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Activation and waste management considerations of fusion materials
EJ.vSEViER
Journal af Nuclear Materials 212-215 (1994) 621-627
Activation and waste management considerations of fusion materials E.T. Cheng a, G. Saji b ’ TSI Research, Inc., 225 StevensAvenue, Soiana Beach, CA 92075, USA b ITER San Diego Go-Center, 11025 N. Torrey PinesRoad, La Jo&z, CA 92037, USA
Abstract Inconel-625 (Ni625), SS316, Ti-6A.L4V (T&4), ferritic steel (IS), reduced activity ferritic steel (RAFS), manganese steel (Mn-steel), and VSCrSTi (V5.51, were examined for a near-term experimental D-T fueled fusion power reactor with respect to waste management. Activation calculations for these materials were performed assuming one year continuous operation at I MW/m’ wall loading. The results show that the blanket com~~ents made of V55, T&l, Mn-steel, and FS will be allowed for transfer to an on-site dry storage facility after 10 years of cooling after discharge. To transport the discharged blanket components to a permanent disposal site, the cooling time needed can be within 10 years for Ti64 and V.55, provided that the impurities (mainly Ni, Nb and MO) be controlled to an acceptable level. The RAPS and Mn-steel wiil need about 30 y cooling time because of its Fe and Mn contents. Ni625,316SS, and FS, however, will require more than 5~~ y cooling time because of their Nb and MO contents. The RAFS, Mn-steel, Ti64 and V55 can be sh~low-land wastes if the irnpu~~ level for Nb and MO is dropped below 10 ppm. 1. lntraduction As fusion energy development approaches the construction of a deuterium-t~tium (D-T) fueled experimental power reactor, it becomes urgently important that the management scenario of decommissioned reactor components is determined prior to the selection of reactor materials. The waste management scenario should be as practical as possible and should take into ~nsideration the characteristic of activated reactor ~~~~ents. For the near-team experimental power reactor, the important issues relevant to waste management are mainly centered on the handling of decommissioned components. These issues are: (1) on site storage of discharged reactor ~mponents; (2) t~a~~rtation of these components to the permanent storage site; and (3) permanent disposal. On site storage is needed to allow the induced radioactivity to decay. During the storage period, the decay heat has to be properly removed to allow the decommissioned components to maint~n below an acceptable temperature. When the disch~~ed components are cooled off to the extent
that the decay power density and contact dose rate are lower than certain specified limits, transportation of these ~rn~o~~nts to a pe~anent storage site can be facilitated. At the permanent storage site, waste management option, either permanent underground burial disposal or recycling of reusable materials, can then be exercised immediately or after further cooling. In this paper, we explore the activation characteristics of several candidate structural materials in the first wall of a D-T expe~mental power reactor. The cooling times needed for these materials were examined against satisfying a set of predetermined limits on decay power density and contact dose rate for the transportation of the decommissioned blanket components. Comments on permanent waste management option of these components are also provided in this paper.
2. Quantitative limits for waste management A set of limits on activity related quantities was chosen as an attempt for the purpose of comparing the
00223115/94/$07.00 0 1994 Elsevier Science B.V. All rights reserved WX 0022-3 115(94)00185-Q
candidate structural materials. This set of limits should be revisited further in detail and revised for the practical application to licensing the construction of an experimental power reactor.
1.0tz+15 1.oftc14 l.OEt13
2.1. Decay power density
l.OE+12 l.OE+ll
A limit of 50 W/m3 was chosen based on the assumption that the temperature rise in a dry stack of waste containers with a nominal diameter of 5 m should not exceed 55°C [l]. Heat removal by natural convection is allowed and is probably the only mechanism. Volume averaging may be considered if sufficient heat conduction within the component is possible. At this decay power density, the discharged ~mponents can be transferred to an on-site dry storage facility where they can be made ready for transporting to the permanent storage site if the contact dose rate meets the required limit. 2.2. Corttactdose rate The legal contact dose rate limit to enable transportation is assumed to be 2 mSvfh, or 0.1 mSv/h at 1 m from the surface f2]. The contact dose rate should be the maximum value, without volume averaging. 2.3. Waste disposal there is no global consensus regarding the limit of radioactive quantities for low-level waste (lowlevel waste is presumably the goal of a majority of fusion energy development projects), two limits were adopted and examined here based on the study summarized in Ref. [3]. One of them is a limiting activity of 3 Ci/m3 (or about 0.5 Ci per metric ton of solid activated material). The other is a set of specific activity limits evaluated for shaliltow-land burial (Class C) waste disposal by USNRC lKFR61 fir] and supplemented by its equivalent, Fetter evaluations [S]. Since
l.OE+iO i.OE+OQ t.OE+OS l.OE+07 l.OE+06 1.ofc+05 i,QE+O4 l.cmO3 t.OE+OZ l.OE+ol l.OE+OO
Neutron
Energy
(eV)
Fig. 1. The energy-dependent neutron Box (spectrum1 at the first wall of a Na-cooled fusion blanket. The flux is normalized to 1 MW/m2 14 MeV neutron wail loading,
1 l&W/m*. As seen in Fig. 1, the neutron flux is steadily reducing as the energy is decreased. This is typical of an enhanced neutron absorbing medium because the lithium breeder is enriched, with 20% 6Li. Since only the first wall neutron flux will be analyzed to study the activation characteristics of different structural materials, the above model with FS structure should be adequate for a ~lf~~istent comparison. For the locations inside the blanket and shield, the flux calculations should be carried out for each structural material. 3.2. Neutron wail loading and fruence
3. Neutmnics and activation calculations 3.1. Geometry, blanket model, and material compositions The neutronics model is a one-dimensional cylinder with the plasma and scrape-off region in the center. It consists of a 5 mm first wall, a 0.5 m blanket, and a 0.5 m shield. The blanket and shield are based on ferritic steel as the structure and cooled by Na. The breeding material is liquid lithium. The blanket consists of 5% Na, 86% Li, and 5% structure, all by volume. The shield is primarily made of ferritic steel with 5% Na as the coolant. Fig. 1 depicts the energy-dependent spectrum of the neutron flux at the first wall normalized to
The neutron wall loading in all calculations was assumed to be 1 MW/m’, and the irradiation time was set to be one year continuously to reach a neutron fluence of 1 my/m* at the first wall. The neutron fluence is typical for an experimental reactor such as ITER 161.
Neutron flu calculations were performed using the one-dime~ion~ ~~rete~r~nates transport code, ANISN [71, with P3, and St3 approximations. The nuclear data library used is the MATXSS library processed at Los Alamos National Laboratory based on
E. T. Cheng, G. Suji/Jo~i
of Nuclear Materials212-215
Table 1 Elemental compositions of structural materials (in at%) SS316 C - 0.2837; Si - 1.3005; S - 0.0175; Cr - 17.3927; Fe - 65.9313; Nb - 0.0308; Ta - 0.006; N - 0.0411; P - 0.037; V - 0.1676; Mn - 0.1871; Ni - 13.1573; MO - 1.4413; W - 0.0061 Ni625 (Inconel625) C - 0.2487; Si - 0.5313; S - 0.014; Cr - 24.6704; Fe - 2.6708; Ni - 63.0325; MO - 5.5984; Al - 0.4425; P - 0.0145; Ti - 0.2493; Mn - 0.2715; Co - 0.5062; Nb - 1.1561; Ta - 0.5940 Ti64 (Ti-6AI-4V) Ti - 90.0; Al - 60, V - 4.0 Impurities (see * ) FeCrMn (manganese steel) C - 1.1329; Si - 0.4183; S - 0.0171; Cr - 12.0382; Fe - 65.6882; MO - 0.0006; W - 0.2976; N - 0.0202; P - 0.0363; V - 0.1644; Mn - 20.1707; Ni - 0.0096;
Ta - 0.0059 FS (ferritic steel - 9CrlMo) C - 0.4597; Si - 0.3108, S - 0.017; Cr - 9.594; Fe - 87.9395;Nb - 0.048, Ta - 0.003; N - 0.1999, P - 0.036; V - 0.2199; Mn - 0.3997; Ni - 0.1899, MO - 0.5797; W - 0.003 RAFS (reduced activation ferritic steel - 9CrWV) C - 0.4609, Si - 0.3116; S - 0.017; Cr - 9.6195; Fe - 88.173; MO - O.ooo6, W - 0.6012; N - 0.02; P - 0.0361; V - 0.2756; Mn - 0.4509; Ni - 0.0095; Ta - 0.0241
623
&J94~621-427
vanadium alloy WSCrSTi), the impurities derived from Ref. [lo] were used. Table 1 also shows these impurities and their levels. Note that these impurities were primarily those indentified based on the consideration for the generation of long-lived radioactivity, and should be adequate for the assessment of waste management issues.
5. Results
The results derived from the shutdown activities are summarized in Figs. 2-5. Observations over these results are given below: 5.1. Significance of impurities (and minor alloying elements)
For the illustration purpose, the shutdown activities, decay power densities, and contact dose rates were examined for the V5Cr5Ti alloy identifying the contributions from main alioying elements (V, Cr, and Tij, and impurities. It is clearly demonstrated in this investigation that the initial domination of the main alloying elements is effectively only to about 100 y for the shutdown activity and decay power, and to about 3 y for the contact dose rate. Thereafter, the impurities dominate. Fig. 2 depicts, as an example, the respective contact dose rates from the main alloying elements and impurities, as well as the total. If impurities are not
VSCrSTi V - 90.0; Cr - 5.0, Ti - 5.0 Impurities (see *I Impurities (ppm) for Ti64 and VSCrTi* ~-2~Ni-3;~u-5;Co-O.~~-S~Mo-l5~ Ag - 0.3; Cd - 0.2; Tb - 0.1; ELI- 0.1; Hf - 5.6; Dy - 0.3
1.
i.OE+03 f
.OEt02
1
.OE+Ol
1.
ENDF/B-V evaluations [8]. The activation calculations were completed by the REAC code and REAC*2 decay and cross section libraries [9].
ofi+
lsOE+04
DE+00
1 .OE-01 1 .OE-02 1 .OE-03 1 .OE-04
4. Structural
materials and their elemental composi-
tiOllS The alloying elements and impurities in each of the structural materials studied are shown in Table 1. For the cases of SS316, Inconel-625 (Ni625), maganese steel (FeCrMn), ferritic steel (FS: 9Cr), and reducedactivity ferritic steel (ILLS: 9CrWV), the compositions used include all minor and dominating impurity elements. For the titanium alloy (Ti64: Ti-6Al-4V) and
1 .OE-05 1 .OE-06
Time After Shutdown WVCrTi
*Vintp
*Total
(y)
V-alloy
Fig. 2. Comparison of residual contact dose rates in V5Cr5Ti due to the main alloying elements and impurities. The total contact dose rate is also shown.
624
E. T. Cheng, G. Saji /J~i~rnaL of Nuclear ~~ateria~ 212-215
1.OE+OS
l_OE+05
t .OE+Ol
l.OE+07
l.OE+03
ii
l.OE+06
d
0 -
1 .OE+05
_ .t;
l.OEt04
.?i
.c .
1.OE+02
2_
1 .OWOl
8
1
2
t;
i.OE+03 z
g ,ii
B
l.OE+02
z a
10
1
Time After
100
1000
Shutdown
+RAFS
-FS
-cTiS4
%-SS316
*FeCrMn
--VSCr5Ti
(y) +Ni625
DECAY POWER DENSITIES hGV/m2; ONE
YEAR
OPERATION)
l.OE+05 l.OE+04 l.OE+03 z
l.OE+02 1.OE+ol 1 .OE+OO
8
g a.
1
10
100
1000
10000
Time After Shutdown (y)
considered, the activity related quantities can be underestimated by more than two orders of magnitude when the cooling time is longer than 10 y. For VSCrSTi
(1
0.1
10000
Fig. 3. Comparison of residual activities (a/m31 in several candidate structural materials, namely Ni625, SS316, FS, RAFS, Ti64, FeCrMn, and VSCrSTi, after 1 y exposure at 1 MW/m2 neutron wah loading.
b
1 .OE-02
1 .OE-05 0.1
x
1.OE-01
1 .OE-04
1 . OE+OO
.t; ii
ofi+
i .OE-03 1 . OE+Ol
s
(1994) 621-627
-tRAFS
-FS
+Ti64
-W_SS316
*FeCrMn
eV5Cr5Ti
*NM25
Fig. 5. Comparison of residual contact dose rates in several structural materials, namely Ni625, SS316, FS, RAFS, Ti64, FeCrMn, and VSCrSTi after 1 y exposure at 1
candidate
MWfm2 neutron wall loading.
the decay power density level of 50 W/m3 may still be dominated by the main alloying elements since the needed cooling time to reach that level is within 10 y. However, the activity and contact dose rate at 1 Ci/m3, and 2 mSv/h level, respectively, are significantly influenced by the presence of impurities. The cooling time beyond that the impurities dominate, generally depends on the types of activity related quantities and on the kind of alloying elements, and impurities in the structural materials, and will vary for different structural materials. The impli~tion of this observation is that it is absoiutely required to include all minor and impurity elements in any assessment of management and long-term disposal of blanket and shield materials discharged from a D-T fusion reactor.
1 .OE-01
5.2. Residual activity
1 .OE-02 1 .OE-03 1 .OE-04
f3.1
10
1
100
1000
Time After Shutdown *RAFS
-FS
-5SS316 *FeC!-Mn
+Ti64
lOOOf
(y) *NH25
+V5CrtrTi
Fig. 4. Comparison of decay power densities (W/m3) in several candidate structural materials, namely Ni625, SS316, FS, RAFS, Ti64, FeCrMn, and VSCrSTi after 1 y exposure at 1 MW/m2 neutron wall loading.
Fig. 3 displays the induced activities from 1 to 1000 y after shutdown. At the first year after shutdown, the activities from RAFS, FS, Ni625, FeCrMn and SS316, are higher than Ti64 and VSCrSTi, and are around 1-2 x 10’ Ci/m3 (3.7-7.4 X 10' GBq,/m3) level. The activities from Ti64 and V5CrSTi alloys at 1 y after shutdown are 1.3 X lo5 and 4.4 X lo5 Ci/m3 (4.8 X lo6 and 1.6 x lo7 GBq/m3), respectively. The activities at 100 y after shutdown and beyond from SS316 and Ni625 are 1300 and 6600 Ci/m3 respectively, and are higher than FS, Ti64, VSCrSTi, FeCrMn, and RAFS by
E.T. Cheng, G. Saji/Joumal of Nuclear Materials 212-215 (1994) 621-627
one or two orders of magnitude. At 200 y after cooling, Ti64, VSCrSTi, FeCrMn, and RAFS, may have the specific activity lower than 3 Ci/m3 (111 GBq/m31. For Ni625, SS316, and FS, flux attenuation factors of 104, 103, and lo’, respectively, will be needed to reduce the specific activity below 3 Ci/m3 beyond 200 y after shutdown. 5.3. Decay power density Fig. 4 displays the post shutdown decay power densities. At the first year after shutdown, the decay power densities at the first wall range from 36 kW/m3 for Ni625, the highest, to 130 W/m3 for VSCrSTi, the lowest among the structural materials compared. At 10 y after shutdown, the decay power densities are reduced and are 3600, 720, 50, 0.27, 0.46, 41, and 52 W/m3, respectively, for Ni625, SS316, FS, VSCrSTi, Ti64, RAFS, and FeCrMn. At 100 y after shutdown, the decay power densities are reduced significantly further to 1.7, 0.19, 0.07, 0.006, 0.012, 0.002, and 0.002 W/m3, respectively, for Ni625, SS316, FS, VSCrSTi, Ti64, RAFS, and FeCrMn. It appears that all structural materials considered here with the present alloying and impurity specifications, except Ni625 and SS316, may satisfy the 50 W/m3 power density limit within 10 y of cooling times. SS316 and Ni625 will need more cooling times, 30 and 50 y, respectively, to satisfy the 50 W/m3 limit. 5.4. Contact dose rate The contact dose rates from the various structural materials are depicted for comparison in Fig. 5. At 1 y after shutdown, the contact dose rate from Ni625, 72000 Sv/h, is significantly higher than any alloy compared in this study because of its Ni content. It is followed by FeCrMn, 8200 Sv/h due to Mn. It is then followed by SS316 (2300 Sv/h), FS, and RAFS (1400 Sv/h each) because of their Fe content. The VSCrSTi alloy has the lowest contact dose rate at this time after shutdown, about 40 Sv/h. The Ti64 alloy has a contact dose rate of 580 Sv/h. At 10 y after shutdown, the contact dose rates are significantly reduced and are 940, 180, 5.6, 3.6, 1.1, 0.019, and 0.017 Sv/h, respectively, for Ni625, SS316, FeCrMn, FS, RAFS, Ti64, and VSCrSTi. Further reduction is expected as the cooling continues. At 100 y after shutdown, th contact dose rates are expected to be 0.23, 0.016, 0.01, 0.0029, 0.00105, 0.000014, and 0.000014 Sv/h, for Ni625, FS, SS316, Ti64, VSCrSTi, FeCrMn, and RAFS. To reach below 2 mSv/h, the cooling times required are 140 000, 60000, 46000, 20 000, 42, 32, and 30 y, respectively, for Ni625, FS, SS316, Ti64, RAFS, FeCrMn and VSCrSTi. Note that only VSCrSTi, FeCrMn, and RAFS will have
lower than 2 mSv/h 50 y after shutdown.
625
contact dose rates readily within
5.5. Waste disposal rating The lOCFR61 Class C waste disposal ratings were estimated to be 500, 24, 22, 5, 2, 0.1, and 0.07, respectively, for Ni625, FS, SS316, Ti64, VSCrSTi, RAFS, and FeCrMn. Note that only FeCrMn and RAFS will qualify as Class C shallow-land burial wastes. Ti64 and VSCrSTi are not Class C wastes unless the present impurity levels, primarily Nb and MO, are reduced, as discussed below.
6. Discussions A detailed analysis was made examining the dominating radionuclides relevant to waste management aspect for the structural materials compared in this paper. The contents of these contributing natural elements can be obtained in Table 1. Discussions are given below for each material. 6.1. Ni625 The dominating radionuclides are @?o (half-life 5.27 y) and Nb (half-life 20000 y). “Co is due to Ni (main alloying element - 63 at%) and Co (minor alloying element - 0.51 at%). 94Nb is due to Nb (minor alloying element - 1.16 at%) and MO (minor alloying element - 5.6 at%). It is thus difficult for Ni625 to reduce further the induced activity because the causing elements are either major or minor alloying elements.
6.2. SS316 The important radionuclides are also @?o and “Nb. They are due to the main alloying element Ni (for “Co), and minor element MO (for “Nb). The impurity element Nb (308 ppm in SS316) is also a parent element for “Nb. Like Ni625, SS316 is unlikely to reduce the long-term induced activity. 6.3. Ferritic steel - 9Cr (FS) The dominating radionuclides are 6oCo, 54Mn (half-life 312 d), and %Nb. Again, 6oCo is due to the main alloying element Ni, and ?Jb due to the minor alloying element MO. 54Mn is due to the main alloying element Fe (87.9 at%) and minor alloying element Mn (0.4 at%). It actually decays away beyond 10 y after shutdown due to its relatively shorter half-life. The
626
E.T. Cheng, G. S~ji/Jou~l
~~~~clear ~~ate~a~ 212-215
impurity element Nb (480 ppm) also contributes to the production of 94Nb. Because of Ni and MO, FS is also not an alloy capable of reducing the long-term activity. 6.4. FeCrMn The dominating radionuclides are s4Mn and 94Nb. “Nb is primarily due to the impurity MO (6 ppm). 54Mn, again is due to the two main alloying elements Mn (20.2 at%) and Fe (65.7 at%). It will decay away in a few more years. The manganese steel investigated here can be an attractive blanket material if the impurities can be controlled. 6.5. Ti64 The dominating radionuclides for the Ti-6Al-4V alloy are 6oCo, 26Al (half-life 740 000 y), and 94Nb. ‘?Jo and 94Nb are, again, due to the impurities Ni, and Nb and MO (Ni: 3 ppm; Nb: 50 ppm; and MO: 150 ppm). %A.l is induced in Al, a major alloying element. However, the 26Al activity itself is still lower than the desirable limit. Thus if the ~pu~ties can be improved further (about 5 ppm level for Nb and MO, and 0.1 ppm level for Ni), Ti64 should be an ideal alloy for the experimental D-T fusion reactor. After the impurity control, the cooling time to reach below 2 m&/h contact dose rate is only about 5 y, instead of 20000 y as estimated with higher level impurities.
6.6. RAFS (9CrWVl The dominating radionuclides are 54Mn, *Co, and 94Nb. Again, 54Mn is due to the major element Fe, and minor element Mn, and will decay away if additional cooling years are provided. ?Zo and “Nb are due to the impurities, Ni (95 ppm), and MO (6 ppm), respectively. The RAFS is an attractive structural material since further reduction in Ni and MO impurities can essentially be available. 6.7. V5CrSTi The dominating radionuclides are @)Co and 94Nb. These radionuclides, like those for RAFS, are primarily due to the impurities Ni (3 ppm), and Nb (50 ppm> and MO (150 ppm). These impurities can be reduced further as discussed in Ref. [lo]. The contact dose rate will drop below the 2 m&/h value at 10 y after shutdown if the Ni impurity level can be reduced to 0.1 ppm. Thus VSCrSTi, like Ti64 and RAFS, can essentially be very attractive if management of discharged reactor components needs to be done shortly after decommissioning.
(1994) 621-627
7. Conclusions We have investigated several candidate structural materials for a near-term D-T fueled experimental power reactor regarding the management of discharged reactor components. We found that the blanket components made of VSCrSTi, Ti-6Al-4V, manganese steel, ferritic steel (9Cr) and the reduced activity ferritic steel, can be transferred to an on-site dry storage facility within 10 y after discharge. For Inconel alloy and stainless steel, it will take longer cooling times, 50 and 30 y, respectively, to enable this procedure. To transport the decommissioned blanket components to a permanent storage site, the cooling time needed can also be within 10 y for Ti-6Al-4V and VSCrTi, provided the impurities (particularly Ni and Nb) are controlled to an acceptable level. Nickel was found to be the most important impurity element when the cooling time is between 10 and 30 y, because of the half-life (5.27 y) of its activation product 6oCo. To achieve the 10 y cooling time for transportation, the required Ni content is about 0.1 ppm. The reduced activation ferritic steel and m~ganese steel will need about 30 y cooling time because of its Fe and Mn contents. Inconel, stainless steel, and ferritic steel will require more than 50000 y cooling time because of their Nb and MO contents. Inconel, stainless steel, and ferritic steel will not qualify as lOCFR61 shallow-land burial wastes after 1 my/m* exposure, due to their Nb and MO contents. The titanium and vanadium alloys can be shallow-land burial wastes if the impurity level for Nb and MO is dropped to 10 ppm. The reduced activity ferritic steel and manganese steel will qualify as the shallow-land wastes because the impurity control is already taken into account in the development of these alloys.
References [l] D.C. Kocher and A.G. Croff, A Proposed Classification System for High-level and Other Radioactive Wastes,
ORNL/TM-10289, Oak Ridge National Laboratory (1987). [2] Code of Federal Regulations, Title 10 - Energy, lOCFR20, Standards for Protection Against Radiation, US Federal Register, Revised January 1, 1991. [3] J. Raeder et at., ITER Safety, ITER Documentation Series, NO. 36 (IAEA, Vienna, 19911, see VIII. 4. for details. 141 Code of Federal Regulations, Titie 10 - Energy, lOCFR61, Licensing Requirements for Land Disposal of Radioactive Waste, US Federal Register, revised January 1, 1991. [S] S. Fetter et al., Fusion Eng. Des. 13 (1990) 239. [6] ITER Conceptual Design Report, ITER Documentation Series, No. 18 (IAEA, Vienna, 1991).
[7] W.W. Engle, Jr., A User’s Manual for ANISN, A OneDimensional Discrete-Ordinates Transport Code with Auisotropic Scattering, K-1693, Oak Ridge Gaseous Diffusion Plant (1967). [8] R. MacFarlane, Los Alamos National Laboratory, private cxnnmunication (1983).
[9] F.M. Mann, REAC*2: Users Manual and Cede Description, Westinghouse Hanford Company Report, WHCEP-0282 (1989). [lo] E.T. Cheng et al., Fusion Technol. 21 (1992) 2001.
Journal af Nuclear Materials 212-215 (1994) 621-627
Activation and waste management considerations of fusion materials E.T. Cheng a, G. Saji b ’ TSI Research, Inc., 225 StevensAvenue, Soiana Beach, CA 92075, USA b ITER San Diego Go-Center, 11025 N. Torrey PinesRoad, La Jo&z, CA 92037, USA
Abstract Inconel-625 (Ni625), SS316, Ti-6A.L4V (T&4), ferritic steel (IS), reduced activity ferritic steel (RAFS), manganese steel (Mn-steel), and VSCrSTi (V5.51, were examined for a near-term experimental D-T fueled fusion power reactor with respect to waste management. Activation calculations for these materials were performed assuming one year continuous operation at I MW/m’ wall loading. The results show that the blanket com~~ents made of V55, T&l, Mn-steel, and FS will be allowed for transfer to an on-site dry storage facility after 10 years of cooling after discharge. To transport the discharged blanket components to a permanent disposal site, the cooling time needed can be within 10 years for Ti64 and V.55, provided that the impurities (mainly Ni, Nb and MO) be controlled to an acceptable level. The RAPS and Mn-steel wiil need about 30 y cooling time because of its Fe and Mn contents. Ni625,316SS, and FS, however, will require more than 5~~ y cooling time because of their Nb and MO contents. The RAFS, Mn-steel, Ti64 and V55 can be sh~low-land wastes if the irnpu~~ level for Nb and MO is dropped below 10 ppm. 1. lntraduction As fusion energy development approaches the construction of a deuterium-t~tium (D-T) fueled experimental power reactor, it becomes urgently important that the management scenario of decommissioned reactor components is determined prior to the selection of reactor materials. The waste management scenario should be as practical as possible and should take into ~nsideration the characteristic of activated reactor ~~~~ents. For the near-team experimental power reactor, the important issues relevant to waste management are mainly centered on the handling of decommissioned components. These issues are: (1) on site storage of discharged reactor ~mponents; (2) t~a~~rtation of these components to the permanent storage site; and (3) permanent disposal. On site storage is needed to allow the induced radioactivity to decay. During the storage period, the decay heat has to be properly removed to allow the decommissioned components to maint~n below an acceptable temperature. When the disch~~ed components are cooled off to the extent
that the decay power density and contact dose rate are lower than certain specified limits, transportation of these ~rn~o~~nts to a pe~anent storage site can be facilitated. At the permanent storage site, waste management option, either permanent underground burial disposal or recycling of reusable materials, can then be exercised immediately or after further cooling. In this paper, we explore the activation characteristics of several candidate structural materials in the first wall of a D-T expe~mental power reactor. The cooling times needed for these materials were examined against satisfying a set of predetermined limits on decay power density and contact dose rate for the transportation of the decommissioned blanket components. Comments on permanent waste management option of these components are also provided in this paper.
2. Quantitative limits for waste management A set of limits on activity related quantities was chosen as an attempt for the purpose of comparing the
00223115/94/$07.00 0 1994 Elsevier Science B.V. All rights reserved WX 0022-3 115(94)00185-Q
candidate structural materials. This set of limits should be revisited further in detail and revised for the practical application to licensing the construction of an experimental power reactor.
1.0tz+15 1.oftc14 l.OEt13
2.1. Decay power density
l.OE+12 l.OE+ll
A limit of 50 W/m3 was chosen based on the assumption that the temperature rise in a dry stack of waste containers with a nominal diameter of 5 m should not exceed 55°C [l]. Heat removal by natural convection is allowed and is probably the only mechanism. Volume averaging may be considered if sufficient heat conduction within the component is possible. At this decay power density, the discharged ~mponents can be transferred to an on-site dry storage facility where they can be made ready for transporting to the permanent storage site if the contact dose rate meets the required limit. 2.2. Corttactdose rate The legal contact dose rate limit to enable transportation is assumed to be 2 mSvfh, or 0.1 mSv/h at 1 m from the surface f2]. The contact dose rate should be the maximum value, without volume averaging. 2.3. Waste disposal there is no global consensus regarding the limit of radioactive quantities for low-level waste (lowlevel waste is presumably the goal of a majority of fusion energy development projects), two limits were adopted and examined here based on the study summarized in Ref. [3]. One of them is a limiting activity of 3 Ci/m3 (or about 0.5 Ci per metric ton of solid activated material). The other is a set of specific activity limits evaluated for shaliltow-land burial (Class C) waste disposal by USNRC lKFR61 fir] and supplemented by its equivalent, Fetter evaluations [S]. Since
l.OE+iO i.OE+OQ t.OE+OS l.OE+07 l.OE+06 1.ofc+05 i,QE+O4 l.cmO3 t.OE+OZ l.OE+ol l.OE+OO
Neutron
Energy
(eV)
Fig. 1. The energy-dependent neutron Box (spectrum1 at the first wall of a Na-cooled fusion blanket. The flux is normalized to 1 MW/m2 14 MeV neutron wail loading,
1 l&W/m*. As seen in Fig. 1, the neutron flux is steadily reducing as the energy is decreased. This is typical of an enhanced neutron absorbing medium because the lithium breeder is enriched, with 20% 6Li. Since only the first wall neutron flux will be analyzed to study the activation characteristics of different structural materials, the above model with FS structure should be adequate for a ~lf~~istent comparison. For the locations inside the blanket and shield, the flux calculations should be carried out for each structural material. 3.2. Neutron wail loading and fruence
3. Neutmnics and activation calculations 3.1. Geometry, blanket model, and material compositions The neutronics model is a one-dimensional cylinder with the plasma and scrape-off region in the center. It consists of a 5 mm first wall, a 0.5 m blanket, and a 0.5 m shield. The blanket and shield are based on ferritic steel as the structure and cooled by Na. The breeding material is liquid lithium. The blanket consists of 5% Na, 86% Li, and 5% structure, all by volume. The shield is primarily made of ferritic steel with 5% Na as the coolant. Fig. 1 depicts the energy-dependent spectrum of the neutron flux at the first wall normalized to
The neutron wall loading in all calculations was assumed to be 1 MW/m’, and the irradiation time was set to be one year continuously to reach a neutron fluence of 1 my/m* at the first wall. The neutron fluence is typical for an experimental reactor such as ITER 161.
Neutron flu calculations were performed using the one-dime~ion~ ~~rete~r~nates transport code, ANISN [71, with P3, and St3 approximations. The nuclear data library used is the MATXSS library processed at Los Alamos National Laboratory based on
E. T. Cheng, G. Suji/Jo~i
of Nuclear Materials212-215
Table 1 Elemental compositions of structural materials (in at%) SS316 C - 0.2837; Si - 1.3005; S - 0.0175; Cr - 17.3927; Fe - 65.9313; Nb - 0.0308; Ta - 0.006; N - 0.0411; P - 0.037; V - 0.1676; Mn - 0.1871; Ni - 13.1573; MO - 1.4413; W - 0.0061 Ni625 (Inconel625) C - 0.2487; Si - 0.5313; S - 0.014; Cr - 24.6704; Fe - 2.6708; Ni - 63.0325; MO - 5.5984; Al - 0.4425; P - 0.0145; Ti - 0.2493; Mn - 0.2715; Co - 0.5062; Nb - 1.1561; Ta - 0.5940 Ti64 (Ti-6AI-4V) Ti - 90.0; Al - 60, V - 4.0 Impurities (see * ) FeCrMn (manganese steel) C - 1.1329; Si - 0.4183; S - 0.0171; Cr - 12.0382; Fe - 65.6882; MO - 0.0006; W - 0.2976; N - 0.0202; P - 0.0363; V - 0.1644; Mn - 20.1707; Ni - 0.0096;
Ta - 0.0059 FS (ferritic steel - 9CrlMo) C - 0.4597; Si - 0.3108, S - 0.017; Cr - 9.594; Fe - 87.9395;Nb - 0.048, Ta - 0.003; N - 0.1999, P - 0.036; V - 0.2199; Mn - 0.3997; Ni - 0.1899, MO - 0.5797; W - 0.003 RAFS (reduced activation ferritic steel - 9CrWV) C - 0.4609, Si - 0.3116; S - 0.017; Cr - 9.6195; Fe - 88.173; MO - O.ooo6, W - 0.6012; N - 0.02; P - 0.0361; V - 0.2756; Mn - 0.4509; Ni - 0.0095; Ta - 0.0241
623
&J94~621-427
vanadium alloy WSCrSTi), the impurities derived from Ref. [lo] were used. Table 1 also shows these impurities and their levels. Note that these impurities were primarily those indentified based on the consideration for the generation of long-lived radioactivity, and should be adequate for the assessment of waste management issues.
5. Results
The results derived from the shutdown activities are summarized in Figs. 2-5. Observations over these results are given below: 5.1. Significance of impurities (and minor alloying elements)
For the illustration purpose, the shutdown activities, decay power densities, and contact dose rates were examined for the V5Cr5Ti alloy identifying the contributions from main alioying elements (V, Cr, and Tij, and impurities. It is clearly demonstrated in this investigation that the initial domination of the main alloying elements is effectively only to about 100 y for the shutdown activity and decay power, and to about 3 y for the contact dose rate. Thereafter, the impurities dominate. Fig. 2 depicts, as an example, the respective contact dose rates from the main alloying elements and impurities, as well as the total. If impurities are not
VSCrSTi V - 90.0; Cr - 5.0, Ti - 5.0 Impurities (see *I Impurities (ppm) for Ti64 and VSCrTi* ~-2~Ni-3;~u-5;Co-O.~~-S~Mo-l5~ Ag - 0.3; Cd - 0.2; Tb - 0.1; ELI- 0.1; Hf - 5.6; Dy - 0.3
1.
i.OE+03 f
.OEt02
1
.OE+Ol
1.
ENDF/B-V evaluations [8]. The activation calculations were completed by the REAC code and REAC*2 decay and cross section libraries [9].
ofi+
lsOE+04
DE+00
1 .OE-01 1 .OE-02 1 .OE-03 1 .OE-04
4. Structural
materials and their elemental composi-
tiOllS The alloying elements and impurities in each of the structural materials studied are shown in Table 1. For the cases of SS316, Inconel-625 (Ni625), maganese steel (FeCrMn), ferritic steel (FS: 9Cr), and reducedactivity ferritic steel (ILLS: 9CrWV), the compositions used include all minor and dominating impurity elements. For the titanium alloy (Ti64: Ti-6Al-4V) and
1 .OE-05 1 .OE-06
Time After Shutdown WVCrTi
*Vintp
*Total
(y)
V-alloy
Fig. 2. Comparison of residual contact dose rates in V5Cr5Ti due to the main alloying elements and impurities. The total contact dose rate is also shown.
624
E. T. Cheng, G. Saji /J~i~rnaL of Nuclear ~~ateria~ 212-215
1.OE+OS
l_OE+05
t .OE+Ol
l.OE+07
l.OE+03
ii
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d
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.?i
.c .
1.OE+02
2_
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8
1
2
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g ,ii
B
l.OE+02
z a
10
1
Time After
100
1000
Shutdown
+RAFS
-FS
-cTiS4
%-SS316
*FeCrMn
--VSCr5Ti
(y) +Ni625
DECAY POWER DENSITIES hGV/m2; ONE
YEAR
OPERATION)
l.OE+05 l.OE+04 l.OE+03 z
l.OE+02 1.OE+ol 1 .OE+OO
8
g a.
1
10
100
1000
10000
Time After Shutdown (y)
considered, the activity related quantities can be underestimated by more than two orders of magnitude when the cooling time is longer than 10 y. For VSCrSTi
(1
0.1
10000
Fig. 3. Comparison of residual activities (a/m31 in several candidate structural materials, namely Ni625, SS316, FS, RAFS, Ti64, FeCrMn, and VSCrSTi, after 1 y exposure at 1 MW/m2 neutron wah loading.
b
1 .OE-02
1 .OE-05 0.1
x
1.OE-01
1 .OE-04
1 . OE+OO
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s
(1994) 621-627
-tRAFS
-FS
+Ti64
-W_SS316
*FeCrMn
eV5Cr5Ti
*NM25
Fig. 5. Comparison of residual contact dose rates in several structural materials, namely Ni625, SS316, FS, RAFS, Ti64, FeCrMn, and VSCrSTi after 1 y exposure at 1
candidate
MWfm2 neutron wall loading.
the decay power density level of 50 W/m3 may still be dominated by the main alloying elements since the needed cooling time to reach that level is within 10 y. However, the activity and contact dose rate at 1 Ci/m3, and 2 mSv/h level, respectively, are significantly influenced by the presence of impurities. The cooling time beyond that the impurities dominate, generally depends on the types of activity related quantities and on the kind of alloying elements, and impurities in the structural materials, and will vary for different structural materials. The impli~tion of this observation is that it is absoiutely required to include all minor and impurity elements in any assessment of management and long-term disposal of blanket and shield materials discharged from a D-T fusion reactor.
1 .OE-01
5.2. Residual activity
1 .OE-02 1 .OE-03 1 .OE-04
f3.1
10
1
100
1000
Time After Shutdown *RAFS
-FS
-5SS316 *FeC!-Mn
+Ti64
lOOOf
(y) *NH25
+V5CrtrTi
Fig. 4. Comparison of decay power densities (W/m3) in several candidate structural materials, namely Ni625, SS316, FS, RAFS, Ti64, FeCrMn, and VSCrSTi after 1 y exposure at 1 MW/m2 neutron wall loading.
Fig. 3 displays the induced activities from 1 to 1000 y after shutdown. At the first year after shutdown, the activities from RAFS, FS, Ni625, FeCrMn and SS316, are higher than Ti64 and VSCrSTi, and are around 1-2 x 10’ Ci/m3 (3.7-7.4 X 10' GBq,/m3) level. The activities from Ti64 and V5CrSTi alloys at 1 y after shutdown are 1.3 X lo5 and 4.4 X lo5 Ci/m3 (4.8 X lo6 and 1.6 x lo7 GBq/m3), respectively. The activities at 100 y after shutdown and beyond from SS316 and Ni625 are 1300 and 6600 Ci/m3 respectively, and are higher than FS, Ti64, VSCrSTi, FeCrMn, and RAFS by
E.T. Cheng, G. Saji/Joumal of Nuclear Materials 212-215 (1994) 621-627
one or two orders of magnitude. At 200 y after cooling, Ti64, VSCrSTi, FeCrMn, and RAFS, may have the specific activity lower than 3 Ci/m3 (111 GBq/m31. For Ni625, SS316, and FS, flux attenuation factors of 104, 103, and lo’, respectively, will be needed to reduce the specific activity below 3 Ci/m3 beyond 200 y after shutdown. 5.3. Decay power density Fig. 4 displays the post shutdown decay power densities. At the first year after shutdown, the decay power densities at the first wall range from 36 kW/m3 for Ni625, the highest, to 130 W/m3 for VSCrSTi, the lowest among the structural materials compared. At 10 y after shutdown, the decay power densities are reduced and are 3600, 720, 50, 0.27, 0.46, 41, and 52 W/m3, respectively, for Ni625, SS316, FS, VSCrSTi, Ti64, RAFS, and FeCrMn. At 100 y after shutdown, the decay power densities are reduced significantly further to 1.7, 0.19, 0.07, 0.006, 0.012, 0.002, and 0.002 W/m3, respectively, for Ni625, SS316, FS, VSCrSTi, Ti64, RAFS, and FeCrMn. It appears that all structural materials considered here with the present alloying and impurity specifications, except Ni625 and SS316, may satisfy the 50 W/m3 power density limit within 10 y of cooling times. SS316 and Ni625 will need more cooling times, 30 and 50 y, respectively, to satisfy the 50 W/m3 limit. 5.4. Contact dose rate The contact dose rates from the various structural materials are depicted for comparison in Fig. 5. At 1 y after shutdown, the contact dose rate from Ni625, 72000 Sv/h, is significantly higher than any alloy compared in this study because of its Ni content. It is followed by FeCrMn, 8200 Sv/h due to Mn. It is then followed by SS316 (2300 Sv/h), FS, and RAFS (1400 Sv/h each) because of their Fe content. The VSCrSTi alloy has the lowest contact dose rate at this time after shutdown, about 40 Sv/h. The Ti64 alloy has a contact dose rate of 580 Sv/h. At 10 y after shutdown, the contact dose rates are significantly reduced and are 940, 180, 5.6, 3.6, 1.1, 0.019, and 0.017 Sv/h, respectively, for Ni625, SS316, FeCrMn, FS, RAFS, Ti64, and VSCrSTi. Further reduction is expected as the cooling continues. At 100 y after shutdown, th contact dose rates are expected to be 0.23, 0.016, 0.01, 0.0029, 0.00105, 0.000014, and 0.000014 Sv/h, for Ni625, FS, SS316, Ti64, VSCrSTi, FeCrMn, and RAFS. To reach below 2 mSv/h, the cooling times required are 140 000, 60000, 46000, 20 000, 42, 32, and 30 y, respectively, for Ni625, FS, SS316, Ti64, RAFS, FeCrMn and VSCrSTi. Note that only VSCrSTi, FeCrMn, and RAFS will have
lower than 2 mSv/h 50 y after shutdown.
625
contact dose rates readily within
5.5. Waste disposal rating The lOCFR61 Class C waste disposal ratings were estimated to be 500, 24, 22, 5, 2, 0.1, and 0.07, respectively, for Ni625, FS, SS316, Ti64, VSCrSTi, RAFS, and FeCrMn. Note that only FeCrMn and RAFS will qualify as Class C shallow-land burial wastes. Ti64 and VSCrSTi are not Class C wastes unless the present impurity levels, primarily Nb and MO, are reduced, as discussed below.
6. Discussions A detailed analysis was made examining the dominating radionuclides relevant to waste management aspect for the structural materials compared in this paper. The contents of these contributing natural elements can be obtained in Table 1. Discussions are given below for each material. 6.1. Ni625 The dominating radionuclides are @?o (half-life 5.27 y) and Nb (half-life 20000 y). “Co is due to Ni (main alloying element - 63 at%) and Co (minor alloying element - 0.51 at%). 94Nb is due to Nb (minor alloying element - 1.16 at%) and MO (minor alloying element - 5.6 at%). It is thus difficult for Ni625 to reduce further the induced activity because the causing elements are either major or minor alloying elements.
6.2. SS316 The important radionuclides are also @?o and “Nb. They are due to the main alloying element Ni (for “Co), and minor element MO (for “Nb). The impurity element Nb (308 ppm in SS316) is also a parent element for “Nb. Like Ni625, SS316 is unlikely to reduce the long-term induced activity. 6.3. Ferritic steel - 9Cr (FS) The dominating radionuclides are 6oCo, 54Mn (half-life 312 d), and %Nb. Again, 6oCo is due to the main alloying element Ni, and ?Jb due to the minor alloying element MO. 54Mn is due to the main alloying element Fe (87.9 at%) and minor alloying element Mn (0.4 at%). It actually decays away beyond 10 y after shutdown due to its relatively shorter half-life. The
626
E.T. Cheng, G. S~ji/Jou~l
~~~~clear ~~ate~a~ 212-215
impurity element Nb (480 ppm) also contributes to the production of 94Nb. Because of Ni and MO, FS is also not an alloy capable of reducing the long-term activity. 6.4. FeCrMn The dominating radionuclides are s4Mn and 94Nb. “Nb is primarily due to the impurity MO (6 ppm). 54Mn, again is due to the two main alloying elements Mn (20.2 at%) and Fe (65.7 at%). It will decay away in a few more years. The manganese steel investigated here can be an attractive blanket material if the impurities can be controlled. 6.5. Ti64 The dominating radionuclides for the Ti-6Al-4V alloy are 6oCo, 26Al (half-life 740 000 y), and 94Nb. ‘?Jo and 94Nb are, again, due to the impurities Ni, and Nb and MO (Ni: 3 ppm; Nb: 50 ppm; and MO: 150 ppm). %A.l is induced in Al, a major alloying element. However, the 26Al activity itself is still lower than the desirable limit. Thus if the ~pu~ties can be improved further (about 5 ppm level for Nb and MO, and 0.1 ppm level for Ni), Ti64 should be an ideal alloy for the experimental D-T fusion reactor. After the impurity control, the cooling time to reach below 2 m&/h contact dose rate is only about 5 y, instead of 20000 y as estimated with higher level impurities.
6.6. RAFS (9CrWVl The dominating radionuclides are 54Mn, *Co, and 94Nb. Again, 54Mn is due to the major element Fe, and minor element Mn, and will decay away if additional cooling years are provided. ?Zo and “Nb are due to the impurities, Ni (95 ppm), and MO (6 ppm), respectively. The RAFS is an attractive structural material since further reduction in Ni and MO impurities can essentially be available. 6.7. V5CrSTi The dominating radionuclides are @)Co and 94Nb. These radionuclides, like those for RAFS, are primarily due to the impurities Ni (3 ppm), and Nb (50 ppm> and MO (150 ppm). These impurities can be reduced further as discussed in Ref. [lo]. The contact dose rate will drop below the 2 m&/h value at 10 y after shutdown if the Ni impurity level can be reduced to 0.1 ppm. Thus VSCrSTi, like Ti64 and RAFS, can essentially be very attractive if management of discharged reactor components needs to be done shortly after decommissioning.
(1994) 621-627
7. Conclusions We have investigated several candidate structural materials for a near-term D-T fueled experimental power reactor regarding the management of discharged reactor components. We found that the blanket components made of VSCrSTi, Ti-6Al-4V, manganese steel, ferritic steel (9Cr) and the reduced activity ferritic steel, can be transferred to an on-site dry storage facility within 10 y after discharge. For Inconel alloy and stainless steel, it will take longer cooling times, 50 and 30 y, respectively, to enable this procedure. To transport the decommissioned blanket components to a permanent storage site, the cooling time needed can also be within 10 y for Ti-6Al-4V and VSCrTi, provided the impurities (particularly Ni and Nb) are controlled to an acceptable level. Nickel was found to be the most important impurity element when the cooling time is between 10 and 30 y, because of the half-life (5.27 y) of its activation product 6oCo. To achieve the 10 y cooling time for transportation, the required Ni content is about 0.1 ppm. The reduced activation ferritic steel and m~ganese steel will need about 30 y cooling time because of its Fe and Mn contents. Inconel, stainless steel, and ferritic steel will require more than 50000 y cooling time because of their Nb and MO contents. Inconel, stainless steel, and ferritic steel will not qualify as lOCFR61 shallow-land burial wastes after 1 my/m* exposure, due to their Nb and MO contents. The titanium and vanadium alloys can be shallow-land burial wastes if the impurity level for Nb and MO is dropped to 10 ppm. The reduced activity ferritic steel and manganese steel will qualify as the shallow-land wastes because the impurity control is already taken into account in the development of these alloys.
References [l] D.C. Kocher and A.G. Croff, A Proposed Classification System for High-level and Other Radioactive Wastes,
ORNL/TM-10289, Oak Ridge National Laboratory (1987). [2] Code of Federal Regulations, Title 10 - Energy, lOCFR20, Standards for Protection Against Radiation, US Federal Register, Revised January 1, 1991. [3] J. Raeder et at., ITER Safety, ITER Documentation Series, NO. 36 (IAEA, Vienna, 19911, see VIII. 4. for details. 141 Code of Federal Regulations, Titie 10 - Energy, lOCFR61, Licensing Requirements for Land Disposal of Radioactive Waste, US Federal Register, revised January 1, 1991. [S] S. Fetter et al., Fusion Eng. Des. 13 (1990) 239. [6] ITER Conceptual Design Report, ITER Documentation Series, No. 18 (IAEA, Vienna, 1991).
[7] W.W. Engle, Jr., A User’s Manual for ANISN, A OneDimensional Discrete-Ordinates Transport Code with Auisotropic Scattering, K-1693, Oak Ridge Gaseous Diffusion Plant (1967). [8] R. MacFarlane, Los Alamos National Laboratory, private cxnnmunication (1983).
[9] F.M. Mann, REAC*2: Users Manual and Cede Description, Westinghouse Hanford Company Report, WHCEP-0282 (1989). [lo] E.T. Cheng et al., Fusion Technol. 21 (1992) 2001.