.: 0.1Ci/m3.

50

3.4.2. Selected ion exchange resins – characteristics

For choosing the most efficient ion exchangers, there were tested both strongly acid cations but also

strong and weak basic anions. The experiments were carried out at laboratory level and the ion exchanger

resins were selected based on the results and then tested at pilot scale.

There were tested cations NRW160, NRW1600, C-100H type, the anions NRW5050, NRW5070,

A600, A-500, A-100 type and bed mixt NRW-3550.

Based on the results obtained at laboratory scale, there were selected the ion exchangers NRW 1600

and NRW5050 type to be tested at pilot scale.

The results obtained at pilot scale recommend using the couple of ion exchangers tested for

implementation within the treatment technology by ion exchange of liquid radioactive waste from

Cernavoda NPP. The major characteristics of these exchangers are shown in the Analysis Certificates

provided by the producer.

There were selected the strongly acid cationit and strongly basic anion because the exchange

capacity is kept to the maxim value regardless of the ion strength of the wated and in a wide range of the

pH.

3.4.3. Establishing parameters which control the ion exchange resins performances

The performances of ion exchangers are governed of the physico-chemical properties of the resin

and the operating conditions. The physico-chemical properties include the exchange capacity, the degree

of crosslinking and the size of beads, the selectivity, the process kinetics and the chemical and thermal

stability.

The operating conditions include the chemical form of the chemical species, the temperature, pH,

the influent flow rate, the dept of resin beds.

1. The size of resin beads affects the kinetics of the ion exchange in that the beads with small

diameters and big specific surface determine a higher velocity of exchange, so that a higher flow of

supply is allowed. The disadvantages using the resins with small granules (gel type) consist of the

significantly reducing of the flow rate, clogging the resin bed, difficulties in operating the installation and

the risk of contaminating the effluents by entrained the resins particles. Therefore in the nuclear field it is

preferred the ion exchangers of macroporous type, with beads of a diameter >0.4mm. The selected resins

for the technology elaborated are between 0.4 and 1.1 mm diameter.

2. The resin bed length imposes the velocity and the flow rate of the waste through the column, but

the bed geometry may favor the formation of preferential flow path in the resin bed, leading to changes in

the waste flow profile. In the design and operation stage of the resin beds, there must be a balance

between the the flow rate and the bed depth. The kinetics of the ion exchange process is controlled by

51

diffusion of ions in solution to the active groups and in the resin beads. A high value of the flow rate

reduces the residence time necessary for ions diffusion, which require deeper beds that induce the

formation of channels and changes in the flow profile. Following the laboratory-scale experiments, for the

pilot scale test it was chosen a 30 to 50 cm length of the resin bed.

3. The resin-waste contact time necessary to achieve the ionic exchange is a very important

parameter since, depending on the reaction rate, part of the process parameters are established. The time

required to obtain a maximum distribution coefficient, considered the optimum time for the waste passing

through the ion exchange columns, was determined at the laboratory scale on different types of

radioactive waste. The necessary time to reach sorption equilibrium ranges between 5-10 minutes. The

necessary liquid residence time on the resin layer, must be correlated with the flow rate and with the

appropriate depth of the resin bed to avoid the preferential path flow formation and liquid flow

perturbation. Following the pilot scale experiments, it was chosen a 30-50 cm depth for the resin bed and

a flow rate of ~1 L/h, corresponding to a flow velocity of 5 cm/min.

4. The waste processing temperature through ionic exchange is the room temperature; high

temperatures, although have the role to increase the process efficinency, cause the resin decay.

5. The resin volume used at pilot scale was of 10 L cationite and 6 L anionite. Depending on the

waste composition and on the available exchange capacity, the anionite/cationite (v/v) ratio varies up to

3/1.

3.4.4. Influent flow rate and flow velocity through ion exchange resin bed

The waste flow rate and flow velocity through the ion exchangers columns were calculated taking

into account the columns section, the minimum time necessary to achieve the ionic exchange and the

depth of the resin bed. From the experiments, for the pilot scale installation the optimum flow rate was

established at ~1 L/h, corresponding to a flow velocity of 5 cm/min.

52

4. RECOMMENDATIONS ON THE TYPE, DIMENSIONING OF SPECIFIC EQUIPMENT ANDCONSTRUCTION MATERIALS

Theoretical basis of the equipment necessary for application of the technologyExperimental work presented in previous reports doubled by the mathematical modeling work offer

important information concerning the equipment type recommended to be used in the industrial process,

the materials used and the scale-up procedure.

First of all, an overview about the wastewater treatment technology improvement implies a first

step with estimation of the strategy to be followed in the equipment designing. Two important ways could

be imagined: one linked to small storage capacities and low processing installations capacities (that way

imply low investment costs but probably high operating costs versus other way) and the second one

implying high storage capacities of the contaminated wastewater doubled by high processing installation

capacities. Obvious, since we talk about wastewater contaminated with some radionuclide it is preferable

to diminish to maximum possible the residence time of the radionuclide in the industrial plant (with

respect to the safety or specific laws for nuclear activity plants – linked with wastewater treatment). That

means a fast water treatment; the recovery of important species and the water recycle in the system. But

our recommendation is to take a decision about the wastewater treatment capacity in accordance with

available data concerning the wastewaters quantities in time available from the industrial plant. So not

only technical and safety reasons should be considered but also economical criterion should be taken into

account.

Applying the results given by the experimental and theoretical research means to follow two steps:

the scale-up laboratory devices and the detailed designing of all industrial devices needed in the

wastewater treatment plant.

A start-up point of scale-up activity it should be the conception of the technological scheme (at

industrial scale). Some devices (like storage vessels, pumps, pipes) are not present during the laboratory

experiments. So a first draft of a possible technological scheme (with possible options) will give a first

image of this issue. A secondary path of the installation, designated to the resin change in the system

obviously should be considerate and also the periodically clean-up of all devices and pipes.

The considered storage vessel capacity should be in accordance with the wastewater input

accounted and the installation treatment capacity. Also, another important point liked to this device

should be the mixing device of the storage tank. Since the experimental work at laboratory scale is done

at constant concentration of the radionuclide in the input wastewater, the same procedure should by apply

at industrial scale. Any sedimentation process that could occur during the storage time (between two

processing periods) must be eliminated. A simple mechanical stirring device or a recycle path of the

53

wastewater from the used pump for column feeding could be considered. The storage vessel could be

considered in two situations:

· Atmospheric Tanks The term atmospheric tank as used here applies to any tank that is designed

to be used within plus or minus several hundred Pascals of atmospheric pressure. It may be either open to

the atmosphere or enclosed. Minimum cost is usually obtained with a vertical cylindrical shape and a

relatively flat bottom at ground level.

· Elevated Tanks These can supply a large flow when required, but pump capacities need be only

for average flow. Thus, they may save on pump and piping investment. They also provide flow after

pump failure, an important consideration for fire systems.

Calculation of Tank Volume. A tank may be a single geometrical element, such as a cylinder,

a sphere, or an ellipsoid. It may also have a compound form, such as a cylinder with hemispherical ends

or a combination of a toroid and a sphere. Each geometrical element of the tank usually must be

calculated separately to establish the total volume. Calculations for a full tank are usually simple, but

calculations for partially filled tanks may be complicated.

In the case of a partially filled horizontal cylinder refer to figure 35

Figure 35. Volume of partially filled horizontal tanks calculation. H = depth of liquid; R = radius; D =

diameter; L = length; α = half of the included angle; and cos α = 1 − H/R = 1 − 2H/D.

The angle α is calculate in degrees. Any units of length can be used, but they must be the same for

H, R, and L. The liquid volume will be:

2V L R sin cos57.30

aæ ö= × × - a × aç ÷è ø

(3.1)

This formula may be used for any depth of liquid between zero and the full tank, provided the

algebraic signs are observed. If H is greater than R, sin α cos α will be negative and thus will add

numerically to α/57.30. The volumes of heads must be calculated separately and added to the volume of

the cylindrical portion of the tank. The four types of heads most frequently used are the standard dished

54

head, torispherical head, ellipsoidal head, and hemispherical head. A partially filled horizontal tank

requires the determination of the partial volume of the heads. A formula for partially filled heads is [3]:

( )2V 0.215 H 3 R H= × × × - (3.2)

where V = volume, R = radius, and H = depth of liquid. Some simplifying assumptions which affect

the volume given by the equation are made in literature, but the equation is satisfactory for determining

the volume as a fraction of the entire head.

Container Materials and Safety. As it is known, storage tanks are made of almost any structural

material. Steel and reinforced concrete are most widely used. Plastics and glass-reinforced plastics are

used for tanks up to about 230 m3. Resistance to corrosion, light weight, and lower cost are their

advantages. Plastic and glass coatings are also applied to steel tanks. Nonferrous metals are used only

when their special properties are required. When expensive metals such as tantalum are required, they

may be applied as tank linings or as clad metals. Small tanks containing nontoxic substances are not

particularly hazardous and can tolerate a reduced factor of safety. Tanks containing highly toxic

substances and very large tanks containing any substance can be hazardous. The designer must consider

the magnitude of the hazard. The possibility of brittle behavior of ferrous metal should be taken into

account in specifying materials. The container material chooses should respect all rules imposed by the

nuclear industry. Also if clean-up process during periodically revisions implies the use of strong acids or

basic solutions that should influence the container material choose.

Pump selection. Selecting pumps for any service imply the necessity a lot of data about the liquid

to be handled, the total dynamic head, the suction and discharge heads, and, in most cases, the

temperature, viscosity, vapor pressure, and specific gravity. The task of pump selection is frequently

further complicated by the presence of solids in the liquid and liquid corrosion characteristics requiring

special materials of construction. Solids may accelerate erosion and corrosion, have a tendency to

agglomerate, or require delicate handling to prevent undesirable degradation. Since radioactive liquids are

involved in transport procedure a solution in pump selection could be choose a kinetic special pump [3]

gas lift type to reduce the nuclear contamination of the device.

Range of operation. Because of the wide variety of pump types and the number of factors which

determine the selection of any one type for a specific installation, the designer must first eliminate all but

those types of reasonable possibility.

Pump Materials of Construction In the chemical industry, the selection of pump materials of

construction is dictated by considerations of corrosion, erosion, personnel safety, and liquid

contamination. The experience of pump manufacturers is often valuable in selecting materials.

55

Presence of solids. If a liquid containing suspended solids, must be transported there are unique

requirements which must be considered. Adequate clear-liquid hydraulic performance and the use of

carefully selected materials of construction may not be all that is required for satisfactory pump selection.

Dimensions of all internal passages are critical. Must be avoided pockets and dead spots, areas where

solids can accumulate. Close internal clearances are undesirable because of abrasion. Flushing

connections for continuous or intermittent use should be provided. Chemical pump installations that

require annual maintenance costing 2 or 3 times the original investment are not uncommon. In most cases

this expense is the result of improper selection [7,9].

Concerning the ion exchange column, data about the ion exchange recommended resin were above

mentioned. Some theoretical considerations are useful for the scale-up activity.

Ion exchangers classification. Ion exchangers are classified according to (1) their functionality and

(2) the physical properties of the support matrix. Cation and anion exchangers are classified in terms of

their ability to exchange positively or negatively charged species. Strongly acidic and strongly basic ion

exchangers are ionized and thus are effective at nearly all pH values (pH 0–14). Weakly acidic

exchangers are typically effective in the range of pH 5–14. Weakly basic resins are effective in the range

of pH 0–9. Weakly acidic and weakly basic exchangers are often easier to regenerate, but leakage due to

incomplete exchange may occur [9].

Except in very small-scale applications, ion-exchangers are used generally in cyclic operations

involving sorption and desorption steps. A typical ion-exchange cycle for water-treatment applications

involves (a) backwash—used to remove accumulated solids obtained by an upflow of water to expand

(50–80 percent expansion is typical) and fluidize the exchanger bed; (b) regeneration—a regenerant is

passed slowly through the used to restore the original ionic form of the exchanger; (c) rinse—water is

passed through the bed to remove regenerant from the void volume and, in the case of porous exchangers,

from the resin pores; (d) loading—the fresh solution to be treated is passed through the bed until leakage

begins to occur.

A complete deionization with ion-exchange columns is the classical method of producing ultrapure

water for boiler feed, in electronics manufacture, and for other general uses in the chemical and allied

industries. For deionization will be necessary two exchangers with opposite functionality to remove both

cations and anions. These can be in separate columns, packed in adjacent layers in the same column, or,

more frequently, in a mixed bed. In the mixed bed case, the two exchangers are intimately mixed during

the loading step. For regeneration, backwashing separates the usually lighter anion exchanger from the

usually denser cation exchanger. The column typically has a screened distributor at the interface between

the two exchangers, so that they may be separately regenerated without removing them from the column.

56

Differences in selectivity for different species can be used to carry out separations during the

displacement [4] Multibed cycles are also used to facilitate integration with other chemical process

operations.

General Design of the ion exchange devices

A typical fixed-bed ion exchanger consists of a vertical cylindrical vessel of lined steel or stainless

steel. Linings are usually of natural or synthetic rubber. Spargers are provided at the top and bottom, and

frequently a separate distributor is used for the regenerant solution (if case). The resin bed, consisting of a

meter or more of ion-exchange particles, is supported by the screen of the bottom distributor.

Externally, the unit is provided with a valve manifold to permit downflow loading, upflow

backwash, regeneration, and rinsing. For deionization, a two exchanger assembly comprising a cation and

an anion exchanger, is a common configuration. Characteristic design calculations are presented and

illustrated by Applebaum [9]. Large amounts of data are available in published literature and in bulletins

of manufacturers of ion exchangers. In general, however, laboratory testing and pilot-plant work is

advisable to determine usable exchanger capacities, regenerant quantities, exchanger life, and quality of

product for any application not already thoroughly proven in other plants for similar conditions [9]. For

larger-scale applications, a number of continuous or semi continuous ion-exchange units are also

available. The Higgins contactor was originally developed to recover uranium from leach slurries at the

Oak Ridge National Laboratory [9]. More recently, it has been adapted to a wide variety of applications,

including large-volume water softening. The Asahi process [9] is used principally for high-volume water

treatment.

Another continuous ion-exchange system is described by Himsley and Farkas [9]. Such a system is

used to treat 1590 m3/h of uranium-bearing copper leach liquor using fiberglass- construction columns 3.7

m in diameter. The adsorption column is divided vertically into stages. Continuous transfer of batches of

resin occurs from stage to stage without any interruption of flow. This is accomplished by pumping

solution from one stage (A) to the stage (B) immediately above by means of external pumping in such a

manner that the net flow through stage B is downward, carrying with it all of the resin in that stage B.

When transfer of the ion exchanger is complete, the resin in stage C above is transferred downward in a

similar manner. The process continues until the last stage (F) is empty.

General considerations about column scale-up. Sometimes passing from laboratory scale to

industrial scale is a difficult step. The scale-up procedure should take into account various aspects

concerning the process and corresponding devices. The main recommendation about this are presented

further.

There are many different aspects related to scale-up, some of them technical and others not.

57

Economical aspect

Financially seen it looks more attractive to build one big apparatus rather than several parallel ones.

The investment costs (Inv) go up with the production capacity (Cap) and for this it holds roughly the 0.6-

rule for process equipment:

Inv~Cap 0.6

This can be simply seen by onizing that Inv~A (surface area of equipment)~L2

and Cap~V

(volume of equipment)~L3. From these two it follows Inv~Cap

2/3.

The exponent 2/3 is called the degression coefficient and changes from case to case. As long as the

degression coefficient is smaller than unity there is an economic incentive to increase the size of

equipment rather than to multiply the equipment. The degression coefficient depends, of course, on the

type of equipment. Also the effect of ‘standard’ sizes is important here. If, for example, a valve is made in

different standard sizes than it could be cheaper to buy two smaller ones than to have a large one

especially made.

General aspects of scale-up. The main aspects involved in scale-up activity are:

• Economical aspects (as it was already mentioned):

I ~ C0.6

I ~ A (area)

C ~ V (volume)

• Energy, raw materials, and environment

• Vulnerability of the production to:

- Maintenance

- Sabotage

• Logistics:

- Transportation of parts

- Supply of raw materials

- Discharge of products

• Social aspects:

- Employability, labor conditions

• Mechanical and other technical restrictions

• Start-up, management, maintenance of a plant

• Safety

58

Scale-up considerations. The scale-up of the installation could be made in two ways: multiplying

the number of pilot scale installation (by number) or by devices size enlargement (by size enlargement).

The advantages and disadvantages of each option are briefly presented as follows:

by number:

•fast design procedure: gain of time

•limited chance on practical constraints

•easier construction: small mechanical problems

•good possibility of sub-maximal production

•easier repair/cleaning; cheaper spare parts

•logistical problems in terms of distribution

•extra risk problems due to interference

by size enlargement:

•slower design procedure: loss of time

•equipment limitations due to practical constraints

•construction problems: mechanical constraints

•limited possibility for sub-optimal production

•difficult repair; limitation in spare parts

•logistical problems in terms of large quantities

•risk problems due to bulk quantities

Similarity. In the process industry all kinds of physical and chemical processes take place which in

principle can be described mathematically by means of momentum, heat, and mass balances. The

resulting description is usually too complex to allow general analytical solutions. Although these coupled

partial differential equations can be solved numerically after some computer time, the obtained solutions

are very specific and hardly suitable for generalization and analysis. In such cases it might be useful to

replace the straightforward process description by a comparative method in which the processes at two

different scales (sizes) of the process equipment are compared. From the conditions for process similarity,

scale-up criteria with respect to that equipment can then be derived. So scale-up is actually designing

using “foreknowledge”; this foreknowledge can be the result of laboratory experiments or experience

with a similar industrial process, possibly at a different scale.

Two kinds of similarity are distinguished:

· apparatus similarity, in which the dimensions of the apparatus are being scaled-up with a common

scale-up factor; afterwards it is analyzed what effect scale-up has on the processes in the apparatus.

59

· process similarity, in which a certain process variable (production, energy consumption) is being

scaled-up out of process considerations. Process requirements (such as product quality) result at process

similarity in demands for the dimensions of the apparatus. So now scale-up of the dimensions is

secondary.

Process equipment companies that produce relatively simple equipment often prefer apparatus

similarity for production reasons. Also, a series of similar apparatuses (such as pumps) ‘looks’ better.

However, often only the original design of such a series of apparatuses may be optimal and some (much)

of that quality has been sacrificed in case of the ‘derived’ apparatuses. This is (one of) the reason(s) why

the scale-up factor is never chosen far from unity and is seldom bridging one decade. With more complex

equipment and for whole production processes (factories) apparatus similarity is never considered and

only process similarity is used. Therefore, from now on, the word similarity will be used in the meaning

of ‘process similarity’. It should be mentioned that similarity between model and prototype means that the

deformation, flow, temperature and concentration profiles of model and prototype are congruent in

dimensionless form.

There exist more levels of process similarity. Moreover, there is also a certain hierarchy; each

underlying form of similarity assumes the presence of the one above. We distinguish them as: geometric

similarity (shape), mechanical similarity s: static (mechanical) deformation profiles or d: dynamic

(mechanical) similarity velocity profiles, t: thermal similarity temperature profiles, c: chemical similarity

concentration profiles. To derive the conditions for the different levels of process similarity it should be

used momentum, heat, and mass balances; these balances will be formulated such that the analogy

between these three types of transport is maximal.

General conclusions about similarity are:

• At each level of process similarity it is possible to indicate the conditions: equal dimensionless

numbers at M- and P-scale

• Soon the number of conditions is that large that these conditions can only be satisfied with a scale-

up factor equal to one

• Fortunately not all dimensionless numbers are important at all scales

• A method is needed to determine which dimensionless groups are really important and which are

allowed to vary (hopefully at a modest scale). Regime analysis provides such a method.

Finally: only seldom full process similarity is possible (or desirable). If full similarity is not possible

then partial process similarity must be accepted. The dimensionless groups, which have to be kept

constant at both scales, form together the scale-up criterion.

60

Regime analysis and scale-up criteria

Attempts to scale a process fully similar often fail due to the large number of dimensionless groups

that have to be kept constant. Reasons to partly give up full process similarity can be of :

(1) mathematical,

(2) technological or

(3) economical nature.

In the first case the conditions, which constrain the system by keeping the dimensionless groups

constant, are mathematically contradictory.

In the second case certain technological boundaries are exceeded (gas velocities near the speed of

sound, vibrations near the characteristic frequencies, etc.).

In the third case the design is becoming too expensive (special design, extreme costs of

transportation, high costs for foundations or safety).

In all cases it should give up full process similarity and have to be satisfied with partial similarity:

not all dimensionless groups required for full similarity will show in the scale-up criterion, but some will

and some do not. Clearly it is a problem of choice, entering the field of regime analysis.

Hydraulic considerations [10,11]

In literature there are some correlations for the evaluation of particle Peclet number Pep and liquid

holdup %h (as a percentage of the total bed volume) for materials that are frequently used in adsorption

and ion exchange systems

Rep

npPe N= ×

Where

N is 0.52 for upflow and 0.05 for downflow

n is –0.65 for upflow and 0.48 for downflow

Rep is particle Reynolds number p sp

d uRe

× ×r=

h

where h is the dynamic viscosity of the liquid phase,

r – liquid density;

dp – particle diameter;

us – superficial velocity.0.28% 21 99.72 sh u= + ×

Liquid holdup equation holds for particle size of 1.2-1.3 mm. Then, the bed Peclet number can be

evaluated using the following equation [12]

61

L pp

LPe Ped

= ×

The dynamic liquid holdup can be evaluated for other particle diameters following the equation:

( )( )

1 2

12

%

%

mp p

pp

h d ddh d

æ ö@ ç ÷ç ÷

è ø

Where the exponent m is 0.72 [13,14]

The constant part in liquid holdup correlation (equal to 21) is the static liquid holdup. This part is

between (0.11/ε) and (0.072/ε) for particle size between 0.2 and 2 mm, where ε is the bed voidage In the

typical case of ε=0.4-0.5 static holdup is between 14 and 27% [11].

Several authors [15-17] precise that certain geometrical analogies should be kept within the

following limits in order to avoid large-scale mal distribution of the flow

5LD

³

12 30p

Dd

³ -

50 150p

Ld

³ -

where L is the bed length, D the bed diameter and dp is the particle diameter.

Also, Ruthven [16] precises that a maximum linear velocity should not be exceeded in order to

avoid extended friction between the packing materials in both down and upflow operations. This velocity

is 0.8 times the minimum fluidization velocity for upflow operation and 1.8 times the same velocity for

downflow operation. Using this maximum working velocity and the contact time during pretreatment

(which is generally higher than the treatment one) the maximum bed high could be determined as follows:

FS FSF a V= ×

where FS stands for “full scale”

F is the volumetric flow rate and V is the bed volume

The number a is the number of bed volume per hour.

,maxFS FS

FS FSFS FS

F a Vu uA A

×= = £

A is the bed cross sectional area

umax is the maximum allowable linear velocity

Considering that VFS= AFS·LFS, it result that

,maxFSFS

uL

a£

62

Could be used a bed with a length which represent 85 % off this value.

Those considerations did not take into account the possible compression of the particles at the

bottom of the bed, under the bed weight and under the hydrostatic pressure of the fluid

But Inglezakis [13] showed that the beds of 1.4 – 1.7 mm particle size can be very high (50 m).

Also, in Inglezakis et al. [11] is developed the following relation to estimate the full scale flow rate:

( )2104

FSFS LS

a LF D × p ×£ × ×

where LS stands for laboratory scale

The bed porosity should be the same. It is a function of D/dp, but for dp/D values (lower than 0.1) it

is practically constant.

Maintaining fluid flow in piping systems

The primary purpose of any piping system is to maintain free and smooth flow of fluids through the

system. Another purpose is to ensure that the fluids being conveyed are kept in good condition (i.e., free

of contamination). Piping systems are purposely designed to ensure free and smooth flow of fluids

throughout the system, but additional system components are often included to ensure that fluid quality is

maintained. Piping system filters are one example, and strainers and traps are two others. It is extremely

important to maintain free and smooth flow and fluid quality in piping systems, especially those that feed

vital pieces of equipment and machinery. Consider the internal combustion engine, for example.

Impurities such as dirt and metal particles can damage internal components and cause excessive wear and

eventual breakdown.

To help prevent such wear, the oil is run continuously through a filter designed to trap and filter out

the impurities. Other piping systems need the same type of protection that the internal combustion engine

does, which is why most piping systems include filters, strainers, and traps. These filtering components

may prevent damage to valves, fittings, the pipe, and to downstream equipment/machinery. Chemicals,

various types of waste products, paint, and pressurized steam are good examples of potentially damaging

fluids. Filters and strainers play an important role in piping systems, protecting both the piping system

and the equipment that the piping system serves.

Scaling

Since sodium and calcium hypochlorite are widely used in water and wastewater treatment

operations, problems common in piping systems feeding this chemical are of special concern.

Maintaining the chlorine in solution (used primarily as a disinfectant), sodium hydroxide (caustic) is used

to raise the pH of the hypochlorite; the excess caustic raises the shelf life. A high pH caustic solution

raises the pH of the dilution water to over pH 9.0 after it is diluted. The calcium in the dilution water

reacts with dissolved CO2 and forms calcium carbonate. Experience has shown that 2-in. pipes have

63

turned into ¾-in. pipes due to scale buildup. The scale deposition is greatest in areas of turbulence such as

pumps, valves, rotameters, backpressure devices, etc. If lime (calcium oxide) is added (for alkalinity),

plant water used as dilution water will have higher calcium levels and generates more scale. While it is

true that softened water will not generate scale, it is also true that it is expensive in large quantities. Many

facilities use softened water on hypochlorite mist odor scrubbers only. Scaling also often occurs in

solution rotameters, making flow readings impossible and freezing the flow indicator in place. Various

valves can freeze up and pressure sustaining valves freeze and become plugged. Various small diffuser

holes fill with scale. To slow the rate of scaling, many facilities purchase water from local suppliers to

dilute hypochlorite for the return activated sludge (RAS) and miscellaneous uses. Some facilities have

experimented with the system by not adding lime to it. When they did this, manganese dioxide (black

deposits) developed on the rotameter’s glass, making viewing the float impossible. In many instances,

moving the point of hypochlorite addition to downstream of the rotameter seemed to solve the problem. If

remedial steps are not taken, scaling from hypochlorite solutions can cause problems. For example, scale

buildup can reduce the inside diameter of pipe so much that the actual supply of hypochlorite solution

required to properly disinfect water or wastewater was reduced. Because of the scale buildup, the

treatment system itself will not function as designed and could result in a hazardous situation in which the

reduced pipe size increases the pressure level to the point of catastrophic failure. Scaling, corrosion, or

other clogging problems in certain piping systems, are far from an ideal situation [18-20] Reducing

substances deposition in the pipe system will reduce the cost of major revisions of the installation.

64

5. RECOMMENDATIONS ICN ON THE WORK SAFETY MEASURES

ALARA (As Low As Reasonably Achievable) is a radiation safety principle for minimizing

radiation doses and releases of radioactive materials by employing all reasonable methods.

The ALARA principle includes the proper use of shielding and dosimetry combined with

contamination control techniques. All employees bear a responsibility for their own personal safety in

such work areas as:

- Awareness of potential radiation hazards, exposure levels and safety controls in their work areas.

- Awareness of operating and emergency procedures.

- Awareness of practices that do not seem to follow the ALARA philosophy.

- Compliance with reporting incidents and possibly unsafe working conditions to their supervisors.

- Compliance with wearing personnel dosimetry and ensuring it’s return at the proper exchange

frequency.

- Compliance with providing bioassay samples as needed.

The three major principles to assist with maintaining doses ALARA are:

a) Time – minimizing the time of exposure directly reduces radiation dose.

b) Distance – doubling the distance between your body and the radiation source will divide the

radiation exposure by a factor of 4.

c) Shielding – using absorber materials such as Plexiglas for beta particles and lead for X-rays and

gamma rays is an effective way to reduce radiation exposures.

The following practices are effective for reducing potential internal exposures :

a) Good hygiene techniques that prohibit the consumption of food and drinks in the contaminated

area.

b) Control contamination with absorbent paper and spill trays, properly labeled waste containers,

equipment, etc. And prompt decontamination of any detected contamination.

c) Use fume hoods for materials which could become airborne (e.g., vapors, dust, aerosols, etc.)

and present an inhalation hazard to workers.

d) Use proper protective equipment (PPE) such as disposable gloves, safety glasses, lab coats, etc.

To reduce the possibility of ingestion or absorption of radioactive materials.

The following practices are effective for reducing potential risk at ion exchange resins handling:

· Wear protective eyeglasses or chemical safety goggles.

· Wear gloves impervious to this material to prevent skin contact. Gloves should be removed and

replaced immediately if there is any indication of degradation or chemical breakthrough.

· Good general ventilation should be sufficient. No respiratory protection should be needed.

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· Never eat, drink, or smoke in work areas. Practice good personal hygiene after using this material,

especially before eating, drinking, smoking, using the toilet, or applying cosmetics.

· Avoid contact with strong oxidizing agents, particularly concentrated nitric acid. Before using

strong oxidizing agents, consult sources knowledgeable in handling such materials.

Do not pack column with dry ion exchange resins. Dry beads expand when wet. This expansion can

cause a glass column to shat.

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6. RECOMMENDATIONS ON THE WORK SAFETY MEASURES AT INRNE

· Minimizing the buildup of radioactive material to reduce the activity of chemical and radioactive

waste and to reduce the activity of any planned discharges to the environment.

· Rational organization of the work processes and measures for radiation protection of the personnel

under normal operating conditions in the laboratory.

· Assessment of the nature and degree of the radiation risk

· Specialized training, instructions and control of the observation of the rules for safe work with

sources of onizing radiation

· Use of protective barriers against the spread of radioactive substances, protective screens, devices

for remote manipulation of sources of ionizing radiation, as well as limiting the time of work with sources

of ionizing radiation and optimization of the distance to the sources of ionizing radiation

· Use of means for personal radiation protection, organisation of radiation technological and

dosimetric control at the site

· For works of I or II class with unsealed sources of ionizing radiation the area per worker in a

given room shall be minimum 10 square meters.

· The technological equipment, protection devices and working facilities in the rooms at the site

shall be with smooth surface, of simple design and with coatings with low sorption, which facilitate the

removal of radioactive contamination and which can stand the effects of the working materials,

substances, solutions and chemical agents used.

· The floors and ceilings of the rooms for operations shall be coated with materials of low sorption,

which are durable to cleansing means. It is recommended to paint the rooms belonging to different classes

and areas of operations in different colours

· The edges of the floor coatings shall be glued to the walls, bent in advance at 10 cm height from

the floor. If special sewage system is available, the floors of the rooms shall slope down to the facilities

for draining of the water. The corners of the rooms shall be rounded and the doors and window frames

shall be of simplified design.

· In the case of activities with unsealed sources of ionising radiation of class I and II, the common

electrical distribution boards and the controls of the common systems for heating, ventilation, water

supply, gas supply and pressurised air shall be located outside the basic working rooms of the site.

· Operations with unsealed sources of ionising radiation of class I in chambers and boxes shall be

performed by remote controlled equipment or by gloves, airtight sealed to the front wall of the chambers

or boxes.

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· Radioactive substances shall be moved from one room to another after performing of radiation

control

· The quantity of radioactive substances at the workplaces shall be the minimum required by the

particular activity. Radioactive substances with the minimum toxicity, solutions instead of powders, and

solutions with the least specific activity shall be selected.

· The number of operations, which may create a possibility for radioactive contamination of the

rooms and the environment (transferring of liquids from one container into another, sifting of powders,

etc.) shall be reduced to a minimum. Manual operations with radioactive solutions shall be performed

with the application of proper protective means.

· The air from the ventilated rooms boxes, chambers, sealed cabinets shall be cleaned from

radioactive contamination, before it is released in the atmosphere, and the activity of the released

radioactive substances shall be monitored.

· At worksites ventilation tubes (chimneys) for release of the air into the atmosphere shall be

constructed; their height shall ensure diminishing of the volume concentration of the released radioactive

substances at the point of reaching the ground to values, which, on a conservative estimation, shall not

exceed the established dose quotes for the population.

· A system for hot and cold water supply and sewage system shall be available at the laboratory.

· The water taps of the sinks shall be with hot-cold water mixing facility, which is pedal or elbow

controlled or by other methods, which do not require physical contact by the personnel. Hand dryers shall

be installed in the washing rooms for drying of the hands after washing.

· The surface radioactive contamination of rooms, means for individual protection and equipment at

work sites shall not exceed the established limits in accordance with the currently valid norms of radiation

protection.

· After completion of work the workplaces shall be cleansed and, in case of necessity, the external

surfaces of the used vessels, tools and equipment shall be deactivated to the respective limits. Specially

selected and trained personnel shall perform these operations.

· Persons working with unsealed sources of ionising radiation, or persons visiting sites, where work

is performed with unsealed sources of ionising radiation, shall be provided with means for individual

protection in accordance with the type and class of work with these unsealed sources of ionising radiation.

· At work sites, an emergency reserve of means for individual protection and individual dosimeters

shall be provided and maintained

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· At sites, where sources of ionising radiation are operated, radiation monitoring shall be performed

of the basic characteristics of the working and surrounding environment for determination and assessment

of the exposure of the personnel and the population in the respective areas

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7. SUMMARY OF THE RESULTS AND CONCLUSIONS

7.1 Technology efficiency

The physico-chemical and radiological properties of the aqueous radioactive waste transfered from

Cernavoda NPP led to the development of a selection scheme for appropriate tretament methods. For the

treatment of aqueos radioactive waste, the ion exchange method was selected using synthetic organic ion

exchangers.

The main benefits of radioactive waste treatment using organic ion exchangers are:

- High decontamination factors, on average 103

- High volume reduction factors for low salt content waste

- Acceptable cost price.

Organic ion exchange treatment methods limits are:

- Inefficiency of treatment for the waste with a high inactive salts content, as it leads to the rapid

saturation of the ion exchange resin.

- Relatively low stability to radiation and temperature action.

Proposed ion exchange material evaluation was preformed taking into account the following

criteria:

I. Efficiency (ion exchange capacity, selectivity, ion exchange kinetics, decontamination factor-FD,

chemical and radiological stability, hydromechanical strength )

II. Volume reduction for the secondary waste (volume reduction factor – FRV)

III. Implementing possibility (saturated material disposal possibility, ion exchanger availability).

The presented tests for aqueous radiaoctive waste treatment were performed with synthetic organic

ion exchangers, manufactured in Romania, by Purolite Company.

Considering that the aim of the decontamination process is compliance to the MEL (Monitor Liquid

Effluent Monitor) from Cernavoda NPP limit value of 5 kBq/l, the volume reduction factors obtained

from the experimental steps range from 150 to 820, and the maximum decontamination factor was

~3x104 for NRW1600 resin, at a 10 BV/h flow rate in the column.

Based on the results obtained, it can be concluded:

1. The selectivity of the tested ion exchange resins for the radionuclides identified in the waste was:

Cs-137 > Ag-110m > Co-60.

2. The ionic exchange kinetics study for the tested cationic rasins, recommends the NRW1600

resin.

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3. For the aqueous radioactive waste treatment, nuclear grade resins produced by Purolite can be

used, with very good results obtained: strong acid cationit NRW1600, in the R-H form and strong base

anionite NRW5050, in R-OH form.

4. In order to increase the efficiency of the ionic exchange process, it is necessary the filtration step

of the waste.

5. The process parameters established for the aqueous radioactive waste treatment are:

• Waste characteristics:

-total disolved salts (TDS): maximum 2.5g/l

-suspension material after filtration: max. 10 mg/l

-pH: 1-8

-maximum activity conc.: 0.1 Ci/l

• Resin type:

-strong acid cationit: NRW1600

-stong base anionit: NRW5050

-cationit/anionit ratio: 1/1 ÷ 1/3

• Operating parameters:

- flow rate : 10-20 BV/h

- flow velocity: 1-5 m/h (good mass transfer)

- column aspect ratio, H:W = 2:1 – 4:1

- resin-waste contacting time: 3-6 min

- Operating system: gravitational flow

- operating pressure: room presure

- operating temperature: room temperature

- Alloy type for plant components: stainless steel V4A

6. In order to establish the process performances and control, the following analysis are required to

be performed:

• conductivity (mS/cm)

• pH

• radioactivity (Bq/l)

• total dissolved salt content (mg/l)

7. The performances obtained were:

• volume reduction factor: 350 - 800

• decontamination factor: max. 104

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8. The ionic mass specific consumption varies as a function of the dissolved substances content of

the radioactive waste. Thus, in the tests performed at laboratory scale, the specific cationit NRW1600

consumption was 0.001 liter/waste liter. At a concentration higher then 2.5 g/l dissolved substances, the

method becomes inefficient due to the high ion exchange consumption.

9. From the waste treatment process, it will result solid waste composed from spent ion exchangers,

spent filtration cartridges, settled slurry and eventually the organic phase. The secondary waste will be

conditioned for final disposal, using technologies available at ICN Pitesti for this type of waste.

10. The decontamination waste that uses concentrated solutions are imposible to treat efficiently at

an accessible cost price, through any method, due to the high content of complexing agents. Since these

methods are less and less used, and the generated waste volumes are small, it is recommended the

separate collection, oxidation of organic compounds, neutralization and possibly supernatant treatment

through ionic exchange and direct conditioning of the slurry.

The treatment plant for aqueous radioactive waste arising from CNE Cernavoda must be built from

stainless steel corrosive resistant media; it must be tight in order to eliminate the risk of some damages

that can compromise the security during the treatment process.

7.2 Cost consideration

Before selecting a particular ion exchange treatment process, a complete economic evaluation

should be conducted. The main cost components to be considered are:

- The cost of capital

- The initial cost of the ion exchange media

- The operating costs

- The costs associated with the treatment and disposal of the spent ion exchanger.

The aqueous radioactive waste treatment technology elaborated in the framework of WATER

Project recommends the installation a new ion exchange system in view of decreasing costs with meeting

the effluent quality requirements.

The total cost of operating process depends on the desired treatment rate and the decontamination

factor. But the cost alone is not always the most important factor in the selection process, and more

expensive options may be selected to meet other technical requirements, such as the availability of

materials or conformity with spent ion exchanger treatment systems existing on-site. So, we didn’t choose

high selective synthetic inorganic materials because the spent ion exchanger treatment and conditioning

systems available on-site are not proper for these materials. Ideally, the secondary radwaste should be

72

treated through existing on-site radioactive waste treatment systems. If new facilities are required, the

cost of providing such facilities must be taken into consideration.

Because the nuclear grade resins from these applications are disposed of as radwaste, high capacity

and long resin bed life are critical to minimizing radwaste disposal cost and volume. For most users,

radwaste disposal cost will often exceed resin purchase cost, so high resin capacity directly translates into

savings in these non-regenerable nuclear applications.

Furthermore, longer bed life means fewer bed change-outs, less work, less resin handling, and less

chance for radiation exposure.

RATEN ICN Pitesti applies the treatment of aqueous radioactive waste by ion exchange, for waste

generated in the decontamination processes. The technology used was licensed in 2000, being designed

for treatment of radioactive liquids from the decontamination centre of Cernavoda NPP. Nowadays, this

technology, based on the pair of ion exchangers (cationite and anionite): C100H Purolite/A600 Purolite,

is implemented in a mobile, compact treatment plant, for processing small volumes of waste with variable

composition, from Cernavoda NPP.

The treatment efficiency obtained with the pair of resins NRW1600/NRW5050 was compared with

the same parameter obtained for the pair C100H/A600. This leads to the conclusion that by using the

nuclear grade resins recommended by us, the volume reduction factor can be increased with about 40%.

The analysis of the economic efficiency shown that by changing the ion exchangers with nuclear grade

types, despite of the increased price of the last one, an overall reduction of costs might be achieved due to

the reduction of the secondary waste production which will lead to a decrease of the disposal costs.

Changing the aqueous waste treatment technology existent at RATEN ICN Pitesti, in accordance

with the new technology, elaborated under WATER Project, leads to a 200% increase in the price of the

resin. In spite of this, generally speaking, the technology developed is technically-economically more

efficient, due to the fact that the costs associated with the treatment and disposal of the spent ion

exchanger decrease with 40%, costs that at this moment are approximately eight times the resin

acquisition cost. Thus, an approximate calculation of the integrated process cost for aqueous radioactive

waste treatment, spent resins treatment by using on-site radioactive waste treatment systems and their

disposal show that this cost will be reduced by approximately 16% through the application of the

technology developed under WATER Project, maintaining the effluent quality and reducing the final

volume for the final disposal step.

ACKNOWLEDGEMENTS

The authors gratefully acknowledge the financial support provided by Europe Union, Romania-Bulgaria

Cross Border Cooperation Programme, in the form of the research project.

73

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