technical report - nagra · 2018. 9. 25. · for controlling important radionuclide concentrations...

Nagra Nationale Genossenschaft fOr die Lagerung radioaktiver Abfalle

Cedra Soch~te cooperative nationale pour I'entreposage de dechets radioactifs

Cisra Societa cooperativa nazionale per I'immagazzinamento di scorie radioattive

TECHNICAL REPORT 86-28

Preliminary Study on the Use of Correlations for Controlling Important Radionuclide Concentrations in Waste Packages

c. Degueldre S. Huwyler

December 1986

Swiss Federal Institute for Reactor Research, Wurenlingen

Parkstrasse 23 5401 Baden/Schweiz Telephon 056/2055 11

Der vorliegende Bericht wurde im Auftrag der Nagra erstellt.

Die Autoren haben ihre eigenen Ansichten und Schlussfolge

rungen dargestellt. Diese müssen nicht unbedingt mit den

jenigen der Nagra übereinstimmen.

Le présent rapport a été préparé sur demande de la Cédra.

Les opinions et conclusions présentées sont celles des

auteurs et ne correspondent pas nécessairement à celles

de la Cédra.

This report was prepared as an account of work sponsored

by Nagra. The viewpoints presented and conclusions reached

are those of the author(s) and do not necessarily represent

those of Nagra.

NAGRA NTB 86-28 - 1-

ABSTRACT

Studies concerning correlations between "key" nuclides easily detected by high resolution gamma spectrometry and safety relevant radionuclides defined for a reference type B repository are reviewed. This survey allows an evaluation of the control of radionuclide concentrations in the waste from nuclear power plant operation by defining the detection limits of the relevant nuclides. These limits are a function of the "key" nuclide detection limits, which are themselves dependent on waste package, canister and detection characteristics: that is, isotopic and waste package compositions, shielding, distance between canister and detector, acquisition time and counting efficiency. On the basis of calculations carried out for typical packaged operational wastes from various nuclear power plants, recommendations for the use of correlations and high resolution gamma spectrometry are suggested. In all cases, the determination of the relevant nuclides is possible by correlation because the indirect detection limits of these nuclides are always lower than the concentration limits defined for the Swiss repository for low- and intermediate level activity waste considered in "Project Gewaehr 1985".

NAGRA NTB 86-28 - II -

RESUME

En vue de réaliser le contrôle final de qualité des fûts contenant les déchets radioactifs de maintenance des centrales nucléaires suisses, les études de corrélation entre nuclides-clefs aisément détectables par spectrométrie gamma de haute résolution et les nuclides critiques définis pour le dépôt de reference du type B sont évaluées en vue de définir les limites de détection des nuclides importants. Ces dernières sont fonction des limites de détection des nuclides-clefs qui elles-mêmes dépendent du fût, c'est à dire de la composition du matériel de remplissage et des sources isotopiques, du blindage et de la détection, à savoir: de la distance fût/détecteur, du temps de comptage et de l'efficacite du détecteur. L'étude, sur base des calculs réalisés pour différents exemples typiques de fût de déchets provenant de diverses centrales nucléaires, énonce quelques recommandations quant à l'utilisation des corrélations et du dosage des nuclides-clefs par spectrométrie gamma de haute résolution. Dans tous les cas, la détermination des nuclides importants est possible, les limites de détection définies par corrélation étant inférieures aux concentrations limites de ces nuclides critiques pour le dépot suisse de basse et moyenne activité considéré dans project "Gewaehr" 1985.

NAGRA NTB 86-28 - 111 -

ZUSAMMENFASSUNG

Es wird eine Literaturübersicht über Studien gegeben, die Korrelationen zwischen Leitnukliden, die mit hochauflösender gamma-Spektrometrie leicht bestimmt werden können, und sicherheitsrelevanten Radionukliden behandeln, die für ein ReferenzendIager Typ B definiert wurden. Diese Übersicht erlaubt die Beurteilung der Kontrolle der Radionuklidkonzentrationen in radioaktiven Abfällen aus dem Betrieb von Kernkraftwerken, indem die Nachweisgrenzen der kritischen Nuklide definiert werden. Diese Nachweisgrenzen sind abhängig von denjenigen der Leitnuklide, die ihrerseits von den Charakteristiken der Abfallpackung abhängen, d.h. der Isotopenzusammensetzung, Zusammensetzung der Verpackung, Abschirmung, Abstand zwischen Detektor und Behälter, Messzeit und Wirkungsgrad des Detektors. Aufgrund von Berechnungen, die für typischen Betriebsabfall aus Kernkraftwerken durchgeführt wurden, können Empfehlungen für den Gebrauch von Korrelationen und die Anwendung der hochauflösenden gamma-Spektrometrie gemacht werden. In allen Fällen ist die Bestimmung der relevanten kritischen Nuklide mittels Korrelation möglich, da die indirekten Nachweisgrenzen dieser Nuklide immer kleiner sind als die für ein schweizerisches Endlager für schwach- und mittelaktiven Abfall in "Projekt Gewaehr 1985" definierten Grenzkonzentrationen.

NAGRA NTB 86-28 - IV -

CONTENTS

1. Introduction 1

2. Radionuclide concentration limits 2

3. Scaling factor evaluation and discussion 4 3.1 Calculation approach 6 3.2 Experimental approach 7

4. Determination of the "key" nuclide detection limits 11 4.1 Determination of the background BG 14 4.2 Considerations of the counting efficiency EF 15 4.3 Determination of the proportional factor PF 15 4.4 Calculation of the detection limit SA(DL/KN) 16

5. Results and discussion 17 5.1 DL calculation and correlation, example: BA-KKB1-1B 18 5.1.1 System description 18 5.1.2 SA(DL/Co-60), SA(DL/Cs-137), SA(DL/Ce-144) results

and correlations 19 5.2 Spectrum calculation, example: BA-KKMl 20 5.2.1 Activity, flux and counting spectrum propagation 20 5.2.2 SA(DL/Co-60), SA(DL/Cs-137), SA(DL/Ce-144) results

and correlations 24 5.3 Effect of acquisition time on SA(DL/KN),

example BA-KKL-1c 25 5.3.1 SA(DL/Co-60) , SA(DL/Cs-137) , SA(DL/Ce-144) results

and correlations 26 5.3.2 Effect of the acquisition time on SA(DL/RR) 26 5.4 Effect of detection dIstance on SA(DL/KN),

example: BA-KKG-2 27 5.5 Most pessimistic case for the detection limits 29

6. Conclusions and recommendations

7. List of symbols and parameters

References

30

32

33

NAGRA NTB 86-28 Page 1

1 INTRODUCTION

Operation of nuclear power plants results in production of radioactive waste. Before disposal of this waste to a final storage site, it is necessary to verify its radionuclide content. "Project Gewaehr 1985" provides tentative maximum allowable specific activities or concentrations of nuclides in the waste packages which are derived from the safety analyses of the repository considered in this Project. However, many of these nuclides are difficult or impossible to measure by direct counting techniques. Their accurate determination would require sampling, chemical separation, and expensive analyses. Since correlations between measurable gamma emitting nuclides, called "key" nuclides (KN), and these safety relevant radionuclides (RR) have been observed and could be used for assessment or control, they have been studied by several organisations during recent years.

The idea of using "key" nuclides for assessing concentrations of other nuclides in the radioactive waste from nuclear power plants has been developed in the U.S. and a good description of the rationale is given by <LIE 83>: "During the operation of a nuclear reactor, large quantities of radioactive materials are formed in the fuel, coolant and structural components, and any reactor which has operated for any length of time at a high power level contains a large inventory of radioactive materials. Under ordinary operating conditions, most (>99.9%) of the fission products are maintained within the fuel itself and do not become available for potential shipment to a shallow land burial ground. That small portion «0.1%) which does escape into the coolant and that portion of the

- structural materials which dissolve into the coolant, are free to circulate throughout the entire primary system, including those components of the reactor coolant system designed to remove them. Most of this circulating material is held within the reactor coolant system, and the major means of removal of radioactive material therefrom are the filters, resins and evaporator bottoms, waste streams designed to be the major sources of waste shipments from a reactor ... While some isotopes (for example Cs-137 and Co-60) are readily measurable with little difficulty, others (such as Tc-99, 1-129, Sr-90, C-14) are much more difficult and expensive to measure. Therefore, it would be highly desirable to develop an acceptable procedure by which readily measured "key" isotopes were analyzed in the reactor coolant, and the less readily measured isotopes ratioed to them. The ratios determined for the reactor coolant would be used throughout the system unless there is a good physical reason to do otherwise. Then the key isotopes would be measured in all streams and the other isotopes of concern would be calculated through the use of these ratios or scaling factors."


The first goal of the present study is to evaluate the literature available on this topic. The second is to assess the applicability of the "key" method in the Swiss case using the range of published scaling factors, i.e. on the basis of preliminary radionuclide concentration limits for disposal and typical waste package description. Calculations have been carried out in order to determine the detection limits of the "key" nuclides in these waste packages, and to apply correlations so that a comparison can be made between detection limits of relevant nuclides and their allowable concentration using the proposed methodology.

2 RADIONUCLIDE CONCENTRATION LIMITS

The development concentration of reduced if

involved relevant

in using scaling factors to estimate the radionuclides in waste packages can be greatly

- the level and the required accuracy of the estimates are fixed - the radionuclide spectrum, matrix composition and package characteristics

affecting the key nuclide measurement are known.

The radionuclide concentration limits used are taken from Project Gewaehr 1985 <NGB 85>. This project is a feasibility study of safe disposal of the radioactive waste produced in Switzerland. The type B repository for lowand intermediate-level waste considered in this study was planned at one of the three potential sites under current investigation (Oberbauenstock). The doses predicted by the safety analyses after repository sealing were converted into radionuclide concentration limits. They correspond conservatively to the dose limit of 10 mrem/year set for disposal in Switzerland. These concentration limits could change with the final choice of the repository site and also with design optimization at Oberbauenstock and are summarized in Table I.


Table I: Concentration limits (CL) of relevant radionuclides for a Project Gewaehr 1985 type B repository <NGB 85>

------------------------------------------------------------------

Nuclide HL/a HL/s CL/Ci.mE-3 CL/Bq.cmE-3 ------- ---------- -----------

C-14 5.7 10E+3 1.8 10E+11 3.7 10E+3 1.4 10E+8 Cl-36 3.0 10E+5 9.5 10E+12 4.4 10E-4 1.6 10E+1 Ni-59 7.6 10E+4 2.4 10E+12 2.1 10E+6 7.8 10E+10 Se-79 6.5 10E+4 2.0 10E+12 7.9 10E-2 2.9 10E+3 Zr-93 1.5 10E+6 4.7 10E+13 5.9 10E+6 2.2 10E+11 Tc-99 2.1 10E+5 6.7 10E+12 1.5 10E+1 5.5 10E+5 Pd-107 6.5 10E+6 2.0 10E+14 1.2 10E+1 1.6 10E+10 Sn-126 1.0 10E+5 3.1 10E+12 4.0 10E+0 1.5 10E+5 1-129 1.6 10E+7 4.9 10E+14 1.1 10E-4 4.1 10E+0 Cs-135 2.3 10E+6 7.2 10E+13 2.7 10E-1 1.0 10E+4 Ra-226 1.6 10E+3 5.1 10E+10 6.3 10E+2 2.3 10E+7 U-235 7.0 10E+8 2.2 10E+16 1.7 10E-4 6.3 10E+0 U-236 2.3 10E+7 7.4 10E+14 3.1 10E-1 1.1 10E+4 U-238 4.5 10E+9 1.4 10E+17 2.0 10E-3 7.4 10E+1 Np-237 2.1 10E+6 6.7 10E+13 1.4 10E-1 5.2 10E+3 Pu-239 2.4 10E+4 7.6 10E+11 4.6 10E+0 1.7 10E+5 Pu-240 6.5 10E+3 2.1 10E+11 1.1 10E+3 4.1 10E+7 Pu-241 1.4 10E+1 4.5 10E+08 2.3 10E+4 8.5 10E+8 Am-241 4.3 10E+2 1.4 10E+10 6.3 10E+2 2.3 10E+7

The concentration limits refer to the waste matrix.


3 SCALING FACTOR EVALUATION AND DISCUSSION

A scaling factor between two radionuclide specific activities SA can be defined for any phase in a decontamination process. In the case of correlation between an RR (safety relevant radionuclide) and a KN (key nuclide) the scaling factor SF for a given phase can be defined as follows:

[SA(RR)] [SF(RR/KN)]

[SA(KN)]

Although this definition is rather general, the scaling factor has been found useful for characterization of "machine-made" wastes. However, as far as we know, nothing has been done about the "hand-made" wastes. Hence, in this study, only operational wastes from PWRs and BWRs will be considered.

Since nuclides from roughly the whole isotopic chart are generated during reactor operation, the presence of RR and KN in the primary system or in the coolant is the result of the following processes:

1. -Generation of KN and RR as fission products, -diffusion through fuel pellets and claddings

(mainly through defects), -transfer from the coolant to a decontamination phase, -possible transfer from one decontamination phase to another.

2. -Generation of KN and RR by neutron activation, -in the cladding followed by transfer into the coolant

(corrosion), -in the coolant either with the coolant itself or with corrosion product traces and with nuclides that have already been transferred.

The activity in the coolant. is a function of the operation processes including:

-operation time and burnup -core temperature -fuel performance and quality -quality of cladding -cleanup processes

The activity from the coolant is collected phases (Figure I) as well as on parts classification of operational waste (BA summarized as follows:

-filters (BA-3, BA-4) -resins (BA-1) -dry active phases (BA-2)

into various decontamination of the system. The Nagra

Betriebsabfaelle) can be

-fuel element supports, probes and parts from BWR (BA-7, BA-8, BA-9)

-miscellaneous (BA-5, BA-6)

NAGRA NTB 86-28

Figure I: The decontamination processes for BWR and PWR primary systems.

[CF(1) 1

SIMPLIFIED SCHEMATIC OF PWR PRIMARY SYSTEM

Filter

[FF]

Overhead to Recycle or Discard as Liquid Waste

[DF(i,4)]

eves Evaporator

[DF(i,3)] Bottoms to Recycle or Discard as Solid Waste

SIMPLIFIED SCHEMATIC OF BWR PRIMARY SYSTEM

[DF(i,2)] [DF(i, 1)]

Condenser

'SJAE: steam jet air ejector AOG: au~mcnted off-gas system Resin to Solid Waste

[DF(i,4)] SJAE Exhaust toAOG'

Page 5


3.1 Calculation approach

Starting from the calculation of the radionuclide build-up rate in the coolant, the scaling factors are derived as follows. For a series of decontamination steps (Figure I), at the j-th step, from the phase j and through the phase j+1, one can define <LIE 83> a decontamination factor DF(i,j) for an isotope i. Introducing DF(i,j), the mass balances of RR and KN can be described, starting from the specific activity SA(i,c) of the isotope i in the coolant c, on the basis of an equation of build-up rate BR such as:

[SA(i,c»).[CF].[FF] 1 [BR(i,n)] --------------------.{ 1 - ---------}

~ [DF(i,j)] [DF(i,n)] 1<j<n-1

where the decontamination factor DF concerns the nuclide i at the jth decontamination step of the cleanup with a coolant flow rate CF and a flow fraction FF. Since nuclide transfers and decays can be quantified, we obtain the activity:

[BR(i,n») [A(i,n)] . { 1 - exp-[DC(i)].T }

[DC(i)]

where DC(i) is the decay constant of the isotope i and T the process run time. The scaling factor SF of the phase n can then be expressed with the scaling factor in the coolant and two corrective factors:

1 [SA(RR,c)] {1 - ----------}.TT[DF(KN,j)]

[DF(KN,n)] 1<j<n-1 [SF(RR/KN,n)]=---------- . -----------------------------

1 [SA(KN,c)] {1 - ----------}.II[DF(RR,j)]

[DF(RR,n)] l<j<n-l

[SF(c)] correction for difference in DF

[DC(KN)].{1-exp-[DC(RR)

[DC(RR)].{l-exp-[DC(KN)

correction for differen in DC

Since the RR and KN have long or relatively long half-lives, the correction factor for the DC differences is seldom significantly different from unity (for typical run time e.g. 300 d). However, if there is a difference in the DF values, then the question is how does SF vary from one decontamination step to another? Hence, some interim scaling factor modifiers SFM(n) were proposed (Table II) in order to correct the reactor coolant scaling factor SF(c) by a simple multiplication to obtain the scaling factor in the phase n:

[SF(n)] = [SF(c)] . [SFM(n)]


3.2 Experimental approach

Scaling factors can be determined by measuring SA(KN) and SA(RR) in various waste phases. As a result of our literature survey, we have summarized the scaling factors available for the operational wastes in Table III.

Table II : Proposed interim scaling factor modifiers SFM(n).

Vaste stream (n) SFM(n) for Cs-137

1-129

PWR wastes

Primary cleanup filter 1 Primary cleanup resin 1 CVCS evaporator bottoms 1

Dirty waste filter 1 Dirty waste evaporator bottoms 1

Steam generator blowdown filter 0.1 Steam generator blowdown resin 1 Steam generator condensate resin 1

Trash 1

BWR Vastes

Primary cleanup filter/resin Condensate cleanup resin

High purity filter High purity resin

Low purity filter Low purity resin

Chemical evaporator bottoms

Trash

1 10

1 1

1 1

10

1

Other nuclides

10 1 0.1

10 1

10 1 1

1

2 2

10 1

10 1

1

1

All proposed interim scaling factor modifiers SFM(n) are 1 for Co-60.

NAGRA NTB 86-28

Table III : Scaling or correlation factors

For all potential RR from <NGB 85) versus Co-60, Cs-137 and Ce-144. In Table III the references are simplified as follows:

Table III references Report references

<1> <LIE 83) <2> <STA 84)

(*) (3) <CLI 85> (average values) (*) <4> <VIL 81>

<5> <DAM 77> (*) <6> <MIL 84>

<7> <NTB 84> (*) <8> <BES 85>

<9> <CLI 83) <10> <MIL 85)

(*) Studies reporting their own experimental data

Page 8


Table IlIa : SF for PWR

Isotope Reactor Resins Dry active wastes coolant and trash

------ ------- ------ -----------------

vs. Co-60 ---

C-14 1 10E-1<3> 1 10E-1<3> 1 10E-1<3> 2 10E-4<4> 1 10E-4<4> 1 10E-3<4> 1 10E-4<7> 3 10E-3<5> 1 10E-2<10>

Ni-59 6 10E-4<4,8> 2 10E-2<3> 2 10E-2<3,9> 4 10E-3<7> 6 10E-4<4> 6 10E-4<4>

Pu-239 1 10E-4<10>

vs. Cs-137

C-14 4 10E-3<4> 4 10E-3<4> Cl-36 6 10E-6<7> 6 10E-6<7> Sr-90 6 10E-4<1> 3 10E-3<2> 3 10E-3<2>

1 10E-2<3> 3 10E-2<3> 3 10E-2<3> 5 10E-3<6,7,10> 8 10E-3<4> 5 10E-2<9>

5 10E-2<9> Tc-99 1 10E-7<1> 1 10E-4<2,6> 1 10E-4<2,6>

1 10E-4<2> 1 10E-4<3> 4 10E-5<4> 4 10E-5<4> 4 10E-5<4> 7 10E-4<7> 8 10E-4<10>

1-129 3 10E-7<1,7> 1 10E-6<2> 1 10E-6<2> 1 10E-6<2> 8 10E-6<3> 2 10E-3<3> 1 10E-4<4> 3 10E-5<5> 1 10E-4<4> 2 10E-4<10> 3 10E-7<7>

Cs-134 7 10E-1<1> 7 10E-1<1> 7 10E-1<1> 6 10E-1<7>

Cs-135 4 10E-5<4> 6 10E-6<3> 6 10E-6<3> 3 10E-6<7> 3 10E-6<7> 4 10E-5<4>

Pu-239 4 10E-4<10> Pu-241 5 10E-3<2> 5 10E-3<2> 5 10E-3<2>

3 10E-4<7> Cm-242 2 10E-4<2> 2 10E-4<2> 2 10E-4<2>

3 10E-6<7>

vs. Ce-144

Pu-238 1 10E-4<7> 1 10E-2<3> 1 10E-2<3> Pu-239 1 10E-5<7> 1 10E-2<3,9> 1 10E-2<3,9>

3 10E-3<10> Pu-241 2 10E-2<7> 4 10E-1<3> 4 10E-1<3> Am-241 1 10E-4<7> 7 10E-3<3> 7 10E-3<3> Cm-242 2 10E-4<7> 2 10E-3<3> 2 10E-3<3>


Table IIIb : SF for BYR

Isotope Reactor Resins Dry active waste coolant and trash

------ ------- ------ ----------------

vs. Co-60 ---

C-14 1 lOE-4<7> 3 lOE-2<3> 3 lOE-2<3> 6 lOE-4<lO> 4 lOE-3<4>

Ni-59 6 lOE-4<4> 3 lOE-3<3> 3 lOE-3<3> 4 lOE-3<7> 6 lOE-4<4> 3 lOE-4<9>

3 lOE-4<9> Pu-239 5 lOE-5<lO>

vs. Cs-137

C-14 6 10E-4<4> 6 10E-4<4> 6 10E-4<4> Cl-36 6 lOE-6<7> 6 10E-6<7> Sr-90 1 10E-2<1,10> 2 lOE-2<2> 2 lOE-2<2>

5 10E-3<7> 7 lOE-2<3> 1 10E-2<3> 5 lOE-2<7,9> 5 lOE-2<9>

Tc-99 1 lOE-4<2> 1 10E-4<2> 1 10E-4<2> 2 10E-3<4> 7 10E-5<3> 3 10E-4<3> 7 10E-4<7> 2 10E-3<4> 1 10E-4<6> 9 lOE-5<10> 1 10E-4<6>

1-129 5 10E-5<2> 5 10E-5<2> 5 10E-5<2> 4 10E-5<4> 7 10E-4<3> 5 10E-4<3> 3 10E-7<7> 4 10E-5<4> 1 lOE-4<4> 7 10E-5<10> 2 10E-4<9>

Cs-134 4 10E-l<1> 4 10E-1<1> 4 lOE-l<l> 6 lOE-l<7>

Cs-135 4 lOE-5<4> 4 10E-5<4> 4 10E-5<4> 3 10E-6<7> 3 lOE-6<7> 4 10E-5<4>

Pu-239 3 10E-4<10> Pu-241 2 10E-4<7> 8 10E-3<1> Cm-242 3 10E-6<7> 3 10E-4<1> 2 10E-4<1>

vs. Ce-144

Pu-238 1 10E-4<7> 3 10E-2<3> 3 10E-2<3> Pu-239 1 10E-5<7> 2 10E-2<3> 2 10E-2<3>

5 10E-3<10> Pu-241 2 10E-2<7> 7 10E-1<3> 7 10E-1<3> Am-241 1 lOE-4<7> 2 10E-2<3> 2 10E-2<3> Cm-242 2 10E-4<7> 4 10E-2<3> 4 10E-2<3>


4 DETERMINATION OF THE "KEY" NUCLIDE DETECTION LIMITS

As far as we know, the most widely used and practical way to determine the concentration of the "key"'radionuclides (KN) in the waste package is by means of high resolution gamma spectrometry.

The intrinsic detection limits listed in Table IV are reported from quoted references.

Table IV : Detection limits for the "key" nuclides

Nuclide Detection limit Observations Ref. ------- --------------- ------------

Co-60 5 10E-7 Ci.mE-3 H20/NaI det <SAS 83> 2 10E-3 Ci.mE-3 can/GeLi det <HOU 82>

Cs-137 1 10E-2 Ci.mE-3 resin <TOI 81> 2 10E-3 Ci.mE-3 cement <HOU 82>

bitumen <HOU 84>

Ce-144

Pu-239 1 g / drum GeLi gamma scan <HOU 82> 0.1 g /200L can GeLi gamma scan <HAR 86>

Since the effective detection limit is correlated to background and other specific activities present, it will be necessary to calculate its value for each canister.

The purpose of this section is to apply a gamma transport code to a reference measurement installation (spectrometer and package) to calculate the detection limit DL on the basis of its IUPAC definition:

count(DL/KN) = 3 . v-[BG] where BG is the background counted from a background flux F(BG). It corresponds to the number of counts under the peak of the KN in the gamma spectrum. Count(DL/KN) should be the smallest number of counts assigned to a peak.


For a given geometry and energy, the lowest detectable gamma flux F from the key nuclide is given for the efficiency EF and for an acquisition time T by:

3 . V [BG] F(DL/KN)

T • [EF]

Between this flux and the "key" nuclide specific activity in the waste package: SA(KN) , there is a proportional factor PF which is related to the detection geometry and the material of the package and there is also the gamma abundancy AB of the KN. Then, we have:

3 . V [ BG ] [ PF ] SA(DL/KN)=

T • [EF] [AB]

PF is the ratio of the specific activity of the "key" nuclide for a given gamma decay (1): SA(KN/1) to the flux through the energy window of the peak: F(KN/1) , i.e.:

SA(KN/1) [PF]

F(KN/1)

In order to calculate the specific activity SA(DL/KN) corresponding to the detection limit DL of the key nuclide KN in a given waste package assayed in a given detection geometry, one first has to compute the proportional factor PF and the background BG.

The geometry of the measurement system is cylindrical <BIR 77>. It can be seen from Figure II and from the mathematical formula presented that PF is dependent only on the detection geometry and the waste package materials.

In this study, the waste package is assumed to be homogeneous. The AF(j)/cmE-1 attenuation factors of the material through which the gammas pass are considered to be constant for each medium j.

NAGRA NTB 86-28

Figure II : Study of gamma penetration through the systems: waste package (active waste matrix, shielding, drum wall)

air - detector

Page 13

Note: The shielding can be represented by inactive matrix or drum inner shielding or both.

Counts [EF).T.[AB].SA(KN). integral ~m 00'n c

integral = J J J ~~~~=~~~~~~~=~T=T~~~~~~=~~=~~~=~~~·dX.d«.dt o 0 b

for a homogeneous waste package

J exp{- AF(i).[x-x(j)]}

counts = [EF].T.[AB]. SA(KN/z,r,e).-------~------------- dV 4·II·xE+2

V(waste) for a heterogeneous waste package.

Sou~ce (active waste matrix)

H

R3

I I I I " I I I 1/ .....

r I, . / I'--n~ ..... 1 I I I I I . I

Shielding 11.---- Drum

SA (r,B,zl

0.5 H

! _I _ : t~~-f-::~f

,/ ..,/

Rl

R2

Detector


As the proportional factor PF is energy-dependent, it will be calculated for the following energy groups (Code MELANIE <LAN 86»:

group (i)

11-n 10-n 9-n 8-n 7-n 6-n 5-n 4-n 3-n 2-n 1-n

E(i)-E(i+1)/MeV

0.05-0.01 0.10-0.05 0.20-0.10 0.30-0.20 0.40-0.30 0.60-0.40 0.80-0.60 1.00-0.80 1.33-1.00 1.66-1.33 2.00-1.66

where n is the flag of the most energetic gamma group.

The calculation starts from the specific activities in the waste matrix and leads to a histographic spectrum of the gamma flux through the detector surface. It considers the source parameters (i.e. the radius of the source package R1/cm and its composition N(i)/At.cmE-3), the shielding and canister parameters (i.e. the external radius of the shielding R2/cm and of the canister R3/cm and their respective compositions N(j)/At.cmE-3), the distance between the axes of the cylindrical canister and the detector R4/cm (at mid-height), the air composition N(h)/At.cmE-3, and the [SA). [AB)/gamma.Bq.cmE-3 listed for the various energy groups i.

4.1 Determination of the background BG

For a given KN gamma decay and its detection window EW located between E(i) and E(i+1), one can calculate BG taking the BG as the result of the degradation of the gammas(i) stronger than the gamma from the "key" nuclide. The natural background component can be added here (if needed).

NAGRA NTB 86-28

EW 1 1

1 1

1

1

1

1 1

J\---F-~' ,--- 1

1

I 1

-----------+---------+------ --+--------+--------+--- -------1 E E(i) E(i+1) Group 1 2 i

Page 15

F(E)

From the specific activities of the stronger gamma emitters than the KN, one can calculate (MELANIE) the flux of the gamma emitted in the group (i) where the detection window EW is located without considering the KN itself. This flux F(i) for position R4 of Figure II can be used to calculate BG corresponding to a selected window <DEG 84>:

[EW] . F(i) . [EF] . T [BG] ---------------------

[E(i) - E(i-1)]

Where EF is the counting efficiency for the group (i).

4.2 Consideration of the counting efficiency EF

The counting efficiency is a function of energy. To a first approximation and for E>O.l MeV, one can accept the following relationship:

log[EF] = log[EF/1MeV] - K.log[E/MeV]

In this treatment, we consider EF as an average efficiency for each group, e.g.: [EF/1MeV] = 0.01 and K = 1.0 . However, as far as EF is associated with the acquisition time, we can consider [EF].T as the product to be selected.

4.3 Determination of the proportional factor PF

For a given specific acitvity SA(KN/1) of the gamma energy line 1 of the key nuclide KN in the waste package, one can calculate (MELANIE) the flux F(l) (between E(l) and E(2» and F(2) in the next group (between E(2) and E(3».

NAGRA NTB 86-28

EW 1 1

1

gamma peak 1

1

1

background 1

~~~~-------------------------I ~ 1

-------------+-- , +---------------+------------w----I E E(l) E(m)E(f) E(2) E(3) Group 1 2 3

Page 16

F(E)

The flux F(KN/1) corresponding to the detected gamma peak is then given by:

or:

I[E(f) - E(2)] + 0.5 [EW] I F(KN/1) F(l) - F(2) . 1------------------------1

I [E(2) - E(3)] I

F(1).[E(2)-E(3)] - F(2).[E(m)-E(2)] F(KN/1) = ----------------------------------

[E(2) - E(3)]

The proportional factor PF is then obtained by dividing SA(KN/1) by F(KN/1) as defined above in section 4.

4.4 Calculation of the detection limit SA(DL/KN)

Substituting PF and BG in the SA(DL/KN) equation, results in:

9 . [EW] . F(i) SA(KN) . [E(2)-E(3)] SA(DL/KN) = • ---------------------------------------

[EF] . T . [E(i)-E(i+1)] [AB].{F(1).[E(2)-E(3)]-F(2)[E(m)-E(2)]}

stronger ga~ma factor KN alone factor

The detection limit SA(DL/KN) can be directly determined knowing EW, EF, T, E(i),E(i+1), using F(i) calculated for the stronger gamma emitters than line 1 of KN and for SA(KN) and AB, using F(l) and F(2) calculated for the KN alone and knowing E(l), E(m), E(2) and E(3). Caution: the flag energy group changes when passing from the stronger gamma factor to the KN alone factor.


5 RESULTS AND DISCUSSION

Since the number of waste packages to be considered is rather large, we have selected typical packages including various reactor wastes (BA types) with or without internal shielding and with cement, bitumen or plastic matrices (Figure III). All calculations were carried out using the following values for the various parameters (unless otherwise mentioned):

spectrum [EV] E(m) [AB] E(m) [AB] E(m) [AB] [EF] . T K

energy resolution 6 KeV

1,333 KeV 0.999

662 KeV 0.846

134 KeV 0.108

10E+2 1

= 1 KeV (average window, for Co-60 for Co-60 for Cs-137 for Cs-137 for Ce-144 for Ce-144 at 1,000 KeV

see <DEG 84»

Basic detection considered: through 1 cmE2. The detection is carried out at 5 cm from the canister. Vith a 10,000 second detection time T, the detector efficiency EF at 1 MeV is about 0.01.

The photon emission from electron and positron interactions in the waste package is considered to be negligible for E > 0.5 MeV. In addition, the natural background is considered and is expected to be smaller than that produced by gamma energy degradation.

Figure III: BA waste packages considered in this study

Type 1

Type 5 and 6

Type 1

for BA-KKBI-IB for BA-KKM-IA

for BA-KKG-2 for BA-KKL-1C

Type 5 Type 6

// //// // 0.

(

O<X

)(

~.

:% ~~~

"" "- _'\.

plasticised waste package cemented waste package

bituminized waste package cemented waste package

'<


5.1 DL calculation and correlation, example BA-KKB1-1B

The following example is treated in detail. It concerns the waste package BA-KKB1-1B as described in <NTB 84>. Emphasis will be placed on how to use the SA(DL/KN) definition.

5.1.1 System description

The geometry and composition parameters of the whole package and of the detection assembly are as follows:

system

waste matrix

shielding

canister

air

additional data

type of package

type of waste

type of reactor

radius, composition

R1/cm C IAt.cmE-3 H lAt. cmE-3

R2'lcm Fe/At.cmE-3

R2/cm Ca/At.cmE-3 Si/At.cmE-3 AI/At.cmE-3 Fe/At.cmE-3 H IAt.cmE-3 o I At. cmE-3

R3/cm Fel At. cmE-3

R4/cm o IAt.cmE-3

V(can)/L

type

1.870 10E+1 5.8 10E+22 5.8 10E+22

1.900 10E+1 7.6 10E+22

2.200 8.5 2.5 8.8 2.9 2.6 3.0

10E+1 10E+21 10E+21 10E+20 10E+20 10E+22 10E+22

2.210 10E+1 7.6 10E+22

2.71010E+1 3.0 10E+19

102

5 (Fig. III)

resin from primary coolant.

PWR

The waste package gamma activity is then inventoried gamma line per gamma line and then energy group per energy group on the basis of the specific activities given for the waste package <NTB-84>. This inventory includes the "key" nuclides and other significant gamma emitters as well as the radionuclides whose gamma energies are higher than the line 1 (1.33 MeV) of the KN Co-60.

NAGRA NTB 86-28

5.1.2 SA(DL/Co-60) , SA(DL/Cs-137) , and SA(DL/Ce-144) results and correlations

For this waste package (BA-KKB-1B) SA(Co-60) 5.22 10E+6 Bq.cmE-3. The KN detection limit is then:

SA(DL/Co-60) = 2.3 Bq.cmE-3 or: SA(DL/Co-60) = 6.3 10E-5 Ci.mE-3

The most useful correlations and hence the RR detection limits are:

Page 19

Nuclide SF(RR/KN) SA(DL/RR)/Ci.mE-3 CL/Ci.mE-3 Remarks

C-14 Ni-59

1 10E-l 2 10E-2

6 10E-6 1 10E-6

Now, with a SA(Cs-137) 6.81 10E+5 Bq.cmE-3,

SA(DL/Cs-137) = 1.1 10E+2 Bq.cmE-3 or: SA(DL/Cs-137) = 2.9 10E-3 Ci.mE-3

3.7 10E+3 2.1 10E+6

The most useful correlations and RR detection limits are:

Nuclide SF(RR/KN) SA(DL/RR)/Ci.mE-3 CL/Ci.mE-3

+ +

Remarks -----------------------------------------------------------------Cl-36 6 10E-6 2 10E-8 4.4 10E-4 + Tc-99 1 10E-4 3 10E-5 1.5 10E+l + 1-129 3 10E-5 9 10E-8 1.1 10E-4 + Cs-134 7 10E-l 2 10E-3 pot. Cs-135 6 10E-6 2 10E-8 2.7 10E-1 + Pu-241 5 10E-3 1 10E-5 2.3 10E+4 +

Finally, with SA(Ce-144) = 1.74 10E+3 Bq.cmE-3:

SA(DL/Ce-144) = 1.4 10E+3 Bq.cmE-3 or: SA(DL/Ce-144) = 3.8 10E-2 Ci.mE-3

and the most useful correlations and RR detection limits are:

Nuclide

Pu-238 Pu-239 Pu-241 Am-241

SF(RR/KN)

1 10E-2 1 10E-2 4 10E-l 7 10E-3

SA(DL/RR)/Ci.mE-3 CL/Ci.m-3 Remarks

4 10E-4 4 10E-4 2 10E-2 2 10E-4

+ 4.6 10E+O + 2.3 10E+4 + 6.2 10E+O +

KN

In all cases, the detection limits are much lower than the concentration limits CL of the safety-relevant radionuclides.

M


5.2 Spectrum calculation, example BA-KKM-1A

5.2.1 Activity, flux and counting spectrum propagation

This example for BA-KKM-1A waste deals with the spectrum propagation. On the basis of the specific activity inventory of the waste package, the specific activities are initially plotted as a spectrum of gamma energies (Fig. IV).

-

0 0 CD CD , , 0 0 u u

I -1 w -E

(.J ....... ,...., m « '7 - 2

-3

-4

-s

-

-

-

o

.."

.." ..-, :J

l.LJ

...:

...: I ..,

U

I

,.... ,..., ...:~ !:2", 'U

'" U

...:

.."

~ ...: :J ,..., l.LJ

I

'" U .." N -, .D If)

I I

0.5

...: .."

"7 ...: :J .." l.LJ "7 :J

W

...: ,..., -, '" u

E ...: ...: 0

,..., ~ ::: -;-

'" ,

I U :J C7'I

<{ l.LJ

E 0

, E C7'I 0 <{ ,

C7'I <{

I I I I I

to 1.5 2.0 E/MeV Figure IV : Specific activity spectrum of the waste package: BA-KKM-1A

Reference: <NTB 84>


The gamma flux per group can be calculated for a given distance from the canister (i.e. 5 cm) on the basis of the geometric and composition data from <NTB 84> listed as follow:

system radius, composition

waste matrix

shielding

R1/cm Ca/At.cmE-3 Si/At.cmE-3 Al/At.cmE-3 Fe/At.cmE-3 C/At.cmE-3 0/At.cmE-3 H/At.cmE-3

R2/cm Ca/At.cmE-3 Si/At.cmE-3 Al/At.cmE-3 Fe/At.cmE-3 01 At. cmE-3 H/At.cmE-3

canister R3/cm Fe/At.cmE-3

air R4/cm 0/At.cmE-3

additional data: V(can)/L

type of package: type

2.855 5.7 1.7 5.7 1.4 2.1 1.9 3.8

3.063 8.8 2.7 8.9 2.2 2.9 2.7

10E+1 10E+21 10E+21 10E+20 10E+20 10E+22 10E+22 10E+22

10E+1 10E+21 10E+21 10E+20 10E+20 10E+22 10E+22

3.075 10E+1 7.6 10E+22

3.575 10E+1 3.0 10E+19

217

6 (Fig. III)

type of waste: resin from primary coolant

type of reactor: BWR

By application of the code MELANIE, the flux spectrum is calculated and its histogram is given in Fig. V. Hence, for a given acquisition time (10E+4 s) and an efficiency given as log[EF] = -2 - log[E/MeV] , one can calculate the spectrum shown in Fig. VI. For the drawing of these last spectra the peaks are assumed to be triangular symmetric histograms with a 6 KeV basis.


Figure V : Flux spectrum of the BA-KKM-l waste package for a distance of 5 cm

4 -

<:) <:)

CD CD . . 0 0 u U

r-.. - M ~ 3 .

-.r", ~u -.r

M t --.r '"

. -.r U '" -.r ...- ~u V"I

t -.r V"I -.r ...-cu V"I r"I -:- V"I

I

U /"'J ...- :J ...- :J ...- . w . w

- '" :J .D U W

~ w (j)

N 2 lfl

I I l

1

. w E u --..,

~ -.r - ~ V"I

~

:> OJ ~

'" - ..... U :J -- W E

I ~ LL

...- E (J)

0 en C)

<1: ...-~

en - <1 0

!-

-1 -

-2 I I I I I I I I I

o O.S 1.0 1.5 2.0 E/MeV


Figure VI : Gamma spectrum produced by a commercial GeLi detector situated at a distance of 5 cm from a BA-KKM-lA waste package

7

6

,.--, U)

~

wI.. 0 -4J

C ::J 0 U

L...J

en ~3

2 • • • •

• • •

•

a a 0,5

• •

-I

11\

U

• .. • • •

•

1.0

C)

w

•

•

...,z l")

I 11\

U

E S?

I

0"1 ~

Conditions: Counts/l KeV, peak EW : 6 KeV, acquisition time: 1.0 10E+4 s,

~ LJ") .-

I

::l W

E a

I

0"1 ~

1.5

log[EF] = log{[EF(lMeV)]=O.Ol} - 1.log(E/MeV)

2,0

. : typical GeLi background / 1.0 10E+4 s from <RUE 86>

E/MeV


5.2.2 SA(DL/Co-60), SA(DL/Cs-137) and SA(DL/Ce-144) results and correlations

For the waste package BA-KKM-1A, we obtain:

SA(DL/Co-60) SA(DL/Co-60) SA(DL/Cs-137) SA(DL/Cs-137) SA(DL/Ce-144) SA(DL/Ce-144)

1.1 10E-1 Bq.cmE-3 or: 3.0 10E-6 Ci.mE-3 . 2.5 10E+0 Bq.cmE-3 or: 6.7 10E-5 Ci.mE-3 . 3.3 10E+1 Bq.cmE-3 or 8.9 10E-4 Ci.mE-3 .

For Ce-144, the detection limit of the specific activity is higher than its actual specific activity and, since Ce-144 is not detectable, the correlation is not calculable.

The most useful correlations with Co-60 and hence the RR detection limits are:

Nuclide SF(RR/KN) SA(DL/RR)/Ci.mE-3 CL/Ci.mE-3 Remarks -------------------------------------------------------------------C-14 3 10E-2 9 10E-8 3.7 10E+3 + Ni-59 3 10E-3 9 10E-9 2.1 10E+6 +

While for Cs-137 the results are:

Nuclide SF(RR/KN) SA(DL/RR)/Ci.mE-3 CL/Ci.mE-3 Remarks -------------------------------------------------------------------CI-36 6 10E-6 4 10E-10 4.4 10E-4 + Tc-99 2 10E-3 1 10E-7 1.5 10E+1 + 1-129 7 10E-4 5 10E-8 1.1 10E-4 + Cs-134 4 10E-1 3 10E-5 pot. KN Cs-135 4 10E-5 3 10E-9 2.7 10E-1 + Pu-241 8 10E-3 5 10E-5 2.3 10E+4 +


5.3 Effect of acquisition time on the SA(DL/KN), example BA-KKL-1C

This example evaluates the effect of the acquisition time T on the detection limit of the relevant radionuclides in order to determine the optimum counting time from an economic point of view. Here again the fluxes are calculated for a Scm distance from the canister BA-KKL-1C. The corresponding geometric and composition data are given in the following table.

system radius, composition -----------------------------------------------waste matrix R1/cm 2.800 10E+1

Ca/At.cmE-3 5.7 10E+21 Si/At.cmE-3 1.7 10E+21 Al/At.cmE-3 5.7 10E+20 Fe/At.cmE-3 1.4 10E+20 C/At.cmE-3 2.1 10E+22 O/At.cmE-3 1.9 10E+22 H/At.cmE-3 3.8 10E+22

shielding: R2/cm 2.800 10E+1

canister: R3/cm 2.815 10E+1 Fe/At.cmE-3 7.6 10E+22

air: R4/cm 3.315 10E+1 O/At.cmE-3 3.0 10E+19

additional data: V(can)/L 215

type of package: type 1 (Fig. III)

type of waste: resin

type of reactor: BWR

On the basis of the specific activities of the gamma emitters in the waste package, the fluxes per energy group for the whole source and for each of Co-60 alone (2 lines), Cs-137 alone and Ce-144 can be calculated as previously described.


5.3.1 SA(DL/Co-60) , SA(DL/Cs-137) , SA(DL/Ce-144) results and correlations

Calculations are carried out for a 10E+4 s acquisition time. Ve obtain:

KN

Co-60 Cs-137 Ce-144

SA(KN)I Bq.cmE-3

6.6 10E+3 1.1 10E+3 3.5 10E+0

F(i), F(l) and F(2)1 sE-1.cmE-2

3.3 10E3 2.5 10E4 5.2 10E3 1.3 10E4 3.2 10E3 1.3 10E3 1.8 10E4 7.7 10EO 1.6 10EO

SA(DL/KN)I Bq.cmE-3

6.1 10E-2 2.0 10E+O 2.2 10E+1

Here again Ce-144 is not detectable. From these results, the correlations can be applied as follows for deriving the RR detection limits:

RR SF SA(DL/RR)/Ci.m-3 CL/Ci.m-3 Remarks

vs Co-60 C-14 3 10E-2 5 10E-8 3.7 10E+3 + Ni-59 3 10E-3 5 10E-8 2.1 10E+6 +

vs Cs-137 Cl-36 6 10E-6 3 10E-10 4.4 10E-4 + I-129 7 10E-4 4 10E-8 1.1 10E-4 + Cs-135 4 10E-5 2 10E-9 2.7 10E-1 +

However on the basis of a 10E+4 s acquisition time, the correlations wi th Ce-144 are not possible.

5.3.2 Effect of the acquisition time on SA(DL/RR)

The acquisition time effect on the RR detection limits is presented in the next table. The SA are given in Ci.mE-3:

Tis SA(DL/C-14) SA(DL/Ni-59) SA(DL/Cl-36) SA(DL/I-129) SA(DL/Cs-135)

10E+5 10E+4 10E+3 10E+2 10E+1

2 10E-8 5 10E-8 2 10E-7 5 10E-7 2 10E-6

2 10E-9 5 10E-9 2 10E-8 5 10E-8 2 10E-7

1 10E-10 3 10E-10 1 10E-9 3 10E-9 1 10E-8

1 10E-8 4 10E-8 1 10E-7 4 10E-7 1 10E-6

5 10E-10 2 10E-9 5 10E-9 2 10E-8 5 10E-8

The acquisition time required to detect Ce-144 is T > 3.8 10E+5 s. For the other safety-relevant nuclides, an aquisition time of 10 to 100 s would suffice to check that their concentrations are below the concentration limits CL.


5.4 Effect of detection distance on SA(DL/KN), example BA-KKG-2

This example deals with the effect of the distance between the canister and the detector (R4-R3, see Fig. II) on the detection limit with a standard acquisition time of 10E+4 s. Here, the fluxes are calculated groupwise for the canister (BA-KKG-2) using the geometric and composition data reported in the following table and the average source compositions given in <NTB 84>.

system radius, composition -----------------------------------------------waste matrix R1/cm 2.800 10E+1

C/At.cmE-3 2.5 10E+22 O/At.cmE-3 1.1 10E+22 H/At.cmE-3 6.6 10E+22 Na/At.cmE-3 5.2 10E+21

shielding: R2/cm 2.800 10E+1

canister: R3/cm 2.815 10E+1 Fel At. cmE-3 7.6 10E+22

air: R4/cm 3.315 10E+1 O/At.cmE-3 3.0 10E+19

additional data: V(can)/L 213

type of package: type 1 (Fig. III)

type of waste: concentrate

type of reactor: PWR

The results for Ce-144, Cs-137, Co-60, C-14, Cl-36 and 1-129 are given in Fig. VII and compared with the concentration limits CL of Cl-36 and 1-129.

NAGRA NTB 86-28

Fig. VII: SA(DL) as a function of detection distance (R4-R3)

Acquisition time: 10E+4 s, K = 1, [EF(l MeV)] = 0.01

,.., M

I

l.J.J

E

-2

-3

-4

-5

t3 -6 -a:::: a:::: Lo

Z :::.:: -is -7

« V> w C'l .8

-8

-9

·10

-

...

-

-

-

-

-

..

o

-. Ce·144 •

-e Cs.137 .-

• Co·60 •

• C-14 •

• Cl·36 •

• 1-129 •

I

5 25

f-

Cl·36 )

1-129 () -

-

l-

-

I-

l-

-

R4/cm

-2

·3

·4

-5

-6 ,....., M

I

l.J.J E u --7 rr. a::::

-8

-9

.....J U w C'l .8

-10

Page 28


As can be seen from Fig. VII, there is only a small increase in the detection limit DL when increasing the distance between canister and detector. While these results are derived from a canister of infinite height and from a 2 11 steradian collimation, we conclude that experimental results would follow the same trend in the case of a real canister. Hence, the 5 cm standard distance between canister and detector is appropriate.

5.5 Most pessimistic case for the detection limits

The most pessimistic case corresponds to the waste with the highest gamma background under the KN peak. This situation is given for the package with the highest SA. Comparing the DL for the typical cases treated previously, one can see that the most critical case is the BA-KKB1-1B with:

SA(DL/Co-60) = 6 10E-5, SA(DL/Cs-137) = 3 10E-3, SA(DL/Ce-144) = 4 10E-2

All these specific activities are given in Ci.mE-3 (These values correspond to averaged activities as reported in <NTB 84».

If it is assumed that the specific activities of the source considered here correspond to the maximum values also stated in <NTB 84>, one can then calculate:

SA(DL/Co-60) = 9 10E-5, SA(DL/Cs-137) = 4 10E-3, SA(DL/Ce-144) = 5 10E-2.

If the effect of a scaling factor modifier (e.g. a factor of 10) needs to be considered, then, even in this case, determination by correlation is possible because SA(DL/RR) < CL(RR) (see section 5.1.2 for comparison).

Finally, the decay effect of the KN should be taken into consideration. The following correction can be made:

SA(T) = SA(O).exp{-DC(KN).T}

which would be relevant only for Ce-144 because of its short half-life.


6 CONCLUSIONS AND RECOMMENDATIONS

With an average production of 2000 reactor waste packages per year in Switzerland, a final check of the individual waste package activities could be scheduled on the basis of a one-hour measurement. Key nuclide activities, expected range of scaling factors and configuration of the waste packages would allow assessment of the concentrations of the safety-relevant radionuclides for the purpose of checking that they are below the concentration limit for disposal. This holds for type B repositories with similar concentration limits to those considered in this study and for the particular type B repository in "Project Gewaehr 1985".

In all the cases studied in this report, the correlation of the most important safety-relevant radionuclides, i.e. <ALD 85> along with the others <NGB 85)

allows detection CI-36 and I-129

The results of the study show that, in limits for disposal (for instance could be controlled as a consequence detection limits and concentration value of the published scaling factors

most cases, much lower concentration for a sub-surface type A repository) of the large differences between limits. Moreover, the conservative has been used.

This study was not devoted to the validation of the scaling factors for the different waste sorts and the different nuclear plants, but is in the nature of a preliminary study simply based on published data. The problem of using specific scaling factors (validation of the correlations and accuracy of the scaling factors for the different cases) has to be investigated further if the method is to be implemented.

The problems involved and the caution required in using correlations and scaling factors are best described in <LIE 83>: "In the course of this study an effort was made to unearth as much data as possible which might be useful in verifying the calculations reported here. The results were not totally satisfactory. The problem is several-fold. As mentioned before, existing data are seldom adequately identified as to exact source and they are seldom traceable back to the reactor coolant concentration which led to their formation. Some of the nuclides which Part 61 requires be identified exist only in very low concentrations to begin with, and are difficult to measure. Perhaps the most important part of the problem is the difficulty of obtaining a representative sample. This is particularly true at both ends of the waste spectrum. At the low end, dry active waste cannot be sampled in any practical way. At the other, high activity resin samples are very difficult to obtain, frequently result in excessive personnel exposure, and often have to be kept to the size of a few beads (out of several hundred cubic feet) in order that they may be analyzed. Solutions to the problem of representative sampling are not immediately obvious. In summary, the existing data are very limited in their usefulness. Verification of theoretical relationships such as those developed in this report is currently only marginally possible."

<eLI 85) also stresses that correlation factors have to be used with care: "A baseline for activity concentrations in solid waste and scaling factors in the radwaste streams should be established for each plant or station. A maintenance verification program should then be conducted to assure that the waste character has not undergone changes requIrIng the use of different scaling factors. Data should be acquired on most radwaste


streams. Substantive changes, requiring additional sampling, could occur in the base line scaling factor following large fuel failures or differences in equipment or operation of the radwaste system. Some changes in the scaling factors of dry-active-waste could also occur during extensive maintenance operations such as replacement of recirculation piping or steam generator."

The level of the difficulties mentioned in these two quotations depends strongly on the range and the accuracy of the radionuclide concentration determination required.

The recommendations which can be derived from this work are as follows:

- The assay of the safety-relevant radionuclides by way of correlation should be investigated further for quality control of radioactive waste packages. The outlook for such control obtained on the basis of this study is promising.

- The gamma counting should be carried out with - an average acquisition time between 0.5 and 3 h - an average distance between canister and detector of 5 to 10 cm

- In this paper we studied only high resolution gamma detection without scanning of the canister assumed to contain homogeneous waste. The effect of inhomogeneous waste and canister scanning should be investigated further.

- Comparison of effectively measured spectra for the waste sorts considered in this study with the theoretical spectra calculated here will narrow the basis for further investigation.

- As an additional potential key nuclide we suggest Cs-134 mainly for gamma-spectroscopic investigation of reprocessing waste which contains considerably more beta-emitters <BRO 84>.

Acknowledgements are due to M. Lanfranchi for running the code MELANIE for the typical cases treated.


7 LIST OF SYMBOLS AND PARAMETERS

A AB AF(i) BA BG BR BWR CF CL c DF

DC DL E E+n E(f) E(i) E(m) EF EY F FF FR H HL i IUPAC j K KKB KKG KKL KKM KN N(i)

PWR PF R RR SA

SF T

activity at time T / Bq or Ci gamma decay abundancy / dimensionless gamma attenuation factor in material i / cmE-l Betriebsabfaelle or reactor wastes background build-up rate / Ci.dayE-l or Bq.sE-l boiling water reactor cleanup or coolant flow rate / mE3.dayE-l or cmE3.sE-l concentration limit (SA limit) for disposal from <NGB 85> refers to the coolant phase decontamination factor between two phases / dimensionless decay constant / sE-I or dayE-l detection limit photon energy / KeV n-th power low energy limit for a photopeak I KeV group or window energy limit / KeV energy of a photopeak maximum detection efficiency energy window / KeV flux (gamma) / photon.cmE-2.sE-1 flow fraction / dimensionless flow rate / mE3.dayE-I or cmE3.sE-l canister height / cm half-life / s or a refers to specific isotope, group or element International Union of Pure and Applied Chemistry refers to specific process unit or element dimensionless constant in detection efficiency formula Kernkraftwerk Beznau Kernkraftwerk Goesgen Kernkraftwerk Leibstadt Kernkraftwerk Muehleberg "key" nuclide composition or concentration of i in a phase / At.cmE-3 pressurised water reactor proportional factor (SA. FE-I) / cmE-I phase radius from canister axis / cm safety-relevant radionuclide specific activity or concentration / Bq.cmE-3 or Ci.mE-3 scaling factor (SA ratio) / dimensionless time / s

NAGRA NTB 86-28

References

ALD 85

BES 85

BIR 77

BRO 84

CLI 83

CLI 85

CRO 80

DAM 77

DEG 84

Page 33

Alder, J.C.; Houriet, J.P. Concepts for experimental control of radioactive waste to be disposed of in Switzerland. in: Merz, E.; Odoj, R.; Warnecke, E. (eds)

Radioactive waste products - suitable for final disposal. Juel-Conf--54, June 1985, p. 54-62

Best, W.T.; Miller, A.D. Radionuclide correlations in low-level radwaste. EPRI-NP-4037 (June 1985)

Monitoring of plutonium contaminated solid waste streams. Chap. III: Passive gamma assay. Applicability and limitations. EUR 5637 (1977)

Brodzinski, R.L.; Perkins, R.W.; Rieck, H.G.; Wogman, N.A. Measurement of radionuclides in waste packages. US Pat.Appl. No 6-649-625 (12.9.1984)

Cline, J.E.; Coe, L.J. Long-lived radionuclides in low-level waste. SAI-83/107S (Aug. 1983)

Cline, J.E.; Noyce, J.R.; Coe, L.J.; Wright, K.W. Assay of long-lived radionuclides in low-level wastes from power reactors. NUREG/CR-4101, April 1985

Croff, A.G. ORIGEN2 - A revised and updated version of the Oak Ridge Isotope Generation and Depletion Code. ORNL-5621 (1980)

Characterization of selected low-level radioactive waste generated by four commercial light-water reactors. ORP/TAD-77-3 (Dec. 1977)

Degueldre, C.A. Neptunium neutron activation analysis. EIR internal report TM-42-84-24 (30.9.84)


HAR 86 Hartmann Private communication to C. Degueldre during his visit to the IAEA-Laboratory in Seibersdorf, April 1986

HOU 82 Houriet, J.-Ph. Monitoring des dechets radioactifs. Nagra NTB 82-02 (Aug. 1982)

HOU 84/1 Houriet, J.-Ph. Monitoring des dechets radioactifs. Mise a jour 1984. Nagra NTB 84-10 (Mai 1984)

HOU 84/2 Houriet, J.-Ph. Controle de la teneur en radionucleides des dechets radioactifs. Nagra NTB 84-39 (Nov. 1984)

LAN 86 Lanfranchi, M. Code MELANIE Swiss Federal Inst. for Reactor Research, Yuerenlingen, 1986

LIE 83 Lieberman, J.A.; McIlvaine, J.B.; Miller, A.D.; Rodger, Y.A. Methodology for the classification of low-level radioactive wastes from nuclear power plants. AIF/NESP-027, Dec. 1983

MIL 83 Miller, A.D. Classification of low-level waste under 10CFR61. Trans.Am.Nucl.Soc. 45(1983)

MIL 84 Miller, A.D.; Leventhal, L. Data base for 10CFR61. Trans.Am.Nucl.Soc. 46(1984)

MIL 85 Miller, A. D.; Shaw, R. A. Low-level waste: processes and technical correlations. Trans. Am. Nucl. Soc. 50 (1985) 89.

NGB 85 Projekt Gewaehr 1985 Radioaktive Abfaelle: Eigenschaften und Zuteilung auf die Endlager. Nagra NGB 85-02 (Jan. 1985)

NTB 84 Inventar und Charakterisierung der radioaktiven Abfaelle in der Schweiz. Nagra NTB 84-47 (Dez. 1984)


RUE 86 Ruethi, M. Private communication about EIR-SU's GeLi detector background to C. Degueldre (May 1986)

SAS 83 Sasaki, T.; Ishizuka, A.; et ale

STA 84

TOI 81

WIL 81

High sensitivity on-line monitor for radioactive effluent. J. of Nucl. Sci. and Techn. 20(4), p. 317-321 (1983)

Stanford, R.E.L. Classification of low-level radioactive wastes from nuclear power plants. in: Proceedings of the sixth Annual Participants' Information Meeting DOE Low-level Waste Management Program, Denver, Colorado, Sept. 11-13, 1984 CONF-8409115, p. 389-398

Toivola, A.; Treatment, Monitoring and on Site Storage of Solid Waste at the Olkiluoto BVR Nuclear Power Plant. Proceedings of the Seminar on the Management of Radioactive Waste from Nuclear Power Plants, Karlsruhe, October 5-9, 1981, IAEA-SR-57/47.

Wild, R.E.; Oztunali, 0.1.; et ale Data base for radioactive waste management. Waste source options report. NUREG/CR-1759 Vol. 2 (Nov. 1981)

technical report - nagra · 2018. 9. 25. · for controlling important radionuclide concentrations...

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