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Contributions towards the development of a packaging concept for the final disposal of spent HTGR pebble bed fuel
Nuclear Engineering and Design 118 (1990) 107-113 North-HoUand
107
CONTRIBUTIONS TOWARDS THE DEVELOPMENT OF A PACKAGING CONCEPT FOR THE FINAL DISPOSAL OF SPENT HTGR PEBBLE BED FUEL H.U. B R I N K M A N N 1, R. D U W E 1, B. G A N S E R 2, A.-W. M E H N E R 2 a n d A. R E B M A N N 3 i KFA J~lich GmbH., J~lieh, Fed` Rep. Germany 2 NUKEM GmbH., Hanau, Fed. Rep. Germany 3 DBE GmbH., Fed. Rep. Germany
Received: first version 18 November 1988, revised version 20 March 1989
In accordance with decisions of the Government of the German Federal Republic on the treatment of spent fuel from nuclear reactors, the spent spherical fuel elements from the German pebble bed High Temperature Gascooled Reactors (HTGRs), AVR (Arbeitsgemeinschaft Versuchs Reaktor) and THTR (Thorium HochTemperatur Reaktor) shall be directly disposed of in the national repository. Owing to their radiological and thermal properties, the spent HTGR fuel may be treated similar to heat generating medium active waste for which the disposal concept is the emplacement in deep boreholes of a salt mine. Packaging concepts for the final disposal of spent HTGR fuel shall follow the general design and the radiological requirements for the borehole emplacement technique, but credit may also be taken of advantageous properties of the HTGR fuel elements. Especially their built-in barriers against fission product release, designed for stringent operational, upset or even accident conditions in service, can help in the development of a safe and economic packaging and disposal concept for spent HTGR fuel. Evaluation of the mechanical behaviour of HTGR spent fuel elements in experimental testing under near repository conditions has shown that the standard HTGR fuel particle can generally survive the crushing of the elements under rock pressure, but, for safety reasons, a backfill matrix such as quartz sand or cement should be introduced in the case of thin walled disposal canisters to prevent element breakage as a whole at rock pressure.
1. Introduction In accordance with the decisions of the Government of the German Federal Republic on the treatment of spent fuel from nuclear reactors, spent spherical fuel elements from the operating German pebble bed HTGRs, AVR (Arbeitsgemeinschaft Versuchs Reaktor) and THTR (Thorium HochTemperatur Reaktor) shall be disposed of directly in a salt mine, the national repository for heat generating nuclear waste [1]. Final storage of nuclear waste requires efficient natural and technical barriers to protect the biosphere against radioactive nuclides, e.g. the fission products contained in the spent H T G R fuel. Prior to final disposal, nuclear waste must be conditioned and put into a form suitable for acceptance in the repository, i.e. into a form which meets all radiologic and operational requirements laid down in legal rules and specifications. Because of the complex saltmine related boundary conditions involved and their combination with the radiologic requirements, it has generally been accepted
that satisfactory solutions and final specifications for packaging designs and emplacement techniques can be achieved by iteration only and must be accompanied by careful analysis work. Disposal in large selfshielding and corrosion resistant containers, under development for spent LWR fuel bundles of full length, was proposed as the provisional reference disposal concept for spent T H T R fuel. However, owing to both their radiological and thermal properties, spent fuel elements from pebble bed HTGRs could likewise be treated as heat generating medium active waste (MAW) from LWR fuel reprocessing. For these waste forms, zircalloy hulls, feed solution residues etc., the reference disposal concept is the emplacement of 400 1 volume packages in deep boreholes of the salt mine repository [2]. Adoption of this diposal concept looks rather more attractive for the H T G R spent fuel, especially if more than the two ~xisting H T G R plants need to be considered. Investigations into the mechanical behaviour of spent HTGR fuel elements were carried out. The results ob-
0 0 2 9 - 5 4 9 3 / 9 0 / $ 0 3 . 5 0 © Elsevier Science Publishers B.V.
108
H. U. Brinkmann et al. / Development of a packaging concept
tained were evaluated in a subsequent systems analysis together with a compilation of spent fuel data, technical boundary conditions and accident scenarios, known at present. Favourable fuel element properties such as the capability of longterm retention of relevant nuclides by barriers within the fuel elements were accounted for. The M A W concept adopted here involves waste stabilization by cementing and the use of not gas tight, thin walled containers as the standard packaging design. Selfsupporting containers, withstanding the rock pressure and possibly sealed against gas release have recently been proposed as a backup solution. For the M A W packages - mainly because of their stabilizing matrix component - the repository local temperature shall not exceed 100°C, preferably 90°C. The spent H T G R fuel, however, does not require such a temperature limit. Acceleration of a proper enclosure of the packages by converging salt rock is being sought by shifting the location for H T G R fuel disposal without changing the emplacement technique from the M A W area into the vicinity of the high active waste (HAW) borehole fields, where the temperature horizon is expected to exceed the 1 0 0 ° C level.
2. S y s t e m s a n a l y s i s c o n s i d e r a t i o n s
A variety of conceptual designs of packagings for the final storage was assessed: - canisters with loose spheres, - canisters with bonded spheres (cement, graphite compound), - canisters with backfilled spheres (quartz sand, salt grout). The pertinent fuel element properties and boundary conditions for the planned repository leading to the main requirements for the packaging concepts are given in table 1. The assessment of available boundary conditions for the planned repository revealed that the safety related requirements are not yet completely defined [3-6]. Therefore all statements based on these requirements are of preliminary status. The main criteria for comparison of the packaging design variants are as follows: - Can given requirements be met by technical measures and methods within the present state of the art? - Is the gas generation from radiolysis possible? - Which are the additional costs for conditioning?
Table 1 Design base for packaging concepts Fuel dement properties - Small dimensions: 600 mm dia - Low decay heat production: ~<0125 W / B E - High fission product retention provided by intact coating layers - Coated fuel particles are already embedded in structural matrix - High corrosion resistance of matrix and particle ioadings
Relmsitory boumdary conditions
Normal operation - Staple forces: - R o c k pressure:
5 MPa for 300 m hole 25 MPa for 1000 m depth - Temperature: 90 ° C (MAW-field) 200 * C (HAW-field) - Tolerable gas release from radiolysis: Maximum values not yet defined Accident conditions
Dropping during transport or into the borehole - Fire in the mine works - Brine contact in the closed borehole -
Design base for p ~ k a g l n g concepts for spent H T R fuel elements
- Packaging design: - Staple height:
- Resistance against rock pressure: - Temperature resistance: - Release of radionuclides: - Design against accident condition:
400 1 canister 8m 25 MPa 90 °/200 * C H-3 0.41x 109 Bq/canister- year Kr-85 1.10 x 1 0 9 Bq/canister- year not yet defined
109
H. U. Brinkmann et al. / Development of a packaging concept Table 2 Assessment matrix for different packaging designs Ser. No.
1
2
3
4
5
6
Back-fill material
none
normal cement
quartz sand
crushed rock salt
graphite compound
dehydrated cement
Resistance against rock pressure/MPa
< 10
> 50
100
Temperature resistance/°C
> 200
~<90
> 200
State of the art
Original canister
Pilot scale
Gas release by radiolysis
no
Relative Costs (investment, operating expenses) Suited for storage in HAW-field in MAW-field
> 50
> 30
~< 200
< 100
> 200
Pilot scale
Lab scale
Lab scale
Lab scale
(yes)
no
no
no
(no)
1
1,3
1,2
1,2
2,4
2,3
yes yes
no yes
yes yes
yes yes
no yes
yes yes
T h e assessment m a t r i x (table 2) using these criteria shows that only fuel elements b a c k filled with c e m e n t or q u a r t z sand c a n reasonably resist the design rock pres-
•
17-26
sure of 25 M P a and, besides, c a n b e realized in pilot scale c o m p o n e n t s i.e. b y techniques already available with reasonable effort.
/
13i ~"
'~
t0~r~
Temperature/°C
0,59
1]
I
f
-/
0,87
),S
m
75
VI Fig. 1. Particle failure in crushed V 1 Single spheres unirradiated no backfill matrix V 2 Model package 21 spheres, unirradiated no backfill matrix V 3 Model package 21 spheres, unirradiated cement matrix
V2
v3
v4
v5
ve
Ser. No.
model packages containing spheres with TRISO coated particles and different backfill materials. V 4 Model package 21 spheres, unirradiated salt grout V 5 Model package 21 spheres, unirradiated quartz sand V B Single spheres irradiated no backfill matrix.
110
1t. U. Brinkmann et al. / Development of a packaging concept
3. E x p e r i m e n t a l investigations
The experimental part of this work consisted of mechanical tests with single spheres and with model arrangements with and without a backfill matrix. Experimental conditions were resembling the converging rock salt. The impacts by canister dropping within the borehole and loads occurring during stapling were not taken into account. Test conditions were derived from boundary conditions for a conceptual design of a borehole situated in a depth of about 800-1000 m. The expected maximum rock pressure is 25 MPa, the maximum temperature is 90°C for storage in a MAW-field and at maximum 200°C when HTR spent fuel disposal is envisaged in the vicinity of boreholes with HAW canisters. The latter would exclude the use of a cement matrix as this material would rapidly deteriorate at this temperature. A variety of pressing tools were used to simulate the load to be experienced by a single sphere or in packing configurations. The effect of backfill materials was tested with unirradiated material only. 3.1. Experiments with unirradiated single spheres and model configurations at N U K E M
The main targets of these experiments were first to qualify the pressing tools, to determine the effect of various backfill materials, to look for any differences between spheres with BISO and TRISO coated particles and to study possible scale up effects for configurations with 7 and 21 spheres. All experiments were performed either with loose spheres or with additional backfilling using cement, quartz sand or rock salt grout as the filler matrix. The experimental temperatures ranged from room temperature to 200°C. Preliminary crushing tests were performed using matrix spheres without fuel particles. Different point loads on single spheres were simulated by using configurations with 1, 7 or 21 spheres. The degree of damage was assessed by a grain size analysis of the crushed material. No dependence upon the pressure ram material whether it was the original metal piece or moulded rock salt as an interlayer could be detected. Temperature had no marked influence in the range from room temperature up to 200 o C. Configurations with 7 spheres in one layer backfilled with cement or quartz sand did not show any failure. Configurations with 21 spheres in three layers behaved similarly. Slight failures were observed in both configurations when backfilled with rock salt grout. This was checked as due to densification of the salt
grout, i.e. a sintering effect leading to imperfections in the matrix continuity. Based on the results for crushed matrix spheres an experimental program with fueled spheres was started. Failed particles of TRISO coated type were detected by the burn leach test. In this test the matrix graphite and outer pyrocarbon layer is burned off. In the case of failed SiC-layers the inner pyrocarbon is burned away too and the exposed fuel kernel can be dissolved and the uranium analysed [7]. For B1SO coated particle fuel with no protective SiC interlayer a gas leach procedure has to be used where the fuel material is leached from defect particles by hot chlorine gas [8]. Loose single spheres were tested in a mould with 25 MPa (7.1 t maximum load) between metal or rock salt pistons. Temperature varied between room temperature and 200 o C. The temperature of the mould was controlled by an oil heater. Proper sample temperature was established by preheating in a drying furnace for 24 h. Temperature levels were the same as in single sphere tests. The recipe for the backfill cement slurry was a 1 : 1 mixture of blast furnace slag cement and quartz sand. For all samples a setting time of about one month was used. The same quartz sand was used for the backfill variant "sand". Common salt with a grain size between 0.2-0.8 mm was used as the backfill variant "salt". The evaluation of the experiments was with special emphasis on the coated fuel particle integrity. Quantitative results are given in fig. l. The packages without backfill material exhibited a considerable volume reduction with nearly all spheres crushed. Inspite of most of the spheres crumbling little damage on the fuel particles themselves was observed. Backfilling with quartz sand or cement, however, fully stabilized the initial volume with hardly any sphere damaged. Backfilling with salt grout, especially at increased temperatures led to additional sintering of the salt and hence, to considerable damage of the spheres but also in this case particle failure fraction did not show up to the same proportion.
3.2. Pilot scale experiments at N U K E M
For these test canisters of 180 1 volume and a high pressure ram capable of a maximum load of 1500 t were used. Four containers with 800 graphite matrix spheres each were prepared as follows: - one container with the spheres loosely packed,
H.U. Brinkmann et al. / Development of a packaging concept
- one container with the spheres backfilled with quartz sand, - two containers with the spheres backfilled with cement. The crushing tests were executed at room temperature with a final load of 520 t rendering a pressure of 25 MPa. The package containing loose spheres was compressed and thereby considerably densified when apply-
~
~ ii
~
111
ing the 25 MPa pressure. Analysis of the contents revealed 4 intact pebbles only: 91% of the spheres had crumbled into small pieces. These observations are in agreement with the results from the three layer tests. Pressing the two canisters, backfilled with cement, at 25 MPa final pressure did not result in a detectable volume change. One of the canisters was cut in two halves. The backfill matrix was such that all the initial
i il ~i~ ii~
Fig. 2. Crushing tests with 180 1 waste canisters at room temperature (final load 25 MPa). (Top-left) 1500+pressing device, (top-fight) package with loose spheres, (bottom-left) backfilled with cement test, (bottom-right) left package cut open: still intact.
112
H.U. Brinkmann et al. / Development of a packaging concept
void volume was totally consumed. N o failure marks except some very thin cracks or scratch marks were observed (fig. 2). The canister backfilled with quartz sand also underwent no remarkable volume change. Some (2.7%) failed spheres in the upper region were detected, however. A thorough analysis led to assume that the upper crimp of the canister had caused the failures. Besides this effect, no significant differences between the results of these tests and those of the three layer experiments with graphite matrix spheres were observed. The second package, backfilled with cement was eventually taken to a pressure level of 72 MPa, the maximum possible load with the pressing device. This condition is far beyond any design rock pressure. A volume reduction of about 3% was observed, the deformation starting at pressures beyond 35 MPa. This canister, too, was cut into two halves. Optical inspection showed 63% of the spheres to have cracked severely, 30% showed large, 6% still visible but thin cracks. Only 0.8% showed no detectable failures. F r o m this result it can be concluded that failure in cemented packages starts far beyond expected loads in the repository, i.e. that a fair margin of strength is at hand. 3.3. Experiments with irradiated single fuel spheres at KFA The integrity of the fuel particle coatings governs the mobilization of radionuclides at the attack of brine which, according to the model might reach the fuel immediately after emplacement and proper closure of the repository location. A remotely operable device allowing load pressures up to 30 MPa onto single irradiated spheres was
l-lat~ l
Fig. 3. Flow sheet of hot cell device to investigate release of Kr-85 during crushing of single irradiated spheres. constructed and installed in a hot cell (fig. 3). The equipment consists of an outer cylindrical steel vessel which houses the sample container, the working platon and the pressure ram. The vessel can be swept with an inert gas and is mounted on a movable rail of a test machine. The Kr-85 release during particle failure is monitored by means of an ionization chamber. The carrier gas for such Kr-85 release is air which, by a diaphragm pump, is swept through the crushing device and from there through a filter and a rotameter into the ionization chamber. The fuel element is placed into the sample container and crushed by moving the rail hydraulically with the pressure ram. First the fuel element cracks and then it is densified until it completely fills up the cylindrical sample container space. 10 fuel elements irradiated in the A V R and containing 15000-20000 particles each were investigated. The experimental parameters and the results are given in table 3. An example for the time dependence of a measurable Kr-85 activity is given in fig. 4. The activity release
Table 3 Number of failed particles in single spheres crushed after irradiation (final load 25 MPa) No.
Fuel type
Coating type
Burnup (~ FIMA)
Fract. rel. Kr-85 ( × 10 4)
Crushing strength (kN)
1 2 3 4 5 6 7 8 9 10
GO-1 GO-1 GO-3 CrO-3 GO-3 GO-3 GO-3 GLE-3 GLE-3 GLE-3
BISO BISO BISO BISO BISO BISO BISO TRISO TRISO TRISO
16.7 17.9 10.6 11.0 11.2 11.2 12.5 4.0 6.6 7.2
2.30 0.14 0.25 0.16 0.12 0.16 0.71 0.19 1.40 2.90
27.0 24.6 27.8 32.5 24.0 28.5 28.7 31.0 28.1 33.0
Number of failed part. ~<13 ~<3 ~<4 <~10
H.U. Brinkmann et al. / Development of a packaging concept
113
Crushing tests with spent fuel elements of full burnup will be extended to configurations with packs containing up to 9 spheres under repository relevant temperatures. These tests will render even more realistic results.
K r - 85 bursts
2~
Acknowledgements >
Pariicle failure
Fig. 4. Characteristic behaviour of single spheres crushed after irradiation showing spontaneous Kr-85 release (burst) from failing particles.
comprises a small continuous fraction coming from the matrix contamination and in our example two short bursts. These bursts can be assigned to two failing particles. Tests with single fuel particles cracking under mechanical load had proven the sensitivity of such a measuring device for Kr-85 release [9]. The evaluation of the experiments revealed: - Particle may fail under the given experimental conditions, and if so, the failure of a single particle is detectable. - P a r t i c l e failure depends on the coating design. However, for both the BISO and T R I S O coated particles the failure fraction observed remained below 10 -3 . - The first breakage of a fuel element in this device did not lead to any particle failure. - Variation of the loading speed by a factor of 20 was not influencing the test result i.e. the particle failure fraction.
4. Conclusions Results obtained so far show that, inspite of fuel elements crushing under converging rock pressure, there appears to be little damage to fuel particles. The present evaluation and assessment for different backfill matrices was done at a time where a complete set of boundary conditions for the repository and the emplacement technique for borehole disposal was not yet available. As a consequence, further experiments are needed with the increasing knowledge about the repository and appropriate boundary conditions becoming available.
The work presented here was performed in the H B K Project of the Entwicklungsgemeinschaft H T R , a cooperation between K F A , H O B E G / N U K E M , I N T E R A T O M , H R B and S I G R I / R i n g s d o r f f - W e r k e for H T R fuel and graphite development sponsored by the German Federal G o v e r n m e n t and the State G o v e r n m e n t of Northrhine-Westphalia. The N U K E M contribution was done under contract K W A 3601 from the Federal Minister of Research and Technology.
References [1] Pressemitteilung 8/85 des BMFT vom 22. Jan. 1985, Beschluss der Bundesregierung vom 23. Jan. 1985 zur abschliessenden Beurteilung der Anderen Entsorgungstechniken. [2] H. B~cher et al., Versuch zur Einlagerung von mittelaktivem Abfall und HTR-Brermelementen im Salzbergwerk ASSE, Kerntechnik 50, Nr. 1 (1987) 40-44. [3] B. Ganser et al., FuE-Arbeiten zur Endlagerung von HTRBrennelementen, Statusseminar HochtemperaturreaktorBrennstoffkreislauf, KFA-Jiil Conf-61 (1987) 147-160. [4] P. Brennecke und E. Warnecke, Anforderungen an endzulagernde radioaktive Abf~llle (Vod~iufige Endlagerungsbedingungen, Stand November 1986), PTB-Bericht SE-16, Braunschweig (1987). [5] Sicherheitskriterien for die Endlagerung radioaktiver Abf~Ulein einem Bergwerk, Bundesanzeiger 35, Nr. 2 (1983) 45-46. [6] "Modellans'~tze und Ergebnisse zur Freisetzung yon Radionukliden aus einem Modell-Salzstock", Projekt "Sicherheitsstudien Entsorgung", Abschlussbericht PSEII, Fachband 16 (Januar 1985). [7] NUKEM, priv. communication. [8] G. Stolba, G. Reitsamer, G. Falta und W. Schertk, Chlorgaslaugung unbestrahlter und bestrahlter HTR-Brennelemente, Jahrestagung Kerntechnik des DAtF (1980) Berlin, Tagungsberichte, pp. 586-589. [9] R. Duwe and A. Rebmann, KFA internal report.
107
CONTRIBUTIONS TOWARDS THE DEVELOPMENT OF A PACKAGING CONCEPT FOR THE FINAL DISPOSAL OF SPENT HTGR PEBBLE BED FUEL H.U. B R I N K M A N N 1, R. D U W E 1, B. G A N S E R 2, A.-W. M E H N E R 2 a n d A. R E B M A N N 3 i KFA J~lich GmbH., J~lieh, Fed` Rep. Germany 2 NUKEM GmbH., Hanau, Fed. Rep. Germany 3 DBE GmbH., Fed. Rep. Germany
Received: first version 18 November 1988, revised version 20 March 1989
In accordance with decisions of the Government of the German Federal Republic on the treatment of spent fuel from nuclear reactors, the spent spherical fuel elements from the German pebble bed High Temperature Gascooled Reactors (HTGRs), AVR (Arbeitsgemeinschaft Versuchs Reaktor) and THTR (Thorium HochTemperatur Reaktor) shall be directly disposed of in the national repository. Owing to their radiological and thermal properties, the spent HTGR fuel may be treated similar to heat generating medium active waste for which the disposal concept is the emplacement in deep boreholes of a salt mine. Packaging concepts for the final disposal of spent HTGR fuel shall follow the general design and the radiological requirements for the borehole emplacement technique, but credit may also be taken of advantageous properties of the HTGR fuel elements. Especially their built-in barriers against fission product release, designed for stringent operational, upset or even accident conditions in service, can help in the development of a safe and economic packaging and disposal concept for spent HTGR fuel. Evaluation of the mechanical behaviour of HTGR spent fuel elements in experimental testing under near repository conditions has shown that the standard HTGR fuel particle can generally survive the crushing of the elements under rock pressure, but, for safety reasons, a backfill matrix such as quartz sand or cement should be introduced in the case of thin walled disposal canisters to prevent element breakage as a whole at rock pressure.
1. Introduction In accordance with the decisions of the Government of the German Federal Republic on the treatment of spent fuel from nuclear reactors, spent spherical fuel elements from the operating German pebble bed HTGRs, AVR (Arbeitsgemeinschaft Versuchs Reaktor) and THTR (Thorium HochTemperatur Reaktor) shall be disposed of directly in a salt mine, the national repository for heat generating nuclear waste [1]. Final storage of nuclear waste requires efficient natural and technical barriers to protect the biosphere against radioactive nuclides, e.g. the fission products contained in the spent H T G R fuel. Prior to final disposal, nuclear waste must be conditioned and put into a form suitable for acceptance in the repository, i.e. into a form which meets all radiologic and operational requirements laid down in legal rules and specifications. Because of the complex saltmine related boundary conditions involved and their combination with the radiologic requirements, it has generally been accepted
that satisfactory solutions and final specifications for packaging designs and emplacement techniques can be achieved by iteration only and must be accompanied by careful analysis work. Disposal in large selfshielding and corrosion resistant containers, under development for spent LWR fuel bundles of full length, was proposed as the provisional reference disposal concept for spent T H T R fuel. However, owing to both their radiological and thermal properties, spent fuel elements from pebble bed HTGRs could likewise be treated as heat generating medium active waste (MAW) from LWR fuel reprocessing. For these waste forms, zircalloy hulls, feed solution residues etc., the reference disposal concept is the emplacement of 400 1 volume packages in deep boreholes of the salt mine repository [2]. Adoption of this diposal concept looks rather more attractive for the H T G R spent fuel, especially if more than the two ~xisting H T G R plants need to be considered. Investigations into the mechanical behaviour of spent HTGR fuel elements were carried out. The results ob-
0 0 2 9 - 5 4 9 3 / 9 0 / $ 0 3 . 5 0 © Elsevier Science Publishers B.V.
108
H. U. Brinkmann et al. / Development of a packaging concept
tained were evaluated in a subsequent systems analysis together with a compilation of spent fuel data, technical boundary conditions and accident scenarios, known at present. Favourable fuel element properties such as the capability of longterm retention of relevant nuclides by barriers within the fuel elements were accounted for. The M A W concept adopted here involves waste stabilization by cementing and the use of not gas tight, thin walled containers as the standard packaging design. Selfsupporting containers, withstanding the rock pressure and possibly sealed against gas release have recently been proposed as a backup solution. For the M A W packages - mainly because of their stabilizing matrix component - the repository local temperature shall not exceed 100°C, preferably 90°C. The spent H T G R fuel, however, does not require such a temperature limit. Acceleration of a proper enclosure of the packages by converging salt rock is being sought by shifting the location for H T G R fuel disposal without changing the emplacement technique from the M A W area into the vicinity of the high active waste (HAW) borehole fields, where the temperature horizon is expected to exceed the 1 0 0 ° C level.
2. S y s t e m s a n a l y s i s c o n s i d e r a t i o n s
A variety of conceptual designs of packagings for the final storage was assessed: - canisters with loose spheres, - canisters with bonded spheres (cement, graphite compound), - canisters with backfilled spheres (quartz sand, salt grout). The pertinent fuel element properties and boundary conditions for the planned repository leading to the main requirements for the packaging concepts are given in table 1. The assessment of available boundary conditions for the planned repository revealed that the safety related requirements are not yet completely defined [3-6]. Therefore all statements based on these requirements are of preliminary status. The main criteria for comparison of the packaging design variants are as follows: - Can given requirements be met by technical measures and methods within the present state of the art? - Is the gas generation from radiolysis possible? - Which are the additional costs for conditioning?
Table 1 Design base for packaging concepts Fuel dement properties - Small dimensions: 600 mm dia - Low decay heat production: ~<0125 W / B E - High fission product retention provided by intact coating layers - Coated fuel particles are already embedded in structural matrix - High corrosion resistance of matrix and particle ioadings
Relmsitory boumdary conditions
Normal operation - Staple forces: - R o c k pressure:
5 MPa for 300 m hole 25 MPa for 1000 m depth - Temperature: 90 ° C (MAW-field) 200 * C (HAW-field) - Tolerable gas release from radiolysis: Maximum values not yet defined Accident conditions
Dropping during transport or into the borehole - Fire in the mine works - Brine contact in the closed borehole -
Design base for p ~ k a g l n g concepts for spent H T R fuel elements
- Packaging design: - Staple height:
- Resistance against rock pressure: - Temperature resistance: - Release of radionuclides: - Design against accident condition:
400 1 canister 8m 25 MPa 90 °/200 * C H-3 0.41x 109 Bq/canister- year Kr-85 1.10 x 1 0 9 Bq/canister- year not yet defined
109
H. U. Brinkmann et al. / Development of a packaging concept Table 2 Assessment matrix for different packaging designs Ser. No.
1
2
3
4
5
6
Back-fill material
none
normal cement
quartz sand
crushed rock salt
graphite compound
dehydrated cement
Resistance against rock pressure/MPa
< 10
> 50
100
Temperature resistance/°C
> 200
~<90
> 200
State of the art
Original canister
Pilot scale
Gas release by radiolysis
no
Relative Costs (investment, operating expenses) Suited for storage in HAW-field in MAW-field
> 50
> 30
~< 200
< 100
> 200
Pilot scale
Lab scale
Lab scale
Lab scale
(yes)
no
no
no
(no)
1
1,3
1,2
1,2
2,4
2,3
yes yes
no yes
yes yes
yes yes
no yes
yes yes
T h e assessment m a t r i x (table 2) using these criteria shows that only fuel elements b a c k filled with c e m e n t or q u a r t z sand c a n reasonably resist the design rock pres-
•
17-26
sure of 25 M P a and, besides, c a n b e realized in pilot scale c o m p o n e n t s i.e. b y techniques already available with reasonable effort.
/
13i ~"
'~
t0~r~
Temperature/°C
0,59
1]
I
f
-/
0,87
),S
m
75
VI Fig. 1. Particle failure in crushed V 1 Single spheres unirradiated no backfill matrix V 2 Model package 21 spheres, unirradiated no backfill matrix V 3 Model package 21 spheres, unirradiated cement matrix
V2
v3
v4
v5
ve
Ser. No.
model packages containing spheres with TRISO coated particles and different backfill materials. V 4 Model package 21 spheres, unirradiated salt grout V 5 Model package 21 spheres, unirradiated quartz sand V B Single spheres irradiated no backfill matrix.
110
1t. U. Brinkmann et al. / Development of a packaging concept
3. E x p e r i m e n t a l investigations
The experimental part of this work consisted of mechanical tests with single spheres and with model arrangements with and without a backfill matrix. Experimental conditions were resembling the converging rock salt. The impacts by canister dropping within the borehole and loads occurring during stapling were not taken into account. Test conditions were derived from boundary conditions for a conceptual design of a borehole situated in a depth of about 800-1000 m. The expected maximum rock pressure is 25 MPa, the maximum temperature is 90°C for storage in a MAW-field and at maximum 200°C when HTR spent fuel disposal is envisaged in the vicinity of boreholes with HAW canisters. The latter would exclude the use of a cement matrix as this material would rapidly deteriorate at this temperature. A variety of pressing tools were used to simulate the load to be experienced by a single sphere or in packing configurations. The effect of backfill materials was tested with unirradiated material only. 3.1. Experiments with unirradiated single spheres and model configurations at N U K E M
The main targets of these experiments were first to qualify the pressing tools, to determine the effect of various backfill materials, to look for any differences between spheres with BISO and TRISO coated particles and to study possible scale up effects for configurations with 7 and 21 spheres. All experiments were performed either with loose spheres or with additional backfilling using cement, quartz sand or rock salt grout as the filler matrix. The experimental temperatures ranged from room temperature to 200°C. Preliminary crushing tests were performed using matrix spheres without fuel particles. Different point loads on single spheres were simulated by using configurations with 1, 7 or 21 spheres. The degree of damage was assessed by a grain size analysis of the crushed material. No dependence upon the pressure ram material whether it was the original metal piece or moulded rock salt as an interlayer could be detected. Temperature had no marked influence in the range from room temperature up to 200 o C. Configurations with 7 spheres in one layer backfilled with cement or quartz sand did not show any failure. Configurations with 21 spheres in three layers behaved similarly. Slight failures were observed in both configurations when backfilled with rock salt grout. This was checked as due to densification of the salt
grout, i.e. a sintering effect leading to imperfections in the matrix continuity. Based on the results for crushed matrix spheres an experimental program with fueled spheres was started. Failed particles of TRISO coated type were detected by the burn leach test. In this test the matrix graphite and outer pyrocarbon layer is burned off. In the case of failed SiC-layers the inner pyrocarbon is burned away too and the exposed fuel kernel can be dissolved and the uranium analysed [7]. For B1SO coated particle fuel with no protective SiC interlayer a gas leach procedure has to be used where the fuel material is leached from defect particles by hot chlorine gas [8]. Loose single spheres were tested in a mould with 25 MPa (7.1 t maximum load) between metal or rock salt pistons. Temperature varied between room temperature and 200 o C. The temperature of the mould was controlled by an oil heater. Proper sample temperature was established by preheating in a drying furnace for 24 h. Temperature levels were the same as in single sphere tests. The recipe for the backfill cement slurry was a 1 : 1 mixture of blast furnace slag cement and quartz sand. For all samples a setting time of about one month was used. The same quartz sand was used for the backfill variant "sand". Common salt with a grain size between 0.2-0.8 mm was used as the backfill variant "salt". The evaluation of the experiments was with special emphasis on the coated fuel particle integrity. Quantitative results are given in fig. l. The packages without backfill material exhibited a considerable volume reduction with nearly all spheres crushed. Inspite of most of the spheres crumbling little damage on the fuel particles themselves was observed. Backfilling with quartz sand or cement, however, fully stabilized the initial volume with hardly any sphere damaged. Backfilling with salt grout, especially at increased temperatures led to additional sintering of the salt and hence, to considerable damage of the spheres but also in this case particle failure fraction did not show up to the same proportion.
3.2. Pilot scale experiments at N U K E M
For these test canisters of 180 1 volume and a high pressure ram capable of a maximum load of 1500 t were used. Four containers with 800 graphite matrix spheres each were prepared as follows: - one container with the spheres loosely packed,
H.U. Brinkmann et al. / Development of a packaging concept
- one container with the spheres backfilled with quartz sand, - two containers with the spheres backfilled with cement. The crushing tests were executed at room temperature with a final load of 520 t rendering a pressure of 25 MPa. The package containing loose spheres was compressed and thereby considerably densified when apply-
~
~ ii
~
111
ing the 25 MPa pressure. Analysis of the contents revealed 4 intact pebbles only: 91% of the spheres had crumbled into small pieces. These observations are in agreement with the results from the three layer tests. Pressing the two canisters, backfilled with cement, at 25 MPa final pressure did not result in a detectable volume change. One of the canisters was cut in two halves. The backfill matrix was such that all the initial
i il ~i~ ii~
Fig. 2. Crushing tests with 180 1 waste canisters at room temperature (final load 25 MPa). (Top-left) 1500+pressing device, (top-fight) package with loose spheres, (bottom-left) backfilled with cement test, (bottom-right) left package cut open: still intact.
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H.U. Brinkmann et al. / Development of a packaging concept
void volume was totally consumed. N o failure marks except some very thin cracks or scratch marks were observed (fig. 2). The canister backfilled with quartz sand also underwent no remarkable volume change. Some (2.7%) failed spheres in the upper region were detected, however. A thorough analysis led to assume that the upper crimp of the canister had caused the failures. Besides this effect, no significant differences between the results of these tests and those of the three layer experiments with graphite matrix spheres were observed. The second package, backfilled with cement was eventually taken to a pressure level of 72 MPa, the maximum possible load with the pressing device. This condition is far beyond any design rock pressure. A volume reduction of about 3% was observed, the deformation starting at pressures beyond 35 MPa. This canister, too, was cut into two halves. Optical inspection showed 63% of the spheres to have cracked severely, 30% showed large, 6% still visible but thin cracks. Only 0.8% showed no detectable failures. F r o m this result it can be concluded that failure in cemented packages starts far beyond expected loads in the repository, i.e. that a fair margin of strength is at hand. 3.3. Experiments with irradiated single fuel spheres at KFA The integrity of the fuel particle coatings governs the mobilization of radionuclides at the attack of brine which, according to the model might reach the fuel immediately after emplacement and proper closure of the repository location. A remotely operable device allowing load pressures up to 30 MPa onto single irradiated spheres was
l-lat~ l
Fig. 3. Flow sheet of hot cell device to investigate release of Kr-85 during crushing of single irradiated spheres. constructed and installed in a hot cell (fig. 3). The equipment consists of an outer cylindrical steel vessel which houses the sample container, the working platon and the pressure ram. The vessel can be swept with an inert gas and is mounted on a movable rail of a test machine. The Kr-85 release during particle failure is monitored by means of an ionization chamber. The carrier gas for such Kr-85 release is air which, by a diaphragm pump, is swept through the crushing device and from there through a filter and a rotameter into the ionization chamber. The fuel element is placed into the sample container and crushed by moving the rail hydraulically with the pressure ram. First the fuel element cracks and then it is densified until it completely fills up the cylindrical sample container space. 10 fuel elements irradiated in the A V R and containing 15000-20000 particles each were investigated. The experimental parameters and the results are given in table 3. An example for the time dependence of a measurable Kr-85 activity is given in fig. 4. The activity release
Table 3 Number of failed particles in single spheres crushed after irradiation (final load 25 MPa) No.
Fuel type
Coating type
Burnup (~ FIMA)
Fract. rel. Kr-85 ( × 10 4)
Crushing strength (kN)
1 2 3 4 5 6 7 8 9 10
GO-1 GO-1 GO-3 CrO-3 GO-3 GO-3 GO-3 GLE-3 GLE-3 GLE-3
BISO BISO BISO BISO BISO BISO BISO TRISO TRISO TRISO
16.7 17.9 10.6 11.0 11.2 11.2 12.5 4.0 6.6 7.2
2.30 0.14 0.25 0.16 0.12 0.16 0.71 0.19 1.40 2.90
27.0 24.6 27.8 32.5 24.0 28.5 28.7 31.0 28.1 33.0
Number of failed part. ~<13 ~<3 ~<4 <~10
H.U. Brinkmann et al. / Development of a packaging concept
113
Crushing tests with spent fuel elements of full burnup will be extended to configurations with packs containing up to 9 spheres under repository relevant temperatures. These tests will render even more realistic results.
K r - 85 bursts
2~
Acknowledgements >
Pariicle failure
Fig. 4. Characteristic behaviour of single spheres crushed after irradiation showing spontaneous Kr-85 release (burst) from failing particles.
comprises a small continuous fraction coming from the matrix contamination and in our example two short bursts. These bursts can be assigned to two failing particles. Tests with single fuel particles cracking under mechanical load had proven the sensitivity of such a measuring device for Kr-85 release [9]. The evaluation of the experiments revealed: - Particle may fail under the given experimental conditions, and if so, the failure of a single particle is detectable. - P a r t i c l e failure depends on the coating design. However, for both the BISO and T R I S O coated particles the failure fraction observed remained below 10 -3 . - The first breakage of a fuel element in this device did not lead to any particle failure. - Variation of the loading speed by a factor of 20 was not influencing the test result i.e. the particle failure fraction.
4. Conclusions Results obtained so far show that, inspite of fuel elements crushing under converging rock pressure, there appears to be little damage to fuel particles. The present evaluation and assessment for different backfill matrices was done at a time where a complete set of boundary conditions for the repository and the emplacement technique for borehole disposal was not yet available. As a consequence, further experiments are needed with the increasing knowledge about the repository and appropriate boundary conditions becoming available.
The work presented here was performed in the H B K Project of the Entwicklungsgemeinschaft H T R , a cooperation between K F A , H O B E G / N U K E M , I N T E R A T O M , H R B and S I G R I / R i n g s d o r f f - W e r k e for H T R fuel and graphite development sponsored by the German Federal G o v e r n m e n t and the State G o v e r n m e n t of Northrhine-Westphalia. The N U K E M contribution was done under contract K W A 3601 from the Federal Minister of Research and Technology.
References [1] Pressemitteilung 8/85 des BMFT vom 22. Jan. 1985, Beschluss der Bundesregierung vom 23. Jan. 1985 zur abschliessenden Beurteilung der Anderen Entsorgungstechniken. [2] H. B~cher et al., Versuch zur Einlagerung von mittelaktivem Abfall und HTR-Brermelementen im Salzbergwerk ASSE, Kerntechnik 50, Nr. 1 (1987) 40-44. [3] B. Ganser et al., FuE-Arbeiten zur Endlagerung von HTRBrennelementen, Statusseminar HochtemperaturreaktorBrennstoffkreislauf, KFA-Jiil Conf-61 (1987) 147-160. [4] P. Brennecke und E. Warnecke, Anforderungen an endzulagernde radioaktive Abf~llle (Vod~iufige Endlagerungsbedingungen, Stand November 1986), PTB-Bericht SE-16, Braunschweig (1987). [5] Sicherheitskriterien for die Endlagerung radioaktiver Abf~Ulein einem Bergwerk, Bundesanzeiger 35, Nr. 2 (1983) 45-46. [6] "Modellans'~tze und Ergebnisse zur Freisetzung yon Radionukliden aus einem Modell-Salzstock", Projekt "Sicherheitsstudien Entsorgung", Abschlussbericht PSEII, Fachband 16 (Januar 1985). [7] NUKEM, priv. communication. [8] G. Stolba, G. Reitsamer, G. Falta und W. Schertk, Chlorgaslaugung unbestrahlter und bestrahlter HTR-Brennelemente, Jahrestagung Kerntechnik des DAtF (1980) Berlin, Tagungsberichte, pp. 586-589. [9] R. Duwe and A. Rebmann, KFA internal report.