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ITER divertor heat transfer system and loss of vacuum accident sequence analyses overview
Fusion Engineering and Design 42 (1998) 95 – 101
ITER divertor heat transfer system and loss of vacuum accident sequence analyses overview G. Cambi a,*, M.T. Porfiri b, H. Jahn c, D.G. Cepraga d, H.-W. Bartels e a
Physics Department, Bologna Uni6ersity, Via Irnerio 46, I-40126 Bologna, Italy b Associazione Euratom – ENEA sulla Fusione, C.P. 45, 00044 Frascati, Italy c Gesellschaft fu¨r ReaktorSicherheit, Garching, Germany d ENEA, INN-FIS, Via Don Fiammelli 2, 40129 Bologna, Italy e ITER JCT, San Diego Co-Center, San Diego, CA, USA
Abstract The paper presents the analyses performed by the European Union ITER Home Team for five ITER reference accident sequences: two loss of flow and two loss of coolant accidents on the divertor primary heat transfer system, and one loss of vacuum accident. Tritium and activated materials ground and stack environmental releases have been evaluated for each one of the accident sequences, up to 7 days from the accident start. The ATHENA and the INTRA codes are used for the accident transient thermal – hydraulic analysis and for the containment response to coolant leaks, respectively. The NAUA Mod 5M code has been used to assess the radioactive material transport, deposition and leakage inside and through the containment. © 1998 Elsevier Science S.A. All rights reserved.
1. Introduction A set of 25 reference accident sequences [1] has been analysed for ITER and documented in the non-site-specific safety report (NSSR-1) [2]. They include: operational events (which are named category I events in Ref. [2]), likely sequences (category II events, expected frequency greater than 10 − 2 per year), unlikely sequences (category III events, expected frequency from 10 − 4 to 10 − 2 per year) and extremely unlikely sequences (category IV events, expected frequency from 10 − 6 to 10 − 4 per year). Five of these accidents have been as-
* Corresponding author. Tel.: + 39 51 6098297; fax: + 39 51 6098062; e-mail: [email protected]
sessed by the European Home Team (EU-HT): two loss of flow accidents (LOFA) on the divertor primary heat transfer system (DV PHTS), two loss of coolant accidents (LOCA) on the DV PHTS and one loss of vacuum accident (LOVA). They are: DV PHTS pipe leakage (LOCA category II) Pump trip in DV PHTS (LOFA category II) Large DV PHTS ex-vessel coolant leak (LOCA category III) Pump trip in DV PHTS and failure of the active plasma shut down (LOFA category III) Loss of vacuum through one vacuum vessel (VV) penetration line (LOVA category IV) The specification for the analyses was defined by the ITER Joint Central Team (JCT) in an accident analysis specifications (AAS) report [3].
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The design information is documented in the safety analysis data list (SADL) report [4]. For each one of the accident sequences, the initiating event is assumed to occur during reactor operation at 110% (1.65 GW) of the nominal fusion power with the PHTS in steady-state conditions. The analyses have been performed according to the methodology approach described in Cambi et al. [5], up to the evaluation of the tritium and the activated material releases to the environment, using ATHENA [6], INTRA [7] and NAUA Mod 5M [8,9] codes. For each of the five accidents, some parametric analyses have been also performed, as specified in the AAS. A detailed description of the assessment is given in Ref. [10].
2. DV pipe leakage (LOCA category II) A small diameter pipe of a cooling loop of the DV PHTS is assumed to break (equivalent crossarea size 3 cm2) inside the room (i.e. the lower vault) that houses the divertor primary heat transfer system. A low DV PHTS pressuriser level signal triggers the fusion power shutdown system (FPSS) about 150 s from the accident start. The signal also triggers the isolation of the HTS vaults in 30 s. The coolant leak pressurises the lower and upper HTS vaults, which are connected by four 27-m length pipes (total cross-area 4 m2). The upper vault houses the first wall and blanket primary heat transfer system. The intervention of the fast train of the FPSS stops plasma burn resulting in a plasma disruption. Once plasma burn stops, the decay heat of the divertor plate materials is considered as the heat source for the DV PHTS. All the intact cooling loops are functioning: they cool down the in-vessel components. The VV cooling system is in the forced circulation mode during and after the accident: it is kept at the normal operating temperature (100°C). The coolers of the heating, ventilation and air conditioning (HVAC) system are able to restore the vault’s negative pressure in 6 h from a vault overpressure of 140 kPa. The ATHENA code is used to model a DV primary cooling loop and to assess the accident transient. The thermal effect of the disruption is taken into account by assuming a constant uniform
heat load of 15 MW m − 2 on the plasma exposed components of the involved DV loop for 1 s from the start of plasma disruption, which corresponds to 1.5 GJ of energy delivered to the DV, including a peaking factor of 3. Conservatively, it has been assumed that the decay heat to be removed by the DV PHTS is due to a full coverage of the DV in-vessel components by a 1-cm thick tungsten protective layer. The higher DV copper temperatures (550°C on the plasma side and 280°C on the coolant side) are reached on the DV lower wing after 155 s from the accident start. The Cu plasma-side temperature drops to 200°C in about 20 s after plasma shut down. Hence, no structure melting is induced inside the plasma chamber, i.e. no in-vessel LOCA results in this accident. A 36 volume INTRA model representing all relevant rooms (VV, upper and lower vaults, a generic room, the plant building and the pressure suppression system) has been set up to evaluate the containment response to the accident. The divertor region is modelled by 22 heat structures (heat slabs): 13 DV heat structures face the plasma and nine the vacuum vessel. The maximum vault pressure after the accident is 104 kPa, lower than its design pressure. The pressure stays constant for about 5 h up to the loop emptying, then the HVAC coolers reduce pressure to subatmospheric value in the following 10 min. The long-term thermal behaviour of the DV structure is investigated with the COSMOS code [11]. The maximum divertor temperature reaches 580°C after half a day. The maximum temperature falls to 245°C after 10 days. The radiological source terms involved are the tritium and corrosion products in the coolant of the failed loop. The total coolant inventory is conservatively considered released from the DV failed loop. The tritium content per DV PHTS loop is 10.9 g (concentration 0.2 g m − 3). The corrosion product inventory per loop is 10 kg (0.1% in the suspended form and 99.9% as wall deposits). The activated corrosion product (ACP) particle/droplet size is 2 mm. The tritium mobilised into the vault depends on the percentage of water flashing to steam. According to INTRA, this corresponds to 0.28 g tritium and 26 g ACP mobilised into the
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lower vault. The release to the environment is governed by three time windows: (1) 0 sB t B 180 s (Time needed to isolate the HTS vaults): the release occurs through an open ventilation duct (200 m length, 10 cm2 cross-section). About 0.2 mg ACP and 0.11 mg tritium are released at the stack during this period. (2) 180 sBt B5 h: During this period, the HTS vaults are pressurised up to 104 kPa, at t=600 s, and they stay at constant pressure up to 5 h. The total tritium environmental releases are 0.5 mg at ground level and 1.5 mg at stack. A NAUA calculation provides the ACP releases in the same time interval: 7.5 mg at ground level and 25 mg stacked. (3) 5 hB t B7 days: Negative pressure is reestablished in the HTS vaults by coolers in about 10 min, then no uncontrolled out-leakage occurs. The HTS atmosphere is assumed to be sent to the stack passing through aerosol filters (95% efficiency), dryers and molecular sieves (99% efficiency for HTO), and ventilation detritiation system VDS (99% removal efficiency for HT). The total stack controlled releases are then 0.3 mg tritium and 5 mg ACP. For this accident, the margins against the design guideline of releases (see Ref. [2]), given as the ratio between the release limits and the actual releases (both for tritium and activated products), are in the range 60 – 500 (see Table 1).
3. Large DV ex-vessel coolant leak (LOCA category III) A large-diameter pipe of a cooling loop of the DV PHTS is assumed to break (double-ended guillotine break) inside the HTS lower vault. The lower and upper vaults are pressurised and a high lower-vault atmosphere pressure signal triggers the FPSS and the isolation of the HTS vaults. The fast train of the FPSS stops plasma burning in 3 s and this results in a plasma disruption which delivers 1.5 GJ of energy to the DV. Once plasma burn stops, the decay heat of the divertor plate materials is considered as the heat source for the DV PHTS.
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The HTS vaults are isolated in 30 s from the high-pressure signal. The coolers of the HVAC system are used to restore negative pressure inside the vaults. The pressuriser in the failed loop empties in 13 s from the accident start and the loop is quasi-empty in about 60 s. The total coolant released is about 40000 kg. During the coolant blow down, the in-vessel components of the failed DV loop are cooled down with the residual cooling water present before complete drainage. After the loop emptying, the divertor of the failed loop is cooled by radiation to the surrounding cooled in-vessel components and by heat conduction with the divertor in-vessel structure (i.e. the cassette body). The lower vault pressure rises to 105 kPa in 0.8 s from the LOCA. The highest DV copper tube temperatures (750°C on the plasma side and 615°C on the coolant side) are reached on the lower wing after 6 s from the accident start. The highest Cu temperature drops to about 500°C in a few seconds. Hence, no structure melting is induced inside the plasma chamber, i.e. no in-vessel LOCA results in this accident. The long-term temperature transient driven by decay heat is the same as that discussed in Section 2. The maximum vault pressure after the accident is 0.16 MPa, lower than the vault design pressure. The HVAC coolers reduce pressure to subatmospheric value in about 2 h 40 min. According to INTRA, the total amount of steam in the vaults is 6.2 m3, which corresponds to a total airborne tritium of about 1.2 g. A release of 106 g ACP Table 1 Margins against the ITER design guideline of releases (ratio between the release limits and the actual releases) Event
2.4 3.3 4.5 a b
Stack releases
Ground releases
a
b
a
b
500 6750 85
200 300 30
100 2500 500
65 1000 385
Tritium as HTO. Activated products.
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occurs into the lower vault in about 60 s (about 55 g in the first 30 s). The releases to the environment are: (1) 0 sBtB 30 s: The average leak rate through the stack calculated by INTRA is 0.06 m3 s − 1 during the first 30 s. Thus, the environmental releases are: 0.12 mg tritium and 5.5 mg ACP. (2) 30 sBt B2 h 40 min: The HTS vaults are pressurised up to 160 kPa. The vault leak is 20% vault volume per day at 0.14 MPa overpressure, scaling with the square root of pressure. The resulting tritium release is 8 mg (2 mg at ground level and 6 mg stacked). A NAUA calculation provides the total ACP release to the environment, namely, 100 mg (25 mg at ground level and 75 mg stacked). (3) 2 h 40 minBt B 7 days: About 3 h from the accident start, negative pressure is restored inside the vaults by the HVAC coolers; thus, no uncontrolled release occurs. From that time the HTS atmosphere is assumed to be sent to the stack, passing through aerosol filters, dryers and molecular sieves. The total controlled stack releases are 1.2 mg tritium and 3 mg ACP. For this accident the environmental releases are at least two orders of magnitude lower than the design release limits foreseen for category III events (see Table 1).
4. Pump trip divertor (LOFA category II) A pump trip in one of the four divertor primary cooling loops results in a coolant flow reduction with a pump coast down from 100% speed to 0% in 147 s. A low flow signal triggers the FPSS after about 3 s and this results in a plasma disruption with 1.5 GJ of energy delivered to the DV. Once plasma burn stops, the decay heat of the divertor plate materials is considered as the heat source for the DV PHTS. The highest DV copper temperature is reached on the inboard wing after 7.5 s from the accident start. The maximum temperature value is 560°C on the first Cu layer (plasma side) and 296°C on the last layer (coolant side). No structural damage results from the calculation and no loss of coolant will result as a consequence.
Due to the DV PHTS integrity, no radiological release results from this accident.
5. Pump trip in divertor HTS (LOFA category III) A pump trip in one of the four divertor primary cooling loops results in a coolant flow reduction with a pump coast down of 147 s. The failure of the FPSS maintains the plasma burning condition. A loss of off-site power is considered as aggravating failure at the time of the divertor pump trip. As a consequence, a pump trip is assumed in all cooling loops and the VV cooling loop is assumed to be switched to the natural circulation mode. Emergency power will feed HTS and HVAC pumps 30 s after the loss of off-site power. The failed cooling DV PHTS loop heat up and the copper heat sink of the involved divertor cassette reaches its melting temperature (71.4 s from the accident start). An in-vessel loss of coolant therefore occurs, leading to the plasma termination with a disruption. A total of 1.5 GJ of energy is supposed delivered to the divertor in-vessel components of the failed loop. The steam reacts with the hot tungsten and CFC protective materials of the failed DV PHTS loop. The flashing of steam will lead to a rapid pressurisation of the VV. Once 200 kPa differential pressure is reached between the plasma chamber and VV suppression system, the rupture disks will open (87 s from the accident start) and the pressure is released to the suppression system. The VV pressure becomes subatmospheric at t= 365 s and no VV failure occurs. In the first 100 s after the rupture disk actuation, about 45000 kg water/ steam are discharged to the suppression system. The chemical reactions between steam and protective layer materials produce in the first hour of the accident about 0.01 g H2. Steam ingress into the plasma chamber and thermal conduction to colder parts of the divertor in-vessel components will reduce the highest divertor temperature to below 800°C after about 50 s from the in-vessel break. After this initial time period, decay heat will determine the further temperature transient starting from about 100 s. The long-term decay
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heat removal is performed by the intact cooling loops and the VV. The long-term transient is calculated with COSMOS. After approximately 2 h, the tungsten temperature drops to about 564°C. Within the next 6 h, the temperature increases to a maximum of about 590°C. The temperature settles at this level for about half a day. After this time, the maximum tungsten temperature decreases to about 230°C at 10 days and 215°C at 30 days. Steam would provide convection cooling but would also chemically react with tungsten. Approximately 40 g H2 is produced, most of it within the first day following the accident. A safety factor of 2 was used for all reaction rates. Hydrogen will also be produced by the reaction of steam and dust. Assuming an effective CFC surface of 36000 m2 kg − 1 [12], all dust located at hot surfaces would react during the transient. The assumed amount of dust at hot surfaces is [13]: 20 kg for Be, 40 kg for W, 80 kg for CFC. Since just one quadrant is affected by the accident, the maximum resulting hydrogen production amounts to 4.6 kg, which is below the flammability limit of 10 kg for hydrogen in the vacuum vessel. The radioactive inventories involved in the accident are the tritium and the activated corrosion products of the failed loop, the tritium in the co-deposited layer, in PFC bulk, in cryopumps and the activated dusts inside the VV. The total tritium mobilised is about 1220 g, while 100% of the suspended ACP and 20% of the deposited ACP are released from the loop in case of break. The initial amount of corrosion product aerosols (droplet) mobilised is 1.3% of the water spilled. The amounts of activated dust mobilised inside the VV in this accident are: 100% of the 10 kg small size (0.1 mm), 20% of the 30 kg medium size (4 mm) and of the 60 kg large size (100 mm) particles. A large amount of the mobilised dusts, ACP and tritium is transferred to the suppression system: about 24 kg of activated material (dust and ACP) in 50 s. Moreover, at that time about 4 kg of activated material is deposited on the VV by gravitational settling. The airborne mass into the VV then decreases drastically in a few seconds: 100 s after the rupture disks opening, it amounts to only 6 g (total for ACP and dust). After 1 h,
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the total amount of airborne activated material is 0.07 g. After the water/steam discharge to the suppression system, about 100 g tritium are contained in the VV atmosphere as HTO. In the following 1 h, the vapour partial pressure inside the VV, as calculated with the INTRA code, drops from 131 to 10 kPa. The airborne HTO is consequently reduced: the total amount of tritium in the VV atmosphere at that time is about 8 g. The release to the environment is due to the leakage from the vacuum vessel, that is 1% volume per day at design pressure (0.5 MPa), scaling linearly with the square root of the pressure. Because the pressure inside the VV is at subatmospheric value at t= 365 s, the environmental releases due to the leakage are negligible.
6. Loss of vacuum through one VV penetration line (LOVA category IV) Air from a generic by-pass room (GBR) is postulated to enter the plasma chamber through one VV penetration line (breach equivalent crosssection is 0.02 m2) and this terminates fusion power, inducing a plasma disruption: about 1.5 GJ are delivered to the DV. The signal of loss of primary vacuum triggers the isolation valve on the failed penetration line to close. The valve fails to close. Air ingress continues and it reacts with the PFC. Due to the assumed loss of off-site power all the PHTS pump trip, excluding the PHTS of the VV which is in its normal forced circulation mode. The emergency power is available after 30 s. The in-vessel tritium and dusts are mobilised by the air ingress. The GBR (its volume is 30000 m3) is isolated in 30 s from the loss of vacuum. An operator will initiate vacuum vessel pumping 1 h after the accident, stopping release from VV and GBR. The thermal effect of the disruption is taken into account by increasing uniformly the initial 1-cm thick plasma facing material by 30°C. So far, the initial first wall and FW coolant temperatures for INTRA calculations (end of plasma disruption) are 240 and 152°C, respectively. The effect of the emergency cooling is taken into account by decreasing the coolant temperature of the in-vessel components by 30 K
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h − 1. Due to the air inlet flow, the VV pressure rises to about 95 kPa in about 600 s and it reaches 98 kPa about 1 h after the accident start. No outlet flow from the VV occurs up to 2 h from the accident. In any case, because the VV and the room pressures equalise after about 600 s from the accident, a 1% of the VV atmosphere volume per day is considered exchanged with the generic room (to envelope diffusion processes) until the start of the vacuum pumps re-establishes subatmospheric pressure conditions. Due to the small amount of air entering the vacuum vessel and the low PFC temperatures, no significant chemical reactions occur. The radioactive inventories involved in the accident are the tritium in the co-deposited layer, in PFC bulk in cryopumps and the activated dusts inside the VV. The total tritium mobilised is 1216 g. The activated dust mobilised inside the VV in this accident are: 100% of the 10 kg small size (0.1 mm), 20% of the 30 kg medium size (4 mm) and of the 60 kg large size (100 mm) particles. The release to the environment is governed by the following time windows: (1) 0 s B t B30 s (Time needed to isolate the generic room from the environment): because the pressure inside the VV is lower than that in the generic room, no release occurs. (2) 30 sBt B 600 s: If the pressure inside the VV and the generic room stays below 100 kPa, then no environmental release will result. (3) 600 sBt B 1 h: During this period, the VV and the room pressures are at equilibrium; therefore, about 0.5 g tritium and 8 g dust are transferred from the VV to the room. Conservatively, 100% of the room atmosphere per day is considered to leak at ground level to envelope the leakage processes. The total tritium release (ground level) is 0.02 g. The NAUA calculation indicates a dust ground level release of 0.13 g. (4) t\ 1 h: During the VV atmospheric cleanup period, 99% of the tritium and 99.9% of the aerosols will be captured by VDS and filters. Therefore 1.2 g tritium and 16 g aerosols are released to the environment during this period. This contribution dominates the environmental release. It might be further reduced by cleaning up the VV atmosphere in recirculation mode.
For this accident the environmental releases are at least one order of magnitude lower than the design release limits foreseen for category IV events (see Table 1). The lowest margin can easily be increased as mentioned above.
7. Conclusions The EU-HT assessment of five ITER reference accidents shows a strong robustness of the ITER design. Radioactive releases associated with two divertor primary heat transfer system loss of flow accidents, two divertor ex-vessel loss of coolant accidents, and with the loss of primary vacuum accident are well below the ITER design limits fixed for the corresponding categories of accidents.
References [1] H.-W. Bartels, A. Poucet, G. Cambi, et al., ITER reference accidents, Fusion Eng. Des. 42 (1998) 13 – 19. [2] ITER non-site specific safety report (NSSR-1), vol. VII, S 84 SF 2 97-02-03 F 1, ITER San Diego JWS, CA, 1997. [3] H.-W. Bartels (Ed.), Accident analysis specifications for NSSR-1 (AAS), S 81 RE 4 96-03-12 W1.2, ITER San Diego JWS, CA, 1996. [4] H.-W. Bartels (Ed.), Safety analysis data list (SADL), S 81 RE 3 96-03-05 W1.2, ITER San Diego JWS, CA, 1996. [5] G. Cambi, G. Cavallone, M. Costa, Environmental source terms during a few reference accident sequences (RAS) of NET/ITER plant, J. Fusion Energy 12 (1/2) (1993) 139 – 143. [6] K.E. Carlson, P.A. Roth, V.H. Ransom, ATHENA Code Manual, Volumes 1 and 2, EGG-RTH-7397, EG & G Idaho Inc., Idaho Falls, ID, 1986. [7] H. Jahn, INTRA, In-Vessel Transient Analyses Code, Manual and Code Description, Draft, GRS, Garching, Germany, January 11, 1996. [8] H. Bunz, M. Koyro, W. Schoeck, NAUA Mod 4, A code for calculating aerosol behaviour in LWR core melt accidents, Code Description and Users Manual, KfK 3554, Kernforschungszentrum Karlsruhe, Germany, 1983. [9] H. Bunz, M. Koyro, W. Schoeck, NAUA Mod 5 und NAUA Mod 5-M, Zwei Computerprogramme zur Berechnung des Aerosolverhaltens in Containmentsystem eines LWR nach einem Kernschmelzunfall, KfK 4278, Kernforschungszentrum Karlsruhe, Germany, 1987. [10] G. Cambi, M.T. Porfiri, H. Jahn, EU accident sequence analyses for ITER NSSR-1, ENEA ERG-FUS/TECN S+E TR 27/96, 1996.
G. Cambi et al. / Fusion Engineering and Design 42 (1998) 95–101 [11] H.-W. Bartels, K. Jongeward, Accident analysis of decay heat driven transients for NSSR-1, S 81 RE 5 9606-18 W1.2, ITER, San Diego JWS, CA, 1996. [12] C.H. Wu, Highlights of EU R and D on co-deposition and dust in tokamaks, presented to ITER Dust Coordi-
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nation Meet., ITER San Diego JWS, CA, June 12–14, 1996. [13] S.J. Piet, ITER white paper on integrated picture of in-vessel tritium and dust, S 81 RI 13 96-06-28, ITER San Diego JWS, CA, 1996.
ITER divertor heat transfer system and loss of vacuum accident sequence analyses overview G. Cambi a,*, M.T. Porfiri b, H. Jahn c, D.G. Cepraga d, H.-W. Bartels e a
Physics Department, Bologna Uni6ersity, Via Irnerio 46, I-40126 Bologna, Italy b Associazione Euratom – ENEA sulla Fusione, C.P. 45, 00044 Frascati, Italy c Gesellschaft fu¨r ReaktorSicherheit, Garching, Germany d ENEA, INN-FIS, Via Don Fiammelli 2, 40129 Bologna, Italy e ITER JCT, San Diego Co-Center, San Diego, CA, USA
Abstract The paper presents the analyses performed by the European Union ITER Home Team for five ITER reference accident sequences: two loss of flow and two loss of coolant accidents on the divertor primary heat transfer system, and one loss of vacuum accident. Tritium and activated materials ground and stack environmental releases have been evaluated for each one of the accident sequences, up to 7 days from the accident start. The ATHENA and the INTRA codes are used for the accident transient thermal – hydraulic analysis and for the containment response to coolant leaks, respectively. The NAUA Mod 5M code has been used to assess the radioactive material transport, deposition and leakage inside and through the containment. © 1998 Elsevier Science S.A. All rights reserved.
1. Introduction A set of 25 reference accident sequences [1] has been analysed for ITER and documented in the non-site-specific safety report (NSSR-1) [2]. They include: operational events (which are named category I events in Ref. [2]), likely sequences (category II events, expected frequency greater than 10 − 2 per year), unlikely sequences (category III events, expected frequency from 10 − 4 to 10 − 2 per year) and extremely unlikely sequences (category IV events, expected frequency from 10 − 6 to 10 − 4 per year). Five of these accidents have been as-
* Corresponding author. Tel.: + 39 51 6098297; fax: + 39 51 6098062; e-mail: [email protected]
sessed by the European Home Team (EU-HT): two loss of flow accidents (LOFA) on the divertor primary heat transfer system (DV PHTS), two loss of coolant accidents (LOCA) on the DV PHTS and one loss of vacuum accident (LOVA). They are: DV PHTS pipe leakage (LOCA category II) Pump trip in DV PHTS (LOFA category II) Large DV PHTS ex-vessel coolant leak (LOCA category III) Pump trip in DV PHTS and failure of the active plasma shut down (LOFA category III) Loss of vacuum through one vacuum vessel (VV) penetration line (LOVA category IV) The specification for the analyses was defined by the ITER Joint Central Team (JCT) in an accident analysis specifications (AAS) report [3].
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The design information is documented in the safety analysis data list (SADL) report [4]. For each one of the accident sequences, the initiating event is assumed to occur during reactor operation at 110% (1.65 GW) of the nominal fusion power with the PHTS in steady-state conditions. The analyses have been performed according to the methodology approach described in Cambi et al. [5], up to the evaluation of the tritium and the activated material releases to the environment, using ATHENA [6], INTRA [7] and NAUA Mod 5M [8,9] codes. For each of the five accidents, some parametric analyses have been also performed, as specified in the AAS. A detailed description of the assessment is given in Ref. [10].
2. DV pipe leakage (LOCA category II) A small diameter pipe of a cooling loop of the DV PHTS is assumed to break (equivalent crossarea size 3 cm2) inside the room (i.e. the lower vault) that houses the divertor primary heat transfer system. A low DV PHTS pressuriser level signal triggers the fusion power shutdown system (FPSS) about 150 s from the accident start. The signal also triggers the isolation of the HTS vaults in 30 s. The coolant leak pressurises the lower and upper HTS vaults, which are connected by four 27-m length pipes (total cross-area 4 m2). The upper vault houses the first wall and blanket primary heat transfer system. The intervention of the fast train of the FPSS stops plasma burn resulting in a plasma disruption. Once plasma burn stops, the decay heat of the divertor plate materials is considered as the heat source for the DV PHTS. All the intact cooling loops are functioning: they cool down the in-vessel components. The VV cooling system is in the forced circulation mode during and after the accident: it is kept at the normal operating temperature (100°C). The coolers of the heating, ventilation and air conditioning (HVAC) system are able to restore the vault’s negative pressure in 6 h from a vault overpressure of 140 kPa. The ATHENA code is used to model a DV primary cooling loop and to assess the accident transient. The thermal effect of the disruption is taken into account by assuming a constant uniform
heat load of 15 MW m − 2 on the plasma exposed components of the involved DV loop for 1 s from the start of plasma disruption, which corresponds to 1.5 GJ of energy delivered to the DV, including a peaking factor of 3. Conservatively, it has been assumed that the decay heat to be removed by the DV PHTS is due to a full coverage of the DV in-vessel components by a 1-cm thick tungsten protective layer. The higher DV copper temperatures (550°C on the plasma side and 280°C on the coolant side) are reached on the DV lower wing after 155 s from the accident start. The Cu plasma-side temperature drops to 200°C in about 20 s after plasma shut down. Hence, no structure melting is induced inside the plasma chamber, i.e. no in-vessel LOCA results in this accident. A 36 volume INTRA model representing all relevant rooms (VV, upper and lower vaults, a generic room, the plant building and the pressure suppression system) has been set up to evaluate the containment response to the accident. The divertor region is modelled by 22 heat structures (heat slabs): 13 DV heat structures face the plasma and nine the vacuum vessel. The maximum vault pressure after the accident is 104 kPa, lower than its design pressure. The pressure stays constant for about 5 h up to the loop emptying, then the HVAC coolers reduce pressure to subatmospheric value in the following 10 min. The long-term thermal behaviour of the DV structure is investigated with the COSMOS code [11]. The maximum divertor temperature reaches 580°C after half a day. The maximum temperature falls to 245°C after 10 days. The radiological source terms involved are the tritium and corrosion products in the coolant of the failed loop. The total coolant inventory is conservatively considered released from the DV failed loop. The tritium content per DV PHTS loop is 10.9 g (concentration 0.2 g m − 3). The corrosion product inventory per loop is 10 kg (0.1% in the suspended form and 99.9% as wall deposits). The activated corrosion product (ACP) particle/droplet size is 2 mm. The tritium mobilised into the vault depends on the percentage of water flashing to steam. According to INTRA, this corresponds to 0.28 g tritium and 26 g ACP mobilised into the
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lower vault. The release to the environment is governed by three time windows: (1) 0 sB t B 180 s (Time needed to isolate the HTS vaults): the release occurs through an open ventilation duct (200 m length, 10 cm2 cross-section). About 0.2 mg ACP and 0.11 mg tritium are released at the stack during this period. (2) 180 sBt B5 h: During this period, the HTS vaults are pressurised up to 104 kPa, at t=600 s, and they stay at constant pressure up to 5 h. The total tritium environmental releases are 0.5 mg at ground level and 1.5 mg at stack. A NAUA calculation provides the ACP releases in the same time interval: 7.5 mg at ground level and 25 mg stacked. (3) 5 hB t B7 days: Negative pressure is reestablished in the HTS vaults by coolers in about 10 min, then no uncontrolled out-leakage occurs. The HTS atmosphere is assumed to be sent to the stack passing through aerosol filters (95% efficiency), dryers and molecular sieves (99% efficiency for HTO), and ventilation detritiation system VDS (99% removal efficiency for HT). The total stack controlled releases are then 0.3 mg tritium and 5 mg ACP. For this accident, the margins against the design guideline of releases (see Ref. [2]), given as the ratio between the release limits and the actual releases (both for tritium and activated products), are in the range 60 – 500 (see Table 1).
3. Large DV ex-vessel coolant leak (LOCA category III) A large-diameter pipe of a cooling loop of the DV PHTS is assumed to break (double-ended guillotine break) inside the HTS lower vault. The lower and upper vaults are pressurised and a high lower-vault atmosphere pressure signal triggers the FPSS and the isolation of the HTS vaults. The fast train of the FPSS stops plasma burning in 3 s and this results in a plasma disruption which delivers 1.5 GJ of energy to the DV. Once plasma burn stops, the decay heat of the divertor plate materials is considered as the heat source for the DV PHTS.
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The HTS vaults are isolated in 30 s from the high-pressure signal. The coolers of the HVAC system are used to restore negative pressure inside the vaults. The pressuriser in the failed loop empties in 13 s from the accident start and the loop is quasi-empty in about 60 s. The total coolant released is about 40000 kg. During the coolant blow down, the in-vessel components of the failed DV loop are cooled down with the residual cooling water present before complete drainage. After the loop emptying, the divertor of the failed loop is cooled by radiation to the surrounding cooled in-vessel components and by heat conduction with the divertor in-vessel structure (i.e. the cassette body). The lower vault pressure rises to 105 kPa in 0.8 s from the LOCA. The highest DV copper tube temperatures (750°C on the plasma side and 615°C on the coolant side) are reached on the lower wing after 6 s from the accident start. The highest Cu temperature drops to about 500°C in a few seconds. Hence, no structure melting is induced inside the plasma chamber, i.e. no in-vessel LOCA results in this accident. The long-term temperature transient driven by decay heat is the same as that discussed in Section 2. The maximum vault pressure after the accident is 0.16 MPa, lower than the vault design pressure. The HVAC coolers reduce pressure to subatmospheric value in about 2 h 40 min. According to INTRA, the total amount of steam in the vaults is 6.2 m3, which corresponds to a total airborne tritium of about 1.2 g. A release of 106 g ACP Table 1 Margins against the ITER design guideline of releases (ratio between the release limits and the actual releases) Event
2.4 3.3 4.5 a b
Stack releases
Ground releases
a
b
a
b
500 6750 85
200 300 30
100 2500 500
65 1000 385
Tritium as HTO. Activated products.
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occurs into the lower vault in about 60 s (about 55 g in the first 30 s). The releases to the environment are: (1) 0 sBtB 30 s: The average leak rate through the stack calculated by INTRA is 0.06 m3 s − 1 during the first 30 s. Thus, the environmental releases are: 0.12 mg tritium and 5.5 mg ACP. (2) 30 sBt B2 h 40 min: The HTS vaults are pressurised up to 160 kPa. The vault leak is 20% vault volume per day at 0.14 MPa overpressure, scaling with the square root of pressure. The resulting tritium release is 8 mg (2 mg at ground level and 6 mg stacked). A NAUA calculation provides the total ACP release to the environment, namely, 100 mg (25 mg at ground level and 75 mg stacked). (3) 2 h 40 minBt B 7 days: About 3 h from the accident start, negative pressure is restored inside the vaults by the HVAC coolers; thus, no uncontrolled release occurs. From that time the HTS atmosphere is assumed to be sent to the stack, passing through aerosol filters, dryers and molecular sieves. The total controlled stack releases are 1.2 mg tritium and 3 mg ACP. For this accident the environmental releases are at least two orders of magnitude lower than the design release limits foreseen for category III events (see Table 1).
4. Pump trip divertor (LOFA category II) A pump trip in one of the four divertor primary cooling loops results in a coolant flow reduction with a pump coast down from 100% speed to 0% in 147 s. A low flow signal triggers the FPSS after about 3 s and this results in a plasma disruption with 1.5 GJ of energy delivered to the DV. Once plasma burn stops, the decay heat of the divertor plate materials is considered as the heat source for the DV PHTS. The highest DV copper temperature is reached on the inboard wing after 7.5 s from the accident start. The maximum temperature value is 560°C on the first Cu layer (plasma side) and 296°C on the last layer (coolant side). No structural damage results from the calculation and no loss of coolant will result as a consequence.
Due to the DV PHTS integrity, no radiological release results from this accident.
5. Pump trip in divertor HTS (LOFA category III) A pump trip in one of the four divertor primary cooling loops results in a coolant flow reduction with a pump coast down of 147 s. The failure of the FPSS maintains the plasma burning condition. A loss of off-site power is considered as aggravating failure at the time of the divertor pump trip. As a consequence, a pump trip is assumed in all cooling loops and the VV cooling loop is assumed to be switched to the natural circulation mode. Emergency power will feed HTS and HVAC pumps 30 s after the loss of off-site power. The failed cooling DV PHTS loop heat up and the copper heat sink of the involved divertor cassette reaches its melting temperature (71.4 s from the accident start). An in-vessel loss of coolant therefore occurs, leading to the plasma termination with a disruption. A total of 1.5 GJ of energy is supposed delivered to the divertor in-vessel components of the failed loop. The steam reacts with the hot tungsten and CFC protective materials of the failed DV PHTS loop. The flashing of steam will lead to a rapid pressurisation of the VV. Once 200 kPa differential pressure is reached between the plasma chamber and VV suppression system, the rupture disks will open (87 s from the accident start) and the pressure is released to the suppression system. The VV pressure becomes subatmospheric at t= 365 s and no VV failure occurs. In the first 100 s after the rupture disk actuation, about 45000 kg water/ steam are discharged to the suppression system. The chemical reactions between steam and protective layer materials produce in the first hour of the accident about 0.01 g H2. Steam ingress into the plasma chamber and thermal conduction to colder parts of the divertor in-vessel components will reduce the highest divertor temperature to below 800°C after about 50 s from the in-vessel break. After this initial time period, decay heat will determine the further temperature transient starting from about 100 s. The long-term decay
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heat removal is performed by the intact cooling loops and the VV. The long-term transient is calculated with COSMOS. After approximately 2 h, the tungsten temperature drops to about 564°C. Within the next 6 h, the temperature increases to a maximum of about 590°C. The temperature settles at this level for about half a day. After this time, the maximum tungsten temperature decreases to about 230°C at 10 days and 215°C at 30 days. Steam would provide convection cooling but would also chemically react with tungsten. Approximately 40 g H2 is produced, most of it within the first day following the accident. A safety factor of 2 was used for all reaction rates. Hydrogen will also be produced by the reaction of steam and dust. Assuming an effective CFC surface of 36000 m2 kg − 1 [12], all dust located at hot surfaces would react during the transient. The assumed amount of dust at hot surfaces is [13]: 20 kg for Be, 40 kg for W, 80 kg for CFC. Since just one quadrant is affected by the accident, the maximum resulting hydrogen production amounts to 4.6 kg, which is below the flammability limit of 10 kg for hydrogen in the vacuum vessel. The radioactive inventories involved in the accident are the tritium and the activated corrosion products of the failed loop, the tritium in the co-deposited layer, in PFC bulk, in cryopumps and the activated dusts inside the VV. The total tritium mobilised is about 1220 g, while 100% of the suspended ACP and 20% of the deposited ACP are released from the loop in case of break. The initial amount of corrosion product aerosols (droplet) mobilised is 1.3% of the water spilled. The amounts of activated dust mobilised inside the VV in this accident are: 100% of the 10 kg small size (0.1 mm), 20% of the 30 kg medium size (4 mm) and of the 60 kg large size (100 mm) particles. A large amount of the mobilised dusts, ACP and tritium is transferred to the suppression system: about 24 kg of activated material (dust and ACP) in 50 s. Moreover, at that time about 4 kg of activated material is deposited on the VV by gravitational settling. The airborne mass into the VV then decreases drastically in a few seconds: 100 s after the rupture disks opening, it amounts to only 6 g (total for ACP and dust). After 1 h,
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the total amount of airborne activated material is 0.07 g. After the water/steam discharge to the suppression system, about 100 g tritium are contained in the VV atmosphere as HTO. In the following 1 h, the vapour partial pressure inside the VV, as calculated with the INTRA code, drops from 131 to 10 kPa. The airborne HTO is consequently reduced: the total amount of tritium in the VV atmosphere at that time is about 8 g. The release to the environment is due to the leakage from the vacuum vessel, that is 1% volume per day at design pressure (0.5 MPa), scaling linearly with the square root of the pressure. Because the pressure inside the VV is at subatmospheric value at t= 365 s, the environmental releases due to the leakage are negligible.
6. Loss of vacuum through one VV penetration line (LOVA category IV) Air from a generic by-pass room (GBR) is postulated to enter the plasma chamber through one VV penetration line (breach equivalent crosssection is 0.02 m2) and this terminates fusion power, inducing a plasma disruption: about 1.5 GJ are delivered to the DV. The signal of loss of primary vacuum triggers the isolation valve on the failed penetration line to close. The valve fails to close. Air ingress continues and it reacts with the PFC. Due to the assumed loss of off-site power all the PHTS pump trip, excluding the PHTS of the VV which is in its normal forced circulation mode. The emergency power is available after 30 s. The in-vessel tritium and dusts are mobilised by the air ingress. The GBR (its volume is 30000 m3) is isolated in 30 s from the loss of vacuum. An operator will initiate vacuum vessel pumping 1 h after the accident, stopping release from VV and GBR. The thermal effect of the disruption is taken into account by increasing uniformly the initial 1-cm thick plasma facing material by 30°C. So far, the initial first wall and FW coolant temperatures for INTRA calculations (end of plasma disruption) are 240 and 152°C, respectively. The effect of the emergency cooling is taken into account by decreasing the coolant temperature of the in-vessel components by 30 K
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h − 1. Due to the air inlet flow, the VV pressure rises to about 95 kPa in about 600 s and it reaches 98 kPa about 1 h after the accident start. No outlet flow from the VV occurs up to 2 h from the accident. In any case, because the VV and the room pressures equalise after about 600 s from the accident, a 1% of the VV atmosphere volume per day is considered exchanged with the generic room (to envelope diffusion processes) until the start of the vacuum pumps re-establishes subatmospheric pressure conditions. Due to the small amount of air entering the vacuum vessel and the low PFC temperatures, no significant chemical reactions occur. The radioactive inventories involved in the accident are the tritium in the co-deposited layer, in PFC bulk in cryopumps and the activated dusts inside the VV. The total tritium mobilised is 1216 g. The activated dust mobilised inside the VV in this accident are: 100% of the 10 kg small size (0.1 mm), 20% of the 30 kg medium size (4 mm) and of the 60 kg large size (100 mm) particles. The release to the environment is governed by the following time windows: (1) 0 s B t B30 s (Time needed to isolate the generic room from the environment): because the pressure inside the VV is lower than that in the generic room, no release occurs. (2) 30 sBt B 600 s: If the pressure inside the VV and the generic room stays below 100 kPa, then no environmental release will result. (3) 600 sBt B 1 h: During this period, the VV and the room pressures are at equilibrium; therefore, about 0.5 g tritium and 8 g dust are transferred from the VV to the room. Conservatively, 100% of the room atmosphere per day is considered to leak at ground level to envelope the leakage processes. The total tritium release (ground level) is 0.02 g. The NAUA calculation indicates a dust ground level release of 0.13 g. (4) t\ 1 h: During the VV atmospheric cleanup period, 99% of the tritium and 99.9% of the aerosols will be captured by VDS and filters. Therefore 1.2 g tritium and 16 g aerosols are released to the environment during this period. This contribution dominates the environmental release. It might be further reduced by cleaning up the VV atmosphere in recirculation mode.
For this accident the environmental releases are at least one order of magnitude lower than the design release limits foreseen for category IV events (see Table 1). The lowest margin can easily be increased as mentioned above.
7. Conclusions The EU-HT assessment of five ITER reference accidents shows a strong robustness of the ITER design. Radioactive releases associated with two divertor primary heat transfer system loss of flow accidents, two divertor ex-vessel loss of coolant accidents, and with the loss of primary vacuum accident are well below the ITER design limits fixed for the corresponding categories of accidents.
References [1] H.-W. Bartels, A. Poucet, G. Cambi, et al., ITER reference accidents, Fusion Eng. Des. 42 (1998) 13 – 19. [2] ITER non-site specific safety report (NSSR-1), vol. VII, S 84 SF 2 97-02-03 F 1, ITER San Diego JWS, CA, 1997. [3] H.-W. Bartels (Ed.), Accident analysis specifications for NSSR-1 (AAS), S 81 RE 4 96-03-12 W1.2, ITER San Diego JWS, CA, 1996. [4] H.-W. Bartels (Ed.), Safety analysis data list (SADL), S 81 RE 3 96-03-05 W1.2, ITER San Diego JWS, CA, 1996. [5] G. Cambi, G. Cavallone, M. Costa, Environmental source terms during a few reference accident sequences (RAS) of NET/ITER plant, J. Fusion Energy 12 (1/2) (1993) 139 – 143. [6] K.E. Carlson, P.A. Roth, V.H. Ransom, ATHENA Code Manual, Volumes 1 and 2, EGG-RTH-7397, EG & G Idaho Inc., Idaho Falls, ID, 1986. [7] H. Jahn, INTRA, In-Vessel Transient Analyses Code, Manual and Code Description, Draft, GRS, Garching, Germany, January 11, 1996. [8] H. Bunz, M. Koyro, W. Schoeck, NAUA Mod 4, A code for calculating aerosol behaviour in LWR core melt accidents, Code Description and Users Manual, KfK 3554, Kernforschungszentrum Karlsruhe, Germany, 1983. [9] H. Bunz, M. Koyro, W. Schoeck, NAUA Mod 5 und NAUA Mod 5-M, Zwei Computerprogramme zur Berechnung des Aerosolverhaltens in Containmentsystem eines LWR nach einem Kernschmelzunfall, KfK 4278, Kernforschungszentrum Karlsruhe, Germany, 1987. [10] G. Cambi, M.T. Porfiri, H. Jahn, EU accident sequence analyses for ITER NSSR-1, ENEA ERG-FUS/TECN S+E TR 27/96, 1996.
G. Cambi et al. / Fusion Engineering and Design 42 (1998) 95–101 [11] H.-W. Bartels, K. Jongeward, Accident analysis of decay heat driven transients for NSSR-1, S 81 RE 5 9606-18 W1.2, ITER, San Diego JWS, CA, 1996. [12] C.H. Wu, Highlights of EU R and D on co-deposition and dust in tokamaks, presented to ITER Dust Coordi-
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nation Meet., ITER San Diego JWS, CA, June 12–14, 1996. [13] S.J. Piet, ITER white paper on integrated picture of in-vessel tritium and dust, S 81 RI 13 96-06-28, ITER San Diego JWS, CA, 1996.