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Activation product transport in a FLiBe-vanadium alloy-HT9 system
261
Journal of Nuclear Materials 149 (1987) 261-265
North-Holland, Amsterdam
ACTIVATION PRODUCT TRANSPORT IN A FLiBe-VANADIUM ALLOY-W
SYSTEM *
A.C. KLEIN Department
of Nuclear Engineering,
Oregon State University, Corvailis, OR 97331, USA
D.-K. SZE Fusion Power Prograrq
Bldg. 205, Argonne National ~~ra~o~,
9700 S. Case Avenue, Argonne, IL 60439, USA
Received April 1986; accepted 2 February 1987
An assessment is made of the gamma radiation hazards likely to be found around a fusion reactor heat transfer and tritium breeding loop which employs a vanadium alloy for the blanket and first wall structure and the ferritic-steel HT9 for the remainder of the loop. The coolant/tritium breeding fluid is the molten metallic salt FLiBe. Since the radiation levels near the primary loop components are found to be less than 100 mR/h 3-5 days after shutdown after three years of continuous full power operation, limited hands-on maintenance could be allowed. The very short half-lives of the pr~o~n~t corrosion products make this result possible and make such a system very attractive.
1. Infmduction Radioactive products will be generated within the blankets of all D-T fusion reactors. A small fraction of these will be removed from the reactor by the processes of corrosion and neutron sputtering into the coolant and tritium breeding fluids. Additionally, material can be removed from the out-of-blanket surfaces by corrosion, deposited in the blanket, become radioactive, and then be released into the coolant. All of these products will then be transported throughout the primary system and can be deposited on out-of-blanket surfaces. The gamma radiation fields which result from these radioisotopes will have an impact on the operation and maintenance tasks to be performed on the primary loop. Such radiation fields could, if they were large enough, preclude direct hands-on rn~ten~~ and require that remote operations be performed on all of the primary coolant loop systems - including purification and tritium removal systems.
* Presented at the Second International Conference on Fusion Reactor Materials (ICFRM-2), Chicago, IL, USA, April 13-17,1986.
It has been shows previously that, for a number of reactor scenarios, the radiation levels around the primary systems of fusion reactors will be too high to allow hands-on maintenance to be performed, even after the fluid has been drained from the system [l]. In contrast, it is found that for a FLiBe-Vanadium alloy system, with adequate FLiBe chemistry control, limited handson contact could be allowed. This paper will describe the analysis of this reactor system and any effects on its operation and maintenance.
2. Prilwy loop description The primary loop is designed to handle between 500 and 600 MW (thermal) using FLiBe (47% LiF-53% BeF,) as the heat transfer material and can be used as a stand alone reactor system or be a part of a larger reactor system. Table 1 contains a number of useful parameters for this analysis. The first wall and blanket are divided into four sections. The temperature for the hot and c&d portions of the blanket and first wall sections are 550°C and 650 ’ C respectively. Approximately 25% of the primary coolant flow is directed through the “first wall’, or for
0022-3115/87/$03.50 0 Elsevier Science Publishers B.V. (North-Holland Physics Publishing Division)
262
A.C. Klein, D.-K. Sze / Actioation product transport
Table 1 Primary
loop components
Node number
Component name
1 2 3 4 5 6 7 8 9 IO 11 12
First walI Blanket First wall Blanket Outlet header Hot leg pipe Steam generator Steam generator Crossover pipe Pump Cold leg pipe Inlet header
Operating temperature
Surface area
(“C)
(cm’ )
550 550 650 650 650 650 650 550 550 550 550 550
1.4+0.06 5.6+0.06 1.4+0.06 5.6+0.06 3.4+0.05 4.7 + 0.05 1.5 + 0.06 7.5 + 0.06 7.8 + 0.04 1.1+0.04 4.7 + 0.05 3.4 + 0.05
this analysis, the high neutron flux region, and the remaining 75% is channeled through the “blanket”, or low neutron flux regin. This four-way split of the blanket is utilized to adequately treat (1) the increase in temperature through the blanket (and its associated corrosion release and deposition), and (2) the rapid decrease in neutron flux (and therefore activation rates) through the blanket. All of these first four components are assumed to be constructed of the vanadium based alloy V-lSCr-5Ti. The outlet header collects the flow from the blanket sections and also is constructed of V-15Cr-5Ti. The inlet header, in the Lc d portion of the loop, is the last remaining component made from V-ISCr-STi. The rest of the reactor loop is constructed of the ferritic steel HT-9. As can be seen in table 1, two steam generator sections are assumed, one for the hot and the other for the cold portions of this component. A few special notes must be mentioned here. First, no purification system has been included in this analysis. Such a system could be added later if the corrosion or deposition rates were found to lead to radiation levels too large to be tolerated, or to plugging of the steam generator tubes. Also, it is assumed that the tritium produced will diffuse through the vanadium first wall and be removed by the plasma gas handling system. Again, a tritium removal system could be included if it were necessary to provide additions tritium removaf capability.
3. Calculational
model
The RARTOR [2,3] computer code was modified for use in this analysis to include the gamma producing
Volume (cm’ ) 3.6+0.06 1.4+0.07 3.6 + 0.06 1.4+ 0.07 4.4 + 0.06 5.9+0.06 2.5 + 0.06 2.5 t 0.06 9.8 + 0.05 1.6 + 0.05 5.910.06 4.4 + 0.06
Pipe diameter
FL% rate
(cm)
(kg/h)
10.0 10.0 10.0 10.0 50.0 50.0 1.0 1.0 50.0 20.0 50.0 50.0
4.5 1.35 4.5 1.35 1.8 1.8 1.8 1.8 1.8 1.8 1.8 1.8
flow
+0.05 + 0.06 i-o.05 + 0.06 +0.06 +0.06 +0.06 +0.06 iO.06 i-O.06 i-O.06 +0.06
Material
V-alloy v-ahoy V-alloy V-alloy v-ahoy HT9 HT9 HT9 HT9 HT9 HT9 V-alloy
radioisotopes likely to be the cause of radiation hazards to operational and m~nten~ce personnel. Five additional reactions have been added to RAFTOR’s original 15 to treat the addition of vanadium and titanium reacting products. Table 2 lists all of the neutron reactions currently available in RAFTOR.
Table 2 Activation
products
Formation reaction
considered
Activation product
in RAPTOR
Half-life
5.26 yr 58Ni(n, p) “Co(n, 2n)
54Fe(n, P) “Mn(n,
2n)
=co 54Mn
71.3 d
303 d
Gamma energies (MeV) 1.173, 1.332 0.511, 0.81, 0.865, 1.67 0.835
50Cr(n,Y) 52Cr(n, 2n) 54Fe(n, a) i
s’Cr
27.8 d
0.32
55Mn(n, Y) “Fe&, yj’
56Mn
2.6 h 45.6 d
0.847,1.811, 2.11 0.143,0.192 1.095, 1.292 0.368, 1.115, 1.481 0.264,0.6845,1.479 0.181,0.372,0.74,0.78 0.889, 1.12 0.16
59Co(n, P)
>
s9Fe
MNi(n, Y)
=Ni
92Mo(n, y) 98Mo(n, y) 41iTi(n, p) 47Ti(n, P) 48Ti(n, p) s’V(n, a) I
Q3Mo *MO %C 47sc
2.6 6.9 66.7 83.8 3.3
‘%c
43.7 h
s’Ti(n,
47Ca
a)
h h h d d
4.54 d
0.983, 1.04, 1.314 0.49, 0.185, 1.297
A. C. Klein, D.-K. Sze /Activation
Since the primary objective of this analysis is to determine the gamma radiation field surrounding the primary loop, no attempt has been made to estimate any degradation of system integrity due to wastage of the blanket tubes or plugging of the steam generator by the depositing corrosion products. It is felt that with adequate chemistry control, the corrosion rates can be kept low enough that the system integrity can be maintained. The corrosion removal rate which is used for this analysis is set at 0.5 gm/yr for all surfaces in contact with the FLiBe. It has been shown that the addition of beryllium metal to the salt allows a corrosion release rate of 0.5 pm/yr for 316 stainless steel at 650 o C [4]. This rate of corrosion release has been assumed for all of the surfaces of the loop. including the vanadium alloy. Further experimental data is needed to substantiate these assumptions and should be obtained as soon as possible. The data in table 1 is used as input into the RAPTOR code. Corrosion release rates, transport and deposition coefficients, erosion coefficients, decay constants and transmutation rates are then determined for each of the components of the reactor system shown in table 1. After solving the resulting set of differential equations, RAPTOR determines the specific deposits of the radioactive nuclei for each loop component and the specific activity for each radioisotope in the coolant/tritium breeding material. These results are then utilized to determine the radiation field at a particular point due to the gamma decay of the radioisotopes which remain attached to the walIs of a cylindrical pipe. A point-kernal method for radiation transport is utilized and allows for varying pipe thickness and distance from the surface of the pipe.
4. Results The gamma dose rates near the primary system during operation consist of three components. The first includes the self-activation of the coolant/tritium breeding material itself. The second involves the radioactive corrosion and neutron sputtering products which are entrained in the fluid and are carried throughout the primary system. The third component of the radiation field involves the activation products which are attached to the component wails as a result of deposition. Each of these will contribute to the radiation levels in the reactor building during operation. However, personnel access to the reactor buiIding and steam generation will probably not be possible during operation for any fu-
263
product transport
sion reactor system and thus these ~ncerns
are mini-
mized.
Table 3 lists the major gamma producing radionuclides, their half-lives, the gammas which are produced, and their equilibrium concentration in the blanket during operation due to fluorine activation [5]. The radiation level around the primary system can be estimated using the point kemal method for a cylindrical volume source with a slab shield [6]. Since each of the radioisotopes in table 3 has a relatively short half-life, a lo-second delay time is assumed for the FLiBe in the blanket to be transported to the hot leg pipe. The most penetrating, and also most available gamma rays produced in the hot leg pipe are the 7.11 MeV and 6.13 MeV gammas resulting from the decay of “N. These provide a radiation field in contact with the hot leg pipe (pipe radius = 25 cm, pipe thickness = 6 cm) of over 4 X lo5 R/h. This radiation level by far exceeds any limit for allowance of hands-on maintenance, or even access into the reactor hall during operation. This radiation level could easily be reduced by either adding copious amounts of shielding or by lengthening the amount of time it takes for the FLiBe to reach the hot leg after it leave the blanket; however, eliminating access to the reactor ball during operation is likely to be much simpler. After shutdown the three components of the radiation field around the primary loop will remain. Since the self-activation products in the FLiBe have such short half-lives, this component can easily be neglected after about an hour. Since the minimum liquidois temperature of FLiBe is relatively high (363*(Z). It can be assumed that the FLiBe will be drained from the loop within a day or two after shutdown rather than attempt to maintain the high temperatures necessary to keep the FLiBe from freezing. Therefore, only the deposited activation products will contribute to the radiation fields during the time when m~ten~ce and repair tasks are being performed. Fig. 1 shows the buildup of the radiation level at a
Table 3 Primary self-activation products producing penetrating gamma rays in a FLiBe coolant Radionuclide
Blanket concentration (Ci/=)
Half-life Gamma energies (s) (Mev)
‘6N
4.13 1.33 0.30
7.13 26.9
7.11,6.13 1.4gl.37
11
1.63
‘90
*OF
264
A.C. Klein, D.-K. Sze /Activation FLIBE-VISCRSTI-HT9
SYSTEM
700 600
-
500
-
HOT
,i_
0
,
,
,
,
5
IO
I5
20
LEG PIPE
Fijx+yj 25
30
35
40
TIME-MONTHS
Fig. 1. Buildup of radiation level in contact with hot leg pipe.
point in contact with the hot leg pipe of the described system. After three years of full time operation the contact gamma radiation field is almost 800 mR/h at shutdown and will preclude hands-on maintenance. However, due to the relatively short lived nature of the dominant gamma decaying radioisotopes, the radiation level decays away rapidly as seen in fig. 2. Even after three full years of operation the dose rate decays to around 100 mR/h after 1 week of shutdown. These dose rates are calculated for a 6 cm pipe thickness. It is important to note the major contributing isotopes to these radiation levels. At shutdown and during the first few days after shutdown, the radiation level is due primarily to 48Sc and 56Mn. The 48Sc is produced in the blanket structural alloy in nearly equal amounts by the 48Ti (n, P)~*SC and “V(n, CX)~~SCreactions. This is then released into the coolant by corrosion to be
FLIBE-Vl5CR5TI-HT9
,000 0 3
6 TIME
9 AFTER
SYSTEM
I2
15
16
21
SHUTDOWN-DAYS
Fig. 2. Decay of contact dose rate as a function of time after shutdown for three operating times for the hot leg pipe.
product transport
deposited in the system. The s6Mn is produced by the 56Fe(n, P)~~MvI~reaction in iron which has been released from the HT9 in the reactor system and has been deposited in the blanket. The 56Mn is then released into the coolant by erosion to be deposited in the primary loop components. After the 56Mn and 48Sc decay away the dominant radioisotopes are &SC and 54Mn. The %c is produced by the &Ti(n, p)%c reaction in the blanket structural material and the 54Mn is produced in a similar manner as the 56Mn by the 54Fe(n, p)54Mn reaction. It can be observed that the isotopes produced by the transport of out-of-blanket material into the neutron flux in the blanket (such as 56Mn, 54Mn, and 59Fe) continue to increase in importance with extended operation. In fact it is the continued buildup of these isotopes in the primary loop which causes the continued increase in the radiation levels seen in fig. 1. The time constants for saturation for these radioactive products is significantly longer than the decay constant saturation exhibited by the in-blanket originating radioisotopes. This leads to the conclusion that the radiation levels will continue to increase throughout the life of the plant. Such behavior is also seen in present light water reactor systems [7].
5. Conclusions Results have been presented which indicate that a fusion reactor blanket system composed of a vanadium alloy structure with the molten salt FLiBe as the tritium breeding and coolant material will be very attractive from a corrosion product transport viewpoint. Due to the low corrosion release rates which may be possible and to the dominance of short-lived gamma producing radioisotopes, hands-on maintenance and repair operations are conceivable within a few days after shutdown. The radioactive corrosion products and the selfactivation of the FLiBe itself will preclude access to the primary system during operation; however, this is standard operating procedure for even present day light water reactors and should pose little problem for such a fusion reactor. Since FLiBe has a very high melting point, it can be assumed that within a day or two after shutdown the FLiBe will be drained from the primary system rather than attempt to maintain the high temperatures necessary to keep the FLiBe from freezing. Thus, only the deposited activation products will contribute to the dose to maintenance personnel. After about three years of continuous operation, the radiation levels at shutdown will be approximately 800 mR/h, and will decay to less
AC. Hein, iX-K. Sre / Actiuation product transport
than 100 mR/h within 5 days. At this level a limited amount of hands-on maintenance could be possible due to the short half-lives of the predominant radioactive corrosion products in the loop. One uncertainty in this analysis is the assumed value for the corrosion product release rate for vanadium alloys in contact with flowing FLiBe. If the release rates are found to be considerably higher than 0.5 pm/yr, then the above conclusions may no longer be valid; however, the addition of a purification system to the loop, an eventually to maintain adequate FLiBe chemistry control anyway, will reduce the amounts of material deposited in the loop. This could be used to mitigate any potential consequences of enhanced corrosion produce release, and also ensure the original conclusions.
A~wl~ment This work was supported by the US Department Energy, Office of Fusion Energy.
of
265
References VI A.C. Klein and W.F. Vogelsang, Nucl. Engrg. Des/Fusion 2 (1985) 355-363. 121 AC. Klein and W.F. Vogelsang, RAPTOR: a computer code to calculate the transport of activation products in fusion reactors, UWFDM-567, University of Wisconsin (Febr. 1984). [31 A.C. Klein, Activation product Transport in Fusion Reactor, Thesis, University of Wisconsin (Aug. 1983). [41 J.R. Keiser, J.D. DeVan and E.J. Lawrence, in: Proc. First ‘Topical Meeting on Fusion Reactor Materials, Miami Beach, FL, Jan. 29-31, 1979, p. 295, PI E.T. Cheng, Personal communication, July 1985. PI A. Foderaro, Photon Shielding Manual, Pennsylvania State University (April 1978). 171 W.E. Berry and R.B. Diegle, Survey of corrosion product generation, transport and deposition in light water nuclear reactors, EPRI NP-522 (Electric Power Research Institute, Palo Alto, CA, March, 1979).
Journal of Nuclear Materials 149 (1987) 261-265
North-Holland, Amsterdam
ACTIVATION PRODUCT TRANSPORT IN A FLiBe-VANADIUM ALLOY-W
SYSTEM *
A.C. KLEIN Department
of Nuclear Engineering,
Oregon State University, Corvailis, OR 97331, USA
D.-K. SZE Fusion Power Prograrq
Bldg. 205, Argonne National ~~ra~o~,
9700 S. Case Avenue, Argonne, IL 60439, USA
Received April 1986; accepted 2 February 1987
An assessment is made of the gamma radiation hazards likely to be found around a fusion reactor heat transfer and tritium breeding loop which employs a vanadium alloy for the blanket and first wall structure and the ferritic-steel HT9 for the remainder of the loop. The coolant/tritium breeding fluid is the molten metallic salt FLiBe. Since the radiation levels near the primary loop components are found to be less than 100 mR/h 3-5 days after shutdown after three years of continuous full power operation, limited hands-on maintenance could be allowed. The very short half-lives of the pr~o~n~t corrosion products make this result possible and make such a system very attractive.
1. Infmduction Radioactive products will be generated within the blankets of all D-T fusion reactors. A small fraction of these will be removed from the reactor by the processes of corrosion and neutron sputtering into the coolant and tritium breeding fluids. Additionally, material can be removed from the out-of-blanket surfaces by corrosion, deposited in the blanket, become radioactive, and then be released into the coolant. All of these products will then be transported throughout the primary system and can be deposited on out-of-blanket surfaces. The gamma radiation fields which result from these radioisotopes will have an impact on the operation and maintenance tasks to be performed on the primary loop. Such radiation fields could, if they were large enough, preclude direct hands-on rn~ten~~ and require that remote operations be performed on all of the primary coolant loop systems - including purification and tritium removal systems.
* Presented at the Second International Conference on Fusion Reactor Materials (ICFRM-2), Chicago, IL, USA, April 13-17,1986.
It has been shows previously that, for a number of reactor scenarios, the radiation levels around the primary systems of fusion reactors will be too high to allow hands-on maintenance to be performed, even after the fluid has been drained from the system [l]. In contrast, it is found that for a FLiBe-Vanadium alloy system, with adequate FLiBe chemistry control, limited handson contact could be allowed. This paper will describe the analysis of this reactor system and any effects on its operation and maintenance.
2. Prilwy loop description The primary loop is designed to handle between 500 and 600 MW (thermal) using FLiBe (47% LiF-53% BeF,) as the heat transfer material and can be used as a stand alone reactor system or be a part of a larger reactor system. Table 1 contains a number of useful parameters for this analysis. The first wall and blanket are divided into four sections. The temperature for the hot and c&d portions of the blanket and first wall sections are 550°C and 650 ’ C respectively. Approximately 25% of the primary coolant flow is directed through the “first wall’, or for
0022-3115/87/$03.50 0 Elsevier Science Publishers B.V. (North-Holland Physics Publishing Division)
262
A.C. Klein, D.-K. Sze / Actioation product transport
Table 1 Primary
loop components
Node number
Component name
1 2 3 4 5 6 7 8 9 IO 11 12
First walI Blanket First wall Blanket Outlet header Hot leg pipe Steam generator Steam generator Crossover pipe Pump Cold leg pipe Inlet header
Operating temperature
Surface area
(“C)
(cm’ )
550 550 650 650 650 650 650 550 550 550 550 550
1.4+0.06 5.6+0.06 1.4+0.06 5.6+0.06 3.4+0.05 4.7 + 0.05 1.5 + 0.06 7.5 + 0.06 7.8 + 0.04 1.1+0.04 4.7 + 0.05 3.4 + 0.05
this analysis, the high neutron flux region, and the remaining 75% is channeled through the “blanket”, or low neutron flux regin. This four-way split of the blanket is utilized to adequately treat (1) the increase in temperature through the blanket (and its associated corrosion release and deposition), and (2) the rapid decrease in neutron flux (and therefore activation rates) through the blanket. All of these first four components are assumed to be constructed of the vanadium based alloy V-lSCr-5Ti. The outlet header collects the flow from the blanket sections and also is constructed of V-15Cr-5Ti. The inlet header, in the Lc d portion of the loop, is the last remaining component made from V-ISCr-STi. The rest of the reactor loop is constructed of the ferritic steel HT-9. As can be seen in table 1, two steam generator sections are assumed, one for the hot and the other for the cold portions of this component. A few special notes must be mentioned here. First, no purification system has been included in this analysis. Such a system could be added later if the corrosion or deposition rates were found to lead to radiation levels too large to be tolerated, or to plugging of the steam generator tubes. Also, it is assumed that the tritium produced will diffuse through the vanadium first wall and be removed by the plasma gas handling system. Again, a tritium removal system could be included if it were necessary to provide additions tritium removaf capability.
3. Calculational
model
The RARTOR [2,3] computer code was modified for use in this analysis to include the gamma producing
Volume (cm’ ) 3.6+0.06 1.4+0.07 3.6 + 0.06 1.4+ 0.07 4.4 + 0.06 5.9+0.06 2.5 + 0.06 2.5 t 0.06 9.8 + 0.05 1.6 + 0.05 5.910.06 4.4 + 0.06
Pipe diameter
FL% rate
(cm)
(kg/h)
10.0 10.0 10.0 10.0 50.0 50.0 1.0 1.0 50.0 20.0 50.0 50.0
4.5 1.35 4.5 1.35 1.8 1.8 1.8 1.8 1.8 1.8 1.8 1.8
flow
+0.05 + 0.06 i-o.05 + 0.06 +0.06 +0.06 +0.06 +0.06 iO.06 i-O.06 i-O.06 +0.06
Material
V-alloy v-ahoy V-alloy V-alloy v-ahoy HT9 HT9 HT9 HT9 HT9 HT9 V-alloy
radioisotopes likely to be the cause of radiation hazards to operational and m~nten~ce personnel. Five additional reactions have been added to RAFTOR’s original 15 to treat the addition of vanadium and titanium reacting products. Table 2 lists all of the neutron reactions currently available in RAFTOR.
Table 2 Activation
products
Formation reaction
considered
Activation product
in RAPTOR
Half-life
5.26 yr 58Ni(n, p) “Co(n, 2n)
54Fe(n, P) “Mn(n,
2n)
=co 54Mn
71.3 d
303 d
Gamma energies (MeV) 1.173, 1.332 0.511, 0.81, 0.865, 1.67 0.835
50Cr(n,Y) 52Cr(n, 2n) 54Fe(n, a) i
s’Cr
27.8 d
0.32
55Mn(n, Y) “Fe&, yj’
56Mn
2.6 h 45.6 d
0.847,1.811, 2.11 0.143,0.192 1.095, 1.292 0.368, 1.115, 1.481 0.264,0.6845,1.479 0.181,0.372,0.74,0.78 0.889, 1.12 0.16
59Co(n, P)
>
s9Fe
MNi(n, Y)
=Ni
92Mo(n, y) 98Mo(n, y) 41iTi(n, p) 47Ti(n, P) 48Ti(n, p) s’V(n, a) I
Q3Mo *MO %C 47sc
2.6 6.9 66.7 83.8 3.3
‘%c
43.7 h
s’Ti(n,
47Ca
a)
h h h d d
4.54 d
0.983, 1.04, 1.314 0.49, 0.185, 1.297
A. C. Klein, D.-K. Sze /Activation
Since the primary objective of this analysis is to determine the gamma radiation field surrounding the primary loop, no attempt has been made to estimate any degradation of system integrity due to wastage of the blanket tubes or plugging of the steam generator by the depositing corrosion products. It is felt that with adequate chemistry control, the corrosion rates can be kept low enough that the system integrity can be maintained. The corrosion removal rate which is used for this analysis is set at 0.5 gm/yr for all surfaces in contact with the FLiBe. It has been shown that the addition of beryllium metal to the salt allows a corrosion release rate of 0.5 pm/yr for 316 stainless steel at 650 o C [4]. This rate of corrosion release has been assumed for all of the surfaces of the loop. including the vanadium alloy. Further experimental data is needed to substantiate these assumptions and should be obtained as soon as possible. The data in table 1 is used as input into the RAPTOR code. Corrosion release rates, transport and deposition coefficients, erosion coefficients, decay constants and transmutation rates are then determined for each of the components of the reactor system shown in table 1. After solving the resulting set of differential equations, RAPTOR determines the specific deposits of the radioactive nuclei for each loop component and the specific activity for each radioisotope in the coolant/tritium breeding material. These results are then utilized to determine the radiation field at a particular point due to the gamma decay of the radioisotopes which remain attached to the walIs of a cylindrical pipe. A point-kernal method for radiation transport is utilized and allows for varying pipe thickness and distance from the surface of the pipe.
4. Results The gamma dose rates near the primary system during operation consist of three components. The first includes the self-activation of the coolant/tritium breeding material itself. The second involves the radioactive corrosion and neutron sputtering products which are entrained in the fluid and are carried throughout the primary system. The third component of the radiation field involves the activation products which are attached to the component wails as a result of deposition. Each of these will contribute to the radiation levels in the reactor building during operation. However, personnel access to the reactor buiIding and steam generation will probably not be possible during operation for any fu-
263
product transport
sion reactor system and thus these ~ncerns
are mini-
mized.
Table 3 lists the major gamma producing radionuclides, their half-lives, the gammas which are produced, and their equilibrium concentration in the blanket during operation due to fluorine activation [5]. The radiation level around the primary system can be estimated using the point kemal method for a cylindrical volume source with a slab shield [6]. Since each of the radioisotopes in table 3 has a relatively short half-life, a lo-second delay time is assumed for the FLiBe in the blanket to be transported to the hot leg pipe. The most penetrating, and also most available gamma rays produced in the hot leg pipe are the 7.11 MeV and 6.13 MeV gammas resulting from the decay of “N. These provide a radiation field in contact with the hot leg pipe (pipe radius = 25 cm, pipe thickness = 6 cm) of over 4 X lo5 R/h. This radiation level by far exceeds any limit for allowance of hands-on maintenance, or even access into the reactor hall during operation. This radiation level could easily be reduced by either adding copious amounts of shielding or by lengthening the amount of time it takes for the FLiBe to reach the hot leg after it leave the blanket; however, eliminating access to the reactor ball during operation is likely to be much simpler. After shutdown the three components of the radiation field around the primary loop will remain. Since the self-activation products in the FLiBe have such short half-lives, this component can easily be neglected after about an hour. Since the minimum liquidois temperature of FLiBe is relatively high (363*(Z). It can be assumed that the FLiBe will be drained from the loop within a day or two after shutdown rather than attempt to maintain the high temperatures necessary to keep the FLiBe from freezing. Therefore, only the deposited activation products will contribute to the radiation fields during the time when m~ten~ce and repair tasks are being performed. Fig. 1 shows the buildup of the radiation level at a
Table 3 Primary self-activation products producing penetrating gamma rays in a FLiBe coolant Radionuclide
Blanket concentration (Ci/=)
Half-life Gamma energies (s) (Mev)
‘6N
4.13 1.33 0.30
7.13 26.9
7.11,6.13 1.4gl.37
11
1.63
‘90
*OF
264
A.C. Klein, D.-K. Sze /Activation FLIBE-VISCRSTI-HT9
SYSTEM
700 600
-
500
-
HOT
,i_
0
,
,
,
,
5
IO
I5
20
LEG PIPE
Fijx+yj 25
30
35
40
TIME-MONTHS
Fig. 1. Buildup of radiation level in contact with hot leg pipe.
point in contact with the hot leg pipe of the described system. After three years of full time operation the contact gamma radiation field is almost 800 mR/h at shutdown and will preclude hands-on maintenance. However, due to the relatively short lived nature of the dominant gamma decaying radioisotopes, the radiation level decays away rapidly as seen in fig. 2. Even after three full years of operation the dose rate decays to around 100 mR/h after 1 week of shutdown. These dose rates are calculated for a 6 cm pipe thickness. It is important to note the major contributing isotopes to these radiation levels. At shutdown and during the first few days after shutdown, the radiation level is due primarily to 48Sc and 56Mn. The 48Sc is produced in the blanket structural alloy in nearly equal amounts by the 48Ti (n, P)~*SC and “V(n, CX)~~SCreactions. This is then released into the coolant by corrosion to be
FLIBE-Vl5CR5TI-HT9
,000 0 3
6 TIME
9 AFTER
SYSTEM
I2
15
16
21
SHUTDOWN-DAYS
Fig. 2. Decay of contact dose rate as a function of time after shutdown for three operating times for the hot leg pipe.
product transport
deposited in the system. The s6Mn is produced by the 56Fe(n, P)~~MvI~reaction in iron which has been released from the HT9 in the reactor system and has been deposited in the blanket. The 56Mn is then released into the coolant by erosion to be deposited in the primary loop components. After the 56Mn and 48Sc decay away the dominant radioisotopes are &SC and 54Mn. The %c is produced by the &Ti(n, p)%c reaction in the blanket structural material and the 54Mn is produced in a similar manner as the 56Mn by the 54Fe(n, p)54Mn reaction. It can be observed that the isotopes produced by the transport of out-of-blanket material into the neutron flux in the blanket (such as 56Mn, 54Mn, and 59Fe) continue to increase in importance with extended operation. In fact it is the continued buildup of these isotopes in the primary loop which causes the continued increase in the radiation levels seen in fig. 1. The time constants for saturation for these radioactive products is significantly longer than the decay constant saturation exhibited by the in-blanket originating radioisotopes. This leads to the conclusion that the radiation levels will continue to increase throughout the life of the plant. Such behavior is also seen in present light water reactor systems [7].
5. Conclusions Results have been presented which indicate that a fusion reactor blanket system composed of a vanadium alloy structure with the molten salt FLiBe as the tritium breeding and coolant material will be very attractive from a corrosion product transport viewpoint. Due to the low corrosion release rates which may be possible and to the dominance of short-lived gamma producing radioisotopes, hands-on maintenance and repair operations are conceivable within a few days after shutdown. The radioactive corrosion products and the selfactivation of the FLiBe itself will preclude access to the primary system during operation; however, this is standard operating procedure for even present day light water reactors and should pose little problem for such a fusion reactor. Since FLiBe has a very high melting point, it can be assumed that within a day or two after shutdown the FLiBe will be drained from the primary system rather than attempt to maintain the high temperatures necessary to keep the FLiBe from freezing. Thus, only the deposited activation products will contribute to the dose to maintenance personnel. After about three years of continuous operation, the radiation levels at shutdown will be approximately 800 mR/h, and will decay to less
AC. Hein, iX-K. Sre / Actiuation product transport
than 100 mR/h within 5 days. At this level a limited amount of hands-on maintenance could be possible due to the short half-lives of the predominant radioactive corrosion products in the loop. One uncertainty in this analysis is the assumed value for the corrosion product release rate for vanadium alloys in contact with flowing FLiBe. If the release rates are found to be considerably higher than 0.5 pm/yr, then the above conclusions may no longer be valid; however, the addition of a purification system to the loop, an eventually to maintain adequate FLiBe chemistry control anyway, will reduce the amounts of material deposited in the loop. This could be used to mitigate any potential consequences of enhanced corrosion produce release, and also ensure the original conclusions.
A~wl~ment This work was supported by the US Department Energy, Office of Fusion Energy.
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References VI A.C. Klein and W.F. Vogelsang, Nucl. Engrg. Des/Fusion 2 (1985) 355-363. 121 AC. Klein and W.F. Vogelsang, RAPTOR: a computer code to calculate the transport of activation products in fusion reactors, UWFDM-567, University of Wisconsin (Febr. 1984). [31 A.C. Klein, Activation product Transport in Fusion Reactor, Thesis, University of Wisconsin (Aug. 1983). [41 J.R. Keiser, J.D. DeVan and E.J. Lawrence, in: Proc. First ‘Topical Meeting on Fusion Reactor Materials, Miami Beach, FL, Jan. 29-31, 1979, p. 295, PI E.T. Cheng, Personal communication, July 1985. PI A. Foderaro, Photon Shielding Manual, Pennsylvania State University (April 1978). 171 W.E. Berry and R.B. Diegle, Survey of corrosion product generation, transport and deposition in light water nuclear reactors, EPRI NP-522 (Electric Power Research Institute, Palo Alto, CA, March, 1979).