ML20004F235
| ML20004F235 | |
| Person / Time | |
|---|---|
| Site: | 05000054 |
| Issue date: | 06/02/1981 |
| From: | George K UNION CARBIDE CORP. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8106160651 | |
| Download: ML20004F235 (34) | |
Text
._ _
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O's UNION CARBIDE CORPORATION -
MEDICAL PRODUCTS DIVISION g _s_l((jb r.o. wx m. ruxcoo. ~tu vonx 109 7 n upsont: m.m.a u s
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]y i JUN 15 081" 9 hus.wg*j%"#
June 2, 1981 U. S. Nuclear Regulatory Commission Division of Licensing Washington, DC 20555 Attn: Chief, Operating Reactors Branch 4 Ref: Docket 50-54
References:
- a. Letter from N.R.C. to Union Carbide Corp.
Dated Dept. 13, 1973 Subj: Change No 11, License R-81
- b. Amendment No. 14 to License R-81 Dated May 17, 1979: Technical Specifications
- c. Letter Union Carbide Corp. to N.R.C.
Dated May 23, 1980 Subj: Renewal, Operating License R-81 Gentlemen:
It is requested that item 3.5.2 c (5) in the Technical Specifi-cations for the Union Carbide Nuclear Reactor, as contained in Refs. b. and c. above, be changed to read as follows:
(5) The iodine inventory of a single capsule shall be limited to 500 Curies I-131 dose-equivalent. :
l A Safety Analysis supporting this request is enclosed. 1 relosae65T 0 1
o U.S.N.R.C.
Divicion of Licencing Juna 2, 1981 The requested change will enable more efficient use to be made of a scarce national energy resource material (uranium-235) in the production of radioisotopes for nuclear medicine applications. The present wording of the Technical Specifi-cations item requires double encapsulation for any quantity over 70 Curies. This results in substantially less effective utilization of the U-235 target material. It also greatly increases the temperature attained by the inner capsule, which is undesirable from those safety considerations principally relating to mechanical strength and to fission product diffusion out of the UO2 into the gas phase. The volume of irradiated waste is also increased, which is undesirable for disposal reasons.
For the record, in the past 12 years a total of approximately 120 singly and 6700 doubly-encapsulated capsules have been irradiated in the reactor with no instance of capsule failure.
We believe this is evidence of good experiment design and quality control.
We propose this requested change as a Class III Amendment and enclose our check for $2000.
Very truly yours, Kenneth D. Geor .
Senior Development Scientist KDG:js 3
Enclosures:
- 1. Safety Analysis
- 2. Appendix A
- 3. Appendix B (13 copies enc.)
STATE OF NEW YORK)
SS COUNTY OF ORANGE)
On this // day of June 1981, before me personally came Kenneth D. George to me known and known to me to be the in-dividual described in and who executed the foregoing instrument and acknowledged that the executed the same.
hatsy PuWic, Stat o' ew Yort [. '
k 0327M Notary Public Calitiec in Orange Ccunty CerMcate filed in Orange County Term Egires March 30,19.#3 1
SAFETY ANALYSIS Single Encapsulation of Capsules Containing Enriched Uranium NOTE: References are listed at the end of this Analysis.
Pertinent extracts from these are included for convenience in an Appendix B.
- 1. Abstract The need to limit single ecapsulation to a maximum capsule content of 70 Curies I-131 dose-equivalent (Ref. 1) is re-examined. Published data on release-limiting physical processes are shown to result in an iodine decontamination factor that is about 500 times greater than that used in the N.R.C. Staff analysis. If a margin of 10 for uncertainties is then applied, the capsule inventory can be safely increased by a factor of 50 without exceeding the acceptable doses calculated by Staff for a 70-Curie (I-131 dose-equivalent) capsule. Other present re-strictions in the Technical Specifications, such as maximum fission power, limit the content of a capsule to 500 Curies I-131 dose-equivalent so the practical incre';e in inventory for a singly-encapsulated capsule would be only a factor of a little over 7.
- 2. N.R.C. Staff's Safety 0 valuation:
The source term bases used by the Staff (Ref. 1) for evaluation of possible radiation doses to personnel both within the containment building (hereafter called "onsite") and offsite at the plant boundary were as follows:
100% of the noble gas and iodine inventory is released from a capsule.
Safcty Annlysis Singlo Encrp2ulction of Capsules Containing Enriched Uraniam June 2, 1981 100% of the noble gases appear in the containment building air.
10% of the iodines appear in the containment building air.
Mixing takes place with only 50% of the building air.
Doses to personnel onsite and offsite for a 500-Curie capsule were then calculated for two different condi-tions of the engineered safeguards, as follows:
- a. Engineered Safeguards Fail.
Thyroid doses both within containment and offsite were shown to be unacceptably high (i.e. 204 and 84 rem respectively), a not unexpected result con-sidering the above source term assumptions.
- b. Engineered Safeguards Function Correctly.
Doses within containment (onsite) were not affected.
Doses offsite were reduced to acceptable levels.
Based on the above evaluation, Staff requested that capsules containing more than 70 Curies I-131 equiva-lent be doubly encapsulated to reduce the probability of occurrence of capsule failure. Tue failure of capsules containing less than 70 Curies I-131 equiva-lent of fission products will result in onsite personnel exposures that do not exceed those permitted in 10 CRF Part 20. This limitation was made a Technical Specification condition.
It was further requested that the engineered safeguards be evaluated for reliability and modified to provide assured performance. This licensee made the evaluations and completed the resulting modifications on a schedule agreed to by Staf f. These changes give assurance that the engineered safeguards will operate properly and, in case of capsule failure and release of all contents, result in acceptable offsite doses for capsule contents up to 500 Curies I-131 dose equivalent.
l
Safety Anclycis Single Enccpaulation sf Capsules Containing Enriched Uranium June 2, 1981
- 3. Re-evaluation of Iodine Release:
It will be shown below that the source term bases of the Staff's 1973 evaluation, as listed above, are excessively conservative for iodines.
Release of fission products to containment atmosphere can be broken down as follows:
3.1 Capsule Damage Mechanisms
- a. Mechanical damage, e.g. crushing
- b. Material defects, e.g. bad welds, and cracks
- c. Improper seals, i.e. loose entry port caps
- d. Capsule melting, e.g. flow blockage, excessive heat flux 3.2 Type of Fission Product Release
- a. Release due to the first three mechanisms above (3.1 a,b,c,) is that of the gas-phase contents of a capsule. This will therefore be similar to the " gap activity" release resulting from cladding failure in power reactor fuel. The inventory fractions so released are small, typically of the order of .001-l%. For example the method of (proposed) ANSI /ANS Standard 5.4 (see Appendix A) gives release fractions of 0.1% for Xa-133 and 0.16% for I-131 from the UO2 in a capsule. These calculated results are roughly confirraed by the measurements of 2.5% and 0.1% respectively made by this licensee in 1973. In any event the source term is below that assumed by Staff by a factor of 40 or more for noble gases, and a factor of 500 or more for iodines.
- b. Release with the fourth mechanism (above,3.1 d) should be similar to that from UO2 at high tempera-ture, in the presence of molten metal (stainless steel or zirconium) and of steam. The inventory fraction released from the fuel itself can approach 100%, as in the Staff's evaluation. In the case of iodines only a small portion of this is, however, available for transport through the pool water to the containment air, as indicated below.
S3foty Analysic
- Single Encapsulation of Capsules Containing Enriched Uranium June 2, 1981 1 f
- 3.3 Release-Limiting Mechanisms-I '
- a. The first type of release above (gas-phase fraction) wili.be subject to absorption in the j 24-f t-thick layer of reactor . pool water, for ,
j capsules in the core. For capsules being trans-ported outside the core, the water depth may be
- less~(but at least 10 ft. because of shielding ,
- requirements).
1
- b. The second type of release (capsule melting) will in the case of iodines be reduced by interaction i with molten capsule material, by chemical combi-l nation with other fission-product elements and by scavenging with condensing steam. Any gaseous i iodine escaping those natural consequence-limiting l' mechanisms will be subject to absorption and dilu-tion in the 24-foot thick layer of reactor pool
[ water before relaase to the building atmosphere.
1 l 3.4 Magnitude of Source-Term Reduction Factors
! a. In the case of the " gap-activity" or gas phase type of release (3.2 a above) , the fraction of the inventory available for release is assumed to be the larger of the values cited above, namely 2.5%
for gases, 0.2% for iodines. Additional evidence i is two experiments (Ref. 2, Table 1.5) in which t Zircaloy-clad UO2 fuel was heated enough to rupture the cladding, but not to melt either the fuel or the clad, that gave iodine release fractions of 0.189% i
! and 0.115% of the contained inventory. Iodines i will be further reduced by pool water absorption. .
Staff assumed a pool decontamination factor of !
l only 10. Using the latter number and a release fraction of 0.2% the iodine reduction factor for this case is about 5000. Using the more realistic decontamination (or trapping) factors discussed below, the iodine reduction f actor would be much greater than 5000.
- b. For the capaule-melting type of release (3.2 b above) , in which it is conservatively assumed that 100% of the iodines (and gases) are released from the UO2, the following considerations apply:
(
Safety Analysis Singlo Encapsulation of Capsules Containing Enriched Uranium June 2, 1981 (1) Scavenging By Steam and Molten Metal Research programs conducted at Oak Ridge National Lab. (and elsewhere) around 1965 on fission product release from UO2 fuel subjected to high temperature damage has produced data that are applicable to the capsule melting situation under discussion here. The fact that the UO2 configuration differs is not relevant because we are con-servatively assuming 100% release of gas and iodines from the fuel matrix. In these research programs U02 fuel clad with stain-less steel or with Zircaloy was heated, electrically or with a reactor transient, sufficiently to melt the cladding and in some cases the fuel as well. The environ-ment around the fuel element was either steam or water. It was found that both the molten cladding metal and the condensing steam are highly effective in absorbing and retaining iodine. Experiments in which SS-clad UO2 was melted in moist air (#3,4, in Table 1 below) resulted in only about 1% of the I-131 inventory being transported from the autoclave by the sweep gas. Two experiments in high-pressure (1000 psi) steam, in which the clad, but not the fuel, melted (#5, 6Z below) showed f airly small I-131 release from the fuel (1-10%) and a small release from the autoclave (#5). A series of four experiments
(#7, 8Z, 9,10Z below) , was conducted in which UO2 was melted under water with low-pressure steam conditions that should approximate those l
in an open pool reactor. Release of I-131 from the fuel itself was large (up to about 75%) but the majority of this deposited on t adjacent surfaces of the autoclave. The
! fraction of I-131 released from the autoclave l through venting of the steam was less than 20%
l in all cases. The important point, however, is that in all cases practically all of this
(
vented I-131 condensed and was retained in water
! traps. Only insignificant fractions were left for absorption in membrane filters or charcoal l
l traps. The results of experiments 3 through 10 are summarized in the following table:
I l
l l
l
TABLE 1 DISTRIBUTION OF I-131, PERCENT OF INVENTORY Released From Condensed In Trapped In d Experiment # Fuel + Clad Autoclave Autoclave Water Traps Filters Reference 3,4 60 38.1 1.1 0.9 .17 .3, Table 2 i
5 99 0.55 -0.45 0.15 0033 3, Table 5 c
6Z" 90 10 O 3, Table 6 7 18 77 5 5 .002 4, Table 1.2 8Z 25 68 ~7.4 7.4 .007 4, Table 1.3 9 15 66 19 18.5 .004 5, Table 1.3 10Z 39 49 11.8 11.8 .0056 5, Table 1.3 NOTES: a. Z denotes Zircaloy cladding
- b. Estimated by difference
- c. No steam release due to blockage
- d. References are listed in Sect. 4 KDG:js 5/28/81
Sciety Analysis
- Single Encapsulation of Capsulcs Containing Enriched Uranium June 2, 1981 Some conclusions from the r4 ults of the last four experiments were (Ref. 5, pp. 19-20):
(a) There was nc. significant effect of cladding material (stainless steel vs Zircaloy) .
(b) Fission product release was similar in all four experiments, the twice larger I-131 transport in #9 and 10Z being ascribed to faster steam release.
(c) Rapid formation of non-volatile water-soluble iodine compounds was indicated by the results of #9, 10Z.
(d) Although 60-80% of the I-131 was released from the melted fuel, only .0006 .005%
was found on charcoal traps or collected as gas.
(e) Condensing steam is effective in trapping particulates.
As further evidence of the efficacy of con-densing steam for removal of iodine, laboratory experiments are cited (Ref. 6, Table 4) in which fission products were released from heated uranium. It was found that 97% of the released I-131 was retained by the condensing steam, leaving only 3% for transport elsewhere.
(2) Trapping In pool Water In connection with suppression systems for power plants, work has been done in which iodine was injected into water-filled pools and the relative concentrations in water and supernatant air measured.
The resulting concentration ratio (or partition coef ficient) is dependent on pool water pH and initial concentration. Experiments (Ref. 7) with pool water at 1000C (an extreme case) , pH = 6, and initial iodine concentration 6.5 x 10-5 g/l (5 x 10-7 M) ,
gave a partition coefficient of 3 x 10 2. Other experiments (Ref. 8, Fig. 11) in which iodine-gas mixtures were bubbled into a water pool at 25 0C with pH = 6 to 7 and iodine concentration 10-6 971
Snfsty Analysis
- Singlo Encapsulation of Capsules Containing Enriched Uranium June 2, 1980 4
gave partition coefficients of > 10 . The initial iodine concentrations in our pool are < 10-8 g/1 even if all fission-produced iodine is assumed dis-solved therein, so our partition coefficient is expected to be more than 10 4 .
(3) Water-to-Air partition:
While this topic is treated above in (2), the emphasis there is or. che f'Irther trapping of the small amount of iodine that ir released from the vicinity of the (assumed) melted capsule. One avenue of iodine release to containment building air that could be postulated, but was not specifically analyzed by Staff, is to assume that 100% of the iodine inventory becomes dissolved in. the pool (50% of pool volume to be conservative) , and to estimate the resulting air concentration above the pool surface. This re-quires values of partition coefficients from water-to-air. Reference 9, Tables 9 and 10, gives parti-tion coefficients, calculated from measured con-centrations, for water at 500C and pH in 5-7 range, of 3 x 103 - 3 x 105 for pool iodine concentrations of < 4 x 10-9 g/1. Reference 10 (Fig. 9) has plots of measured results at 25-80 0C, pH 6-7. At an iodine liquid concentration of 10-4 g/1, equivalent to 10-5 mg/l in gas phase, the partition coefficient is shown as 2 x 103 This work was not extended to lower aqueous concentrations where partition coef-ficients would be expected to be larger still. In Reference 11 (Fig. 1), partition coefficients cal-culated from measured parameters are In water 250C and a concentration of 10~given.
/ g/l (the smallest used), the partition coefficients for pH = 5 and 7 are 10 4 and > 105 respectively.
For our pool water conditions (pH 5-7, total iodine concentration < 4 x 10-9 g/1) it is reasonable to assume a partition coefficient of at least 104 If all radioactive iodine (500 Ci) is dissolved in half the pool water volume of 120,000 gallons che concentration in the water is 1.1 x 10-6 Ci/ml.
Applying a partition coefficient of 104 gives an air concentration (near the pool surface) of 1.1 x 10-10 Ci/ml. The building air concentration calculated by Staff was 50 Curies in half the build-ing volume of 7700 m3, or 1.3 x 10-8 Ci/ml, which is larger by a factor of more than 100.
Safety Analysis Single Encapsulation of Capsules Containing Enriched Uranium June 2, 1981 3.5 Summary of Release-Limiting Factors for Iodine (The most conservative values are chosen below)
- a. Capsule break (gas-Phase contents) : 0.2% of inventory (Sect. 3.4 a).
- b. Capsule melting:
(1) Release from U02 fuel: up to 84% of in-ventory (Table 1)
(2) Release to pool at core region: up to 20%
of inventory (Table 1) . (Includes effect of plateout on metal capsule and housing.)
(3) Release after steam condensation or water trapping in core vicinity: 0.17% (Table 1)
- c. Decontamination through 24-ft-thick water layer:
104 (Sect. 3.4 b (2)
- d. Air conceptration from complete dissolution in pool 1.1 x 10-10 Ci/ml (i.e. 100% iodines in half pool volume). (Sect. 3.4 b (3).
3.6
Conclusions:
Even without invoking the large iodine decontamination due to passage through the 24-ft-deep pool, the amount of iodine available for mixing with building air (0.2%) is a factor of 50 below the quantity assumed 4 by Staff (10%). A further reduction by a factor of 10 is supportable by the evidence described previously.
This leaves 3.5 d above as the avenue more likely to give the greater air concentration in the case of com-plete fuel melting. As stated previously in 3.4 b (3),
100% dissclution gives a building air concentre. tion that is 1/100 of that used by N.R.C. Staff. From the result cited in Sect. 3.5 b (2) above, a further reduction by a factor of 5 (20% release to pool water) can be justi-fied.
Overall it is reasonable to expect that the containment building air-concentration will, under accident condi-tions, be lower than that estimated by Staff by a factor of 500. Allowing an order of magnitude margin for un-certainties leaves a safe reduction factor of 50.
Safety Analysis Singlo Encapsulation of Capsules Containing Enriched Uranium June 2, 1981 While the two reactor systems differ considerably in size and design, there is some corroboration of the above from the results of the TMI-2 power reactor accident. It has been reported (Ref.12) that 60%
of the core I-131 inventory was released to the reactor cooling system, 17% was in the containment building water, and only .007% in the containment building air. These last two figures would indicate a partition coefficient of 2400. Similarly, the cor-responding _ figures for the auxiliary building are 6%
and 2 x 10 4%, for a partition coefficient of 3 x 104 Even closer confirmation of the correctness of these conclusions comes from tests performed in a system similar to the Union Carbide Nuclear Reactor (UCNR).
These are experiments (Ref. 13) done in an open pool reactor (CABRI) in France in which several fuel plates were deliberately melted in order to measure the re-lease of iodine into the water and its transmission through the pool water layer into the room air. The release of iodine into the water was 2-10% of that present in the damaged fuel. It is stated that the fuel, the water, and the materials constituting the interior of the core retain 99.995% of the iodine pre-sent in the damaged portion of the core. For a pool depth above the core of 2.5 - 3.3 m, the transmission factor through the water (ratio of activity in air to that in water) was found to be between 5 x 10 6 and 5 x 10-5 This means that only .0005 .005% of the iodine released into the water reached the room air, which implies g partition between water and air acti-vity of 2 x 10 - 2 x 10 4.
- 4. References 1: Letter N.R.C. to U.C.C., dated Sept. 13, 1973; Docket 50-54; Subj: Change No. 11, License R-81.
2: Nuclear Safety Program. Annual Report for Period Ending 30 December 1970. ORNL-4647 (May 1971).
3: International Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ridge TN., April 1965. CONF-650407, pp. 299-310.
4: Nuclear Safety Program Semiannual Progess Report for Period Ending December 31, 1965. ORNL-3915 l (March 1966).
L __ ._ . ._ _ .-.
Safety Analysis Single Encapsulation of Capsules Containing Enriched Uranium June 2, 1981 5: Nuclear Safety Program Annual Progress Report for Period Ending December 1966. ORNL-4071 (March 1967).
6: Hilliard, R. K., etal, Health Physics 7, 1-10 (1961) 7: Devell L., etal, Nuclear Technology 10, 466-471 (1971).
8: Diffey, R., etal, ibid, CONF-6 50407, (AERE-R4882),
pp. 776-804.
9: Parsly, L. F., Calculation of Iodine-Water Partition Coefficients, OREL-TM-2412, Par. IV, (Jan. 1970).
10: Nishizawa, etal, Nuclear Technology 10, 491 (1971).
11: Eggleton, A.E.J., etal, Atomic Energy Research Establishment, Harwell, AERE-R4887, (Feb. '967).
12: Hull, Andrew P., Nuclear News, April 1981, pp. 61-67.
13: Dadillon, J., Bull Inform. Sci. Tech. (Paris),
No. 112, 13-18, (Feb. 1967).
1 4
F t
I r
i . . _ . . _ . _ . . _ . _ . _
APPENDIX A Volatile Fission Product Release From Oxide Fuel Proposed American National Standard ANSI /ANS-5.4 (Nov. 1979),
as revised to Nov. 1980, recommends methods for calculating oxide-fuel plenum-gas activity. The methods consider high-temperature and low-temperature releases and distinguish be-tween short half-life (less than one year) and long half-life nuclides. The standard requires that releases be calculated with both high and low temperature methods, the larger result being selected.
This standard is applied to calculate releases of iodine and gaseous radionuclides in the analogous case of release into the in'terior of an intact capsule of the type under consider-ation here. The principal difference is that the UO2 ruel in the case of a capsule is in the form of a thin layer.
All the nuclides of interest are short-lived. Comparison of the high and low temperature methods shows releases calculated by the latter to be 10 -105 4 times greater than by the former.
Thus only the low-temperature results are given below.
From Sec. 3.2.2 of the standard, the release fraction F is:
~
F = (10 A.5 + 1.6 x 10-12 p) jA, where: P = specific power = MW/t A = decay constant (s~1)
Results of the calculation are given in the following table.
t I
l l
T APPENDIX A Volatile Fission Product Release From Oxide Fuel TABLE 2 Nuclide Half-Life A (sec. ~ ) F
-6 -4 Xe-133m 2.19 d 3.66 x 10 4.6 x 10
-6 -3 Xe-133 5.24s d 1.53 x 10 *l.24x 10
-5 -5 Kr-85m 4.48 h 4.30 x 10 5.0 x 10
-4 Kr-87 76.3 m 1.51 x 10 1.8 x 10-
-5 Kr-88 2.84 h' 6.78 x 10 3.4 x 10-
-4 -6 Xe-135m 15.29 m 7.56 x 10 5.6 x 10
-5 -5 Xe-135 9.09 h 2.12 x 10 9.2 x 10
-3 -6 Xe-137 3.85 m 3.00 x 10 2.3 x 10
-4 -6 Xe-138 14.17 m 8.15 x 10 5.3 x 10
-7 -3 I-131 8.04 d 9.98 x 10 1.6 x 10
-5 -5 I-132 2.30 d 8.37 x 10 2.9 x 10
-6 -4 I-133 20.8 h 9.26 x 10 1.9 x 10
-4 -5 I-134 52.6 m 2.20 x 10 1.4 x 10
-5 -4 I-135 6.61 h 2.91 x 10 *1.6 x 10
- Corrected for precusors per Section 3.3 of Standard.
4 APPENDIX B For convenience, pertinent extracts from References
- cited in the foregoing Safety Analysis are appended.
J 1
(These extracts are keyed to the corresponding number in the list of References, Section 4.)
4 n
nt
(
O i
l 4
9 9
1 i
4
- - , , e -g--,. -, e , ,, - , _ - -
- . , ~ - , , .,n,,- ,-,, , n.. . . -n,,. ,-,, - . ,, , , , - - ., , - . , - -, , _ ,, ,,,,--,-,wr--.,re,,
- ~~-
T Table 1.5. Fis ion Products ReleaseJ in TRFAT Fuel Rod Failure Experiments FRF 1 and FRF 2 Material Fourd in Eath I.ocation (% of totalin center irradiated ru l)*
e29 7, Rnion Prmluces 1.oca .an '3'l Cs with I.ow Volatility 6 U FRI' l FR F-2 l'R F. I I RF-2 FRFl F RI'-2 FRF1 I RI -2 FRF1 l'RI' 2 Piimary vessel' O.054 0 066 0.046 0.19)
Filter patk Deposition by diffusion 0.076 0.037 0.0004 0.005 0.0001 ~ l.5 x 10 O O O O Deposited with particles 0 046 0.010 0.010 0.090 0.015 < l.5 x 10 O.0009 1.4 x 10 O.00fn24 16 x 10
Condensate ta Unit i 46 x 10 <60 x 10*
Unit 2 4) x 10 <60 x 10
lleated charcoal Unit 1 0 0087 0 0002 Unit 2 0.0018 01:026 Totat teleased 0.189 0.115 0 056 0.288
% i I
81 RI' l center rod was irrailiated 13.1 fuit power days to 645 kiWJ/kif peak burnup at 15.2 kW/ft peak linear power, peak-to-average flus ratio.1.15; total liwion.9 8 x
. FRI' 2 center rod was irradiated 62.7 full power days to 2800 MWJ/kt r peak burnup at 14 kW/ft peak linear power; peak-to-average flux ratio, l.15; total fiwion. 4.3 x Q 6kleJian of s9 3,, 9 sZr. ilu. and Ce.
N 'Primasy vesselIcached oth 0 5 N Nil4 0ll;only soluble materials reported.
d Iotal release of "Kr was 0 094'1 in experiment I:RF-l and 0.48 T. in esperiment I'RF-2; total release of ' 3'Xe was 0.14% in experiment i RI'-l but s.as not determined p
in esperiment i R F-2 lesauw of long decay time. t m
- t g ct~.s ,
p.
x tv 3::= '
r---
(D tt
.i N
v t
\
-.. Appendix B - (Raf. 3)
TABLE 1. SUIDIAIT OF TREAT FlSSION PRGMCT RELIASE IIPERIMENTS I
o O
hE a me =
h
=
gy lug 83
" a e 8 8 aa g Egu 5. l. . aj a- lg B
- a. h h
a e.-
- e. -
a e a a la a 38
=E a
85 a5 dEg! l C e
" 2 e'
- ggn
-: n. . m 8 su wo O
U m>m E Sm PEm m - <
47 40 ARGON S.S. 800 1 0 2 50% 65% 315 3 30% 75% 346 74 43 3% Steam- 5.5. 800 1 0.2 97% Air 4 30% 75% 335 77 53 30% Steam- S.S. 800 1 0.2 70% Air 309 108 39 ARGON 3.S. 800 1 0 1 20% 0 6 10% 0 1 327 78 1000 STEAM Zr-2 350 2 13 5 0.5% 0 i 330 77 1000 STEAM S.S. 310 2 TABLE 2. DISTRIBCTION OF MATERIAL IN TREAT IIPERIitENTS 3 AND 4:
1DW PRESSURE STEAM-AIR. TRACE PREIRRADIATION. 75% MELTED PERCRNTAGE IN EACH IDCATION 131 1 129 To 137 Cs 106Ru 89 3r 95 Zr IDCATION 140 104 Ba Ce UO, 60 79 66 95 98 98.3 UOs FLT.L + 5.5. CLADDING (S.S. CLADDING) (3) (9) (0.8) (0.2) (0.25) (0.24)
ALUMINA HEATER (1000'C) 25 13 28 2.3 0.4 0.43 AUTOCLAVE LINER 6.8 5.6 5.1 1.5 1.1 C.8 FUEL AUTOCLAVE WALIA (150 C) 6.3 0.6 0.7 0.7 0 0.46 GAS TRANSPORT ZONE 1.1 0.15 0.5 0.001 0.5 0.001 FILTERS (MIMBRANE) 0.02 0.01 0 0.01 0 0.0001 GAS-COLLECTION AI?rOCLAVE 0.15 0.001 0.2 0.001 0.05 0.0')00 l
l 30s
f
. Appendix B - (Ref. 3)
TABLE 5. DISTRIBtTTION OF MATERIAL IN TREAT EEPERIMENT 5:
1000 psia STEAM (RELEASED), TRACE PREIRRADI ATION, NOT MELTED PERCENTAGE IN EACH I4 CATION 129 7e 137Cs 89 3r 95 2r DO 3 131 I thCATION 140g , g44 gy 99 99.6 99 9 99.99 99.998 UO3 FUEL
- S.S. CLADDING (3.3) (1.1) (2.1) (0.12) (0.1)
(S.S. CLADDING)
.. . ........................................ 0.0006 0.5 0.3 0.06 0.0014 ALUMINA HEATER (400 C) 0.046 0.004 0.006 0.002 0.0004 FUEL AUTOCLAVE WALIA (300*C)
.........................................- 0.02 0.09 0.0005 0.0009 0.00004 MAIN WATER TRAP WALLS 0.12 0.0007 0.01 0.0009 0 .
UATER IN WATER TRAP 0.008 0.002 0.0008 0.0005 0.00002 SECOND WATER TRAP 0.0001 0.0001 0.0001 0.0001 0.00005 MEMBRANE FILTERS CHARCOAL FILTERS 0.0036 - -
0.C001 0.0001 0.0002 0.0003 0.00001 GAS-COLLECTION AUTOCLAVE s'
TAB 12 6. DISTRIBUTION OF MATER AL IN TREAT EXPERINENT 6:
1000 psia STEAM (RETAINED), TRACE PREIRRADI ATION, FUEL A!TD CLADDING FRAGMENTED PERCENTAGE IN EACH I4 CATION 131 g 129 Te l37 Cs 103 Ru 893 , 95 2r UOa I4 CATION 1
l 106g , 140 Ba I44 Ce l
90 95 77 97.7 98.9 UOa . 3r.2 CLADDING (2.5) (11) (5) (10) (7)
(Zr.2 CLADDING) 4.4 2.6 .Il 2.1 1.7 ALUMINA HEATER (400'C) 4.8 0.2 0.5 0_,11 0.06 METAL HEAT REFLECTORS l
l AtTTOCLAVE WALL? AND 0.07 0.05 SPf.CER (3000C) 0.7 0.2 0.5 0.2 0.004 0.07 0.01 0.0002 UATER IN AITTOCLAYE i
310 l
i i
l l . - ._ . _ . - - - _ . - - . . _ - - - . . , , _- .-. .. - . - - . . , - . .
I l
4 Table 1.2. Distribution of hterial in TREAT Experiment 7 in Which Stainless-steel-clad UO 2 s
Was mited Under hater by Transient Fission Heat Input of 504 cal per Gram of UO2 Location Material Found in Each Location (%)
131 1 129 Te 140Ba 1370s I Ru 8937 95Zr Ce" UD2 I
i Haterial renaining in 95 97 95 99.8 99.6 99.1 99.% 99.90 99.92 fuel autoclave Melted fuel clinker 18 36 18 55 43 41 43 48 45 Autoclave components 77 61 77 45 57 58 57 52 55 including scattered .
fuel H
Haterial released from 5 3 5 0.2 0.4 0.04
- 0.9 0.1 0.08 fuel autoclave Condensate in water 2 0.2 3 0.06 0.08 0.1 5 x 10 O.01 0.01
~
traps Walls of water traps 3 3 2 0.1 0.2 0.04 0.5 0.C') 0.07
! and c.ondenser Membrane filter 0.002 0.001 0.005 9 x 103b 0.02 0.2 1 x 10 O.005 7 x 10
papers % . '
'O Charcoal-loaded 6 x 10 O.01 2 x 108 0.02 0.001 papers 3D Gas-collection 5 x 105 3 x 10 4 0.03 3 x 10 O.06 0.08 5 x 10'7 D 2 x 10~4 tank 7 x 10~7 f.
X
" Average of 14I Ce and 144 Ce. tp b 8 Plus or minus 100%.
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Appendix B - (Raf. 5) e 19 fuel specimen to catch melted fuel and cladding. Small holes allow access for water and stea=. Tne safety-relief rupture disk directly above the primary vessel contains eight radial ports to absorb recoil. The icw-pressure rupture housing on top of the assembly is cennected to the pri-mary vessel by small-diameter tubinC-Conclusions In experiments 7, 8Z, 9, and 1CE at ambient pressure, no si6nificant effect of cladding (stainless steel or Zirealoy-2 material) was evident.
In.all experiments the fuel and cladding residues appeared to have moved about in the fluid state with very little fragmentation or shattering.
The residue was foamy or porous when stainless steel cladding was used, but a nonporous solid was formed when Zircaloy-2 cladding was melted.
For fission heat input of 520 cal per Gram of UO2 , 16 to 24% of the stainless steel cladding and 41 to 49% of the Zircaloy-2 cladding reacted with water to form hydrogen. These values agree generally with the amount of reaction found in the AHL metal-water reaction studies in TREAT.
Fission-product release and transport was similar in all four experi-ments, except that about twice as much of the volatile group (tellurium, i cesiu=, and iodine) was carried out of the fuel autoclave in experiments l 9 and 1CZ (6 to 19%), probably because of faster steam release. Release of UO2 , "Ce, and 95 Zr from the fuel autoclave was less than 0.1%.
In experiments 9 and 10Z, 33 to 44% of the 131 I and 137 Cs were found in distilled water rinses of the scraped fuel autoclave walls, even though l the autoclaves were boiled to dryness (220 to 32S*C) after the transients.
This indicates rapid formation of nenvolatile water-soluble iodine and c2sium ecmpcunds. Tilase elements vers also E;nerally dissolved in water collected in the first water trap.
Although 60 to 80% of the 131 I was released from the melted fuel and cladding, only 0.0006 to 0.005% was found on the charcoal-loaded papers and in the gas-collection tank. All the iodine on the charecal papers cnd in the gas-collection tank was nonreactive or penetrating, based on its serbability en chtrecal. The small amount of nonreactive iodine found i
'..-E L
c M I k b k N N h. bM k(:hME' N kk__ . f_W.,, Ah_ _ .!.*b., %__ s.f,_. 0' &._.$$_- _u N _ eh_ .D. _ 2J%_g&, _d.UU_ h._ _ -5.I$
- -_ 1 h
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- 3..
e.
, . '.. n:-
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R. K. HILLIARD, C. E. LINDEROTH and A. J. SCOTT 9 ,
+-
properties of the collected material. Both strongly dependent on extent of oxidation of the .- ,
I condensed vapors and particulates were prob- uranit m. '
abiy retained on the filter. In steam atmos. Rei me of the volatile elements was greatest in i -
pheres, the filter was maintained at 140 C to air, nex6 greatest in steam and least in helium assure that steam condensation did not occur atmospheres for similar conditions of time and ,
on the Elter st.cface. temperature. >
In the tests with steam atmospheres, the off- All fissicn-product elements except rare gases j gas steam was condensed after leaving the escaping the heated uranium tended to deposit .
furnace tube. It was interesting to note that a heavily on tube walls. From 10 to 98 per cent of [ j a
very large percentage of the released fission the released material deposited in the furnace products was contained in the condensate catch tube used in the present study. When heating I [
t ;
Table 4. Comparathe efectheness of steam condenation ts. fit -in remoting releasedfssion products i l )':- ;
yf Per cent fission product ternosed a Niethod of removal l gm Te" 8 Sr*-* Cs"8-"? Rulo2 Bauo .@,;
Ev " Absolute" filter * '
98 99 l 80 55 92 74 .@N 4
Bv steam condensationt i
l 97 77 l 40 '
S0 50 50 p.e Y e
- Filter teniperature HO*C. ,'
M .Q i
+ Average of six experiments, uranium heated at 1215'C in steam. ,
~ m[(
' . -?f; ;,
4 y ' , -.
borne fission products by condensation of the occurred in steam atmospheres, condensing the 7 %
' fission product laden steam is nearly as effective steam caused 40-97 per cent of the evolved ? ?
in removing the fission produc:s is an " absolute" fission products to remain in the condensate.
filter. " Absolute" filters removed 50-98 per cent of T-N~'*
frf Summarizing the behavice of the released the evolved fission products, the efficiency h, fission products, it can be stated that for the depending on the isotope and the typeofgasused. i &
j F J ,.qi.
'p conditions employed in these tests, deposinon on The test parameters used in these experiments ,
j' cool tuue walls, retention on filters and con- tend to mar.imize the release of fission products. t A ,6 f' densation of steam are effective mechanisms One exception to this is the possiEniy that at '
- j. bf;7 i.
for retaining the evolved radioactise materials higher irradiation Icvels, release could possibly within the reactor shell or building. be greater due to high concentrations of fission { f-(.ql^.F li -
~
4 products and radiation damage to the uranium !
SU31.\tARY This laboratory investigation shows that Release of strontium, barium and zirco.uum definite trends and c:.alations are obtainable fef.T~ h.. .MY .-
from heated uranium was lower than most for applicatiori to the fission-product release previous theoretical estimates, being <0.5 per question. However, the many* complicating F-N.i i
(,}, ' ' 3 cent under all conditions tested, k factors involved necessitate a conservative Cesium and ruthenium behaved as " semi- attitude when applying the data to hazards ClP $
volatile" elements being released from <l per evaluations. 'M % f.
cent to about 15 per cent. Their release was dependent on time, temperature and atmos-REFERENCES
- 1. UsrrEre STATES ATome Estacy Comussios, e'h [ ..
4 ii Phere. g,g pgjg, ,j ,
g 3, p X,enon, iodine and tellurium were volan,le, Adecs in 6p .Wcicar Pourr Pla.m WASH.NO, ' !.
~
their release from heated uranium being de- pp.17-23 (1957: v ^
pendent on time, temperature and atmosphere. 2. G. E. Catzx W. J. AI^ arts and G. W. Panxta, f '
' 'D The release of the three volatile elements was Erreviments on ne Release of F:ssion Produr:s from I '
1flaskT Table 4 shows that the retention ofsteam-31'm Fteaow Furir. ORNL.2616 / I95% !
"* =
k.,
Appendix B - (Ref, 7) i STUDIES ON IODINE TRAPPING BY ~ .<.e.....
WATER SYSTEMS AT STUDSVIK KEYWORDS: fission prodvets, iodine, gases, water, cluemical
'*octions, e e i di t y, perf ,.-
L. DEVELL, R. HESB6L, and E. BACHOFNER* *"*****"'*'""'*"'**9'*T Aktiebolaget Atomenergi, Studsvik, Nyhdping, Sweden Received July 28,1970 Revised October 19,1970 7-
- - r y 77- ' . < presented five years ago by Linder.* In 1966, large-scale ard laboratory runs were performed From large-scale experiments on the iodine to investigate the trapping efficiency of L. boiling remouri efficsency of a water pool at 100*C as water system to remove molecular iodine from wel' as from supporting laboratory runs wsthin the stea m. These studies were run for the Marviken concentration range 10-8 to 10-8 M, it was ob- nuclear power plant project. The results were served that permanent trapping was achieved to a reported
- at an IAEA conference in Vienna 1967.
great cxtent ever at neutral pH conditions. The Based on the findings from those studies and the efficiency was considerably Hgher at icwer iodine theoretical examination of the iodine-water parti-concentration. The fraction permanently trapped tion coefficient made by Eggleton, a mathematic'L1 approximately corresponded to the theoretically model of the washout of iodine by sprays was evaluated equilibrium amounts of hypoiodous acid developed
- during 1967. The model was tested and iodide. against the initial spray results reported from A mathematical model designed for the washout Japan
- and a iair agreement was achieved. The of molecidar iodine in reactor containment atmo- moael gives an explanation of the change in wash-sphere by sprays gives iodine cor. centration in out rate with decreasing iodine concentration.
containment versus spray time. The model takes The model has been used in safety analysts of two '
the degree of hydrolysis of iodine at different pH BWR plants in Sweden. Our work up to now has values and concentrations into consideration. been directed to water sprays without very reac -
Initsal results from spray experiments performed tive additives. The use of such sprays instead of in a 15 m lank 3 at pH = 6 to 7.5 and gas phase very reactive sprays offers obvious advantages concentrations around 4 x 10*', 5 x 10~3, and 1.5 x with respect to simplicity, chemical stability, and 10-' kg/m3 gave half-times due to spraying of low corrosivity. However, the pH value must be about 45, 5, and 2 min. respectively. They con- sufficiently high to reach the necessary spray ef-firm the expected strong influence of iodine con- ficiency.
centration on washout half-time. During 1969 a spray facility was designed and built to test our model and to serve as an experi-
~ ~
Cb 7?"EEE3 mental facility for studies of different removal processes for airborne contaminants. A few ini-tial deposition and spray runs have been carried INTRODUCTION out up to June 1970. The purpose of this paper is i*
summah Mse of ou usuM and tecWes Parts of the program at Studsvik for the inves- that may be of interest for the analysis of the tigation of fission product release, transport, and e ency i dWerent water systems to retain '
removal processes have been devoted to studies gaseous molecular iodine.
concerning trapping of iodine in water system. '
The initial plans of the general program were TRAPPING OF IODINE BY WATER POOLS
- Present address: AKK. Atomic Power Group, Stock- From work i. the U.K.,* it has i een demon-holm. Sweden.
strated that water pools Can act as ef flCient traps W NUCLEAR TECHNOLOGY VOL.10.. APRIL 1971 k
. App ndix B - (Raf. 7)
Devell et al. STUDSVIK STUDIES ON ICDINE for molecular iodine. It was also reported that produced steam, but the products of hydrolysis the iodine wns permanently trapped in the water, are retained despite con tinued boiling. The To investigate the removal efficiency of a specific amount of. iodine permanently trapped approxi-bt.t less favorable system for trappirg, namely an mately corresponded to the amount initially hy-emergency condensor containing boiling pare wa- drolyzed. This observation constitutes one main I
ter, we performed large-scale experiments in a premise in our spray model. Other experiments 8
pool containing 0.5 to 2 m of water. Molecular at 100*C and pH = 3 to 8 in a closed system with-lodine was injected into the 100'C pool water to- out boiling have indicated a decrease during the gether with superheated (175 to 320*C) steam. Io-dine was sampled in the pool water and in the first hocr of the fraction of iodine that was in rexhausted steam.
molecular form and a simultaneous increase of the fraction hydrolyzed. The reduction at low and The decontamination factor at pH about 6 in- high concentration (4 x 10-' and 4 x 10-* M, re-creased from 3 to 300 when the lodine concentra- spectively) was moderate, but for the medium con-tion was decreased from 10-' to 5 x 10M in the centration (4 x 10*' M), a tenfold decrease was water pool. Addition of sodium hydroxide to the observed.
pool in runs at high iodine concentration in-creased the removal efficiency. liowev er, at low iodine t.sncentration the addition of sodium MODEL FOR CAI.CULATING WASHOUT OF hydroxide gave only a small increase of the MOLECULAR IODINE BY SPRAYS decontamination factor in accordance with the initial higher degree of hydrolysis at this condi- For calculation of the washout of molecular tion. The fraction trapped was not released des. iodine in reactor containment atmospheres by
.pite additional steam supply or boiling. These sprayt a mathematical model has been developed.
'results, together with observations from labora- The detailed mathematical treatment will be given tory runs, show that the reversal of the hydrolysis elsewhere.' The model takes the degree of hy-was essked in these studies. This is not in drolysis at different pH values and concentrktions agreement with simple theory of hydrolysis, into consideration and is based on iodine data col-Figure 1 shows the typical behavior of molecu- lected and analyzed by Eggleton.' According to lar iodine in our laboratory boilhg water system. results from calculations, the washout behaves The znolecular iodine is released together with the differently in three ranges of iodine concentration in the gas phase (cf., Fig. 2). At very high con-centration the washout process is liquid-phase-controlled due to the dLfusion of molecular iodine
, , , , , into the water drops. In this first region the half-L
_ :, ; ; Hme can be expressed by e.a:uuuutto : 0.093W i= coworntart 73f ,
KFa h 'E
~
where i
' 8 w 1 .
[ [ totu sooint in vessn ; W = containment volume (m')
{ .
. K = partition coefficient for molecular iodine k -
F = spray water flow rate (m*/sec)
Y 3
)M* tin -
a = degree of saturation of lodine in the drops.
_ For. this high concentration range the expression reflects the analytical treatment given by Grif-l
{ fiths.' In a region of lower concentration the hy-u p q , ,, .
drolyus of iodine within the drops becomes more vtssa ' important and increases the washout rate. The
=casuetoi rate grows higher for lower concentrations and i
ic i
8 'O 20
, , i higher pH values. The mathemstical treatment 30 '8 'O 80
- '"""""E* for this second region gave the following final expression for the iodine concentration in the gas phase:
Fig.1. Iodine release vs boiling tirne for a water sp-tem at pH = 1.3 and iodine concentration = 2 x B B cc, K' _ .B_ K HQ g , (f), 4 _4 - 6 *a 8
10 g/ liter. ,
NUCLEAR TECHNOLOGY Vo L. 10 APRIL 1971
- 467 i
l 1
/ '
Appendix B - (Ref. 8)
/.
h s
10 !
g . .
i rJ E -~
x
- x s's
$ N a 10' -x s -
e N s X 3 -
N X . -
$ \
N
, N -
j Calculated N
.o 3 -
partition N 3 10 coefficient g --
2 Allen (6 ) g c
?
l
/ pH7 s -
w g hi
- he pH 5 s x E M u
us 13 2 -
- s-X r>v "i
8 Theoretical coefficient at 25 *C from solubility -
p and vapour pressure data (5) assuming Henry's law C
~ t.'.s g -
t es e-y g >
10 ' ' I ' ' I ' ' I ' ' "'
7 10-8 1CT 10-8 10-5 10-' ~
~~ e.
TOTAL CONCENTRATION OF IODINE IN WATER g. McL iodine / l a les EFFECT OF IODINE CONCENTRATION ON THE EQUILIBRIUM IE PARTITION COEFFICIENT FOR IODINE AT 25 C. ,
FIGURE 11. A.E.R.E -R. 4882.
i r
i 804
.i.
n i
~
. Te
,r,7
- d. \"
s.
-q .
VI 2 Table 9. Rc= alls of Calculations. t = 50*C. pII = 5.0 h
- E RESULT 5 or 110l w-wg ir ac P.smtlli n C7ErFICIE6T CatCut Afl0N5 t' T E'ePE 14 f UR E . % . t1 9t G f PH = 5.P
- Kl = 2.62r il K! * ..EE '2 43 = 4.9' E-12 K4 = 2.14E-It TnttmlLIfr OF I 2 11 Ja lc et 7 2.e* E-r 1 *ent /L * **
~~ - - " ~ ~ - '--- "*--- -
CONCENTR ATIONS IN MOL5/L %
~ ~ '
CONC 12 IN LIQ CINC HIG CONC H?dit C0 tac In0lDE CONC f3- TOTAL 12 IN LEO LOG TOTAL I ~
?.9 'E * ) 5.62F-15 2.45f '9 2.4LE-05 PARTill'N O CUEFF
~ ' t .4 ie 3.22E-05 2. 8 9E-6 3 -2.539 2. FC E 0 8
- 3 3. 3 5F 'S 1.'4 6E -8. 9 ~ ~ 2.( IE ^ 5 1.35E-75
7.COE-e4 2.12F '5 3.7st-10 1.45E-t1 -2 840 2.71E 01 1.5cE-35 5.33E-06 7.27E-C4 3.5-)E '4 1.47 "? .4.12f-10 1. 21E '. 5
-3.139 2.72E 01 f . 75 E -C 4 2.32E-16 3.66E -0 4 -3.436 2.74E 01
- .55T- 4.17E -11 8.8tE-96 7.43E-?7 1.85E-(4 8.75E-c5 6.62E s 2.8 9E -!O 6.36E-96 2.6 7E-0 7 9.44E-05
-3.732 _ . 2.7 7E c l.
'
- 3 7F.-f 5 ._.'t648-76 -4.C25 2.83E 01 2.19E-r5
. 2.9 2E -l:1 _4.S4E-96 9. 5 3E-C 8 4.85E-f!5 .._ -4. )( 4 _,, ___ ,J e 90 E 01.
1.265-r6 1.426-la 3.2 3E-C 6 3.39E-ne 1.0 9E *' 5 2. EF-*~6 2.52E-r5 -4.599 3.GIE OL
, 1.11E-In 2.29E-06 1.2]E-38 1.32E-05 '-4.878 3.lTE 08
?.47E-i'6 1.62F-16 7.E 9E -I l I.62E-36
- 2. F 3E -( 6 1.15E -e +
4.25E-49 7.19E-r6 -5 149 3.4CE 01 5.CIE-Il 1.85E-:6 1. 5 )E -? 9 3.88E-C6
- 1. 3 7E -0 6 a.llE 7 3.54F-il 9. I t F 4 7
-5.411 3.72E 01 6.84E-s7 5.32E-IO 2.18E-f6 -5.662 4.18E 01
- 5. F3F '? 2.5?E-l1 5. 7 3E-P F .
3.42E ':7 4. 5c -r 7 l .18 E - [E- . l . 2 6E-0 6 -5.905 _.. _ 4.82E.01.
- 1. 7 t E r' 7 2. 9 7E - 17 1.25E-It 2.8 7E-0 7
-6.126 5.73E OL p.54F em 2 . ;30 ' 7 8.45E-12 2.t 3E-C F 2.35E-Il 4.3tE-12
- 4. 5 8 E -P. 7 2.8 8E-L F
-6.339 T.02E.01 E 4.27E-05 1.6 3C -2 7 6.76E-l? 1.4 3E-P 7
-6.540 8.84E 01 2.14E -r 8 1. t E
- F 2 94E-12 1.86E-07 -6.730 l.14E 02 4.43E-l? 1.C IF " 7 1.04E-12 1.23E-07 -6.911 I f TE-J e 7.1FF 6 3.l!E-12 1.5tf 02 5.345-f9 ._F.lTE-08 3.67E-13 . fl. 2 4 E-0 8 ._-7.064.. . . 2.0 3E ,9 2..
5.' TE s 2.2tE-I2 5.u ?E-0 8 1.33E-13 2.6 TE -0 0, . 1. 5 0f ' 1 5.6LE-06 -7.252 2.75E 02
, 1.54E-12 3.58E-E8 4.59E-14 3. 8 5 E -C 8 1.14E-39 2. 9 tF '8 1.llF-12 -7.4t4 3.78E 02 2.5)E-09 1.62E-14 2.67E-08 -7.574 5.23E 02 6.69E-te 1.79E-94 7.92E-13 1.79E-c8 5.74E-15
, _ . _ . l . 6 7 E -J G..
1.27E ?9 2.c3E-15 a .19 E -0 5 -7.8d6 1.02E 03 8.14 E -11 9.06F f o , _ 3.9tE-l} ig9 6 E -t' 9 7. l if - {6 _ 9.lM-Q9_ ,,,. ... -0. 0 4 0 , _ _,_ ,_,, J . 4 J L O )_
6.34*- 79 2.FTE-13 6.340-? 9 2. 5 ".E - 16 ,,, 6.42E-ra 4.)?E-ll_ 4.4 aF "o 1.96E-l3 -8.193 2.C 2E 03 2.?F-il 4.49E-09, R.96E-17 4 72E-09_,, -8,345 .. .2.84E 0)_ >
- 1. l T?-r o 1.19?-13 3.lTE ,9 3.17E-17 1.19E-09
- 2. 24E ' 9 -8.496 4.00E 03 D 1.*4E-II 9.78E-14 2.24E-C 9 l.12E-17 2.25E-09 5.2 2F - 12 1. 58F - 19 6.9tE-14 1.53E-09 3.96E-te
-8.646 5.65E 03 Q _ _ . 7.kl5-12 t.3:E 'l2 g
. l .12E ~ 9 _ .
- 7. 9 2F -l a 4.99F-14 1.4AC-14
. t .12E -G 9 . 1,.41E - 10 .
1.59E-39
._.l .12E-19
-8.799
-5.950 f.98E 03
..__ l .1 M O L g
F.92E-10 4.95E-19 7.93E-IO EL
-9.101 1.59E 04 p.
N , I M e t w m
w m
2: -
e 3> $ l r==== . !
7 t
Talde 10. Results of Calculations.t = 50*C. pli = 7.0 9E5Ut15 08 8 001'eE-wt1En P anilll0N 096 :f IClf NI C ALCUL AT 10N5 T EMPE R Alua F = 5E .(: DE G C PH = F.' -
E l = 2.4 2E c l s' 2 = 4.M F '2 H3 ' 4.R'E-12 K4- ~ ~2. lor-ll =
~~ ~~ ~ '- '-
- SOLU 6IE ITT OF' e 2 I83 T.Alfil' . '2;8 E'-0 3 '90Lil
~ ~
'~
sCONCENTR47 l(1N5 IP MULS/L ,[ ]
CONF'l2'IN LlO CONC H10
~
CONC H20l* CONC IO0iOE CONC 13- TOTAL 12 IN LIO LUG TOTAL I PARitil0N COEFF 2.4DE *3 5 6?E-P4 2.4 5f -19 2.40E-04 3.22E-04 3.68E-33 -2.434 3.45E 01 l .U f'? ]-~ ' -~ ' T.16t 94
't.46E-10 ~ 2.0 lf-C 4 1. 3 5 E-f' 4 I.87E-03 -2.728 3.5GE 01 2.12E *4 9.2 6E -I l I.59E-04 5.33E-35 9.65E-04 -3.0;5 3.6tE 08 7.6 3 E -0 4
~
3.5nE-E 4 1. *
- E -r 4 6.121-11 1.20E-14 2.32E-95 5.10E-04' -3.292 3.82E 01 1.75E-04 9.56E 95 4.17E-Il 9.8 2E-0 5 7.4)E-t6 2.78E-04 -3.556 4.16E 01 8.75E-t5 6.62E 7 5 2 '. n 9 E -I l o.36E-03 2.67E-06 1. 56E-C 4 -3.806 4.68E 01 4.iTE a5 4.64F-15 2 12F-Il ~~ 4.54E-t 5 9.53E-07 9.llE-C5 -4.C41 5.45E 01 2.19 P 5 ~ 1.76T P 5 ~ ~ ' ~ ~ l .4 2 E -l ! ').23E-05 3.49E'-37
~
5.48E-05 -4.261' 6.57E 08 1.39E-n5 2.3"E-r5 1.9tE-Il 2.29E-C5 1.20E-17 3. 41 E -0 5 -4.468 8.86E 01 5.47E-r6 1.62F-05 7.r 4F-12 8 . 62E -0 5 4.25E-08 2.18 E -C 5 -4.662 I.54E 02 2.7 3E-96 1 15F-P5 5. elf-12 1.15E-15 1. 50E -0 8 1 42E-C5 -4.847 1.36E 02 1.37E-06 8.lfE ne' 3.54F-12 1. i lt -C 6 5.32E-09 9.49E-06 -5.023 1.82E 02 6.8 4E-0 7 5.74E-r6 2. 5"E -12 5. 4 4 E-( 6 1.89E-09 6.42E-06 -5.192 2.46E 02
'4. 40 E -( 6
~
T;{2f-TT"- -~ 4.?61 74 - - ~ i T7 7E-12 ' ~ 4'.E 6E -0 6 6.65E-10 -5.357 3.37E 02 g 1.71E-07 2. 8 7F -? 6 1.25E-12 2.ETE-r6 2. 3 5 E -1(- 3.04E -C 6 -5.517 4.66E 02 e 8.54E-38 ' 2. 8. 3E -r 6 8.e5E-s3 2.0 3E-0 6 8.31E-Il 2. l l E-C 6 -5.675 6.48E 02 4.27E 4 1. 4 3E -0 6 6.26E-13 1.4 3E-0 6 2.9'E-Il 1.48E-C6 -5.831 9.06E 02 2.14E-08 1.C IE " 6 4.42E-13 1.6 tt -( 6 1.04E-ll 1.04E-06 -5.985 1.27E 03 1.07E-44 7.17F ' T 3.13E-13 7. l TE-0 7 '~
3.67E-12 7.28E-07 -6.138 I.79E 03 ,
5;RUr4- T.'MUt 1 - ir2 3r ~l s i.f7E-07 T.31 6 12 5. 8 2E -0 7 -6.290 2.58E 03 2.67E-r9 2. 5 9E -4 7 1.56F-13 1.59E-07 4.59E-13 3.6tE-C7 -6.442 3.54E 03
~
- 1. 3'4E -e? 9
~
f . 54E -0 7 1.11E-13 2.54E-37 1.62E-13 '2.55E-07 -6.594 5.00E 03 6.68E-tr 1. 79E -( 7 7.m2E-14 1. 79E -0 7 5.74E-I4 1.80E-07 -6.745 7.G6E 03
~
3.34E-te 1. 2 7E ~ 7 5.53E-14 1. 2 7 E -0 7 2.93E-14 1.27E-07 -6.896 9.99E 03 1.67F-If R.06E
- 9 3.91E-14 8. 9 6E -0 8 7.lTE-15 8.98E-08 -7.047 1.4tf 04 p;}z pI g--- ' ' T)4p e"-' -"2.16f'-14 6.34E-08
' ' ~
^ 2.s4E-15 6. 3 5E-0 8 -7.197 ~ 1.99 L O
4 4.17E-Il 4.4 9E
- 8 1. 9 6E -14 4.40E-OS 8.97E-16 4.49E-08 -7.348 2.82E 04 3.17E '9 1.19E-14 3 .17E -f. 8 3.lTE-16 3.17E-08 -7.499 3.98E 04 2.09F-il 1.948-11 2.24E " 8 9.781-85 2.24E-08 1.12E-16 2.24E-08 -7.649 5.63% 04 >
~5;2 2612 ~ ~ "I .59Ef 8
~
6.91E-15 1. 5 8 E -0 8 ~ ~ 3.96E-17 't.59E-08 -7.80C ~~ 7.96E 04 'O 1.12E-c8 -7.951 3.13E ,0 5 Q 2 die-12 1.30{-!?
1.12E ^ 8 7.928 a9 4.99E-15 3.461-15 1.12E -e n 7.92E-09 I.436-17 4.95E-18 7. 9 2 E -0 9 -8.101 1.59E 05 O 6.52f-13 ,_5. 60 f -C 9
- 3. 9 6E '. 9 2.4 4 E -15 1.71E-15
- 5. 6a E -0 9 3.96E-09 1.75E-18 o.19E-19 5.60F-99 3.9 6E-0 9
-8.252
'8.402
. 25E 05 3.18E 05 h
p.
3.26E-l)
.-.....- I.63E - - - - . - 2.l n1F-8 9 1.2 2E -15 2. P 3 E-0 9 2 19E-19 2 5CE-09 -8.553 4.50E 05 x g .99F
- 9 ' 8.64E-16 1.9FF-09 7.74E-2E I.98E-09 -8 703 6.37E 05
, _ ].74%,-2( . . _ , .-8,854 9.01E 05, tIf 4.17*-14 1. 4 3 -R , _, _f . I l t - l .6 , ) . 40 E-r 9 l.40E-09 2.34E-14 9.97E-te 4.32F-16 9.9'. E - t e 9.6SE-21 9.90E-te -9.C04 1.27E 06 8
(D
';D= ,
r- -
Appandix B - (Ref. 10) i
- g> 1 Nishlzawa et al. SlytAY ITEM:AltCil-JAl'AN half-lives defined above. however, for prolonged Gas-liquid partition coefficients for iodine and spray at lower temperature, rate of decrease of iodides are determined to make cicar the washout iodine concentration has a tendency to saturate, and at higher temperature rate of decrease nearly effect at 25 tofor80*C iodine. The using determination a glass is performed vessel, distilled water, (
follows the rule of the first-order reaction. If the and 8I; the results are shown in Fig. 9. At any /
partition equilibrium between water phase and gas temperature, the partition coefficient # is much /
phase 1:a the vessel is realized, airborne iodine larger at the lower concentration than that at the ?
concentration can be represented by the following higher concentration. At much higher concentra '
equation from material balance: tion near saturation, hcwever, # becomes constant probably due to the realization of Henry's law, M cnu ns a w H ecomes ci = ( H - 1) Q00 + Ft) +1 constant would be higher at temperatures of 50 I,
and 80*C than at room temperature because the equilibrium constant for hydrolysis of iodine in-where creases with increasing temperature.' Since the c = iodine concentration in air, mg/ liter relation between the airborne iodine and H may be represented by straight lines as shown in Fig. 9, co= initial iodine concentration, mg/ liter the following equations are obtained from the H = partition coefficient of iodine, concentra- equations of the straight lines and Eq. (3):
tion in liquid / concentration in gas d
F = spray rate, liter /h -jc =
88 F y c (25*C, ca > 0.5 mg/ liter) (5)
V = volume of vessel, liter t = spray time, h.
~ "
y c (25T, c < 0.5 mg/M W The curve ci in Fig. 8 is the relation between et and t, takin; 100 liter /h as the spray rats 100 as - =
y c!*" (50*C) (7) tl'e partition coefficient, and 4900 liter as the vessel volum e. The curve en will shift to ct' den 39F because 200 liters of water initially charged con- ~7"y c o.., (80'C) . (8) tribute slowly to the gas-liquid equilibrium.
Another model that can be assumed is that the partition equilibrium is realized at al.1 times Equations (5) through (8), with 4 mg/ liter as the between spray drop and gas phase while drop initial iodine concentration and 100 liter /h as the falls, and that iodine transferred to water does not spray rate, are shown in F16 10 as three curves.
return to the gas phase. Then the rate of decrease of the airborne iodine concentration can be repre-sented as follows: t o'
- = C . (3) w ,, _
We have by integration, u . 7e c-'* irs ci H
- S3C** 15o'cl c = ce exp t
, (4) b 0#
NN ' "'
where each symbol corresponds to that of Eq. (2). . as c. @'TdED "'7E" The straight line, c in Fig. 8, shows the relation between c and t when c, is unity.
f Comparing c 25.e, is sonic acio
. so.e,[c snLLED watra
+
Eqs. (2) and (4) with the experimental curves it j# . so c, i l Ne acio -
appaars that the curve for the room temperature so.c, gst to warra can be approximated by Eq. (2), while the curves . so c. n sonic acto for the initial temperature of 125 and 150*C do not 'i r* ir- ir* ir* w* i io indicate a tendency to saturate and they can be sooing couegur3arion in cas ,Hast, e enenti r represented by Eq. (4). The curve for 100*C may snow a transient condition between Eqs. (2) and Fig.9. Partition coefficient vs lodine concentration in (4). gas phase.
NtJCLEAR TECHNOLDCY VOL.10 APRIL 1971 . 491
/
Appendix B - (Ref. 11)
' / A Eri G - d u G */
5 <
j
/ 10 l
i l .
l l
~
pH il j ./
I APPROX.
g ONLY
[ -
/
I w m f a lO' -
l
' )I j,
- u. u.
O O
- ./ .
f ;
g g e e
/ pH 7
/ :
e e -
d 4 l {
w w -
i 5 y /
1 o o 9 9 10 3 /
pH 9 pH 7
[ ,-
/
l cn en 4 / /
=. Y' ./pH 5 i
1 LL
-\
_\ /
w i O \
./ pH 5 U -
\
l z
\
[ ;
9 ' SOLUBILITY ,/ pH 3 4 l b LI MIT */ PH I '
H 1 lO 2
\
^
,q y {
< a : l NOT INCLUDING IODATE REACTION i -
l INCLUDING IODATE REACTION
~
i t
I
-3 gh 0 7 i 2 10 10 10 ICi I IO'I 10 l
AOUEOUS CONCENTRATION (,9 IODINE / LITRE)
FIG. I . PARTITION OF IODINE BETWEEN GAS PHASE ,
AND WATER AT 25 C
-.m- , -..-.-.,-n.w- , - , . - - . , , _ , . , . . _ . -
,,,.,-.-....-e , - - , .
, . . _ - , e
- ' *'ww --pw,'q
~
t t4ty g,+j'e/> '
'%*4
+ _ .. _ _
TEST TARGET (MT-3) l.0 g nu, agg s,yLE l.1 Q" llM -x
! La 1.25 1.4 jg
~/ : 6" >
- %3,, / '
4'
- $tf,,D/y -
, W,,i!)k ,
. a
F dM ww@ ~m=-
- 'k?.$> __
TEST YARGET (MT-3) k.if j' l.0 eIn EL4 1
-!,sugg i
ti
y l s l.25 1.4 I 1.6
-c I
/ 4 6" >
4>,% +<>4 '
"Idh __
k{i).
a . . . . -.
//// , $
b .
NNN\ //
v- l MAGE EVALUATION N TEST TARGET (MT-3) 1.0 EEEBE
'E'yMk 1.1Si'"EE 3 i ! ! =8 l1.25lu g
/ = b -
F .
4 4/> 4%' #4'\
- $ 'f 'N,/ Dj0~%
, n,,/, , .;y 1
u _,
D k W'%-MGMEEEGWSi n g; .:!. .:6'cn m.WH My.y-l- l.e-:mkw G.G'.~QQfMW 7]QQR.$m.
-m::"w c- -*'t&MkW.
- QMy rm .~+.4.,_ .im& Appand2.x B - (Raf. 13 3
,;_g .--opp,w a=u,1--o m.e.ac_m
- Q bcf '$b c]n ensioE cow. ?
l u '[ Or!
j
- x. m . mn ,m mrararPon DF ;.
13 - /.y '
U JUL 151971
_ _blWLbu U L st.,sn + sa,smoc-era,n i
beco,r:e*terk ics37 .i t
i I
f i
I E:ude exascimentale cu comportemen: I DSS proc ui:S Ce :ission en cas c'accice'n:
sur un rsacteur piscine ;
Par J. DADILLON
- i I
[.
I RL t;us : -p i au -
fin disment essentiel et ecpen.lant parinis nigtigs parce que mal connu, dans l'ivaluation des risques nuiliair s anocifs fonctionnemust d'un darteur, c>t !.a evnnaissance, en c== J'arcident, de la nature, de l'importance ei .lu r.mip..1ement Je la ,
, evnisminarian r4ellement liberse, a 11mirieur de renceinte.
Drns 1 1 cas particulier des r4artcurs piscines, le tiu.de de refroidinement s'avire etre une tsarrisre, tres erficace vis-a-sis de la runtrmination. .
au sein m+me du rseur, la ..'
. Trois experiences ant t i d faite dans le rearteur C1.URI au ruur, .lesquelles, a eif provoquse G f ; fusion .le quelques plaques .ritements rombu-titeles. 9 A uusav:
kin the estame, tion af nucl-ur rids ennne.ted arith the running of a reactor un eventini factor, sometimes neglected because ,
lmsufficiently knourn, is the krouled,te of the type, amount and behuniour of the cuntamination actually relewd inside the ';
l.unsuiner in the case of an ursident.
r .'In the special cu<e of swimming pool reactore the conting f uil d pron es to be a very efficient barrier agannst contamination. -
t l
'Three esperiments urre carried out in the reuctor CAllHl. during uhich sei eral fuel element plates nere molted inside the Trur., itself. 2
- Sans prisumer de l'ampleur maximale de la fu- simplifid lor 4qu'il s'ailrese i un type de rEacteur Sion au .eptible de se proiluire dans le emur elu dEtermind.
l r4acteur, il ett primordial de connaitre, en prem.er l lieu, l'ordre de grandeur du taux d' Emir lon a par- En ce qui concerne le rJacteurs piscine <, il a tir iln comburtible, des produit de firsion les plus stJ choi*i de commencer cette Jtnile dans des comli-dangercus et. en <econil lien, le pouvoir ele rJten- tions tri s soi,ine. de la realit4, ce qui rend ruti-
$ ion ein ilnide ele refroiili-rement. Le probleme, ex- lisation des r64ultats beaucoup moine ddlicate.
l trimement comple. se . lane son en-etuhle, re trouve \l alben s ru-e men t, la cin/tique et le- ton.litions l
- crs i. e
. .I' Et n.le- .l.. - Errei. f t...li...=6q ue.
l, l . . - wv. ,me . . . . . . ..~..,r.. -
P00R ORS P
.. y .. .. - - ,- .. -
-u.:mw.
, Mm -cM.Sih. ,S...- c .yw-@wms,mpPqwv.q.rg .or m e m =. m -wCT+2'2"3 m h w-:Mor:: wre q.;n,y .
App:nd2,.x B - (RL,. 13) 44 W..
q i o B.LS.T.
.....wn .a a ro , s u....ny..,
h.Sj Wune fusion acci.lentelle . rune partir du coeur d'un conecrnant rin.le 131 ain-i que les artisites dr, Q rfacteur riequent de ne pa Otre toujnurs rigourcu. autrer iusle et du tellure 132.
I% wement identiques : une estrapida tion peut alors rw s'as Errr nEce<-aire. " If 8"'In on 5 trasne reau Nant tr.1- faihlr comtne ma le s erra. Ic taux ardud--ion .run produit les trois experiences rEalirEe3 dan- le rdacteur i fission A partir du comhu.-tible, est igal au rap.
? CABRI ont permis de determiner le- taux d'dmi3 port le son activirJ dans reau a celle qui ce truu.
sion de riode et du t llure a partir du combu.tible vait dans les plaque, avant leur fusion. Il e-t bien
.h fondu, ain i que le pousoir de ritention de reau. Gvident que la partie, non endommasse du creur.
?i Le niveau, toujours trhs bas ilu taus d' irradiation n' entre pas dans ce calcut
.5 des J16ments combustibles ile la piir (quelques
~
Q dizaines de MWJ/T) et rampleur de la fu. ion vo. l* ' "* 'IE*I'"i , de rinde a varis de 2 5 10 g lontairement limitie a quelqurs plaque, (une pour suir nt rexp rience. Celui du, tellure ele 0.2 5 5 p.
la premi&re expsrience, trois pour les <uivantes),
Nous favon, dej5 dit, le t'aniW;mi3si.m du pro.
.) nous ont contraints a nous limiter aux produits le' duits de fiseion, au cours ifune fusion de combum plus volatils en portant toute notre attention sur gib]e, est un phEnomEne beaucoup trop complexe Q riode, qui e-t le plus dangereut trentre eux. pour qu'il soit possible, actuellement, d'en fixer N Les deux premi&res expiriences ont its faites les valeurs asec certitude, surtout quand il s'agit
% au cours de tents du comportement de la pile dans des produits les plu, solatile. Seule une valeur pro.
certaines condition > de tempGrature et de d6 bit, la hable peut Etre asaneGe : ain>i, dans le cas d'un f.. troi,ihme a 4t4 #pecialement pr6parie pour r tude es eteur a cau, une fusion inhne trGs stendue du
$ de la ditTu.iun de la contamination libsrJe par la emur, mais limitse dans le temp 5 queh ues dizai.
l I fusion du comho-tihte. nes de secondes, ne des rait pas provoquer une i smi
- ion d'ivde et de tellure dans reau, supJrieur-a Lor, de chaque test, la tempdrature moyenne de i 10 9. Par contre, le taux WGmi -ion de> paz me.
j reau stait de 30" C et son dibit. L travers le exur .ur4 de fgou peu preci e S pa,rtir de leur3 de cen.
) de 400 mVh environ. la circulation re faisant tou- dants solides, pourrait asoi iner 100 g. Quant aus j jours de haut en bas. Le niseau Weau a varis de produi 3 solides ou peu volatile, 11. n'ont pas ete 5.30 m 16.10 m, ce qui correspond a des hauteur' ddteetes dan, reau, 5 rissue des pri entes expf-Weau an.dessu du coeur de 2.50 m a 3,30 m.
riences. Cependant, leur taux W6 mission a partir De nombreux prslivements efiectu6s dans reau, #"" C "ib,ustible diterminf e>t beaucoup moin-a didsrents instants aprGs la fusion, ont montr6 I"II"'"es par les e ndition* de fu.-ion e.t derneure.
que la contamination etait trbs homogene. de toute fapn, tr63 faible. Les s aleur, dJterminse4 su cour Wexp riences 5 petite echelle, effectu;e-notamment aux Etats.Unis, peusent stre consid;-
r6es comme parfaitement acceptable 4
- 1. TAUX D'fMISSION DES PRODUlTS DE FISSION L'analy<e des dchantillon< prilesGs sur les pla-ques a fait appara.tre i une perte d uram.um de 1 o r-A PARTIR DE COM B USTIBLE FONDU dre de 70 7c, msme en des emlroits on la gaine n'avait pas fondu. La presence entre le* plaques, Nou, rappelleron- qu'un chargement complet du de coulies Walliage, lai--e <u p po<e r que celui-ci co ur de CABRI se compo-e Wenviron 300 plaqurs peut se trouver en fusion sans que la gaine, plus l
' Walliage u ranium. aluminium den poids 20 g refroidie, ne soit fondne et que, de ce fait. il Wuranium 80 g Waluminium) reprA3entant au .'scoule par les points oh la gaine a rJellement total 4 kg Wuranium enrichi i 93 fc en U.235 fondu.
(90 9. 5 PJpoque un ,'est dJroulJe la prAsente ex.
pGrimentatinn).
I A riyue des deux dernihres experiences, plu- IL PitGEAGE DES PRODUlTS DE FISSION
,ieurs Jehantillon, de comhu-rible ont dts prilesd* DANS L'E A U sur les plaque , tant .lan.- le* parties intacte3 que
, dan- les partie3 fo nd ues. Leur analy c a permis de l dsterminer le poid Walliage di-paru et les quan. Srule la retension de rinde dan- re.m pu Are tit."- Wimir pre-rnte- asant ch ulur insion. Par all. darrininee au cour de en estnb ienen rn rder.
Irur . Ir calcul a fourni de- s alenr- comparatis r- er facicor Jtant i d.- im port.mr. Ic- .ntisitJ. qui l W %.. . g. =.m - a w .m .. -. c
--~ ?00RBR M l,+.. .
w; q .. ?.ac~_-m - m _-r. .my magg.g;c :m.- - - -
App:ndix B - (RSf. 13) M 4
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atteignent le hall.pite sont tr&s faible.. En ce qui Seul, le premier di.-positif a ets utili-s lors de* ,j concerne les gaz, dont la rJtention est moin, deux premiAre, espsrientes ; sa con titution mAme ' .'
~
gramle, les systhmes de pis-eage dont nous pou- ne permettait pas it'obtenir des information, sur '[
vinn, dispo<er, ne permettaient pas de faire des le comportement de l'iode.11 est sentement pos- i . e-
,ible de dire que le papier Delhag retient de- quan. 'h prsiesemen't, sslectifs suffisamment longs.
de celle, ad-orbie, .
I,.io de a traver.
titss led'inde inferiuire, 5 10 %.fierait ilue neolus de I.e facteur de tranonursmn de dans charbon, ce qm. ogm I,cau 3 exprune de la facon cuivante : .. .
10 % ile 1..iode .e trouvent anocies a des ac. rosols.
AH ,w k= Pour la troi iime expsrience, une enceinte de 1:
AE + AH confinement stait plaese au-de .us de la cuve du ~j -
AH activits de l'iode smis dans le hall-pile. rsacteur, reduisant le volume libre, an dessus de l' eau, a 150 m' seulement au lieu de 2.500 m' qui AE activits de l'iode dans l' eau. reprssentent, approximatisement. le volume total Suivant les expsriences, les isotopes et les pesis- du hall. En outre, des ensembles composites staient verr.ents a partir desquels il a st calculs, ce coef- cette fois utilisss. .'
freient se situe entre 5104 et 510-8. Autrement dit, Une assez forte quantits de vapeur apparai sant seule une partie comprise entre 0.0005 7e .t an. des.us du esseteur au cours de la montee en 0,005 % de riode smis dans l' eau pendant la fu-sion du combustible, atteint le hall pile lui meme.
puissance, le taux d'humuh,te relatne atteignait 95 % 5 l'intsrieur de renceinte de confinement.
En outre, une ssrie d'expiriences hors pile (PI-l'5 9""" tile 8 d'iode fixses sur les couches sue.
REE) a n'ne schelle voi,ine de la rsalits, a montrs ces ives de charbon d'un mEme ensemble, staient que, m?me dans le cas on l'iode stait entrains dans s avent du meme ordre, ce phe,nomene ctant du, 2
l' eau par un gaz ous une pression de 20 kg/cm , en premier lieu, au taux d'humidits sless qui a !
S unr tempsrature de 400* C et suivant un dsbit diminus, dans des proportion, considsrables, refti-important de 20 g/s pendant une minute, seule une cacits des charbons. Il est imposibic daus ces con.
partie sgale h 0,1 Tc traverrait une couche d' eau ditions de d6terminer le pousoir d'arrst riel de ces ;
de 3 metres. ensembles, mais les diver cs couche, de charbon ,
Par conse.;aent, dan, le cas d'un accident de ont prssents des efficacits3 qui, ramenses A la va- ,
fu-ion de combustible survenant dans un riacteur leur de 5 cm ile charbon seraient gsnsralement in-piscine', on peut considsrer comme valeur probable Isrieures a 95 9. Le taux d'humidits n'est proba-du coefficient de transmission de l'iode a travers blement pas la seule cause de cette diminution l' eau 0.005 % et comme limite supArieure 0,19 d'efficacits des charbons, la formation de compo- g stant prsei+4 qu's notre sens, on resterait certaine- s s moins bien irrerse que riode motAculaire pour-ment as-ex sloigns de cette valeur limite. rait stre une explication trhs vrainemblable.
i I
111. COMPORTEMENT DE L'lO DE IV. CO NCLI.1510N i DAN 5 LES FILTRES Il apparait que les con, quences d'un accident mettant en cause la fusion d'une partie du ceur l Deux dispositifs expsrimentaux de pisgeage ont d'tm rsacteur piscine, ne seraient pas aus.i :traves i ete utihses au cours de cette expenmentation :
que ce que ron pouvait pen er jusqu'5 prssent. Ces
- des dispositifs simples compo*ss d'u n papier quelques expiriences ont ruontrs que le combus- ;
Delhag suivi de 5 cm de charbon ; tible, reau et les matGriaux constitutifs a Pints.
l rieur du ceur.reenaient 99,9995 7c de rinde prs-
- des ensembles composites comprenant en nom- eent dans la ponion de co-ur emlommagie.11 n'e t i bre variable suivant les cas : . . . pas posible de ge nsraliser encore au cas de fusions
- papiers d, ann. ante, la.me de cuivre actnee a extrsmement brutale, engendrant de grande- quan-1 argent, papier impregnes au charbon, cou- . .
titss ( e sapeur. S..eanmoms d n'est pas de.rai-nn-che, de 1 cm de charbon. .
naga;e i)e pen-cr que dan < la tre, -rande maj.on t-Le3 pa pier, amiante et le charbon staient du ,raccident- entrainant la dersrioration d'une por-tion du cwur -ent 0.19 de, indes serait emis an-mAme type que crus utili es dan. le, in-tallation-dr -u, d.., p i-ci ne-nucleaire, indu trielle.. ,
P00RBRBNAL.
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