Journal Search Engine
Download PDF Export Citation Korean Bibliography PMC Previewer
ISSN : 1738-1894(Print)
ISSN : 2288-5471(Online)
Journal of Nuclear Fuel Cycle and Waste Technology Vol.17 No.4 pp.363-373
DOI : https://doi.org/10.7733/jnfcwt.2019.17.4.363

Validation of Radioanalytical Techniques for Nuclear Waste Characterisation

Phillip E. Warwick
University of Southampton, National Oceanography Centre, Southampton, European Way, Southampton SO14 3ZH, UK
Corresponding Author. Phillip E. Warwick, University of Southampton, E-mail: pew@noc.soton.ac.uk, Tel: +44-23-8059-6600
June 25, 2019 July 3, 2019 October 21, 2019

Abstract


Waste characterisation associated with nuclear site decommissioning relies on radiochemical analysis of a diverse range of sample types, requiring extensive validation of analytical techniques using matrix-matched materials. The absence of relevant reference materials has hindered robust method development and validation. The paper discusses how method validation in support of nuclear waste characterisation can be achieved without using reference materials. The key stages in an analytical procedure are evaluated and a multi-stage approach is proposed with the ultimate aim of determining an operational envelope for an analytical procedure.



초록


    © Korean Radioactive Waste Society. All rights reserved.

    This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/by-nc/3.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited.

    1. Introduction

    The requirement for radionuclide characterisation in support of nuclear decommissioning has resulted in a diverse range of material types requiring analysis. Sample types including ion exchange resins, concretes, desiccants, graphite, plastics, ferrous and non-ferrous metals, insulating material, sludges with varying and often undefined compositions and oils pose a significant challenge to the radioanalytical laboratory [1]. To underpin data quality, the laboratory must confirm that the technique selected for analysis is fit for purpose and all techniques must be validated to confirm this. The validation process is designed to assess the performance of the technique in terms of analytical accuracy, precision, sensitivity and robustness and to compare these parameters with the required performance targets. Historically, the validation assessment has relied on the analysis of reference materials of known composition. Reference materials consist of the analyte at a known concentration associated with a specific matrix that can be used to evaluate the entire analytical separation and measurement process including the efficiency of sample dissolution, chemical separation and measurement accuracy. A number of international programmes have historically supplied characterised reference material representing a range of matrices. However, such programmes have focussed on either environmental reference materials or on very specific sample types for a particular application. With the new demands arising from nuclear decommissioning and the associated range of material types prevalent, the existing range of reference materials do not reflect the matrices now being encountered. A limited number of reference materials have been developed to support nuclear waste characterisation (e.g. [2]) and there are plans for developing more representative reference standards [3]. However, it is improbable that a full range of reference materials covering all potential combinations of radionuclide and matrix could practically be produced. Even for a single matrix there are likely to be variations in composition and associations of the analyte with the matrix depending on the origin of the sample that could potentially impact on the analytical technique. The lack of appropriate reference materials has been identified as a significant risk to delivery of robust method validation [4]. A new overarching approach to method validation is therefore required. One proposed approach is to consider the component stages of the analytical technique and to evaluate each stage separately, assessing the key risk factors that would impact on each stage. Such an approach does not require certified reference materials and provides a more flexible validation procedure that can be applied to any analyte and sample type, taking into account variations in sample composition that may occur even within a single matrix type. By considering sample dissolution and analyte separation independently, matrixrelevant solution standards can be prepared that are more representative of the materials encountered and which can be used to evaluate the separation stages of an analytical process. Independent verification of sample dissolution efficiency can be performed through a combination of supporting research on analyte chemistry and likely association with matrices and experimentally-determined leaching efficiencies using operationally-exposed samples. The approach also permits sample types to be grouped resulting in a more cost-effective validation programme. Where available, reference materials can still be utilised as an additional verification stage and to support the more generic approach proposed.

    2. Factors affecting an analytical procedure

    A radioanalytical technique can be broadly divided into four key stages

    • (i) Sample preparation (storage, homogenisation, subsampling, drying / igniting)

    • (ii) Sample dissolution (acid or alkaline leaching, total digestion, fusion).

    • (iii) Analyte isolation / purification (separation of the analyte from the matrix elements and potential chemical and radionuclide interferences.

    • (iv) Quantification of the analyte.

    All four stages must be assessed to demonstrate the overall performance of the technique and the impact that sample composition may have on the efficiency of each stage (Fig. 1). Of these, stages 2, 3 and to some extent 4 have typically relied on reference materials. However, it is possible to validate each of the stages individually without using a reference material and to provide the necessary data to demonstrate the performance of the technique as a whole.

    3. Sampling / sample storage / preparation– k(prep)

    Procedures for sample collection, storage and preparation must maintain the integrity of the sample matrix and minimise potential analyte loss through volatilisation or precipitation / adsorption to sampling equipment / containers. In addition, knowledge of the impact of storage conditions on sample / analyte stability over time is required to assess whether a sample may have been compromised prior to receipt at the laboratory. Although volatilisation-based techniques can offer an efficient means of radionuclide extraction, uncontrolled losses during sampling and storage can result in significant errors in the analytical measurement. The degree of volatility will depend on the radionuclide and its association with the matrix which in turn will depend on the radionuclide origin (e.g. contamination or insitu production via activation) and matrix composition. Radionuclides prone to loss via volatilisation include 3H, 14C, 99Tc, 106Ru and 129I. Of these, 3H has been identified as being particularly prone to volatilisation during collection and storage of decommissioning samples [5]. However, in concretes and metals, 3H produced in situ via neutron activation of the matrix is relatively non-volatile compared with 3H arising from surface contamination of the matrix [6,7]. Loss of volatile radionuclides from solid samples is effectively minimised through refrigeration or freezing of the sample [5]. For aqueous samples, precipitation and adsorption onto containers can be minimised through effective pH control. However, there are limited data on the stability of radionuclides in decommissioning wastes over time.

    Sample heterogeneity can pose a significant challenge in representatively sampling a material. For solid samples, radionuclides may be present as surface contamination and as such the measured activity concentration will be highly dependent on the quantity of underlying non-active matrix that is co-sampled with the contamination layer. This is particularly noted for metals where contamination is unlikely to penetrate the surface and for painted items where the contamination may be associated predominantly with the paint layer. For aqueous samples, radionuclide partitioning is highly probable between aqueous and particulate phases.

    Sample drying / ignition can result in the loss of volatile radionuclides. Freeze drying can minimise the loss of certain volatile radionuclides but quantitative data are not available. Sample ignition can also render certain radionuclides less soluble. Hou [8] noted that recoveries of both 55Fe and 63Ni decreased significantly if irradiated graphite samples were heated for short periods at 1100℃ or for extended periods (>6 hrs) at lower temperatures (800℃).

    4. Dissolution efficiency, k(sol)

    4.1 Factors affecting dissolution efficiency

    In the majority of radioanalytical techniques, a dissolution stage is required to solubilise the analyte and provide a solution suitable for subsequent separation stages. Typically, the dissolution stage involves either an acid leach of the solid sample (typically using HNO3, HCl or a combination of these acids), total dissolution (whereby the entire matrix of the sample is dissolved, often using HClO4 and / or HF) or a total fusion technique whereby the sample is intimately mixed with a flux and heated to produce a homogeneous melt. Organic-rich matrices such as ion exchange resins can be decomposed using oxidants such as Fenton’s Reagent (a mixture of H2O2 and Fe2+). For certain radionuclides (notably 36Cl and 129I) an alkaline digest or alkaline fusion is used to solubilise the radionuclide and prevent its loss by volatilisation. A summary of dissolution approaches for radionuclide analysis in decommissioning materials was reported by Croudace et al [9].

    The choice of dissolution technique will be dependent on the analyte, the composition of the matrix and the origin of the radionuclide (Table 1). For fission products, the nuclide is most likely to be present as a result of contamination of the material and therefore loosely associated with the matrix. The one exception would be for materials that may contain U and produce fission products in situ following neutron irradiation of the material.

    For activation products, the radionuclide may have been formed in situ as a result of irradiation of the matrix and as such may be more strongly bound within the matrix requiring a total dissolution technique to quantitatively liberate it. Some radionuclides may also require a more aggressive dissolution technique to render them soluble due to their chemical form even if they are not strongly associated to sample matrix (e.g. Pu if present as the refractory PuO2).

    Some information exists in the literature relating to sample dissolution (Table 2). It is therefore possible to use this information to assess the suitability of a proposed dissolution technique. In addition, dissolution of radionuclides formed in situ by irradiation of the matrix can be assessed by determining the dissolution efficiency of the corresponding stable element. For example, the dissolution efficiency of 55Fe in an irradiated ferrous metal can be assessed by measuring the proportion of stable Fe that is recovered during the dissolution process. Such an approach is not appropriate for radionuclides which are present as a result of contamination where the contaminant radionuclide can exhibit substantially different behaviour compared with stable element analogues in the bulk matrix. In general, it could be assumed that a total dissolution-based technique would quantitatively solubilise all radionuclides and only the potential loss of the radionuclide due to volatility would need to be assessed.

    4.2 Concrete

    For structural concretes, radionuclide contamination arises from operational exposure to a contaminated atmosphere / fluid and is likely to be surface bound or adsorbed to mineral phases. In general, leaching of the sample should be sufficient to recover such surface contamination. Pu in bioshield concretes (where the Pu is most likely derived from contamination rather than being formed in situ) has been shown to be leachable and sequential extraction studies have suggested that the majority of the Pu is associated with carbonate and Fe/Mn oxyhydroxide phases [14]. Such phases would be readily dissolved via acid leaching.

    For irradiated concretes (e.g. bioshield concretes), in addition to contamination, radionuclides such as 3H, 14C, 41Ca, 55Fe, 63Ni and 152Eu may be formed in-situ and can be intimately associated with the component mineral phases rendering them less accessible to leaching. Hou [10] compared combustion and H2SO4:HNO3:HClO4 acid digestion techniques for the determination of 3H and 14C in bioshield concretes. Acid digestion data biased low compared with the combustion technique suggesting non-quantitative digestion / analyte recovery. Tritium is produced in-situ via neutron capture of minerogenic 6Li and is strongly bound within mineral phases within the concrete. To liberate radionuclides from such phases, a total digestion procedure is required. These observations are consistent with Kim et al [6] who demonstrated that 3H produced in–situ via neutron activation was non-leachable with water whereas 3H arising from deposition / contamination was readily leachable. However, Hou [8] demonstrated that a significant proportion of activation-derived 55Fe and 63Ni can be leached from bioshield concretes when compared with NaOH / Na2CO3 fusion suggesting, that 55Fe and 63Ni are associated with a more readily soluble phase. Fusion using lithium borate has also been used to achieve complete dissolution of the concrete matrix and has been applied to the analysis of 41Ca [11] Pu, U and 90Sr in concretes [12].

    4.3 Graphite

    Radionuclides including 3H, 14C, 36Cl, 55Fe and 63Ni in graphite arise from neutron activation of the matrix and associated trace contaminants. Hou [8] compared acid leaching with aqua regia, acid digestion using H2SO4: HNO3:HClO4 and ashing / acid digestion with HCl for the recovery of 55Fe and 63Ni from irradiated graphites. The recoveries of both 55Fe and 63Ni were low for acid leaching compared with total dissolution techniques. Comparison of H2SO4:HNO3:HClO4 acid digestion and combustion techniques for the determination of 3H and 14C in reactor graphite confirmed that there was no significant difference between the two techniques, indicating that the acid digestion procedure was effective at recovering the radionuclides from the graphite matrix [10]. Ignition and dissolution of the residue in HCl has been used to solubilise graphite prior to radiochemical analysis of gamma emitting radionuclides, actinides, 41Ca, 55Fe, 63Ni, 90Sr, 93Mo, 93Zr and 126Sn. A similar approach was adopted for carbon bricks, although an insoluble residue resulted which required microwave digestion in a mixture of HBF4, HCl, H2O2 and HNO3 to effect total dissolution [15]. Sealed vessel microwave digestion of graphite using H2SO4 / HNO3 has also been utilised for the analysis of Pu, U and 90Sr in graphite although the approach was not validated on operationally-exposed materials [12].

    4.4 Metals

    A wide range of metals and metal-alloys are encountered during decommissioning with the choice of dissolution procedure being dictated by the sample composition. These include (but are not limited to) ferrous metals (wrought / cast iron, steels, stainless steels), aluminium, cadmium, copper, lead, Zr alloys and Inconel. Radionuclides may be present either as surface contamination or may be formed by neutron activation of the metal and hence integral with the matrix (Table 1). In general, a total dissolution procedure is favoured to ensure quantitative recovery of the analyte. Ferrous metals such as steels can be dissolved in aqua regia prior to analysis of actinides and 90Sr [12]. HNO3 has been used for the dissolution of Cd prior to measurement of neutron-induced 113mCd [13]. 8M HNO3 and 9M HCl have been used to dissolve irradiated lead and aluminium respectively prior to measurement of 55Fe and 63Ni [8]. Sodium metal and NaK alloy have been used in fast breeder reactor programmes. Dissolution of these metals must be undertaken with extreme caution given their high reactivity but can be achieved by dissolving small pieces in methanol or, to further reduce reaction rates, propanol.

    4.5 Resins / desiccants

    Ion exchange resins are routinely used for water circuit purification and therefore activation products tend to be adsorbed to the material. Ignition to decompose organics followed by acid leaching of the residues will effectively recover most transition metal species. Any 3H or 14C would be liberated during the ignition stage and require trapping. Fenton’s reagent has also been used to decompose the organic resin and liberate radionuclides prior to separation. High pressure microwave digestion using AgNO3 + HNO3 has been used to solubilise ion exchange resins prior to 36Cl analysis [16].

    Desiccants, typically alumina used in gas dryers, contain predominantly 3H and 36Cl (and potentially 14C associated with graphite dust from graphite moderated reactors). Extraction of these radionuclides via volatilisation is preferred but where a radiochemical separation of 36Cl is required, the desiccant may be digested using a strong alkaline digest under reflux.

    4.6 Sludges

    The term ‘sludge’ covers a range of materials associated with nuclear reactor operation and fuel reprocessing and is defined more from their physical properties rather than their chemical composition. Acid digestion using a mixture of HNO3, HCl and HF has been used to solubilise radionuclides including 126Sn [17]. Alkaline fusion, using KOH-KNO3 has been used to solubilise solid fractions of diluted ‘feed materials’ [18].

    4.7 Soft wastes

    Soft wastes include gloves, tissues rags, protective clothing etc that become contaminated through contact with loose contamination during routine operations in the nuclear facility. Soft wastes are a particular challenge for sample dissolution as the sample can comprise multiple material types. In general, radionuclides associated with soft wastes are present as surface contamination and not integral to the sample matrix. Contamination may also be highly heterogeneous and care must be taken to ensure representative sampling prior to digestion. In general, ignition followed by acid leaching is used for recovering acidsoluble radionuclides.

    5. Analyte separation – k(sep)

    The separation stage of the analysis is designed to isolate the analyte from the bulk matrix elements, to remove any elemental or radionuclide interferences and to concentrate the analyte into a form suitable for its measurement. During the dissolution stage, a range of matrix elements will be solubilised along with the analyte and these elements may have an adverse affect on the subsequent separation resulting in low analyte recoveries. Some matrix elements may follow the analyte through the separation stage and interfere with the final measurement (e.g. lanthanides in 241Am measurement by alpha spectrometry where the alpha spectrum will be severely degraded if lanthanides are present in the measured source). In addition certain radionuclides may co-separate with the analyte resulting in a radiometric interference.

    Matrix elements can interfere with the efficiency of analyte separation through a number of mechanisms.

    • (i) Co-precipitation of the analyte on an unplanned, matrix associated precipitate (loss of analyte).

    • (ii) Modification of the degree of analyte inclusion, adsorption or occlusion within a planned precipitation stage.

    • (iii) Specific, direct matrix competition with the analyte on extraction media where the chemical properties of the analyte and interferent are similar.

    • (iv) Ion-pair formation / complexation with the analyte and subsequent modification of analyte chemistry.

    • (v) Non-specific matrix competition with the analyte through modification of ionic strength, viscosity and diffusive properties of solution.

    • (vi) Analyte / matrix co-extraction and incorporation into final source affecting the source / measurement characteristics.

    The range and concentration of matrix elements and radionuclides present in the solution will vary depending on sample type, the mass analysed and the chosen dissolution technique. However for each technique, it is possible to determine typical concentrations in digests based on the solubilisation efficiency, km(sol), and set upper limits on elemental composition for classes of matrices. Simulant test solution containing elements at these upper concentration limits can then be prepared and spiked with the analytes and interferents of interest and the separation procedure tested using these simulants. The chemical recovery of the analyte, the concentration of matrix elements in the purified fraction and interference decontamination factors can then be determined providing information on the robustness of the technique. Replicate tests would also provide information on the precision of the technique. Where new matrices are encountered, the composition of the digest can be determined to confirm that the elemental concentrations below the values that would adversely affect the separation. It is assumed that the presence of elements at lower concentrations would not result in any adverse effects unless the element forms an intrinsic part of the separation procedure (e.g. where matrix-derived Fe is used to produce Fe(OH)3 precipitates to co-precipitate actinides). Such instances can be identified and where necessary additional carrier added. Interferent separation can be determined using either added tracers or the stable element analogue arising from the matrix where present. Decontamination factors can then be determined and limits on the concentrations of interference species calculated. These data can be used in defining the operational envelope for the analytical procedure.

    6. Analyte quantification - ɛA

    The accuracy of the final measurement is often assessed through the use of reference materials. However, accuracy can also be readily assessed through participation in proficiency tests and through the analysis of stimulant solutions spiked with a traceable standardised solution of the analyte. For alpha spectrometry, the correlation between the reference solution of the analyte and the tracer provides independent confirmation of the accuracy of the measurement. For liquid scintillation analysis, alternative calibration approaches (such as Ciemat NIST method) can be used to independently confirm instrument calibration. The impact of isobaric / radiometric interferences can also be assessed. The presence of such interferences can be evaluated through a review of the sample composition and likely co-associated radionuclides. An upper limit on the concentrations of interferences can be determined using data on the degree of solubilisation, kI(sol), of the interference during sample dissolution and the degree of separation, kI(sep), during analyte purification. Representative sources containing the interferent can then be prepared and the contribution of the interferent, ɛI, to analyte measurement can be assessed. Assessment of measurement accuracy in the absence of a matrix-matched reference material is therefore readily achievable. Such data can be used in conjunction with data on the separation to define an operational envelope for a measurement

    7. The validation approach

    The following stages are proposed for the assessment of the suitability of a procedure for radioactive waste characterisation (Fig. 2).

    • (i) Collation of information to provide a scientifically underpinned assessment of expected method performance and potential impact of matrix elements.

    • (ii) Assessment of radionuclide volatility and distribution within the matrix to inform sample storage and preparation strategy.

    • (iii) Assessment of the chemical composition of the sample and the recovery of matrix elements during the dissolution stage to provide data on the composition of the solution presented to the chemical separation stages.

    • (iv) Identification of potential isobaric or radiometric interferences arising from the matrix or associated radionuclides.

    • (v) Assessment of the robustness of the separation technique over a defined range of leachate compositions in order to define an operational envelope for the technique. This stage will assess the impact of matrix elements on the separation efficiency and determine the degree of separation of the analyte from potential interferences. Use of experimental design can be utilised to reduce the number of tests required.

    • (vi) Assessment of measurement performance and the impact of isobaric / radiometric interferences

    8. Conclusions

    The diversity of matrices requiring analysis in support of nuclear waste characterisation poses a significant challenge to the radioanalytical laboratory. Careful consideration of the key stages in an analytical procedure and knowledge of the material being characterised is critical in evaluating the performance of a technique. In the absence of matrix-relevant reference materials, much of the method validation can be achieved using simulants that reflect the composition of the sample that is being presented to the separation and measurement stages of the procedure. However, validation of the sample preparation and dissolution stages is more complex. Knowledge of sample composition and analyte association is critical. In particular, radionuclide association will be highly dependent not only on the type of material but also the origin of the radionuclide within that material and this must be carefully evaluated during the validation process. In practice, a multi stage approach to method validation is required to ensure that the impact of sample type on analytical measurement is fully understood. In achieving this, an operational envelope for the procedure can be defined in terms of matrix composition, sample mass and radionuclide activity concentrations and this can be used to assess the suitability of the procedure for samples submitted for analysis.

    Figure

    JNFCWT-17-4-363_F1.gif

    Interaction of analyte, matrix and interferences and impact on analyte separation / detection.

    JNFCWT-17-4-363_F2.gif

    Critical information for multi-stage validation of an analytical procedure.

    Table

    Radionuclide association with matrices

    Review of literature on dissolution techniques

    Reference

    1. P.E. Warwick, I.W. Croudace, and R. Marsh, “Maintaining radioanalytical data quality: The challenges of decommissioning analysis”, Nucl. Future, 7, 44-47 (2011).
    2. A. Harms and C. Gilligan, “Development of a neutron-activated concrete powder reference material”, Appl. Radiat. Isot., 68, 1471-1476 (2010).
    3. S.M. Judge and P.H. Regan, “Radionuclide metrology research for nuclear decommissioning”, Rad. Phys. Chem, 140, 463-465 (2017).
    4. S. Parry, “Quality assurance in the nuclear sector”, Radiochim. Acta, 100, 495-501 (2012).
    5. D. Kim, I.W. Croudace, and P.E. Warwick, “The requirement for proper storage of nuclear anbd related decommissioning samples to safeguard accuracy of tritium data”, J. Haz. Mater., 213-214, 292-298 (2012).
    6. D. Kim, P.E. Warwick, and I.W. Croudace, “Tritium speciation in nuclear reactor bioshield concrete and its impact on accurate analysis”, Anal. Chem., 80, 5476- 5480 (2008).
    7. I.W. Croudace, P.E. Warwick, and D. Kim, “Using thermal evolution profiles to infer tritium speciation in nuclear site metals: An aid to decommissioning”, Anal. Chem., 86, 9177-9185 (2014).
    8. X. Hou, “Determination of 63Ni and 55Fe in nuclear waste samples using radiochemical separation and liquid scintillation counting”, Anal. Chim. Acta, 535, 297 -307 (2005).
    9. I. W. Croudace, B.C. Russell, and P.E. Warwick, “Plasma source mass spectrometry for radioactive waste characterisation in support of nuclear decommissioning: a review”, J. Anal. At. Spectrom., 32, 494-526 (2017).
    10. X. Hou., “Rapid analysis of 14C and 3H in graphite and concrete for decommissioning of nuclear reactor”, Appl. Radiat. Isot., 62, 871-882 (2005).
    11. P.E. Warwick, I.W. Croudace, and D.J. Hillegonds, “Effective determination of the long-lived nuclide 41Ca in nuclear reactor bioshield concretes: Comparison of liquid scintillation counting and accelerator mass spectrometry”, Anal. Chem., 81, 1901-1906 (2009).
    12. D. Zapata-García and H. Wershofen, “Development of radiochemical analysis strategies for decommissioning activities”, Appl. Radiat. Isot., 126, 204-207 (2017).
    13. P.E. Warwick and I.W. Croudace, “Measurement of 113mCd in irradiated cadmium shielding using liquid scintillation counting, In Mobius S., Noakes J. & Schonhofer F. LSC 2008, Advances in Liquid Scintillation Spectrometry”, Radiocarbon, University of Arizona, USA (2009).
    14. J. Qiao and X. Hou, “Frationation of plutonium in environmental and bio-shielding concrete samples using dynamic sequential extraction”, J. Env. Rad., 101,, 244-249 (2010).
    15. B. Bisplinghoff, M. Lochny, J. Fachinger, and H. Brücher, IAEA Technical Committee meeting : Nuclear graphite Waste Management, 153-159, October 18-20, 1999, Manchester, UK.
    16. C. Fréchou and J-P Degros, “Measurement of 36Cl in nuclear wastes and effluents: validation of a radiochemical protocol with an in-house reference sample”, J. Radioanal. Nucl. Chem., 263, 333-339 (2005).
    17. S. Dulanská, B. Remenec, J. Bilohuščin, L. Mátel, and M. Bujdoš, “Development of 126Sn separation method by means of anion exchange resin and gamma spectrometry”, Appl. Radiat. Isot., 123, 128-132 (2017).
    18. PNNL Inorganic and radiochemical analysis of AW- 101 and AN-107 ‘diluted feed’ materials, Pacific North-West National Laboratory, Washington, USA, Report PNWD-2463 Rev 1 (1999).
    1. SEARCH
    2. JNFCWT

      Online Submission

    3. Korean Radioactive
      Waste Society (KRS)

    4. Editorial Office
      Contact Information

      - Tel: +82-42-861-5851, 866-4205
      - Fax: +82-42-861-5852
      - E-mail: krs@krs.or.kr