radiation protection calculations for an encapsulation plant · radiation protection calculations...

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POSIVA OY

Working Report 98-81

Radiation protection calculations for an encapsulation plant

Markku Anttila

December 1998

Mikonkatu 15 A, FIN-001 00 HELSINKI. FINLAND

Tel. +358-9-2280 30

Fax +358-9-2280 3719



Markku Anttila

December 1998

~-!



Markku Anttila

VTT Energy

December 1998

Working Reports contain information on work in progress

or pending completion.

The conclusions and viewpoints presented in the report

are those of author(s) and do not necessarily

coincide with those of Posiva.

""t1T ENERGY

Research organisation and address Customer

VTT Energy, Nuclear energy Posiva Oy P.O. Box 1604 Mikonkatu 15 A FIN-02044 VTT, FINLAND FIN-00 100 Helsinki

Project manager Contact person

Markku Anttila Jukka-Pekka Salo \2!::l Diary code Order reference

ENE4-30T -97 .. 9759/97 /JPS (9696/97 /JPS)

Project title and reference code Report identification & Pages Date

Radiation protection calculations for an ENE4/40/97 19.12.1997 encapsulation plant (N7SU00332; {41NEWCAN-RAD})

Report title and author(s)

RADIATION PROTECTION CALCULATIONS FOR AN ENCAPSULATION PLANT

Markku Anttila

Summary The minimum thicknesses of the concrete walls of an encapsulation plant now in its design phase in Finland have been calculated with the MCNP4B code based on the Monte Carlo technique and with the MARMER code based on the point-kernel method.

The calculations have been carried out in three phases. First, photon and neutron source strengths and spectra have been estimated as a function of discharge burnup and cooling time of the spent fuel. Then. gamma and neutron dose rates outside the final disposal canisters or their radiation shields have been calculated. Finally, the minimum wall thicknesses of the most important rooms of the encapsulation plant have been determined. The main design condition has been that behind a wall the total dose rate must be (clearly) under 2.5 J.LSv/h, i.e. the rooms behind the walls should be uncontrolled area from the point of radiation safety.

According to the present plans, both square BWR fuel bundles coming from the Olkiluoto nuclear power plant and hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant will be handled at the Finnish encapsulation plant. Therefore, there will be two different disposal canisters (in the following 'BWR canister' and 'VVER canister', respectively). In both types of canisters 12 spent fuel bundles can be placed. The geometry of the BWR and VVER canisters are rather similar. The average radioactive properties of the BWR and VVER-440

spent fuel differ from each other less than the uncertainties of the calculation methods. Therefore, all dimensioning calculations have been carried out assuming the BWR spent fuel having a burnup of 60 MWdlk.gU. Before the encapsulation, the fuel have been assumed to be stored for 20 years.

The following cases have been studied: - A fuel bundle with its channel box in the hot cell - A (fully filled) disposal canister in a room - A disposal canister in a cylindrical, 10 cm thick radiation shield made of steel - A disposal canister in a buffer storage - A disposal canister in a cylindrical radiation shield made of a 15 cm thick steel cylinder and a 5 cm thick polyethylene cylinder

It has been assumed that the bundle or the canister with or without a shield is at a distance of 60 cm from the wall (in the buffer storage the distance was 120 cm). According to the calculations the minimum wall thicknesses are from 60 to 120 cm.

Keywords: Encapsulation plant, spent fuel disposal canister, gamma and neutron dose rates, radiation protection calculations

Principal author or Project manager

~r4L ~/I( Senior research scientist Markku Anttila

~~~y ~~ <:::~-~·--- ,/ ~~

Lasse Mattila Research Manager, Nuclear Energy

Re~fJ~-- ryr~ec Riitta Kyrki-Rajamald Group Manager, Nuclear Analysis Availability statement

To be published in Work Reports -series of Posiva Oy

3

ABSTRACT

The minimum thicknesses of the concrete walls of an encapsulation plant now in its design phase in Finland have been calculated with the MCNP4B code based on the Monte Carlo technique and with the MARMER code based on the point-kernel method.

The calculations have been carried out in three phases. First, photon and neutron source strengths and spectr'a have been estimated as a function of discharge burnup and cooling time of the spent fuel. Then, gamma and neutron dose rates outside the final disposal canisters or their radiation shields have been calculated. Finally, the minimum wall thicknesses of the most important rooms of the encapsulation plant have been determined. The main design condition has been that behind a wall the total dose rate must be (clearly) under 2.5 J.LSvlh, i.e. the rooms behind the walls should be classified as uncontrolled area from the point of radiation safety.

According to the present plans, both square BWR fuel bundles coming from the Olkiluoto nuclear power plant and hexagonal VVER -440 fuel assemblies used at the Loviisa nuclear power plant will be handled at the Finnish encapsulation plant. Therefore, there will be two different disposal canisters (in the following 'BWR canister' and VVER canister', respectively). In both types of canisters 12 spent fuel bundles can be placed.

The geometry of the BWR and VVER canisters are rather similar. The average radioactive properties of the BWR and VVER-440 spent fuel differ from each other less than the uncertainties of the calculation methods. Therefore, all dimensioning calculations have been carried out assuming the BWR spent fuel having a burnup of 60 MWdlkgU. Before the encapsulation, the fuel have been assumed to be stored for 20 years.

The following cases have been studied: - A fuel bundle with its channel box in the hot cell -A (fully filled) disposal canister in a room - A disposal canister in a cylindrical, 10 cm thick radiation shield made of steel - A disposal canister in a buffer storage - A disposal canister in a cylindrical radiation shield made of a 15 cm thick steel cylinder and a 5 cm thick polyethylene cylinder

It has been assumed that the bundle or the canister with or without a shield is at a distance of 60 cm from the wall (in the buffer storage the distance was 120 cm). According to the calculations the minimum wall thicknesses are from 60 to 130 cm

In the next phase of the study a few calculations will be carried out in order to check and supplement the present data base. No significant changes of the present conclusions are probable.

Keywords: Encapsulation plant, spent fuel disposal canister, gamma and neutron dose rates, radiation protection calculations

4

KAYTETYN YDINPOLTTOAINEEN KAPSELOINTILAITOKSEN SATEILYSUOJELULLINEN MITOITUS

TIIVISTELMA

Kaytetyn ydinpolttoaineen kapselointilaitoksen betoniseinien sateilysuojelullinen mitoitus on tehty Monte Carlo -tekniikkaan perustavan MCNP4B- ja ns. pisteydinmenetelmaa soveltavan MARMER-tietokoneohjelman avulla.

Mitoituslaskut jakautuivat kolmeen vaiheeseen. Ensiksi laskettiin kaytetyn polttoaineen fotoni- ja neutronituotto poistopalaman ja jaahtymisajan funktiona ORIGEN2.1-ohjelmaa kayttaen. Sitten arvioitiin loppusijoituskapselien ja niiden sateilysuojien ulkopinnalla vallitsevat gamma- ja neutroniannosnopeudet. Lopuksi maariteltiin betoniseinan minimipaksuudet. Sateilysuojelumitoituksen lahtokohtana oli vaatimus, etta seinan tuli vaimentaa kokonaissateilytaso (selvasti) pienemmaksi kuin 2,5 !J.Sv/h.

Nykyisten suunnnitelmien mukaan Suomen ydinvoimalaitosten kaytetty polttoaine sijoitetaan loppusijoituskapseleihin yhdessa laitoksessa. Samoissa tiloissa kasitellaan seka Olkiluodon reaktoreiden geometrialtaan neliolliset BWR-niput etta Loviisan yksikoiden kuusikulmaiset VVER-440-niput. Siksi tarvitaan kaksi erilaista loppusijoituskanisteria ('BWR- ja VVER-kanisteri'). Kanisteriratkaisut ovat kuitenkin varsin samanlaisia pituutta lukuun ottamatta. Lisaksi kaytetyn BWR- ja VVER-440-polttoaineen radioaktiiviset ominaisuudet poikkeavat toisistaan suhteellisen vahan. Siksi seinien mitoituslaskut tehtiin vain BWR-kanisterille. Kaytetyn polttoaineen poistopalamaksi oletettiin 60 MWdlkgU ja sen varastointiajaksi 20 vuotta.

Seuraavat viisi tapausta tutkittiin: - Yksittainen polttoainenippu kuumakammiossa (jossa ntppu siirretaan kuljetusastiasta loppusijoituskanisteriin) - (Taytetty) loppusijoituskanisteri tyotilassa - Kanisteri 10 senttimetria paksun teraksisen sateilysuojan sisalla - Kanisteri puskurivarastossa saannollisessa hilassa - Kanisteri teraksesta ja polyetyleenista valmistetussa sateilysuojassa

Nipun tai kanisterin etaisyydeksi seinasta oletettiin 60 cm, paitsi puskurivarastossa, jossa etaisyys on kaksinkertainen. Laskujen mukaan tarvittavat seinanpaksuudet vaihtelevat valilla 60-130 cm.

Tyon seuraavassa vaiheessa tehdaan tarkistus- ja taydennyslaskuja, jotka mita todennakoisimmin eivat muuta sateilysuojelullista perusmitoitusta.

Avainsanat: Kapselointilaitos, loppusijoituskapseli, gamma- ja neutroniannosnopeudet, betoniseinien sateilysuojelumitoitus

5

TABLE OF CONTENTS

Abstract

Tiivistelma

1 INTRODUCTION

2 COMPUTER CODES AND THEIR DATA LIBRARIES

2.1 2.3 2.3 2.4

ORIGEN2.1 MCNP4B MAR MER Flux-to-dose conversion factors

3 CALCULATION OF PHOTON AND NEUTRON SOURCES

3.1 3.2

Input specifications Photon and neutron sources

4 CALCULATION OF SURFACE DOSE RATES

4.1

4.2 4.3

Geometry and material composition of the final disposal canisters Basic results Impact of backscattering on surface dose rates

5 CALCULATION OF MINIMUM WALL THICKNESSES

5.1 5.2

Introduction Results

6 CONCLUSIONS

REFERENCES

Page

3

4

6

6

6 7 7 7

9

9 9

18

18 21 27

28

28 29

34

35

6

1 INTRODUCTION

The minimum thicknesses of the concrete walls of an encapsulation plant now in its design phase in Finland have been calculated with the MCNP4B code based on the Monte Carlo technique and with the MARMER code based on the point-kernel method.

The calculations have been carried out in three phases. First, photon and neutron source strengths and spectra have been estimated as a function of discharge burnup and cooling time of the spent fuel with the ORIGEN2.1 code. Then, gamma and neutron dose rates outside the final disposal canisters or their radiation shields have been calculated. Finally, the minimum wall thicknesses of the most important rooms of the encapsulation plant have been determined. The main design condition has been that behind a wall the total dose rate must be (clearly) under 2.5 J..LSv/h, i.e. the rooms behind the walls should be uncontrolled area from the point of radiation protection.

2 COMPUTER CODES AND THEIR DATA LIBRARIES

2.1 ORIGEN2.1

ORIGEN is a widely used computer code for calculations of buildup and decay of radioactive nuclides during and after periods of neutron irradiation. The first versions of the programs were written in the late 1960s and early 1970s. The version used in this study, ORIGEN2.1, was released in 1991 (ORIGEN 2.1 1991) with the updated cross section libraries (Ludvig & Renier 1989).

ORIGEN2.1 is a point-depletion code, i.e. no spatial variation during the irradiation periods can be taken into account. The program uses one-group cross sections, which correspond to some specific reactor conditions. In principle, ORIGEN2.1 produces accurate results only, when the case studied is similar to the conditions, on which the cross section library has been generated. To improve the situation to some extent, several cross section sets have been processed.

ORIGEN2.1 solves the system of nonhomogeneous, first-order ordinary differential equations governing changes of material compositions by the so-called matrix exponential method. For short-lived nuclides some appropriate approximations are applied.

The cross section and decay data libraries contain the necessary constants for about 1 300 separate nuclides (of which about 300 are stable) divided into three categories: activations products (720 nuclides), actinides and their daughters (130) and fission products (850). A nuclide can belong to more than one category.

For the ORIGEN2.1 program some new one-group cross section libraries were generated for analyses of both PWR and BWR fuel bundles with the ORIGEN 2.1 program (Ludwig & Renier 1989). In this study, the data sets corresponding to extended irradiation cycles ("PWR-UE" and "BWR-UE") were used.

7

2.2 MCNP48

MCNP4B is the newest version of the Monte Carlo code developed at the Los Alamos National Laboratory. The code was released in Spring 1997.

MCNP4B is according to its User's Manual "a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system" (RSIC CCC-200; Briesmeister 1997). A user can apply the code to quite complicated problems almost without any approximations and get accurate results in a reasonable time when having modem workstations or PCs.

The cross section sets of the standard MCNP4B data library based on the END FIB-VI evaluated data library were used in these calculations, However, for some elements older data sets based on END FIB-V were chosen, because there were no corresponding END FIB-VI cross sections.

2.3 MARMER

MARMER is a point-kernel shielding code having many options not normally available in computer codes of this kind (Kloosterman). With the MARS geometry package one can define quite complicated source and shield structures. A flux integration procedure based on the Monte Carlo technique is used to perform integration over energy range and source volume.

The data libraries of MARMER are based on the JEF 1.1 evaluated data library. The buildup factors are, however, based on the library compiled by the ANSI standard 6.4.3 committee (Kloosterman. p. 6; Trubey&Harima 1987).

2.4 Flux-to-dose conversion factors

The radiation transport and shielding codes calculate first the gamma and/or neutron flux distribution in and around a given source volume. Then the dose rate distributions are generated by multiplying each flux value with a so-called flux-to-dose conversion factor. The conversion factors depend on the type and energy of the radiation. There are several sets of conversion factors, but the set most commonly applied is the American National Standard 6.1.1, which has been updated time to time (ANSI/ANS-6.1.1-1977 ja AN SI/ ANS-6.1.1-1991 ). The dose (rate) calculated according to this standard is the (biological) effective dose equivalent, which is the absorbed dose multiplied by the socalled quality factor.

The flux -to-dose conversion factors can not be very accurate due to difficulties encountered when assessing biological effects of radiation. Regarding gamma radiation, two above mentioned AN SI/ ANS standards differ by about 20% in the energy range from one half to one Me V, which was of the most interest for this study, the updated 1991 version predicting lower dose rates. The gamma dose rates based on the ANSI/ANS-6.1.1-1991 conversion factors are given in this report, if not otherwise specified.

8

The International Commission on Radiological Protection (ICRP) took already in 1985 the position that the so-called quality factors for neutron radiation should be increased by a factor of about two. This recommendation was still under review, when the ANSI/ANS-6.1.1-1991 conversion factors for neutron radiation were approved, and were not taken into account in the standard. Because the recommendation has now been accepted (STUK Guide ST 1.2), the AN SI/ ANS-6.1.1-1991 conversion factors for neutrons were multiplied by a factor of two for this study.

The final modification of the conversion factors for neutron radiation was more complicated than a simple multiplication. Therefore, a third set was defined for MCNP4B calculations taken from the most recent ICRP data compilation (ICRP-74). The neutron dose rates based on these data are given in this report, if not otherwise specified.

I ! 9

3 CALCULATION OF PHOTON AND NEUTRON SOURCES

3.1 Input specifications

The photon and neutron sources and their spectra were calculated with the ORIGEN2.1 program using the PWR-UE cross section library for the VVER-440 fuel and the BWRUE library for the BWR fuel.

Input data were specified mainly as in earlier ORIGEN2.1 calculations (Anttila 1992 and Anttila 1995). The enrichment of the BWR fuel was set to be 3.8% and that of the VVER-440 fuel 4.2%. In the latter case, a bundle with Zirconium spacers (Russian VVER-440 bundle of new type) was studied. In both cases, the photon and neutron sources were calculated for three discharge burnups (40, 50 and 60 MWdlkgU), which were achieved by irradiating the fuel at the constant power density. It was assumed that the Olkiluoto and Loviisa units would operate at the increased power, 2500 and 1500 MWth, respectively.

3.2 Photon and neutron sources

The photon production rates and their spectra of the VVER-440 and BWR spent fuel for three discharge bumup and four cooling times are given in Table 1. The corresponding neutron sources are shown in Table 2.

The ORIGEN2.1 program calculates the total and nuclide-wise photon source strengths in eighteen energy groups. The group-wise results are normalized to the average energy of each group in such a way that the total gamma energy production in each group is preserved. The procedure may not be the best possible for shielding calculations. Therefore, the main source terms ofBa-137m and Eu-154 were renormalised to the correct energies in MCNP4B calculations. In MARMER input files the correction was made only for Ba-137m.

The spontaneous fission of Cm-244 is the most important source of neutrons in the spent nuclear fuel at the cooling times, which are of interest for these calculations. Therefore, the energy spectrum of source neutrons was assumed to correspond to the spectrum of that reaction.

The ORIGEN2.1 concentrations of Cs-137 (the mother nuclide of Ba-137m), Eu-154 and Cm-244 are compared with those calculated with the two-dimensional CASM0-4 fuel assembly bumup code (Edenius, et al. 1995 & Knott, et al. 1995) in Table 3. The spatial details of the fuel bundles (the ATRIUM 1 Ox 1 0-9Q and VVER -440 bundles in this case) can be described in CASM0-4 calculations. Furthermore, the neutron transport equation is solved using many energy groups. Therefore, the effects of irradiation on the fuel composition can be calculated with the code more accurately than with ORIGEN2.1.

Regarding Cs-137 concentrations the CASM0-4 and ORIGEN2.1 results are in good agreement, but in the cases of Eu-154 and Cm-244 the differences are rather large, the ORIGEN2.1 values being larger than the CASM0-4 concentrations. The ORIGEN2.1 source terms were used in all dose rate calculations.

10

Table 1. Photon source strength (lis/tU) and spectrum of the spent WER and BWR fuel at four cooling times according to ORIGEN2.1 calculations.

I Spent VVER fuel

a) Dis~harge bumup 40 MWdlkgU (initial uranium enrichment of 4.2 wt%)

Mean energy of Cooling time (years) the gamma group

(Me V) 10. 20. 30. 40.

0.125 2.581·1014 1.649·1014 1.158·1014 8.537·1013

0.225 2.317·1014 1.707·1014 1.317·1014 1.027·1014

0.375 1.112·1014 7.185·1013 5.512·1013 4.317·1013

0.575 4.063·1015 2.918·1015 2.302·1015 1.825·1015

0.662 (Ba-137m) 3.602·1015 2.859·1015 2.269·1015 1.801·1015

0.850 3.628·1014 6.301·1013 2.884·1013 1.625·1013

1.250 1.284·1014 5.147·1013 2.391·1013 1.163·1013

1.750 3.620·10 12 1.685·1012 8.494·10 11 4.557·1011

2.250 9.260·1010 1.173·108 5.233·107 3.842·107

2.750 5.777·109 5.022·108 4.781·108 4.314·108

3.500 7.130·108 2.098·107 1.404·107 9.780·106

b) Discharge bum up 50 MW d/kgU (initial uranium enrichment of 4.2 wt%)


(Me V) 10. 20. 30. 40.

0.125 3.377·1014 2.086·1014 1.429·1014 1.036·1014

0.225 2.797·1014 2.040·1014 1.566·1014 1.216·1014

0.375 1.333·1014 8.499·1013 6.499·1013 5.084·1013

0.575 5.184·1015 3.627·1015 2.858·1015 2.265·1015

0.662 (Ba-137m) 4.473·1015 3.550·1015 2.818·1015 2.236·10 15

0.850 5.454·1014 9.187·1013 4.068·1013 2.216·1013

1.250 1.934·1014 7.703·1013 3.540·1013 1.693·1013

1.750 5.361·1012 2.469·1012 1.216·1012 6.331·1011

2.250 1.043·1011 2.029·108 1.044·108 7.480·107

2.750 7.620·109 9.350·108 8.435·108 7.576·108

3.500 9.543; 108 5.724·107 3.888·107 2.697·107

11

Table 1 (cont.)

c) Discharge bum up 60 MW dlkgU (initial uranium enrichment of 4.2 wt%)


(Me V) 10. 20. 30. 40.

0.125 4.134·1014 2.493·1014 1.677·10

14 1.202·1014

0.225 3.226·1014 2.333·1014 1.783·10

14 1.382·1014

0.375 1.528·1014 9.638·1013 7.350·1013 5.744·1013

0.575 6.228·1015 4.321·10 15 3.401·1015 2.696·1015

0.662 (Ba-137m) 5.325·1015 4.227·10 15 3.355·1015 2.663·1015

0.850 7.385·1014 1.207·1014 5.229·1013 2.789·1013

1.250 2.585·1014 1.023·1014 4.675·1013 2.213·1013

1.750 7.077·1012 3.242·1012 1.576·1012 8.050·10" 2.250 1.117·10" 3.432·108 1.957·108 1.386·108

2.750 9.358·109 1.448·109 1.360·109 1.161·109

3.500 1.210·109 1.234·108 8.448·107 5.871·107

11 Spent BWR fuel

a) Discharge burnup 40 MWdlkgU (Initial uranium enrichment of 3.8 wt%)


(Me V) 10. 20. 30. 40.

0.125 2.590·1014 1.623·1014 1.122·1014 8.192·1013

0.225 2.206·1014 1.620·1014 1.245·1014 9.690·1013

0.375 1.063·1014 6.777·1013 5.183·1013 4.055·1013

0.575 4.004·1015 2.868·1015 2.262·1015 1.793·1015

0.662 (Ba-137m) 3.540·1015 2.810·1015 2.230·1015 1.770·1015

0.850 3.739·1014 6.604·1013 2.941·1013 1.582·1013

1.250 1.682·1014 6.398·1013 2.805·1013 1.306·1013

1.750 3.912·1012 1.817·1012 9.024·1011 4.746·1011

2.250 6.766·1010 1.784·108 8.293·107 5.475·107

2.750 5.065·109 6.832·108 6.169·108 5.547·108

3.500 6.240·108 3.698·107 2.510·10

7 1.740·107

12

Table 1 (cont.)

b) Discharge bum up 50 MW dlkgU (Initial uranium enrichment of 3.8 %)


(Me V) 10. 20. 30. 40.

0.125 3.347·1014 2.033·1014 1.373·1014 9.878·1013

0.225 2.661·10 14 1.921·1014 1.470·1014 1.140·1014

0.375 1.256·1014 7.949·1013 6.063·1013 4.739·10 13

0.575 5.027·1015 3.545·1015 2.793·1015 2.214·1015

0.662 (Ba-137m) 4.374·10 15 3.471·1015 2.755·1015 2.187·1015

0.850 5.400·1014 9.338·1013 4.056·1013 1.121·1013

1.250 2.381·1014 9.084·1013 3.975·1013 1.833·1013

1.750 5.594·1012 2.579·1012 1.258·1012 6.451·10 11

2.250 7.267·1010 3.079·108 1.641·108. 1.101·108

2.750 6.399·109 1.170·109 1.050·109 9.389·108

3.500 8.160·108 9.337·107 6.391·107 4.429·107

c) Discharge bumup 60 MWd/kgU (Initial uranium enrichment of 3.8 %)


(Me V) 10. 20. 30. 40.

0.125 4.034·10 14 2.402·10 14 1.598·1014 1.138·1014

0.225 3.024·1014 2.185·1014 1.666·1014 1.290·1014

0.375 1.429·1014 8.966·1013 6.823·1013 5.329·1013

0.575 6.042·1015 4.207·1015 3.312·1015 2.625·1015

0.662 (Ba-137m) 5.189·10 15 4.118·10 15 3.269·1015 2.594·10 15

0.850 7.105·1014 1.189·1013 5.074·1013 2.609·1013

1.250 3.044·1014 1.158·1014 5.055·1013 2.317·1013

1.750 7.135·1012 3.274·1012 1.581·1012 7.999·1011

2.250 7.810·1010 5.053·108 2.924·108 1.988·108

2.750 7.805·109 1.742·109 1.554·109 1.382·109

3.500 1.055·109 1.852·108 1.275·108 8.874·107

Table 2.

I

a)

13

Neutron source strength (nls/tU) of the WER and BWR spent fuel at four cooling times according to the ORIGEN2.1 calculations.

VVER spent fuel

The VVER spent fuel with the discharge bum up of 40 MW dlkgU (The uranium enrichment of 4.2 wt%)

10. Cooling time (a) 20. 30. 40.

( a,n)-reaction 8.469·106 8.522·106 8.398·106 8.185·106

Spontaneous fission 2.855·108 1.964·108 1.355·108 9.401·107

Total 2.940·108 2.049·108 1.439·108 1.022·108

Contribution of Cm-244 2.826·108 1.928·108 1.314·108 8.965·107

b) The VVER spent fuel with the discharge bumup of 50 MWd/kgU (The uranium enrichment of 4.2 wt%)

Cooling time (years) 10. 20. 30. 40.

(a,n)-reaction 1.578·107 1.444·107 1.331·107 1.235·107


Total 8.134·108 5.623·108 3.909·108 2.738·108

Contribution ofCm-244 7.929·108 5.395·108 3.679·108

2.510·108

14

Table 2 (cont.)

c) The VVER spent fuel with the discharge bum up of 60 MW d/kgU (The uranium enrichment of 4.2 wt%)


( a,n)-reaction 2.693·107 2.299·107 2.001·107 1.774·107


Total 1.760·109 1.215·109 8.431·108 5.896·108

Contribution ofCm-244 1.712·109 1.168·109 7.962·108 5.430·108

11 BWR spent fuel

a) The BWR spent fuel with the discharge burnup of 40 MWdlkgU (The uranium enrichment of 3.8 wt%)


( a,n)-reaction 1.174·107 1.122·107 1.068·107 1.015·107


Total 5.266·108 3.647·108 2.541·108 1.785·108


15

Table 2 (cont.)

b) The BWR spent fuel with the discharge bumup of 50 MWdlkgU (The uranium enrichment of 3.8 wt%)


( a,n)-reaction 2.186·107 1.913·107 1.701·107 1.535·107


Total 1.336·109 9.216·108 6.391·108 4.462·108


c) The BWR spent fuel with the discharge bumup of 60 MWdlkgU (The uranium enrichment of 3.8 wt%)


( a,n)-reaction 3.622·107 2.995·107 2.535·107 2.193·107


Total 2.655·109 1.829·109 1.270·109 8.882·108


Table 3.

16

Average concentrations of Cs-137, Eu-154 and Cm-244 (gltU) of the spent WER and BWRfuel according to CASM0-4 and ORIGEN2.1 calculations (at the cooling time of20 years)

I Spent VVER fuel (The uranium enrichment of 4.2%)

Cs-137 Eu-154 Cm-244 Burnup 1 2 1 2 1 2 (MWdlkgU)

40 1148 1177 19.9 18.6 25.2 22.7 50 1426 1464 30.3 26.1 70.6 60.5 60 1698 1744 40.7 32.9 152.7 126.9

1 ORIGEN2.1 2 CASM0-4

11 Spent BWR fuel (The uranium enrichment of 3.80 o/o)

A) Cs-137 CASM0-4

Bum up Void history (%) (MWdlkgU) 0. 40. 80. ORIGEN2.1

40 924. 924. 924. 896. 50 1146. 1146. 1146. 1107. 60 1365. 1364. 1363. 1313.

B) Eu-154 CASM0-4

Bumup Void history (%) (MWdlkgU) 0. 40. 80. ORIGEN2.1

40 5.9 6.7 7.9 9.8 50 7.9 8.9 10.7 14.3 60 9.4 10.7 12.9 18.4

Table 3. (Cont.)

C) Cm-244

Bum up (MWd/kGU)

40 50 60

17

CASM0-4 Void history(%)

0. 40.

12.0 15.4 36.0 42.4 83.8 92.0

80.

20.4 51.1

101.7

ORIGEN2.1

31.1 79.2

156.6

18

4 CALCULATION OF SURFACE DOSE RATES

4.1 Geometry and material composition of the final disposal canisters

The horizontal cross sections of the BWR and VVER canisters are shown in Fig. 1. The canister versions are in this respect very similar, the biggest differences being the form and size of the holes in the cast iron insert, in which the spent fuel assemblies will be placed. The BWR canister is also longer than the VVER canister. The canisters have room for 12 bundles. The canisters will be filled with gas.

Following data describe the horizontal layouts of the canisters:

A) Copper mantle:

B) Iron insert:

- Outer radius - Thickness of the mantle - Density of copper

-Outer radius

52.6 cm· 5.0cm 8.96 g/cm3

- Density of nodular cast iron 47.5 cm 7.1 g/cm3

- Composition of cast iron - MARMER calculations: - MCNP4B calculations:

pure 1ron Fe 92.8 wto/o C 3.2 wt% Mg0.05 wt% Si 2.15 wt% Mn0.8 wt% Ni 1.0 wt% (Werme & Ericsson 1995)

The canisters were assumed to be homogenous in the axial direction, when dose rates on the outer surface of the canisters were calculated. Even then some data concerning the whole canisters and the fuel bundles had to be defined:

- Bundle geometry -Bundles in an assembly - Pitch of the rod lattice (cm) - Flow channel box with a bundle -Length of the fuel rod (m)3.68 -Uranium per bundle (kg) 179.5 - Number of fuel rods in a bundle

Canister type BWR VVER

Square 12 12.95 Yes 2.42 120 91

Hexagonal 12 14.4 Yes

126

· The B WR fuel bundle was assumed to be of the ATRIUM 1 Ox 1 0-9Q type.

Figure 1.

BWR -tyyppi

0 1.0 0')

VVER 440 -tyyppi

0 1.0 0')

19

60

58

Radial cross sections of the spent fuel disposal canisters for the BWR (OL1 and OL2) and WER-440 (Loviisa 1 and 2) fuel bundles.

20

The horizontal geometry of the fuel bundles and the canisters were described almost exactly in the basic MCNP4B calculations. The exceptions were that the gap between the fuel pellet and the clad was homogenized with the clad and that the inner water channel box of the ATRIUM bundle was omitted.

Besides the heterogeous' MCNP4B model described above, a homogenous' model of the BWR canister was defined. A ATRIUM 10x10-9Q fuel bundle was assumed to be specified as follows:'

- Radius of the fuel pellet - Inner radius of the clads - Outer radius of the clads - Pitch of the square pin cell - Density of the U02

- Density of the (Zr) clads

0.4335 cm 0.4420cm 0.5025 cm 1.295 cm

10.3 g/cm3

6.55 g/cm3

The 91 fuel rods of the bundle were described by a homogenous cylinder, the volume and mass of which were equal to the regular rod lattice (12.95xl2.95 cm2

). This cylinder was surrounded by a layer of Zirconium having the same mass per unit height as the channel box .

. The homogenized fuel bundle had the following properties:

- The radius of the fuel region - The outer radius of the channel box - The density of fuel region - The composition of the fuel region

- U02

-Zr

7.31 cm 7.56 cm 3.947 g/cm3

83.6 wto/o 16.4 wt%

In all shielding calculations, the fuel was assumed to be fresh uraniumdioxide.

21

4.2 Basic results

The average gamma and neutron dose rates on the outer surface of the final disposal canisters are given in Table 4. These results are from the heterogenous MCNP4B calculations assuming an axially homogeneous, infinitely long canister in a void.

Table 4.

Canister type

VVER BWR

Surface gamma and neutron dose rate (mSvlh) of the WER and BWR disposal canisters according to the heterogenous MCNP4B calculations.

Gamma dose rate 1 2

174 149

136 116

A

10.7 15.8

Neutron dose rate B C

8.3 12.3

9.3 13.7

1 ANSI/ANS-6.1.1-1977 photon flux-to-dose conversion factors 2 ANSI/ANS-6.1.1-1991 photon flux-to-dose conversion factors A AN SI/ ANS-6.1.1-1977 neutron flux -to-dose conversion factors B AN SI/ ANS-6.1.1-1991 neutron flux-to-dose conversion factors multiplied

by a factor of two C ICRP-74 neutron flux-to-dose conversion factors

Note: 1. The canisters filled with 12 spent fuel bundles having the bumup of 60 MWdlkgU and the cooling time of20 years.

The statistical uncertainty ( 1 cr) of the MCNP4B results in Table 4 is about 1% for the gamma dose rates and 0.1% for the neutron dose rates.

In Table 5 and Figure 2, the azimuthal variation of the gamma dose rates in a symmetric quarter are shown (In this report the zero angle is to the right along the horizontal symmetry line of the canisters, see Figures 1 and 4. The angle is increasing counterclockwise).

22

Table 5. Surface gamma dose rate (mSvlh) of the WER and BWR disposal canisters as a function of the azimuthal angle according to the MCNP4B calculations.

Canister type Azimuthal angle BWR VVER

(degrees) 1 2 1 2

0.0 100.4 78.3 108.3 84.6 11.25 168.7 131.5 245.6 191.0 22.50 257.0 199.7 280.2 217.4 33.75 99.5 77.2 90.5 70.4 45.00 19.8 15.5 23.9 18.8 56.25 100.7 78.5 96.1 74.9 67.50 279.5 218.4 299.5 232.9 78.75 167.6 130.7 251.1 195.7

Average 149 116 174 136

1 ANSI/ANS-6.1.1-1977 photon flux-to-dose conversion factors 2 ANSI/ANS-6.1.1-1991 photon flux-to-dose conversion factors

Note: 1. The canisters filled with spent fuel bundles having the burnup of 60 MW d/kg U and the cooling time of 20 years. 2. The MCNP dose rates are the average values over the 11.25-degree sectors, the centre point of which are at the angles given.

The statistical uncertainties (1cr) of sector-wise dose rates in Table 5 are rather large (from 2.5% to about 8%). The gamma dose rates and their azimuthal distributions of the VVER and BWR canisters are very similar.

.....--,..C

~ UJ a ...___.,

Q) ..,.....> (1j >-. Q) (j)

0 '"d Q) () (1j

'+-< >-. ;j

UJ

Figure 2.

Figure 3.

23

BWR and VVER canisters filled with 12 spent fuel bundles Burnup 60 MWd/kgU, cooling time of 20 years

300.0

250.0

200.0

150.0

100.0

50.0

0.0 0= BWR canister (MCNP, heterogenous model) 0 = VVER canister (MCNP, heterogenous model)

0.00 11.25 22.50 33.75 45.00 56.25 67.50 78.75 90.00

Azimuthal sector angle (degrees)

Gamma dose rate at the axial midplane of the outer surface of the BWR and WER canisters as a function of the azimuthal angle.

BWR canister filled with 12 ATRIUM 10x10-9Q bundles Burnup 60 MWd/kgU, cooling time of 20 years

300.0~------------------------------------------~

250.0

200.0

150.0

100.0

50.0

0.0 0= MARMER, homogenized b ndles 0= MCNP, heterogenous bundles 6.= MCNP, homogenized bundles

0.00 11.25 22.50 33.75 45.00 56.25 67.50 78.75 90.00

Azimuthal sector angle (degrees)

Gamma dose rate at the axial midplane of the outer surface of the BWR canister as a function of the azimuthal angle.

24

The gamma dose rates on the outer surface of the BWR canister predicted by three different calculation methods are compared in Table 6 and in Figure 3. The results are in a very good agreement, especially if one takes into account the fact that the MARMER dose rates are point-wise values, of which the average MARMER dose rate is calculated. Therefore, at the angles of 22.5 and 67.5 degrees the MARMER values are much higher than the MCNP4B results. Due to the geometry of the canister (see Fig. 1), the largest gamma dose rates are in the narrow sectors around the angles of 25 and 65 degrees.

Table 6. Surface gamma dose rate (mSvlh) of the BWR disposal canister as a function of the azimuthal angle according to three calculation models.

Computer code MARMER

Azimuthal angle (degrees)

MCNP Heterogenous Homogenous

model model

Average

0.0 11.25 22.50 33.75 45.00 56.25 67.50 78.75

78.3 131.5 199.7 77.2 15.5 78.5

218.4 130.7

116

73.6 87.9 126.5 146.1 216.9 288.2

81.5 84.9 16.1 12.7 84.7 83.7

215.8 296.7 130.6 139.1

118 142

Note: 1. The canister filled with 12 spent fuel bundles having the burnup of 60 MWdlkgU and the cooling time of20 years. 2. The MCNP dose rates are average values over the 11.25-degree sectors, the centre points of which are at the angles given above. The MARMER results are the dose rates at the angles given.

The results in Table 6 prove that the homogenous MCNP4B model produces almost the same dose rates as the heterogenous MCNP4B model. Furthermore, they indicate that MARMER calculations, where the geometry is similar to the homogenous MCNP4B model, can give accurate predictions of gamma dose rates.

The average surface neutron dose rates resulting from heterogenous and homogenous MCNP4B calculations on a BWR canister are given in Table 7. They agree with other almost completely (the differences are about 0.1 %, i.e. about the same as the statistical uncertainties ( 1 cr) of the calculations. The azimuthal variation of the neutron dose rate is relatively small. The maximum values in a 11.25 degree sector are about 9% higher than the average dose rate and the minimum value at the angle of 45° is about 75o/o of the average neutron dose rate.

Table 7.

25

Surface neutron dose rate (mSvlh) of the BWR disposal canisters according to the MCNP4B calculations.

Calculation model

Heterogenous Homogenous

Neutron dose rate A B C

15.8 12.3 13.7 15.8 12.3 13.7

A ANSI/ANS-6.1.1-1977 neutron flux-to-dose conversion factors B AN SI/ ANS-6.1.1-1991 neutron flux-to-dose conversion factors multiplied

by a factor of two C ICRP-74 neutron flux-to-dose conversion factors

Note: 1. The canisters filled with 12 spent fuel bundles having the bumup of 60 MWdlkgU and the cooling time of 20 years.

Radiation shields

According to the present plans, the final disposal canisters will be transferred from a place to another in the encapsulation plant in a radiation shield. There may be two different radiation shields. In the so-called transfer corridor of the plant the canisters will be in a cylindrical, 10 cm thick shield made of steel. When the canisters will be transported by a special vehicle into the repository, they will be placed in a cylindrical two-layer radiation shield. One layer is made of 15 cm thick steel and the other of 5 cm thick (borated) polyethylene (Mayer, Valimaki, Kukkola 1989, Appendix 18).

In this study, the following densities and compositions were used for steel and polyethylene of the radiation shields:

Steel:

Polyethy lene:

Density 7.86 g /cm3, Composition: Iron, 98.1 wto/o; Carbon,

0.2 wt%; Silicon, 0.5 wt% and Manganese, 1.2 wt%, Density 0.92 g/cm3

, Composition: Hydrogen 14.4 wto/o; Carbon, 85.6 wt%.

In the MCNP4B calculations, the systems studied were again assumed to be axially homogenous and infinitely long. In the horizontal cross section the homogenous model was specified. The MARMER solution area was similarly defined, except that it was set to be axially finite (the height being 368 cm). The BWR canister was assumed to be filled with 12 spent fuel bundles having the bumup of 60 MWdlkgU and the cooling time of 20 years.

26

The azimuthal variation of the gamma dose (at the axial midplane) on the outer surface of the radiation shields is given in Table 8. The MCNP dose rates are average values over the 11.25-degree sectors, the centre points of which are at the angles given above. The MARMER results are the dose rates at the angles given.

The differences between MCNP4B and MARMER gamma dose rates seem to increase, when the optical thickness of the radiation shield increases. MARMER predicts larger radiation levels than MCNP4B. However, the overall agreement is still satisfactory.

The corresponding neutron dose rates based on MCNP4B calculations are given in Table 9.

The polyethylene in the radiation shield of the transportation vehicle will most probably be borated in order to decrease neutron dose rates outside the shield. Two MCNP4B calculations were carried out assuming the boron contents of one and five per cent, respectively. In the former case the average surface neutron dose rate was reduced by about 23% and in the latter case by 30%. The results are preliminary, but they indicate that the optimization of a radiation shield is not a straightforward task (see Ueki, et al. 1996)

Table 8. Gamma dose rates (J.!Svlh) on the outer surface of two radiation shields for a BWR disposal canister as a function of the azimuthal angle.

Azimuthal angle (degrees)

Average

0.0 11.25 22.50 33.75 45.00 56.25 67.50 78.75

Radiation shield in the transfer corridor

MCNP4B MARMER

798 863 1142 1217 1693 1921 838 845 239 214 777 823

1603 2006 1152 1223

1030 1140

Radiation shield of the transportation vehicle MCNP4B MARMER

74 102 93 135

113 175 59 92 33 38 60 95

108 181 93 127

79 118

Note: 1. AN SI/ ANS-6.1.1-1991 photon flux-to-dose conversion factors

Table 9.

27

Average neutron dose rate (mSvlh) on the outer surface of the two radiation shields according to MCNP4B calculations.

Radiation shield

In the transfer corridor Transportation vehicle

Neutron dose rate A B C

7.3 5.3 6.3

0.51 0.40 0.46

A ANSI/ANS-6.1.1-1977 neutron flux-to-dose conversion factors B AN SI/ ANS-6.1.1-1991 neutron flux -to-dose conversion factors multiplied

by a factor of two C ICRP-7 4 neutron flux -to-dose conversion factors

4.3 Impact of backscattering on surface dose rates

All the results in Tables 4-7 are from calculations, where a canister (with or without a radiation shield) was assumed to be in the vacuum, i.e. all photons or neutrons coming out through the outer surface were immediately lost. In some test calculations with MCNP4B, the canister was assumed to be surrounded by air with relative humidity of 50% (the composition of air was taken from Ref. (Closs (ed.) 1996, p. 55). Replacing void with air increased gamma and neutron dose rates by a few tenths of per cent.

The impact of backscattering from walls made of concrete was also studied with MCNP4B as a part of wall-thickness calculations described in more detail in Chapter 5.2. If the minimum distance between the outer surface of the canister and the wall was 60 cm, the average gamma dose rate was found to increase by 6.5% due to backscattering on the outer surface of the canister facing the wall. For neutron radiation the corresponding increase was about 7.5o/o.

28

5 CALCULATION OF MINIMUM WALL THICKNESSES

5.1 Introduction

The following cases have been studied when detennining the wall thicknesses of an encapsulation plant, which fulfill the radiation protection criteria:

-a fuel (BWR/ATRIUM 10x10-9Q) bundle with its channel box in the hot cell, -a (fully filled BWR) disposal canister in a working room, - a disposal canister in a cylindrical, 10 cm thick radiation shield made of steel and -a disposal canister in a so-called buffer storage.

The choice of the calculation cases was based on the earlier design report of an encapsulation plant (Mayer, Valimaki, Kukkola 1989).

The main radiation protection condition was that behind a wall the total dose rate must be (clearly) under 2.5 J.!Sv/h, i.e. the rooms behind the walls could be utilized uncontrolled (STUK Guide STI.2).

All calculations were carried out with the fuel bundles with a discharge burnup of 60 MW dlkgU and a cooling time of 20 years.

In all cases, the walls were assumed to be made of concrete, the composition of which was as follows (Anttila, Wasastjema 1989, Table 3):

- Oxygen, 48.7 wt%; Silicon, 31.7 wt%; Iron, 2.57 wt%; Calcium, 7.5 wt%; Potassium, 1.69 wt%; Aluminium, 5.5 wt%; Hydrogen, 0.46 wt%; Magnesium, 1.4 wto/o; Boron 0.21 wt%.

The density of concrete was set to be 2.4 g/cm3 •

It was further assumed that the bundle or the canister with or without a radiation protection shield is at a distance of 60 cm from the wall (in the buffer storage the distance was assumed to be 120 cm).

In MCNP4B calculations, the systems were defined to be axially infinite and homogenous, but in MARMER calculations they were set to be finite (the height of the source region being 368 cm). A typical horizontal cross section of the calculation area is shown in Fig. 4.

29

boundary of the calculation area

I I BWR canister I I

I I I

-t--1

I I I I I

- -centre line

homogenized fuel bundle

L - - - - - - - - - - - - - - - -L-L..-.L.~____.___..

Figure 4.

5.1

concrete wall

Typical horizontal cross section of the solution area in wall-thickness calculations (In the case of the hot cell the BWR canister was replaced by a BWRfuel bundle of ATRIUM 10x10-9Q type).

Results

A bundle in the hot cell

In the hot cell of an encapsulation plant, spent fuel bundles will be removed from their transport cask, dried in a special facility and finally put into disposal canisters.

Gamma and neutron dose rates behind a concrete wall due to a single fuel bundle in the hot cell are given as a function of the wall thickness in Table 10. The axially infinite, homogenous MCNP4B model was used. A 110 cm thick layer of concrete can decrease the total dose rate to a low enough level.

Table 10.

30

Gamma and neutron dose rates (~vlh) behind a concrete wall due to a BWR bundle of ATRIUM 10xl0-9Q type with its channel box at the distance of60 cm from the wall according to MCNP4B calculations.

Average dose rate Maximum dose rate Thickness of the wall ( cni)

Gamma Neutron Gamma Neutron

60 80

100 110

1060 53.6

3.0 0.81

11.1 1.8 0.31 0.13

1240 62.8

3.6 0.97

14.0 2.4 0.45 0.18

Note: 1. The canister filled with 12 spent fuel bundles having the bumup of 60 MW dlkgU and the cooling time of 20 years. 2. The average dose rates are the average values of the 50 cm broad areas on both sides of the centre line. The maximum value is the largest dose rate over a 10 cm broad area. 3. ANSI/ANS-6.1.1-1991 photon flux-to-dose conversion factors and ICRP-74 neutron flux-to-dose conversion factors

When the wall thickness was 80 cm, neutron dose rates inside the wall were also calculated. The average dose rate was 4.4 mSv/h (± 0.1% ). With this results and those given in Table 10 one can calculate the thickness of a concrete layer, which can decrease the neutron dose rate by a factor of two ('one-half thickness'). Over the whole 80 cm thick wall the one-half thickness is 7.1 cm. After a 60 cm thick layer of concrete the corresponding value is 7.6-7.8 cm. For gamma radiation the one-half thickness is about 5 cm.

As an extreme, theoretical case, the neutron and gamma dose rates behind a 110 cm thick concrete wall was calculated in the case, where the source was an infinite lattice of ATRIUM 1 Ox 1 0-9Q bundles. The lattice pitch was set to be 24 cm. The gamma dose rate was 7.0 (±3.6%) f.lSv/h (ANSI/ANS-6.1.1-1991 conversion factors) and the neutron dose rate 3.5 (±1.9%) J.l.Svlh (ICRP-74 conversion factors). By applying the one-half thicknesses given above it can be calculated that an extra 20 cm thick layer of concrete is needed to reduce the total dose rate below the control value of 2.5 J.l.Sv/h.

A BWR canister in a operation room

Gamma dose rates behind a concrete wall caused by a single BWR canister filled with 12 fuel bundles, the discharge bum up and cooling time of which were 60 MW dlkgU and 20 years, respectively, were studied with the MARMER code. As a function of the wall thickness the dose rate is as follows (The distance between the outer surface of the canister and the wall was assumed to be 60 cm):

Wall thickness (cm) 70 80

100

31

Maximum dose rate (J.LSv/h) 3.3 0.8 0.06

According to a MCNP4B calculation the maximum neutron dose rate behind a 80 cm thick wall is 0.74 {±4%) J.LSv/h when using the ICRP-74 conversion factors. The total dose rate is then about 1.6 J.LSv/h, which is less than the specified limit, 2.5 J.LSvlh.

The average neutron dose on the inner surface of the concrete wall was according to the above-mentioned MCNP4B calculation 7.6 (±0.4%) mSv/h. In this case, the one-half thickness of concrete is about 6.0 cm.

A BWR canister inside a radiation shield in the transfer corridor

Gamma dose rates behind a concrete wall caused by a single B WR canister in a cylindrical radiation shield made of 10 thick steel, were calculated with the MARMER code. As a function of the wall thickness the dose rate is as follows (The distance between the outer surface of the canister and the wall was again 60 cm):

Wall thickness (cm) 40 50 70

Maximum dose rate (J.LSvlh) 2.9 0.76 0.06

Neutron dose rates on the inner and outer surface of the 50 cm thick concrete wall were calculated with MCNP4B applying the homogenized canister model. The maximum and average (averaged over a two-meter broad axial area) rates were as follows:

Inner surface Outer surface

Maximun dose rate

4.6 (±0.6%) mSv/h 4.4 (±8.0%) J.LSv/h

Average dose rate

4.0 (±0.5%) mSv/h 3.2 (±1.6%) J.LSv/h

The one-half thickness of concrete is about 5 cm. Therefore, to reduce the total dose rate behind the wall clearly under the limit value of 2.5 J.LSv/h, the wall thickness must be at least 60 cm.

A BWR canister in the buffer storage

According to the present plans there would be a buffer storage in the encapsulation plant. In the buffer storage a few canisters at a time would wait for their transportation to the repository (May er, Valimak.i, Kukkola 1989). In this study. it was assumed that in the buffer storage the canisters would be in a regular square lattice, the pitch of which is 160 cm. The distance of the outer surface of the nearest row of canisters from a concrete wall

32

was set to be 120 cm. The gamma dose rate behind the wall on the axial midplane of the canister was calculated with the MAR}.1ER point-kernel code as a function of the distance from the normal of the wall going along a symmetry line of the canister. The results for two wall thicknesses are given in Table 11.

A total dose rate at a given point can be estimated as a sum of the contributions of the canister itself and its two nearest neigbours. The maximum gamma dose rate behind the wall of the buffer storage would then be about 3 J.J.Sv/h, if the wall thickness were 70 cm. An extra 10 cm layer of concrete would decrease the dose rate by a factor of four, to ea. 0.8 J.J.Sv/h.

The neutron dose rates on the inner and outer surface of an 80 cm thick wall of the buffer storage was estimated by a MCNP4B calculation using the axially infinite, homogenous canister model. It was further assumed that along the wall there were an infinitely long row of canisters. The average neutron dose rate on the inner and outer surface of the wall over a 160 cm broad area were 29.8 (±0.4%) mSvlh and 1.52 (±3.6%) JlSvlh, respectively. According to these results the one-half thickness of the wall for neutron radiation is in this case only about 5.6 cm.

Table 11. Gamma dose rate (J.!Svlh) on the outer surface of a concrete wall of the buffer storage due to a BWR disposal canister according to MARMER calculations

Distance (cm)

0 10 20 30 50 80 100 130 140 150 160

Thickness of the concrete wall (cm) 70 80

2.4 0.64 2.4 0.64 2.3 0.61 2.2 0.58 1.9 0.51 1.4 1.0 0.26 0.6 0.15 0.5 0.4 0.3 0.08

~--

33

The total dose rate behind a 80 cm thick concrete wall of the buffer storage would be about 2.3 !lSv/h, if only the nearest row of canisters is taken into account. There will be a smaller contribution from the second row, too. For gamma radiation that contribution was estimated by a MARMER calculation, where the canisters of the flrst row were assumed to be wholly made of cast iron. The dose rate behind the 80 cm thick concrete wall caused by a single canister in the second row was calculated in the same way as for the canister in the frrst row. The results were as follows:

Distance (cm)

100 110 120 130 140 150 160 170

Gamma dose rate Q.tSvlh)

0.01 0.01 0.02 0.02 0.08 0.12 0.13 0.12

The maximum contribution of a second-row canister to the gamma dose rate behind the wall seems to be at the point, which is at the normal of the wall going through the centre points of the column of canisters nearest to the source canister. The maximum total contribution of the second row is produced by two canisters in the symmetrical positions on both sides of the dose point. It is according to MARMER results ea. 0.26 !lSv/h, i.e. about one third of the gamma dose rate due to the flrst row. By assuming the same relative contribution in the case of neutron radiation, the total dose rate behind the 80 cm thick wall would be about 3 f.!Svlh.

With the conservative assumptions applied in this study, the thickness of the walls of the buffer storage should be 90 cm in order to be sure that the total dose rate behind the wall is less than the limit value of 2.5 J..LSv/h.

34

6 CONCLUSIONS

The minimum thicknesses of the concrete walls of an encapsulation plant now in its design phase in Finland have been calculated with the MCNP4B code based on the Monte Carlo technique and with the MARMER code based on the point -kernel method.

The calculations have been carried out in three phases. First, photon and neutron source strengths and spectra have been estimated as a function of discharge bumup and cooling time of the spent fuel. Then, ganuna and neutron dose rates outside the final disposal canisters or their radiation shields have been calculated. Finally, the minimum wall thicknesses of the most important rooms of the encapsulation plant have been determined.

The minimum wall thicknesses were determined for the following four cases:

a) A fuel bundle with its channel box in the hot cell of the encapsulation plant b) A (fully filled) disposal canister in a operation room of the plant c) A disposal canister in a cylindrical, 10 cm thick radiation shield made of steel (used when the canister is transferred from one room to another inside the plant) and d) A disposal canister in the buffer storage of the plant.

The main design condition was that behind a wall the total dose rate must be (clearly) under 2.5 J..LSv/h, i.e. the rooms behind the walls should be classified as uncontrolled area from the point of radiation protection.

The spent fuel bundl~s were assumed to have a discharge bumup of 60 MWd/kgU and a cooling time of 20 years.

According to this study the minimum wall thicknesses are a) 110 cm, b) 80 cm, c) 60 cm and d) 90 cm. For the hot cell, the minimum wall thickness were 130 cm, if there were a very large storage lattice filled with spent fuel bundles.

In the next phase of this study a few calculations will be carried out in order to check and complement the data base No significant changes of the present conclusions are probable.

35

REFERENCES

AN SI/ ANS-6.1.1-1977, Neutron and gamma-ray flux-to-dose factors. An American National Standard published by the American Nuclear Society.

AN SI/ ANS-6.1.1-1991, Neutron and gamma-ray fluence-to-dose factors. An American National Standard published by the American Nuclear Society (Approved August 26, 1991). '

Anttila, M. 1992, Spent fuel characteristics for TV0-92 safety analysis. Report YJT -92-03 (in Finnish), Nuclear Waste Commission of Finnish Power Companies, Helsinki.

Anttila, M. 1995, Characteristics of Loviisa NPP spent fuel. Report YJT-95-2 (in Finnish), Nuclear Waste Commission of Finnish Power Companies, Helsinki. 45p. ISSN-0359-548X.

Anttila, M., Wasastjerna, F. 1995, Activity inventory of the activated decommissioning waste in the Olkiluoto power plant. Report YJT -89-12, Nuclear Waste Commission of Finnish Power Companies, Helsinki. 18p + 2 appendices. ISSN-0359-548X.

Anttila, M. 1996, Gamma and neutron dose rates on the outer surface of the nuclear waste disposal canisters. Report POSIV A-96-10, Posiva Oy, Helsinki. 30p +Appendices. ISBN 951-652-009-X.

Briesmeister, J.F. (Ed.) 1997, MCNP™-A General Monte Carlo N-Particle Transport Code: Version 4B. Los Alamos National Laboratory, LA-12625-M, Version 4B.

Closs, K.D., (Ed.) 1996, Active handling experiment with neutron sources (AHE). EUR 17124 EN (Final report),_ European Commission, Directorate-General, Science, Research and Development. ISBN 92-827-8673-0.

Edenius. M., et al. 1995, CASM0-4. A Fuel Assembly Burnup Program. User's Manual. Studsvik Report Studsvik/SOA-95-1. Studsvik of America, Inc./ Studsvik Core Analysis AB (Restricted Distribution).

The International Commission on Radiological Protection (ICRP) 1990, 1990 Recommendations of the International Commission on Radiological Protection, Adopted by the Commission in November 1990. ICRP Publication 60.

The International Commission on Radiological Protection (ICRP) 1996, Conversion Coefficients for use in Radiological Protection against External Radiation. Annals of the ICRP, Volume 26, No. 3/4 1996 I ICRP Publication 74. ISBN 0-08-042739-1.

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