ML20070C917
| ML20070C917 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 10/31/1982 |
| From: | Bloom G HANFORD ENGINEERING DEVELOPMENT LABORATORY |
| To: | |
| Shared Package | |
| ML20070C876 | List: |
| References | |
| NUDOCS 8212140388 | |
| Download: ML20070C917 (67) | |
Text
{{#Wiki_filter:* il 4 _ - SumARY REPORT HYDROGEN DISTRIBUTION IN BREEDER REACTOR CONTAINMENTS G. R. Bloom October 1982 . APPLIED TECHNOLOGY Any further distribution by any holder of this document or of the data therein to third parties representing foreign interests, foreign govern-I ments, foreign companies and foreign subsidiaries or foreign divisions of U.S. companies should be coordinated with the Deputy Assistant Secre-tary for Breeder Reactor Programs, Department of Energy. HANFORD ENGINEERING DEVELOPMENT LABORATORY
..,,. Operated by Westinghouse Hanford Company, a :ubsidiary of Westinghouse Electric Corporation, under the Department of .
Energy contract No. DE AC06-76FF02170 P.O. Box 1970, Richland, Washington 99352 l',,7,:L'"!',f . l* .". '.7'#0,.S 8212140388 821207 "" ( PDR ADOCK 05000537 A PDR
t HYDR 0 GEN DISTRIBUTION IN BREEDER REACTOR CONTAINMENTS , G. R. Bloom f i . l ABSTRACT This program addressed the technical issue of hydrogen mixing ( and distribution in breeder reactor containments. Hydrogen concentrations were experimentally and analytically determined for an upper containment region for a simulated environment following a postulated breeder reactor accident. The analytical i work will be discussed in the final report. The experimental data will provide a basis for evaluation and validation of ana-lysis methods used to predict hydrogen distributions. The major work scope was to characterize hydrogen, temperature, and gas velocity distributions in the reactor containment building following a hypothetical beyond the design basis breeder reactor accident. Helium was used as a simulant for hydrogen in the tests for increased experimental safety end reduced cost. Exper-imental tests utilized the large containment vessel at the Hanford Engineering Development Laboratory (HEDL) Containment Systems Test Facility (CSTF). l f 1.0
SUMMARY
AND CONCLUSIONS I Large-scale tests were performed to determine the hydrogen distribution in a I breeder reactor upper containment region after a postulated accident beyond the design basis. Helium was used as a simulant for hydrogen in the experi-mental tests to increase safety and reduce experimental costs. These tests demonstrated the effectiveness of gas mixing mechanisms to pro-vide a relatively uniform hydrogen distribution outside the release plume in the upper containment region following a postulated breeder reactor acci-dent. For all tests the pure helium released from a central vertically orien-ted nozzle was diluted to less than 15 volume percent helium in air along the 1 4
l vessel centerline 10 feet above the nozzle. Excluding the source release plume maximum helium cc . entration differences of 7.5 volume percent helium observed during the source release period decreased to less than 5 volume percent helium within one hour after termination of the simulated hydrogen release. The upper containment region was demonstrated to have significant gas mixing - and no stagnant areas were observed. Even simulated floor vent releases, with flow in the laminar to transitional flow regime, became turbulent, entrained surrounding air, and became diluted.
2.0 INTRODUCTION
l' Large-scale tests were performed to determine the hydrogen distribution in a breeder reactor upper containment region after a postulated accident beyond the design basis. The tests were performed to check the validity of hydrogen distributions assumed for containment safety analyses, and the data will pro-vide a basis for limited validation of analytical models and computer codes used to calculate hydrogen distributions. The hydrogen distribution in containment must be known before conditions resulting from a severe accident can be calculated by such containment codes. Design of an upper containment purge inlet nozzle, outlet vent, and possible hydrogen igniter system would require knowledge of the containment hydrogen l distribution following an accident. The hydrogen release period simulated by these tests is the first 10 hours after the accident when there is insufficient sodium vapor to ignite the hydrogen by autoignition. Hot hydrogen would be released from lower contain . l ment vent pipes in the upper containment floor forming round buoyant jets or plumes during this period. The experimental tests involved a large-scale model upper containment region having a height of 48 ft and a diameter of 25 ft. Similarity theory provided 2
- the basis for applying experimental test data, obtained for the scale model, to a geometrically similar plant containment region. Preliminary analysis indicated that the important mixing p ocesses which apply are: buoyant jet or plume Jaixing, and natural convection flows created by cool walls of the vessel.
The character of the round jet or plume exiting from a vent pipe may be pre-dicted from the Reynolds number (Re) of the release gas in the pipe nozzle where: Re = 00' 1 9 Re = Reynolds Number U = velocity D = pipe diameter ~
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- o = gas density u = gas viscosity.
For fully developed flow with Re <950 the flow of a released buoyant gas is i inicia11y laminar, and gas mixing is by molecular diffusion. However, laminar flow almost never exists very long for large buoyant plumes due to inherent instabilities. Inward propagation of turbulence from the plume boundary, l other flow fields and geometry " trips" will convert the laminar plume flow l into turbulent flow, causing entrainment of surrounding atmosphere and dilu-tion of the release gas. Flow is transitional when the jet or plume Reynolds number is greater than 950 and less than 5000. The flow may be either laminar or turbulent after it leaves the nozzle in the transitional regime. Laminar flow may become turbu-1ent some distance after leaving the nozzle, and that distance depends on the pipe nozzle Reynolds number. Estimates indicate that breeder reactor hydrogen releases from a vent may have Reynolds numbers in the range 300 to 2000. Therefore, the plume flow would be laminar to transitional for the vent release. I For Reynolds numbers greater than 5000, the jet is highly turbulent and is quickly diluted due to entrainment of surrounding atmosphere. Centerline 3 l
concentration, temperature and velocity decreases for such jets are documented, and can be predicted by relatively simple correlations.(1,2,3) Within 50 vent diameters downstream of the release point the jet centerline concentration is reduced to one tenth that of the source gas relative to the. surrounding atmosphere. The radial concentration gradient follows a Gausian distribution from a centerline maximum to a surrounding atmopshere minimum at the edge of the mixing region. Beyond 50 diameters along the centerline the l jet velocity continues to decay and the jet converts into a plume. For a buoyant turbulent jet or plume discharging into a containment volume, the extent of mixing in the far field is governed primarily by geometry, by jet or plume to ambient density ratio, and by the densimetric Froude Number, y; ,(1.2.3) N f oil /2 03 j Lg - l
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U = fluid discharge velocity, g = acceleration due to gravity, ao = difference in density between jet and su*roundings, o = density of jet fluid, L = a characteristic length (jet diameter). The Froude number represents the ratio of momentum force to buoyance force. Momentum forces generate turbulent eddys and convective flows causing entrain-l ment of the surrounding gas and dilution of the jet fluid. Buoyancy effects cause the gas to rise and may generate mixing if turbulent flow results. i Once the gas is at the top of the compartment. positive buoyancy effects tend to cause the gas to stay there. Geometrically similar containment regions having turbulent plume releases with the same densimetric Froude number can be expected to have similar gas distributions in the far field with other effects being equal. For values, of the Froude number greater than 100,the source gas mixes with the surrounding atmosphere due to entrainment. If the Froude number is less than approximately one, buoyancy forces cause the source gas to rise to the 4 _ .. _ _ ~ _. - - .
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- upper region of the receiving compartment and spread laterally. 1he buoyancy of the gas would tend to cause the gas to remain at the top of the compartment.
! Flow patterns caused by the buoyant plume would tend to cause mixing of the gas with the surrounding atmosphere. Estimated Froude numbers are less than one
' for the buoyant round plume of hydrogen that would be released during the first 10 hours .after,a hypothetical, beyond the design basis accident of a breeder reactor.
Although the jet or plume mixing action would be less effective in the case of a low velocity release, additional mixing mechanisms due to natural convection l would generally produce convective flows. These flows are normally sufficient to break up any. buoyant gas layers formed near the top of the compartment and' a well alxed atmospnere still results. Natural convection gas flows occ'ur whenever there is a temperature difference between the wall and the bulk atmos-phere. Gases heated or cooled by the wall will rise or f.all respectively due ' to the density difference between the heated or cooled ga's and the surrounding atmosphere. This buoyant force imparts momentum to the gas and significant turbulent mixing can result. The character of natural convection flow depends
- mainly on the Grashof number (Gr) and the Prandt1 number (Pr).(4,5) i Gr is defined as
2 Gr = g89 aTo * [2] y2 and the Prandtl number is defined as: p [3] Pr = k Definition of the symbols used in equations [2] and [3] are as follows: g = acceleration due to gravity, 6 = fluid coefficient of thermal expansion, L = characteristic *ength, AT = temperature difference between surface and bulk gas, p = fluid density u = fluid <iscosity, C = heat capacity at constant pressure, p k = thermal conductivity. 5
A laminar to turbulent boundary layer flow transition occurs at Rayleigh numbers (Grashof number times ?randt1 number) of roughly 10'. Since the Grashof number varies with the third power of the characteristic length of the vessel, resulting in large Rayleigh numbers, large vessels nearly always have turbulent naturally convected boundary layers. A temperature difference in the boundary layer of a few degrees has resulted in a turbulent boundary layer in the CSTF test vessel in previous experiments. Because bulk gas to wall temperature differences are expected to be greater than a few degrees . in both the tests and in a plant after a hypothetical accident, turbulent boundary layers along the vessel wall can be expected for both cases. Measure-l ments of temperature gradients near the wall were made during the test to allow evaluation of the relative importance of natural convection. 3.0 OBJECTIVES The first objective of these tests was to provide a basis ,for determining the hydrogen distribution in the upper containment region of a sodium cooled breeder reactor plant after a hypothetical beyond the design basis accident. ' Hot helium was released from a pipe oriented vertically upward into a modeled upper containment region to simulate a buoyant hot hydrogen release from the lower cavity region into the upper containment region of a breeder reactor. A pipe in the test system simulated the reactor vent pipe from the lower con-tainment region. The second test objective was to provide experimental data for validation of computer codes used for calculating hydrogen distribution. Hydrogen concen-trations, gas temperatures and local velocities were predicted using the
" TEMPEST" computer code. Predicted hydrogen concentrations, gas temperatures, and gas velocities will be compared with experimental data in a final report to evaluate and validate TEMPEST for application to this geometry and these flow conditions.
4.0 DESCRIPTION
OF TESTS Hydrogen concentration, temperature, and velocity profiles within a simulated upper containment region were detemined for conditions postulated to occur after a hypothetical beyond the design basis accident of a fast breeder reactor. Helium was used i'n these tests to simulate hydrogen. g
i I 4.1 Helium Mixing Tests _ l The helium was released from a vertically oriented pipe in the center of a simulated containment floor region as shown in Figure 1. The l axisymetric geometry provided a two-dimensional problem for hydrogen mixing code validation. TEMPEST code was used to provide a pre-test prediction of concentration, temperature and velocity profiles which - will be compared with the experimental data in the final report. The hydrogen mixing process due to the buoyant plume release in the reactor was modeled by matching estimated densimetric Froude number of the reac-a tor vent release and that of the experimental pipe release. A plant 12-inch vent was simulated with a 4-inch pipe in the tests. i Four tests were perforined over a range of Reynolds numbers and Froude numbers as shown in the test matrix of Table 1. The first test, HEM-1,
' was conducted with a relatively lower natural convec, tion effect. The containment atmosphere and vessel wall were approxim'ately 85"F. The remaining three tests were perforud at approximately 110'F to include the larger natural convection mixing effects expected in the plant.
Froude number is varied over the range from approximately 0.4 to 3.0 during tests HEM-2 through HEM-4 to demonstrate effect of Froude number on helium distribution. The estimated Froude number range for the breeder reactor postulated hydrogen release is.0.04 to 0.8 so the exper-i imental tests correlated well with the upper Froude number range of the postulated reactor releases. The Reynolds number ranged from 500 in test HEM-4 to 4000 in test HEM-3. The predicted flow regime was laminar to transitional for the tests. If gas flow is laminar at the r:ozzle it may go turbulent shortly after leaving the nozzle. Turbulent flow could [ be detected with the gas samples taken approximately 5 and 10 feet above l l the nozzle since the plume centerline concentration would be reduced by i entrained air. 4.2 Test Procedure I The test conditions for the tests.EEM-1 through HEM-4 are summarized in Table 1. The general procedure followed consisted of the following steps: . 7 .
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. 1. Install helium nozzle, equipment and instrumentation.
- 2. Provide pre-test prediction of test HEM-2. .
- 3. Seal vessel and establish initial containment temperatures per Table 1. using steam heater if required.
- 4. Release helium into containment at the rate in Table 1. .
- 5. Take seasurements per Section 4.3 during helium release and for 1 hour after helium source termination.
- 6. Vent vessel and purge with air.
TABLE 1 HELIUM MIXING TEST MATRIX Upper Helium Helium Helium Test Contain. Release Release Release No. Temp. (*F) Temp. (*F) Rate (ft /m) Duration-Min Re Tr HEM-1 85110 400 1 20 4314 60 1000 0.76 HEM-2 110 1 10 400 1 20 4314 60 1000 0.76 HEM-3 110 1 10 400 1 20 172 1 20 15 4000 3.04 HEM-4 110 1 10 400 1 20 22 1 2 120 500 0.38 4.3 Instrumentation The following data was collected during the gas mixing tests to character-ize the test compartment and process conditions.
- He concentrations
- gas temperatures
' gas velocities
*02 e ncentration
- vessel inner wall temperatures
' atmospheric pressure and vessel pressure .
9
Instrument sensor locations are shown schematically in Figure 1 and Figure 2. There were 12 He sample points, 6 velocity probes, and 22 thermoccuples im the gas, 'and 8 thermocouples on the compartment walls. All data was processed by a data acquisition system and recorded on magnetic tape. Test data were also reco.-ded by chart recorders as a back-up system for the tape record. Data was read from the magnetic tape and stored on a disc file where it could be recalled, processed, reduced and/or plotted as a function of time if desired by data pro- - cessing equipmert. Helium concentrations were measured by 12 Teledyne model 225BX thermal conductivity type analyzers. Response time of the analyzers is 90% full scale in SC seconds. A sample flow of 150 cc/ min was taken from a total minimum sample flow of 0.2 ft /3 min. Helium concentrations were recorded or, a 12 point chart recorder to provide a continuous record of the gas concentrations for the 12 sample locations. , Gas velocities were measured with a Thermo Systems Incorporated hot l film anemometer system. Temperature compensated probe models 1332 and 1330 were calibrated in air at 0-10 fps or 0-15 fps. Model 1053B I constant temperature anemometer modules are coupled with a model 1047 averaging module to average local turbulent velocity fluctuations over a 3 second period. Model 1051 power supplies provide power and local monitoring of probe output. Gas velocity probe output voltage was corrected by calculation to account for the helium effect on the gas velt. city sensor. Type K thermocouples 1/8 inch in diameter were used to measure tempera-ture. The thermocouples were directly wired to the Fluke data logger and the temperatures recorded on strip chart printing recorders. - Three Beckman model 7003 oxygen monitors provided oxygen data with a sensor response time of 90% in 20 seconds. An additional sample time of 30 seconds resulted in an overall response time of 5.0 seconds for - oxygen analyses at sample points in the test containment. 10
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l l The data acquisition system consisted of two Fluke 2240 Data Loggers ' and two Tektronixs model 4923 tape recorders. The minimum possible scan rate for the Fluke data logger while providing a paper tape and mag tape record was 30 seconds, and a recording every 45 seconds was taken. 5.0 TEST RESULTS AND DISCUSSION Helium concentrations measured throughout a simulated breeder reactor upper containment region during four modeled hydrogen releases are presented in Figures 3 through 6 for tests HEM-1 through -4 respectively. The helium release period started at time O for each test and terminated 15 minutes to 2 hours later as outlined in Table 1 of Section 4.0. Experimental helium concentrations were determined continuously at the ten upper containment sample points and were recorded at 45 second intervals. ixperimental data was taken for a minimum of 60 minutes after helium source termination to determine gas distribution after eliminatien of the buoyant jet induced mix-ing process. l Analysis prior to the experimental tests concluded that the tests should I preserve the Froude number and natural convection effects of the reactor plant hydrogen jet environment being modeled. The plant jet Froude number for the floor vent was estimated to be 0.8 at 10 hours after start of the postulated hydrogen release. Both tests HEM-l' and HEM-2 had helium jet Froude numbers of 0.8 . Test HEM-2 was performed at the elevated atmosphere temperature of 110 F to include the effects of natural convection expected during the postulated plant accident. Test HEM-1 had a lower atmosphere temperature of 85 F. Figures 3 and 4 show that the measured upper containment helium concentrations during the release period were nearly the same for tests with 85"F and 110 F , atmosphere having the same helium release rate. Somewhat better gas mixing was observ.ed for the 110 F ambient case after the helium source was tenninated. Release of helium at a jet Froude number of 3.0 in test HEM-3 resulted in sig-nificantly higher helium concentrations along the centerline above the jet release nozzle (27 volume percent helium 5 3 ft above the nozzle) and somewhat l higher concentration differences throughout upper containment. The maximum 12
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measured helium concentration difference of 7.5 volume percent helium for all tests, for spatial locations outside the plume, was observed for test HEM-3. The same test had the maximum helium concentration difference of 5 volume percent helium one hour after source jet termination. The helium release at a jet Froude number of 0.4 for test HEM-4 had a nozzle Reynolds number of 500. Even a release at this low Reynolds number; which is in the laminar flow regime, became turbulent as evidenced by the maximum helium concentration of only 12 volume percent measured 5.3 feet above the nozzle along the plume centerline. These tests demonstrated the effectiveness cf gas mixing mechanisms for conditions fol.owing a postulated breeder reactor beyond the design basis accident. Excluding the source release plume, maximum helium concentration differences of 7.5 volume percent helium during the sourci release period decreased to less than 5 volume percent helium within one hour after termin-ation of the simulated hydrogen release. Simulated floor vent release plumes with Reynolds numbers between 500 and 4000, typical of laminar to transitional flow, became turbulent and entrained surrounding air causing the plume center-line concentrations to decrease. 17 l
l
6.0 REFERENCES
1
- 1. C. J. Chen and W. Rodi, "A Review of Experimental Data of Vertical Turbulent Buoyant Jets " Report SFB 80/T/69, University of Karlsruhe, l
!(anover, Germany December 1975. ;
- 2. R. J. Frankel and J. D. Cunning, " Turbulent Mixing Phenomena of j Ocean Outfalls," J. Sanit. Ener. Div. ASCE. H. No. SA2, pp 33-59 (April 1965).
- 3. D. S. Trent and J. R. Welty, " Numerical Computation of Momentum Jets and Forced Plumes," Computer and Fluids,1, pp 331-357 (1973).
~
- 4. K. E. Torrance, L. Orloff, and J. A. Rockett, " Experiments on Natural Convection in Enclosures with Localized Heating From Below "
JJ Fluid Mech., g, pp 21-31 (1969). .
~
- 5. K. E. Torrance and J. A. Rockett, " Numerical Study of Natural Con-vection in Enclosures with Localized Heating From Below -- Creeping Flow to the Onset of Laminar Instability," J. Fluid Mech., H,1969.
'l i I e,' 18
E O Enclosure 4 l t i i t l i t l
_d M P-3, U CL.2 C MMCC PMC CRBRP-3 Vol. 2 Rev. 4 figf2 puSL ModL Mb
- 2. 1 ,
To insure containment atmosphere mixing hedere venting, the purge air
- g' i
shall be 6 L ;;;d int: ;;c.t;5n t ,b;'.[;b:th 0:0' . AosATtB se ps to W Tut. p.tunkc rea, oite4:r aw se mePpAM ohsTo TM. vt,rt udel ,
. 3. The purge syst shall prevent backflow from containment to the outsiae atmosphere.
- 4. The purge s'ystem, in combination with the containment vent and cleanup
- systems, shall maintain containment at a negative pressure after the containment pressure is reduced by the initial venting after 24 hours.
- 5. The purge system operations shall be by remote manual actuation from the control room.
1 2.1. 2. 8 Containment Vent System l. To prevent containment failure by excessive pressure, the vent system shall have a capacity between 24,000 and'26,400 acfm with a containment pressure of 30 psia, a containment atmosphere density of O.'07 lb/ft3 l and a viscosity of 0.06 lb/ft-hr. It shall remain functional if up to 300,000 pounds of aerosol enter the system at a maximum rate of 5,600 lb/hr.
, a.
kh. = m w L e<
- 2. The vent system shall exhaust the containment atmosphere from Mund etS.x t into the containment cleanup system. -
1 l 3. The containment vent system shall be compatitle with the following gases, vapors and aerosols: Ar, N 2
, H2 ' N 0, 00, CO , 0 ' '2 0, Na2 2' 2 2 2 haOH, Na 2 CO 3 , fission products, and compounds resulting from fission product reactions. The system must remain' functional for inlet gas temperatures and pressures given on Figures 2-5 and 2-6, and, beyond 150 hours for temperatures up to 2500F. 4
- 4. The vent system operations shall be by remote manual actuation from the control room. -
1 l 2-7
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3.3 CONCLUSION
S ON THERMAL MARGIN BEYOND THE DESIGN BASE
- ' - -..N . "--
.,.c. -
. 3 73 3 .4. REFERENCES 4-1 4.0 . ASSESSMENT OF RADIOLOGICAL CONSEQUENCES .
4.1 HCDA RADIOLOGICAL SOURCE TERM 4-1 4.1.1 Non-Energetic Core Meltdown . .4-2 4.1.2 Energetic HCDA. 4 .5 4.2 RADIOLOGICAL DOSES FROM ATMOSPHERIC RELEASES
, , ' 4-8 4.2.1 Methods and Data Base 4-8 4.2.2 Radiological' Doses 4-10 4.3 GROUNDWA1ER CONSIDERATIONS 4-11 4.4 "'-"~"""""""^"""a""^""""^"" (%ck*ewD=Neh 4-13
( 4-14
4.5 CONCLUSION
S ON RADIOLOGICAL CONSEQUENCES
4.6 REFERENCES
4-15 5-1 . 5.0 SUMMdRYANDCONCLUSIONS . . t A.0-1
. APPENDICES: -
A.0-2 INDEX TO SENSITIVITY STUDIES IN APPENDICES A-1
'A. DEV'ELOPMENT PROGRAMS SUPPORTING THERMAL MARGIN ASSESSMENTS A-1 A.1 SODIUM-CONCRETE INTERACTIONS DEVELOPMENT PROGRAM A.'1.1 Purpose Al A-1 A.1.2 Program
- A-2 A.1.3 Schedule A-3 l
A.1.4 Criteria of Succes's . A-3 A.1.5 Fallback Position A-4 A.2 HYDROGEN AUTO-CATALYTIC RECOMBINATION Purpose . A-4 , A.2.1 A-4 ! A.2.2 Program A-5 A.2.3 Schedule A-5 A.2.4 Conclusions 4 l 1 ! ii
l CCRP-3 l V31.2. Rev.0 l 4.0 ASSESSMENT OF RADIOLOGICAL CONSEQUENCES This section addresses the radiological consequences associated with an HCDA. These radiological analyses are based on the design described in Section 2 and the thermal and structural analyses presented in Section 3. Section 4.1 discusses the development of the radiological source terms considered in the cases analyzed. A wide range of assumptions on materials initially released to the RCB is used. Section 4.2 provides the results of radiological calculations for atmospheric releases. Section 4.3 considers potential releases to the groundwater. 3 1 ' ' The radiological calculations are based on the homogeneous core design. { Section 4.4 discusses the impact of the heterogeneous core design and I Lconcludes that the radiological consequences for the heterogeneous core are
, t Stund;d b w. ;,;;;;:::::: Or: =:elt .f The overall conclusions on radiological consequences are provided in Section 4.5. The todiological cotesicxtims anct resAts are basect on the heterogeneous core desgrs .
S" " -^ H "W 0-4.1 HCD'A RADIOLOGICAL SOURCE TERM
. 3 The radiological consequences associated with the TMBDB scenario are based on a complete core meltdown. The release of radioactive material from the reactor cavity to the RCB is considered in two parts: an initial release phase, and a sodium boil-up phase. The types and amounts of radioactivity released from the RC deoend on how much damage occurs to the head as a result of the HCDA, which in turn, depends on how much energy is assumed to be
" released as a result of the HCDA. Four cases, representing varying degrees of imediate leakage through the head, were evaluated. The first case represents
- the best estimate consequence of a hypothetical core disruptive accident.
Subsequent cases assume greater initial releases through the reactor vessel head. e 9
\
4-1
~ - ' - , . . _ _ . . _ . ,
~
CCRP-3 VIsl.2, Rev.0 Based on a recent survey (Reference 4-5) of experimental data on liquid carry-over from comercial evaporators and entrainment of solid particles in the vapor stream from an evaporating liquid pool, it was concluded that the decontamination factor (partitioning factor) for plutonium particles would be at least a f actor of 1000. Partitioning of solid fission products in the sodium as it vaporizes is based on the method sumarized in Reference 4-6. The combined partitioning of the fuel and sodium results in a release of 15 of the total non-volatile solid fission product inventory. A more detailed evaluation of the overall solid fission product release is presented in Appendix E. The fuel release during the sodium boilup phase is estimated by considering the two attenuating mechanisms discussed above, i.e.,15% of the fuel particulate remaining in suspension following meltdown and reparticulation,
- and a partition factor of 1000. This would result in approximately@320 grams of plutonium being carried into the RCB atmosphere with the boiling sodium. See Appendix E for a more detailed discussion of plutonium release from the boiling sodium. .
Additional mechanisms for transporting plutonium from the reactor cavity to the RCB have been investigated and found to be negligible in comparison to theMrafnO considered above. These additional mechanisms are also discussed in Appendix E. The initial release phase and boilup phase source terms described above for f , a non-energetic core meltdown are used in Case 1 in Table 4-1. l Post Boil-Dry Phase After the sodium pool in the reactor cavity has evaporated a bare fuel / steel debris bed is lef t. Most of the fission product release is expected to occur prior to boil-off (Reference 4-1). Potential mechanisms for further release of fission products and plutonium from the dry debris bed are: (1) surface vaporization; (2) particle levitation; and (3) gas 1 (') 4-4 %
CRBRP-3 Vol.2, Rev.0 sparging. The first two mechanisms are considered for plutonium in
,j Appendix E and are shown to result in a negligible contribution to the release associated with the boiling sodium pool. The volatile fission products are assumed to have been completely released. The non-volatile fission products have vapor pressures similar to or 1,ower than the vapor pressure of fuel (Reference 4-4). Thus, like the fuel, no significant fraction of the remaining fission products would be released from the l
molten surface due to the first two mechanisms. The release of fission products and plutonium due 'to gas sparging has also beenevaluated(seeAppendixE)'; The results of this evaluation show that those products whose releases are enhanced the most by sparging are the f more volatile products which the analysis already considers to be totally j released. The release of the other less volatile products by sparging is f accounted for by the 1% release fraction assigned to the non-volatile fission products in the boil-up phase source term. Plutonium release from the molten pool by sparging could be on the order of r a several month period and this has been assumed to be released to the RCB. The
- w. . evaluation of this additional plutonium source is discussed in Appendix E.
'I Assuming a 99% filter efficiency and taking credit for aerosol fallout and
~
plate-out, about[gfam's of plutonium could be released to the atmosphere over a several month period beginning at sodium boildry (e 5 days) after
/
the start of the accident. 4.1.2 Energetic HCDA
, In'.tial Release Phase l
The case described in Section 4.1.1 is based on the expected consequence of a hypothetical core disruptive accident; namely a non-energetic condition and consequently, no significant imediate release of sodium or non-volatile fission products through the reactor vessel head. Several variations of the expected case were analyzed using successively more ! pessimistic assumptions on the initial releases through the reactor vessel head. I 4-5 f- * ~ -
CRCRP-3 Vol.2. Rev.0 Fission product and activation product activity levels a're based on the end-of-equilibrium-cycle core inventory identified in Table 12.1-32 of the . PSAR. The beginntnaY. -Nuilibrium-cycle plutonium inventory was used beca'use it results in a slightly higher dose value than C 4 e. of-equilibrium-cycle plutoni h %AkM elm'gerf FFTFgde 6d. w eten.ng Post Boil-Dry Phase The same consideraticns apply here as discussed in Section 4.1.1. 9 e 1 l 4-7
- _ - _ _ _ _ _ l CRBRP-3 Vol.2. Rev.0 Dose Factors I
. 1 r \O '
Dose conversion factors (rem /ci) used in the COMRADEX code to calculate
.' specif t,c organ doses were taken from References 4Mnd 4-@.demme E , ,a tcp udoes. 4 Ge .Aedd r d -e were !
g {mpgo;wkde v40es. Ah ob P were . Containment Modeling .
.The time dependent radiological source term is released directly to the RCB.' The release rate from the RCB is that ca'lculated by the CACECO code. .
~
~
For the first 36 hours of the scenario the RCB atmosphere leaks at a low rate (based on 0.11/ day at 10 psig) to the annulus filter system (described in Section 6.2.5 of the PSAR). During this 36 hour period unfiltered bypass leakage at the rate of 1% of the filtered leakage is considered. After 36 hours the RCB is vented and subsequently purged (Figures 3-13 and 3-14) to maintain the hydrogen concentration at an acceptable level. During this phase filtering is by the 1148 0 8 filter system which is designed for the higher vent rates. The efficiency of the TMBDB filter system is { 99% for solid fission products and fuel and 971 for condensible fission products (halogens, Se, and Sb). Noble gases are assumed to pass through
. the filter system unattenuated. (There is some question of the effectiveness of the filter system to remove Na 2CO3 and the fission products which may be tied up with'this aerosol component. This subject is-addressed in Appendix E.) Because the bypass leakage rate is expected to
~
be so'small relative to the high vent rate af ter 36 hours, the bypass y leakage is not expected to make a significant contribution to the released
. radioactivity and is therefore not considered beyond 36 hours.
The direct gama contribution to t'he whole body dose considers the shielding provided by the steel RCB and the concrete confinement building.
. Figure 4-1 shows the dose rate inside the reactor containment building for Case 2. .
~
O _ 4-9
~..--__ _ _ _ . .. _ .. _ _.
a l 1 CRBRP-3 - Vol. 2. Rev. 4 ,- 4 4.2.2 Radiological Doses , Using the methods described in Section 4.2.1 the radiological doses at~the Exclusion Boundary (0.42 miles) and the Low Population Zone (2.5 miles) were ' calculated for the fe';r different source terms described in Sectio'n 4.1. These doses are summarized in Table 4-3. The 30 day LPZ. doses include the plutonium released after boil-dry to 30 days. Plutonium
- release beyona 30 days could result in an additional Nm to the LPZ O dose. Control room doses were provided in Section 2.2.15. ,
Qg 30 day Theg dose consequences of the four cases that assumed varying, degrees'of severity of the hypothetical accident are all quite low for accidents beyond the design base. For example, the maximum whole body dose is
- predicted to be about M a'nd the maximum thyroid dose would be about 65 Ois rem. "-'---"":
% we.v' e 6 d e dese. 4 ab.dt G.\ w ,
The results also show that t c equences are not strongly sensitive to the degree of severity of the initial release source term. As the initial ' ' ' release to the RCB increases, the rate of aerosol depletion increases which ' N acts as an inverse feedback to limit the release from the RCB. N__ ___ A + Consequently, so long as the initial release does not result ir failure of the containment barrier, the radiological consequences are re'atively insensitive to the magnitude of the release. For the full range of releases considered in Cases 1 through 4.'the RCB pressure and temperatures would not result in failure of the containment barrier. Table 4-4 compares the consequences, in terms of curies released, of a ~ comparable scenario (core meltdown with enough containment leakage to prevent containment f ailure by overpressure) for CRBRP and light water reactors (LWR). The CRBRP values are for the worst of the above four cases. The LWR releases are for the accident scenarios PWR-6 and BWR-4 described in Section 2 of Appendix VI of WASH-1400. This comparison shows the atmospheric releases for CRBRP to be comparable to those for LWRs. Figure 4-2 shows the integrated radioactivity released to the environment for Case 2.
. . . a_ . - - . . . . . . . . . . . . . . . - . . .. .
Insert A The 2 hour dose consequences are somewhat more sensitive to the magnitude of the initial head release. This is because the aerosol depletion effects are a somewhat delayed reaction, and therefore do not reach their full effectiveness until later in the event. The 2 hour dose consequences of all four cases are still quite low. i
._=_r-_-.
_. { CRBRP-3 l Vol.2, Rev.0 l 4.4 [HE1EROGENE005 CORE CONSIDERATION hK4( 762) e llaDB radiological analyses in the previous sections are based on a ho.. geneous reactor core configuration. The dose consequences of an HC invo ng the CRBRP core (heterogeneous) are estimated to be less th those c rently predicted for the homogeneous core. Table 4-6 pro des the plutonium ventory (in kg and curies) for the homogeneous and heteroge'neou cores. Although the heterogeneous core contain ore l kilograms of p tonium, the change in isotopic content resu s in less kilograms of the sclides Pu-238,and Pu-241, which are r ologically most important. The het ogeneous core inventory of Pu-238 a factor of 5 less than the homogen 's core inventory; the Pu-241 nventory is about a factor of 4 less. When e curies of each plutoni isotope are' weighted l by the dose conversion fac rs of Reference 4-9 nd the results are added, the heterogeneous core pluton has less rad logical impact, than the ( homogeneous core plutonium. Si e the fiss' n product inventories are approximately the, same in the two res, ifferences in dose consequences depend on the contribution of pluton to the particular organ dose being n :cnsidered. In general, plutonium ibutes most to the bone dose, has a U minor role in lung and whole body ases, nd has no effect on thyroid doses. In particular, the 2 ho bone dose for cases in which the initial fuel release is greater than equal to 1% a the 30 day bone doses for all cases (because of the jor contribution of arged plutonium to the 30 day Loe dose) will decr se by approximately 50% r the heterogeneous core configuration. is is due to the high degree dependence of the l bone dose on the Pu n the reactor fuel. In general, lung and whole , body doses will .rease by 5-10% due to a lesser depende e on fuel release; these ases are primarily dependent on solid fissi product release and ble gas release, respectively. The thyroid dos is expected l to remain sentially the same, since halogen release (mostly 1 'ill be l the sam in both core configurations. Th . fore, the radiological consequences for the heterogeneous core are unded by the consequences for the homogeneous core presented herein. s O 4-13
CRBRP-3 Vol.2. Rev.0
4.5 CONCLUSION
S ON RADIOLOGICAL CONSEQUENCES O Radiological releases associated with the TMBDB accident scenario have been -l' assessed. The consequences of both atmospheric releases and groundwater releases were considered. To examine the sensitivity of the atmospheric consequences to larger releases than expected, several cases of varying degrees of severity were evaluated. The results of these analyses show the radiological dose consequences to be acceptably low and insensitive to the initial release phase over the range of releases considered. Groundwater contamination levels resulting from reactor cavity melt-through were shown to be lower than the predicted concentrations following an assumed LWR meltdown and even lower than the 10CFR20 MPC values for routine releases. W The radiological consequences reported herein were based on the homogeneous ' core. It has been shown that these results bound the consequences for the heterogeneous core. / It is concluded that the radiological consequences of a hypotheticc1 cerc disruptive accident would be acceptable considering the highly improbable nature of the conditions analyzed. I i l e 4-14
~ .'
CRBRP-3 Vol.2, Rev.0
~ ~ '. )
4.6 REFERENCES
4-1. " Reactor Safety Study - An Assessment of Accident Risks in U.S. Connercial Nuclear Power Plants, Appendix VII, Release of Radioactivity in Reactor Accidents," WASH-1400, Appendix VII, October 1975. ,
. ~
4-2. " Radiological Assessment Models, Fourth Quarterly Report, June-August 1975," GEAP-14034-4, October 1975. ~(Availability: U.S. DOE Technical Information Center).
~
4-3. " Radiological Assessment Models, Fifth Quarterly Report, . September-November 1975," GEAP-14034-5, December 1975. (Availability: U.S. DOE Technical. Information Center). 4-4. L. Baker, Jr., et al., " Post Accident Heat Removal Technology," ANL/ RAS 74-12, July 1974. (Availability: U.S. DOE Technical Information Center). 4-5. " Radiological Assessment Models, Tenth Quarterly Report, December 1976-February 1977," GEFR-14034-10, March 1977. (Availability: U.S.
' DOE Technical Information Center).
4-6. " Radiological Assessment Models, Ninth Quarterly Report, l .- September-November 1976" GEAP-14034-9, December 1976. (Availability: U.S. DOE Technical Information Center). ( , 4-7. " Clinch River Breeder Reactor Plant Environmental Report, Amendaent VIII, Section 2.6.6.1, Calculations," Docket 50-537, U.S. Nuclear Regulatory Commission Washington, D.C., February 1977. >
"CLMck Rive < Breecker Rectche< Plcmt PreHmino rg ScJeg Ancassls Report , Amendment To ,
Se ch<\ 2, 3 3 DocKe t 50- 53 7 3 U.S. Moc le cw Regal cthy Commt s s i on, LJ osh in3% , b. c. , Augas t R62. ( .' 4-15
LMOKr-3 Vol.2,Rev.0 g ReFaraoce asletad. 4-8. U.S. Nuclear Regulatory Commission, " Calculation of Annual Doses tof Man from Routine Releases of Reactor Effluents for the Purpose of et e Evaluating Compliance with 10CFR Part 50, Appendix 1," Regulatory A./ Guide 1.109, Revision 1, October 1977. { ,, 4-9. G. R. Hoenes and J. K. Soldat, " Age-Specific Radiation Dose
. Committnent Factors for a One-Year Chronic Intake," NUREG-0172, November 1977.
4-10. 'E. R. Specht, " Internal Dose Factors for COMRADEX-II," , Il-001-130-051 February 24. 1975. l 4-11. BNWL-1350, "FFTF Site Safety Analysis, Ap,)endix D - Aerosol Behavior Analysis," April 1,1970. l b. E. N nni n3 ,Tr. et al., " E sti mates o f Interna i Dose Eqv: valent to 22 Target organs Tw Racu enucGoles Occorr;n3 (n Routine Re l eases p frerrs Us6 tear Fuel- Cscle Facil;+ies " , Vciv9 e 3, V pure G / c(R- orso , octobe.c 1978. ( l Y l l t M. l ! 4-16 % l l '-- -- --- ~. -_. - _-___-
'~
- 1 CCRP-J Vol.2, Rev.0
- TABLE 4-1 . CORE SOURCE TERMS RELEASED TO THE REACTOR CONTAINMENT SUILDING FOR HYPOTHETICAL ACCIDENT SCENARIOS CONSIDERED Initial Relear* Phase Sodium Boil-Up Phase Case 1 100% Noble Gases 100% Halogens 100% Cs and Rb 100% other Volatile F.P. 15 solid F.P. 0.015% Fuel 1.1 x 106 lb. of Na 100% Noble Gases 100% Halogens Case 2 ' 100% Cs and Rb 100% other Volatile F.P. 1000 lb. of Na with 100 PPB Pu 1% solid F.P. 0.026% Fuel *, Solid F.P., 0.015% Fuel Halogens 1.1 x 106 lb.-of Na Case 3 100% Noble Geses 1% of remaining 99% of solid F.P. 100% Halogens 0.015% Fuel 100% all Volatiles 1.1 x 106 lb. of Na
~
1% Fuel * , 1% Solid F.P. 1000 lb. of Na Case 4 100% Noble Gases 1% of remaining 95% of solid F.P. 100% Halogens 0.015% Fuel 100% all Volatiles 1.1 x 106 lb. of Na ' 5% Fuel
- 5% Solid F.P.
3300 lb. of Na Note: After boil-dry the only significant contribution to the source term is plutonium release due to gas sparging. This additional source amounts to about Q3 ka) of plutonium released from the molten pool, which has
/6been assumed to be freely transmitted to the RCB over a several month N 9*5 period and is considered the same for all four cases.
*Incluaes plutonium in blankets and core.
4-17
CR8RP-3 V;1.2 Hav.0 TABLE 4-2
- r-ATMO5PHERIC DILUTION FACTORS , ,s SOE X/Q Values
- I 3
Exclusion Boundary (0.42 miles) X/0 (sec/M ) 0-2 hours f2.20 x 10-4)T l.o I xio'3 ( Low Population Zone (2.5 miles) 0-2 6ves l.Sclxto-4 2 4 8 hours 2.306 x 10-5 8-24 hours 3.5T % x 10-6 , 1-4 days 2 20.M x 10-b 4-30 days 2. bog x 10-6
*5ee Section 2.3 of the CRBRP PSAR.
0 I 4-18
~
CRBRP-3 . Vol. 2. Rev. 4 - O . TABLE 4-3 DOSE Stif1ARY FOR HYPOTHETICAI, ACCIDENT SCENARIOS CONSIDERED
. Doses in REM an Case 1 Case 2 Case 3 Case 4 Bon 0.0043 0.028 0.93 3.83
~
2 Hour Lung 0.0035 0.0055 0.1 0.39 Exclusion Thyroid 0.0067 0.0096 1 9.51 Bound.ary W. Body 0.16 0.16 .24 0.32 30 Day Bone 55.1 55.1 55.7 56.2 Low Lung 3.95' 3. 3.02 3.07. Population Thyroid 99.2 .2 5.31 1.72
~
Zonc W. Body 3.51 ' 50
. 3.07 2.94 ,4 De\e% d vep%e tddk ressed tcMe 4-3 3
TABLE 4-3 DOSE
SUMMARY
FOR HYPOTHETICAL ACCIDENT SCENARIOS CONSIDERED Doses in REM Organ Case 1 Case 2 Case 3 Case 4 Bone Surface O 027 0.19 6.47 27.0 Red Bone Marrow 0.026 0.040 0.56 2,18 2 Hour Lung 0.020 0.032 0.72 1.77 Exclusion Liver 0.052 0.060 0.44 1.21
~
Boundary Thyroid 0.014 0.020 23.4 19.6 Whole Body 0.8E 0.82 1.09 1.22 Bone Surface 0.92 0.95 2.45 6.07 30 Oay Red Bone Harrow 0.19 0.19 0.27 0.56 Low Lung 1.54 1.55 0.82 1.00 Population Liver 0.36 0.36 0.18 0.32 Zone Thyroid 85.3 85.4 8.13 5.43 Whole Body 2.10 2.09 1.73 1.65
CRSRP-3 .- . . Vol.2. Rev.0 . .
. m,
._J TABLE 4-4 C0t?ARISON OF RADIORJCLIDE RELEASES TO ATMOSPHERE FOR CRORP WITH LWR'S FOR A COMPARABLE MELTDOWN SCENARIO .
Radioactivity Released (curies) . Element CR8RP PWR(3) gwR(3) Xe-Kr 3.4.T x 107 1.0 x 108 2.1 x 108 I . 2.1% x 105 2.0 x 106 1.1 x 106 .
~
Cs, Rb ' 5.2% x IM l'.2 x 104 7.6 x 104 . Te, $b 1lJ @ x 104 2.2 x 105 8.6 x 105 ~ Ba,Sr 1.5 @ x 102 3.3 x 104 2.2 x 105 .. . Ru(1) 2.E C x 103 3.9 x 104 3.3 x 105 , La(2) q)gg x 103 2.9 x 104 2.9 x 105 III Includes: Ru, Rh, Co, Mo, Tc,
~
I2} Includes: Y La, Ir, Nb, Ce, Pr, Nd, Np, Pu, Am, Cm
.. I (3)from WASH-1400 Appendix VI, Calculation of Reactor ' Accident
' Consequences, October 1975. The LWR scenarios used for comparison here are PWR-S and BWR-4 described in Section 2 l ~
~
! of WASH-1400 Appendix VI. 4-20
____c- __ CRBRP-3 o Vol.2, Rev.0 TABLE 4-6 PLUTONIUM INVENTORY IN CORE AND BLANKETS Homogeneous Core - Beginning of Equilibrium Cycle (Used in TMBDB Radiological Heterogen s Core . Isotope Analyses) End of Equi brium Cycle E.2 E K.2 S Pu-238 16 2.70 x 105 3 5.07 x 104 Pu-239 64 8.97 x 104 2 1.39 x 105 Pu-240 37 8.23 x 104 220 4.84 x 104 Pu-241 131 1.33 x 107 32 3.25 x 106 Pu-242 45 . 6 x 102 3 1.17 x 101 Notes:
- 1. The ho:noger.cous core analyses asst. .d the isotopic content of the feed plutonium is 0.8% Pu-238, 72.1% P '9,18.4% Pu-240, 6.5% Pu-241 and 2.2% Pu-242. The isotopic cont to the feed plutonium for the heterogeneous core is 0.1% Pu 8, 86. Pu-239, 11.7% Pu-240, 2.0%
Pu-241 and 0.2% Pu-242. ,
- 2. For the homogeneous core, he limiting dos are associated with the beginning of equilibri , cycle plutonium inv tory. For the heterogeneous core, th limiting doses are as iated with the end of equilibrium cycle pl onium inventory.
De\ete %% C 4-22
CRBRP-3 . . .
~
. . Vol. 2. Rsv. 4 . .
ggs M M . 0 7 _ e - E E
=
5 5 - p 105 i-
. -- 1 1
~
De e e m- ~~ W - Pepkke.uhh . k L Ihk y' s C b. o In S E - O l > -
=
G
- 4 10 e
I I I I I I I I I I I i 4 0 10 20 30 40 50 60 70 80 00 100 120 720 TIME (HOURS) Figure 4 2. Integrated Radioactisity Released to Environment for Case 2 l l 73081 {
=
ke.J eb N\3 ore k~2-.' ff--
~
7 10 - e g - O
- 5>
" 10 6 _
= 2 2.
e
= -
g - s n 5 10 5 __ a _
*g -
= _ ,
ii . . 4 10 103 I I I I I I I I I I I le <
/ /
0 10 20 30 40 50 60 70 80 90 100 120 130 720 TIME (HOURS) Figure 4-2. Integrated Radioactivity Releaed to Emironment for Caw 2 7M.W l
CGRP-3 Vol.2. Rev.0 ! pool of sodium was collected and analyzed for plutonium content. For the sodium containing 'Pu0
, 2
, nine experiments resulted in plutonium release fractions ranging from 1 x 10-6 to 8.7 x 10-5 The average release fraction was 2.0 x 10-5 For sodium-plutooste, the preliminary results I indicate the release fractions are several erders of magnitude less than for Pu0 2
The fractional release of plutonium frm burning sodium used in these analyses is 3 x 10-5 This is consistent with the data in References E-4 and E-5 and may be highly censervative If a substantial amount of the plutonium is in the form of sodium plutonate. If the total amount of oxygen present in the RC and three PHTS cells completely reacts with sodium, the amount of sodium burned would be 1785 pounds. ,
. Based on the particle distributions from the M-Series tests a't ANL, about .
15% of the fuel would be in small enough particulates to be suspended in the sodium. If it is assumed that 15% of the total plutonium inventory @2o% Kg) is uniformly distributed throughout the total primary sodin inventory (1.1 x 106 lbs) then the amount of plutonium contained in the burning ., sodium is (0.15) Mg) (1785/1.1 x 10 ) . gg$s. Applying the 3 x 6 ( 10-5 release fraction gives a plutonium re' lease of 0.015 grams to the RCB. This is an insignificant amount compared to that assmed in the l initial release phase source term. 1 E.3 PLUTONIUM RELEASE FROM A BOILING POOL OF S0DIUM The primary sodium which drains out of the reactor vessel and primary piping .
~ forms a pool in the reactor cavity. The pool contains the fuel debris from the core. The interaction of molten fuel with sodium results in a fuel particle distribution. Based on measurements of particle size distributions l in the ANL M-series tests (Reference E-3), approximately 15% of the fuel could exist in particles small enough to remain in suspension in the sodium pool. The remaining fuel would form a settled bed on the bottom of the sodium pool.
( l l l l E-2 ,
~- %-- - - - ,- - _ , _ ,
; . CRBRp-3 D1.2. Rev.0 The fuel debris heats the sodium to boiling due to decay heat. As the sodium ."', J pool boils, plutonium is released from the pool by vaporization of the plutonium and plutonium particle entrainment in sodium droplets carried from the pool by the sodium vapor. Basedonarecentsurvey(ReferenceE-4)of experimental data on liquid carry-over from commercial evaporators and entrainment of solid particles in the, vapor strees from an evaporating liquid
! pool, it was concluded that the experimental data of Jordan and Ozawa (Reference E-5) is most directly applicable for estimating the potential plutonium release from a boiling pool of sodium. Their results show a minimum l , decontamination factor of about 1000. Thus if 15% of fuel is suspended in the sodium and if 1/1000 of this is released, then the net fuel release is 0.015% of the total inventory. Since there are approximate 1yhNf plutonita in the CRBRP core the plutonium release from the boiling sodium pool to the RC8 sao atmosphere could be about grams. However, because of aerosol depletion and filtering less than one gram would actually be released from the RCB
~
cleanup system. - once the fuel and core debris penetrates the RC liner, carbon dioxide and s steam would be released and bubble through the sodium pool. The effect of this gas sparging on additional plutonium and fission product release has been investigated by Parsly and Fontana (Ref,erence E-6). (Dsing the model and?The anocle.k
-40trtht'; . ;;dfici; ,t: Of ";fe.ec.ce M the effect of gas sparging on '[**((
Most of the fission was used, addition'la releases from CRBRP .... .... ___....... one. of products fall into 3two categories: (1) those with predicted high fractional io d eterm. releases due to sparging, which are the more volatile products that have already been considered totally released either initially or during sodium boiling, (2) those with low release fractions which are conservatively covered by the assumed 1% solid fission product release during the sodium boiling phase. A few isotopes do have sparging release fractions larger than the 1% used in the TISDB analysis; however, these represent a less significant contribution to the dose consequences than does the increased plutonium i release (Reference E-7). Additional plutonium release during the boil-up phase due to sparging is not significant relative to that resulting from the boiling sodium pool. Sparging effects af ter boildry were found to be more significant as discussed below. l E-3
CRSRP-3 Dory b ,h M h U h erQ VMe Rev.0 c.ommuni tatA. E.4 PLUTDNIUM RELEASE AFTER SODIUM BOILDRY Tvtseek h g ( .n__ M ': u. _a,._ ___, 1. . u. _ . . . _ _ _ . . . , t._ . .___..
': t'% r'::n:Cis assumed between the reactor cavity and t.__ m the reactor containment building. This results in a natural convection current through the cavity. The potential mass transport of plutonium via this convection depends on the debris temperature and the convection veloc ity. The debris surface temperature at the top of the crust was determined to be #2500'F. The methods and assunctions related to the calculation of bed surf ace temperature are discussed in Section 3.2.3.1 of this report. The convection veloc'ity was calculated as follows: First, the natural circulation heat transfer rate was calculated (for a specified Pu02 surf ace temperature) using the correlation in Reference E-8 for heat transfer from a hot disk to a large volume of gas. Next, an expression for the temperature rise of the circulating,ga(was derived as a func ion of gas flow rate. Finally, expressions for th.e' form pressure loss were derived and set equal to an expression for the buoyancy pressure driving forces (acceleration and frictional losses were conservatively ignored). This last expression, which contained only flow rate as unknown, was solved and the maximum velocity was calculated as that through the opening between the reactor cavity and the RCB. ,
The analysis of the convection velocity used the following assumptions: 0
'1. The temperature of the debris surf ace was 5500 F. This is highly conservative for the calculation of the convection velocity since the
* ~
j predicted surf ace temperature is #2500 F.
- 2. The temperature of the gas in the reactor containment building (RCB) was 500 F.
*Because mass transfer from the debris surf ace is a function of the convective velocities within the reactor cavity, a 30000F margin was added to the debris temperature to enable the calculation of a higher than expected mass l
j transfer coefficient. This coefficient, when applied at the predicted condition, will result in higher mass transfer rates, providing margin in the calculation. l 4 E-4
;7
. :r s y
< k f
Insert B , s r After the sodium pool in the reactor, cavity has evaporated, a bare fuel / steel debris melt is left. Th'is debris melt continues to decompose the concrete beneath it, releasing carbon dioxide and steam which bubbles through the pool. The remaining concrete constituents , melt and become mixed with the fuel / steel melt as a result of the bubbling action. , The conditions described above give rise to several possible modes of release of plutonium which have been investibated, fThese are: (a) plutonium vapor transport, (b) plutonium particle levitation, and (c)plutoniumgassparging, Each of these plutenium release' m6 des wi1Y be addressed in the following paragraphs. O e 4 9 1 s a l 1 .' l l
'} * >
x k
+
t CRBRP-3 Vol.2. Rev.0
- 3. The temperature 'of the air outside the RCB was 7QDF.
O aa was uniformly distributed over the bottom of the reactor 4 The l
;- cavity.
s
- 5. The natural circulation velocity was based on the open reactor cavity (no reactor head) with the vessel in place.
The resulting peak natural convection velocity is 12 ft/sec. ' ^--- :- '
~
__,.. ___ ___.as._., r__ ,i__,_ ,,- n. _; ,.~>_<(_,i._w_ m . _ _ _ _ _ _ _ . . , ,___ st_ ___ir_,__ __; .- ir_*. g_. ii i-rielem iWLbedeNS The plutonium vapor transportg mode is due to the plutonium.,trying to establish a vapor pressure corresponding to the surface -- '- t pture and the r is est411ske.5 convection current removing the vapor L_m <>m. ...% oncentration gra'dient dowr, which the vapor molecules move. This subject of mass transfer is treated in Reference E-9. Based on a maximum convection velocity of 12 f t/sec and maximum bed surface temperature of 25000F the vapor removal rate
- was cniculated to, tie only 5 x 10-5 grams / hour. At this rate the amount of
..e plutonium released to the RCB in a period of one month is only 0.04 grams.
Plutonium remeval by particles being physically swept up from the surface of the bed by the convection current (levitation) was also considered. The surface cf the debris would be solid, i.e., covered with a layer of steel, or in the form of molten iron oxide. In either case, plutonium particles are not present to be picked up by the convection current. Even if the debris were composed of plutonium particles in a molten steel pool, a velocity of caly 12 ft/ set would not be sufficient to detach particles from the liquid steel. byke / g i m. 1,
=
<.._ z ,...._>
, m C -> .. _,u.
._2:;_ , __; m -
.,_ _-_,_; ,_ . , - > ,2. .. _ ..
- , .. 2 ..... wu_.- ...s__. . <---. .. .-- :- ' - - '2 . Mi :- -
O E-5
Insert C The effect of gas sparging after boildry has also been evaluated. With the molten sodium gone, the dilution of the core materials is reduced, tending to increase the amount of sparging. .The molten con-crete, however, does tend to continue to dilute the core melt, thus reducing the vapor pressure of the dissolved plutonium and reducing somewhat the sparging that could occur. Using the model of reference E-6 as before, coupled with data from reference E-14 to evaluate the released fraction of plutonium, a sparged plutonium mass of 23 grams is predicted during the first 30 days of the event. This amount 4s based on a highly l conservative temperature of 4500*F for the molten concrete and core debris mixture. More realistic temperatures lower than 4500*F would greatly reduce the sparged mass. Beyond 30 days, another 3 grams of plutonium could be sparged. t The 26 grams of plutonium sparged during the entire course of the event would be reduced by aerosol effects within the RCB such that less than half would be released to the TMBDB cleanup system. The release to the environment would be reduced by another factor of ~100 by the 99% efficient cleanup system, leaving only about one-tenth of one gram of sparged plutonium to escape into the environment during the course of . the event. l i L __ -
CO RP-3 Jol.2,Rev.0_ g plutonium release greater than that by all mechanisms before bolldry. The -
- results of the sparging analysis indicate that as much asI l0 kg of plutonium '
could be released to the RC8 in 30 days and an additional 3 kg over the next i several months. An HAA-3 analysis shows that about 70% of the sparged ; I plutonium will remain inside the RC8 due to aerosol plate-out and fallout. Filtering (99% efficiency) further reduces the amount of plutonium released by ! a factor of 100. i E.5 CONCLUSIONS ON PLUTONIUM RELEASE An evaluation of potential plutonium release from the reactor cavity to the RCB during sodium boiling and following boildry has been made. This evaluation indicates that,'in addition to that considered in the initial release phase, about M ra im of plutonium could be released from the boiling sodium, and about @ P.S'i M ;;r: ' frtrd could be released to the RCS over a several month period following boildry due to gas sparging. C 2(, cyc4ms E.6 FISSION PR0 DUCT RELEASE TO THE RC8 r The overall release fraction of fission products from the fuel to the RCB atmosphere is the product of the release fraction from the molten fuel and the release fraction from the sodium pool. References E-10 and E-11 present the results of an evaluation of existing experimental and theoretical data on the volatility of elements in molten fuel. This study developed a list of volatility factors for fission products. These factors represent conservative
-- estimates of the percent release of elements from molten fuel. The volatility of those elements in the categories of noble gases and halogens is 100%.
'Those elements in the category of volatile fission products have volatility factors of 90%. The remaining fission products, in the category of solid fission products, have volatility f actors of 4% or less. Of those in the 4%
group only a few are present in the CRBRP E0EC fission product inventory in sufficient quantities to be significant. Strontium and barium have f actors of
- 25. All other fission products have f actors of 1%. These release fractions from molten fuel are sumarized in Table E-1.
i + 1 l E-6 1
CME ~3 Vs1.2. Rev.0 CO reacted completely with either Na or NaOH, the resulting Ha CO3 2 2 would be no more than 8% of the total aerosol products. This amount of sodium O c'arbonate would not have a significant offect on the overall aerosol behavior and amount of radioactivity reaching the filter system. cause of the limited solubility of sodium carbonate in water, the filter - r 1 efficiency for Na CO3 for a wet filter / scrubber system could be 2 , somewh lower than for the NaOH. If the maximum amount of Na CO 2 3 possible formed and if the removal efficiency of the filter is substantially er (f actor of 10) for Na CO 2 3, then the amount sodium getting through filters and released to the atmosphere uld exceed that predicted in the cur t analysis by about 70%. If it assumed that the other radioactive speci (except noble gases) are nsported with sodium (independent of the chemica form of the sodium) he released radioactivity of these products could then a increase b out 70%. This shows that the radiological results are not high sensi e to the effect of CO2 reactions with airborne NaOH. The effectiveness of the TMBDB ter in re ing Na2 CO 3 will be determined as part of the TP Air Cleaning Sy m Performance Test to be performed at the Contain t System Test Facility the Hanford Engineering Development Laborator . Appendix A.7 provides a descr +1on of these tests. Should these test how the removal efficiency for Na2 3 be comparable to that expec for other aerosol products (99%) then it can concluded , that the f . ation of Na2CO3 presents no increased radiological consequ ce. If however the tests show a significantly lower remov effi ency for Na 00 2 3, then a more detailed evaluation of the extent o 3 C0, formation would be required to assess the radiological impacts. A oe.e m um bM D_ E-9
l Insert . .
\
e Because of the limited solubility of Na CO 2 3in water, more water must be used in a wet filter / scrubber system to remove Na2CO3 than would be needed to remove pure NaOH. Results of the TMBDB Air CTeaning System Performance Tests as described in Appendix A.7 show that removal efficiencies for Na C0 7 3 are comparable to those for other aerosol products (99% or greater), when sufficient quantities of water are used. i Thus, the formation of Na 2CO3 presents no increased radiological consequence. e l l y - - , - -,-- --p----e-- , - - - - , ,m ,m, , , - - -
\
' ~
CRBRP-3 Vol.2 Rev.0 . E-9 R. B. Bird. W. E. Stewart and E. N. Lightfoot, T'ransport Phenomena John
~
Wiley and Sons, New York,1962. E-10 " Radiological Assessment Models, Fourth Quarterly Report June-August 1975 " GEAP-14034-4, October 197,5. (Availability: U.S. 00E Technical Infomation Center). E-11 " Radiological Assessment Models Fif th Quarterly Report. September-November 1975 " GEAP-14034-5, December 1975. (Availability: U.S. DOE Technical Information Cerfter). E-12 A. W. Castleman and'I. N. Tang, " Fission Product Vaporization from Sodium . Systems," BNL-13099, Nove.cber 1968.
- E-13 G. W. Keilho'Itz and G. C. Battle, Jr., " Fission Product Release and Transport in Liquid Metal Fast Breeder Reactors," ORNL-NSIC-37, March 1969.
E- 14 6ctbel n ck3 1 b. and Chasanov ri.6. 3 A Calcula tionn i ' 3 Appecack to the Estirncstion of Fved anct F.srion Produc.t V'o per Pressures anA O s.id ctt;m Stater tu 6 coo K', As> L- 7E(.'7 (oc t ober ,ici i2 ). -
.t I
E-11
~-
CRBRP-3
- Vol.2, Rey,0 G.4 EFFECT OF INITIAL HEAD RELEASE ON TFBDB DOSE CONSEQUENCES AND RCB ATMOSPHERE CONDITIONS O G.4.1 Introduction Section 4 discussed the dose consequences of four cases, representing varying degrees of innediate leakage through the head. These cases covered initial fuel and fission products released from 0 to 5% along with the sodium release )
expected to accompany such fuel releases. The analysis presented here extends - the range of these four cases and evaluates the sensitivity of the dose consequences to (1) changes in fuel release without an accompanying change in
- sodium release and (2) changes in sodium release for a fixed fuel release.
Additional analyses were performed to evaluate the RC8 atmosphere response, and therefore containment integrity, to these various releases.
~
G.4.2 Radiological Consequences Table G.4-1 gives the dose consequences of three cases of initial release of fuel, fission products, and associated sodium assuming containment integrity O (containment integrity is addressed in Section G.4.3). The results indicate that, for releases beyond 5%, the mitigating effect of aerosol depletion is stronger than the effect a larger initial source term might have on increasing tial releases is best seen by. the dose. The overall effect o{ var, igg % to 5% fuel release, theme looking at the 2 hour @ doses. From 1
~
increases about in proportion to the increase in the fuel release. However, y at higher releases, the attenuating effect of the larger sodium releases exceeds the effect of the larger quantity of fuel and the overall 2 hour $ dose decreases as noted in the table. hme/webe. The aerosol depletion effect alone of there releases can be seen by comparison of the thyroid doses. The thyroid dose is the result of the halogen release which is 100% for each of these cases. As the initial source term of fuel and sodium increases, the aerosol depletion rate increases which causes more of the halogen source to f all out leaving less av611able to leak out. O G.4-1
CCRP-3
. Jol,.2,,Rev.0 .
Table G.4-2 'shows the dose consequences of.from ,s ., , 1 to 50% fuel and fission.- ' products released initially without the accompaniment of .the sodium which Here the 2 woulj be expected to be associated with such initial releases.
' M r N c:oc:: r;seen to be in direct proportion to the percent fuel released. The 2 hour thyroid doses show no benefit from aerosol depletion N
during the first two hours. The results in Table G.4-2 show that even for a 50% initial release, with minimal credit for aerosol depl'etion from sodium, the dose consequences are At'::t C vget exce u,'we Ikoew kwwd Finally, Table G.4-3 shows the effect of reduci he sodium in the initial release for the case of a 101 fuel and fission (product release through the l head' assuming an integral contatriment. The" r~; case of no sodium is A $ compared' to the case with 1,000 pounds of sodium. There isgdifference in the 2 hour doses between the 0 and 1,000 pounds of sodiun cases because there l l 1s little aerosol depletion during the first two hours. The 30 day doses are slightly higher with no sodium but are not highly sensitive to the aerosol l effects of the initial release. This is due to the 30 day doses being more dependent on the boilup phase ' ;-- ; ;'; ;' _ : of the scenario. The overall conclusion to be drawn from the results of this analysis is that .
~
the TMBDB scenario is not sensitive to a wide range of initial releases through the head with an integral containment barrier. , , G.4.3 RCB Atmosphere Response To Initial Head Release CACECO Code Model - The CACECO Code model, described in Section 3.2 and Appendix C.1, was modified to simulate the initial release of sodium and fuel through the head. To simulate the initial release of fuel and fission products, the latent heats of the fuel and various fission products were calculated up to the fuel vapor point. The energy calculated was then superimposed on top of the decay curve for a ten second interval. The appropriate partitioning factors were then calculated (i.e., code input that distributes decay heat to a designated cell). Similarly the sodium at 10000F was assumed to be injected to containment and burn completely within a ten second interval. ( l G.4-2 _ ._ = . _ _ - . . .-
~
' CRBRP-3 -
Vol. 2, Rev. 4-
- TABLE 6.4-1 EF CT'0F INITIAL FUEL, FISSION PRODUCT, AND SODIUM RELEASE *-
Dose (rem)
. 1% Fuel & F.P. 5% Fuel & F.P. 7.5% Fu .& f.P.
1000 lb. Na 3300 lb. Na 7000 b. ha Bone 0.93 3.83 . 3.12 . 2 Hour Lung ~ 0.15 0.39 0.30 EB Thyroid 11.3' 9.51 5.19 i Whole Body 0 4 0.32 0.28 Bone 55.7 56.2 55.8 30 Day Lung 3.02 3 2 2.91 - LPZ Thyroid 5.31 1.72 0.83 Whole body 3'.07 2.94 2.90 4 RCB .mosphere ditions** Temperature (DF) 270 580 1030 Pressure (psig) 6 12.7 24.4
- Initial release of noble ases, halogens. and volatile f sion products to yRCB = 100%.
** Peak values for 1 ho .
Dekc. C P CC um cea ta& GA-\ l l l C
TABLE G.4-1 EFFECT OF INITIAL FUEL, FISSION PRODUCT AND SODIUM RELEASE
- Dose (rem) 1% Fuel & F.P. 55 Fuel & F.P. 7.5% Fuel & F.P.
1000 lb. Na 3300 lb. Na 7000 lb Na Bone Surface 6.47 27.0 22.0 Red Marrow 0.56 2.18 1.77 2 Hour Lung .72 1.77 1.42 ' EB Liver 0.18 1.21 0.95 Thyroid 23.4 19.6 10.7 Whole Body 1.09 1.22 1.05 Bone Surface 2.45 6.07 4.90 Red Marrow 0.27 0.56 0.45 30 Day Lung 0.82 1.00 0.88 LPZ Liver 0.18 , 0.32 0.25 Thyroid 8.13 5.43 2.91 Whole Body 1.73 1.65 1.60 t RCB Atmosphere Conditions ** Temperature (*F) 270 580 1030 i i Pressure (psig) 4.6 12.7 24.4
- Initial release of noble gases, halogens and volatile fission products to RCB = 100%.
** Peak values for 1 hour, i
il
A ^ A em. m-- -- m _-m
. CRBRP-3
- Vol. 2. Rev. 4 .
TABLE G.4-2 , EFFECT OF INITIAL FUEL AND FISSION PRODUCT RELEASE ' WITH 50DILM RELEASE FIXED AT 1000 LB.
- Dose (rem) ,
i Initial Release of Fuel and Fiss n Products
- 5% 1 50%
1% 0.93 4.56 9.10 45.4 Bone 0.15 0.46 0.85 3.96 2 Hour Lung 11.3 11.3 11.3 11.3 i EB Thyroid
~
Whole' Body' .24 .35 -0.48 1.53 55.7 58.5 60.6 83.7 Bone l 3.02 3.25 3.35 5.37 30 Day Lung 5.31 4.43 3.53 3.53 LPZ Thyroid ,
- 3. 3.10 3.11 3.68 4 Whole Body RCB At phere Conditions **
r 270 290 310 450 Temperature (DF) 4.6 5.0 5.7 9.5 Pressure (psig) l
- Initial releas' of noble gases, halogens, and volatil fission products to RCB = 100%.
** Peak valu for 1 hour.
De4 cud reda.ce ~ ed\ weuised tokie GA-2_ O . e
TABLE G.4-2 EFFECT OF INITIAL FUEL AND FISSION PROD 0CT RELEASE WITH SODIUM RELEASE FIXED AT 1000 LB. Dose (ren) Initial Release c/ Fuel and Fission Products * ) 1% 5% 10% 50% ! Bone Surface 6.47 32.1 64.2 320.8 . Red Marrow 0.56 2.60 5.15 25.5 Lung 0.72 2.19 3.93 17.8 EB Liver 0.18 1.44 2.70 12.7 Thyroid 23.4 23.3 23.3 23.3 Whole Body 1.09 1.30 1.57 3.71 Bone Surface 2.45 E.17 14.6 70.2 Red Marrow 0.27 0.72 1.24 5.68 Lung 0.82 1.15 1.49 4.75 p Liver 0.18 0.42 0.70 3.10 Thyroid 8.13 7.67 7.21 7.21 Whole Body 1.73 1.75 1.77 2.20 l l RCB Atmosphere Conditions ** Temperature (*F) 270 290 310 450 i i Pressure (psig) 4.6 5.0 5.7 9.5 l
- Initial release of noble gases, halogens and volatile fission products to RCB = 100%.
** Peak values for 1 hour.
~
- CRBRP-3 ,
Yol. 2. Rev. 4 . C TABLE G.4-3 . EFFECT
- SODIUM RELEASE FOR A GIVEN 10% FUEL-FISSION PRODUCT'RELE Dose (rem)
Pounds of Sodium in Initial Release E '1000 Bone .
. 9.10 g,10 .
2 Hour Lung 0.84 0.85 l , EB Thyroid 11.2 11. . Whole Body . 0.48 'og . Bone 5 60.6 30 Day Lung 3. 3.35 LPI Thyroid 5.28 3.53 Whole Body 3.26 3.11 4 RCB Atmosph e Condition
- i Temperature (OF) 60 Pressure (psig) 4.6 5.
- Initial release of nobl ases, halogens, and volatile fi Yonproductsto f RCB = 100%.
** Peak value for 1 ho .
C dtlA DG d c h k r e.v h e d h W G . 4 - 3 C C.4-6 ( -. - _ - ,__ .. -
d TABLE G.4-3 EFFECT OF SODIUM RELEASE FOR A GIVEN 10% FUEL-FISSION PRODUCT RELEASE
- Dose (rem)
Pounds of Sodium in Initial Release 0 1000 Bone Surface 63.4 64.2 2 Hour - Red Marrow 5.09 5.15 EB Lung 3.90 3.93 Liver 2.66 2.70 , Thyroid 23.1 23.3 Whole Body 1.56 1.57 Bone Surface 16.3 14.6 - 30 Day . Red Marrow 1.37 1.24 l LPZ Lung 1 . 61 1.49 Liver 0.78 0.70 Thyroid 8.05 7.21 Whole Body 1.84 1.77 l l RCB Atmosphere Conditions ** Temperature (*F) 260 310 Pressure (psig) 4.6 5.7
- Initial release of noble gases, halogens and volatile fission products to RCB = 100%.
** Peak value for 1 hour.
m I . e Enclosure 5 l O l ) i
Description:
Please provide a similitude analysis of the HEDL CSTF cleanup system and show how system performance changes when scaled up to the CRBRP containment cleanup system. Response: Theoretical correlations currently exist in the technical literature which estimate the filtration performance of components similar to those found in the CRBRP Containment Cleanup System. These correlations are referenced in Chapter 6.3 of HEDL TME 81-1. One of the objectives of the CSTF test at HEDL was to verify the applicability of these correlations to sodium aerosol environments. Table 43 and Figure 39 of HEDL TME 81-1 shows that the theoretical modeling of the HEDL CSTF system (using these correlations) underpredicted the actual measured performance and that the system performance substantially exceeded the specifications. On the basis of the foregoing, the applicability of the theoretical correlations has been established and the actual size (scaled-up) CRBRP Containment Cleanup System was evaluated to determine its estimated performance during the expected operating range of the system using the theoretical correlations. The evaluation indicated thpt the predicted performance of all CRBRP Containment Cleanup System components throughout the operating range of the system is expected to be even higher than the actual performance of the HEDL CSTF cleanup system. On the basis of the above, it could be concluded that the specified performance of the CRBRP Containment Cleanup System is very conservative and it is expected that the system will well exceed the specified perform-ance requirements. The comparison of the CRBRP vs. HEDL system parameters affecting the performance of the system is provided in the attached Table 1. l
TABLE 1 COMPONENT / PARAMETER HEDL CSTF CRBRE Ouanch Tank Gas Flow Rate Max. /s 0.663 25 Gas Flow Rate Min. m /s 0.073 16.6 Liquid Flow 1/s 0.125 31.5 Spray Fall Height m 3.35 5.80 Spray Drop Diameter m 0.00114 0.00114 Relative Particle Velocity m/s 12 12 Particle Diameter Max. microns 6.40 9.4 Particle Diameter Min, microns 1.80 4.6 Residence Time sec. 6.8 6.3 Venturi Scrubber Cas Flow Rate Max. m /s 0.612 11.3 Gas Flow Rate3 Min. m /s 0.056 10.3 Liquid Flow m /s 0.225 63 Venturi Height m 1.71 3.05 Spray Drop Diameter m 0.0014 0.0014 Relative Particle Velocity m/s 1 1 Particle Diameter Max. microns 5.0 9.4 Particle Diameter Min. microns 1.85 4.6 Fiber Bed Scrubber Fiber Fill Coeff. 0.15 0.15 Fiber Bed Thickness em 7.62 7.62 Fiber Diameter cm 0.002 0.002 3 21.2 Face Velocity Max. cm /s 22.2 Face Velocity Min. cm /s 2.1 16.7 Perticle Diameter Max. microns 3.65 9.4
.atticle Diameter Min. microns 0.64 4.6
~
I
Description:
During the presentation of the " dynamic analysis" model of the cleanup system by Peter Fazekas (B&R), equations from the BEDL air cleaning reports were referenced. In HEDL-THE-81-1, there are several equations to correlate pressure differential, gas cooling, and aerosol removal efficiency (e.g., 97, pg. 87, 98, pg. 90, 99, pg. 92, ell, pg. 94, #17, pg. 100, #18, pg. 105) . Many of these equations are based on specific conditions used in the HEDL system. What equations have been used in the CRBR, TMBDB dynamic analysis? Response: The purpose of the CRBRP Containment Cleanup System dynamic analysis was to determine the thermal / hydraulic performance of the scaled-up system. It was not the objective of the analysis to scale-up the filtration' efficiency of the system or to determine the pressure drop thru the staled-up components. 1 The correlations for the containment cleanup system dynamic analysis utilized the test results from the HEDL CSTF AC-1 thru AC-4 tests and they were the following: T QGO = 0.333 L/G + KQ SO (1) and i TvGO " TvSO (2) Where: TQGO = Gas tegp. out of Quench tank ( F) L/G = Liquid gas ratio (SCFM/GPM) ( T o QSO = Quench S luti n temp 3 )ut of tank (F l ! l TVGO = Ggs temp. out of Venturi j ( F) . T Solution ygg = Venturi (gemp. F) out of The dynamic analysis indicated that correlation (1) is valid for cases when the residence time in the quench tank is the same; therefore the CRBRP containment cleanup system quench tank was sized to yield the same residence time at maximum gas
l l l flow as of the HEDL CSTF system quench tank at maximum gas flow. All other equations used in the CRBRP dynamic analysis are fundamental equations. No equation was used from HEDL TME-81-1 for the CRBRP dynamic analysis, s l l l
f D e Enclosure 6
Question CS760.144: . For evaluating containment response to thennal loads generated during the TMBDB scenario, the applicant assumes an axisymmetric distribution of temperatures. Based on the location of the reactor cavity vent system, this does not seem to be a consistent assumption. The applicant needs to either rigorously justify the axisymetric assumption or analyze the containment response to non-axisymmetric thermal distributions. -
Response
- 1. Summary During a TMBDB scenario, the atmosphere is heated by a sodium / hydrogen flame. This heat is transferred to the containment steel shell and then to the annulus cooling system air.
An analysis has been performed to assess the three dimensional temper-ature variations in the RCB atmosphere as a result of the off-center location of the flame. The azimuthal temperature variations of the atmosphere adjacent to the steel dome were detendined 10 assess the potential magnitude of the difference from an axisymmetric assumption. It was concluded that azimuthal variations would not be significant to the containment analyses.
- 2. Analysis A finite diffe'rence fluid flow analysis computer code, TEMPEST (Reference 1), was used to compute the steady state flow fields in a three dimensional model of containment. The boundaries (steel shell and operating floor) were set to 250'F and a heat source of 60 MBtu/hr (Reference 2. Figure 2-8) was applied to a volume of gas 20' x 20' x 60' high located 35' off-center. These conditions were considered to produce representative flow patterns in the containment.
- 3. Results The minimum and maximum containment attrosphere temperatures occurring adjacent to the containment shell were calculated at.each elevation.
The lower 100 feet show an azimuthal variation of less than 10*F. Near the top of containment where the hot gases impinge on the roof the difference is greater, with a maximum azimuthal difference in atmosphere temperature of 91"F at the 35 ft. radius.
- 4. Conclusions
- The azimuthal temperature variation in the lower 100 feet of contain-ment is negligible.
QCS760.144-1
- The stresses associated with the temperature variation near the top of the containment dome have not been analyzed, however, the compound curvature of the dome should be able to adjust to the thermal expan-sion without producing significant stresses.
Therefore, axisymmetric analyses appear to be a valid method of i approximating the themal and stress loads. References l l 1. D. S. Trent, M. J. Budden and L. L. Eyler, " TEMPEST, A Three Dimen-sional Time Dependent Computer Program for !!ydrothemal Analysis," l Pacific Northwest Laboratory, FATE-80-ll4, January 1981.
- 2. "flypothetical Core Disruptive Accident Considerations in CRBRP,"
Volurie 2: Assessment of Thermal Margin Beyond the Design Base," CRBRP-3, March 1980. QCS760.144-2 l l -}}