ML20237D825
| ML20237D825 | |
| Person / Time | |
|---|---|
| Site: | General Atomics |
| Issue date: | 12/22/1987 |
| From: | Asmussen K GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | Rubenstein L Office of Nuclear Reactor Regulation |
| References | |
| 67-1121, NUDOCS 8712240216 | |
| Download: ML20237D825 (26) | |
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AINidEllEINEEEEB1JI5fFEEEDEDIR53RIMJORMDumestmanualDr# Tem CATechnologies musser" IJsatemumar( M3sserA f
G A Technologies Inc.
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$AN DIEGO, CALFORN!A 9 ?136 (619) 455-3000 ,
Decmber 22, 1987 4 67 1121 Mr. Lester Rubenstein. Director '
Standardization and Special Projects Ditwivrate '
Division of tWR Licensing-B ,
Office of Laclear Reactor Regulation U.S. Nuc'inar Rc<Julatory Cmmission Washington, D.C. 29555
} Subjecc: Docknt No. 50-163: License No. R-67; Tudificaticy of
%enali:nic Test Facility as Result of Lite Boundary Chany (3 con es)
\
Dear % Rubenstejl.:
Since early 1935, a c:ntinuous and ongoing Department of Energy '
research' test program for thermionic devices has been successfully conducted in GA Technologies Inc.'s V;A's) Mark F TRIGA reactor facility. We pertinent safety analyss for these tests (experiments) have been predicated on the nearest d te Mun&q being at a diatreme of 350 m from the @proximata center ' af the reactor building. A reactor itself has always had an exclusion area with a minimum radius of 120 feet (36.6 m) frcan tbs approximate center of the TRIGA Mark F reactor fecility building as set forth in the tochnical specifica-tions, Emetime in the future, the site loundary nearest & GA'7 TRIGA ' reactor facility wil3 be rdocated inward to about M6 e ,
coincidentally, an activated <br.rectlre ap will be installed in tp Mark F reactor room. The net 's.ff t.:t will be a reduction it the rSk
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to the health and safety of the pubiic.
The above mentioned of.tivated ?narcoal trap will be installd pik to j the actual change in the " effective" nearest site boundary di.tJiv e.
Such a trap will reduce significantly the potential thyroid dose fic 5 ,
person at the site brandary. (The " effective" site boundary ma,9 rrruin at 350 m for a cmsiderable time alter the official boundn /
change because of significant delays in tre hvelopnent of the sciere (industrial) park real estate.) ,
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The purrose of this suhaittal is twofold. Fint, it 7/tesents for; the closer l te i boundary, and the added activated charcoal filter, an up ;
to-date analysis which demonstrates a reducth. in the consemences ,
for the piblic of a postulated acddental fissim product release from-a thermionic device. Tecond, it requests approval of suggested changes to the subjec' technical specifications which a) require an activated cuarcoal filb: to be available when testing thermionic
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O, devices in the corte whent/cr 'the distance to the nearest site boundary is less than 350 m; b) define surveillance requirements for such a filter system; and c) require ran annual measurement to assure adequate tightness of the reactor n u .1 A safety analysis for the cuabined effects of reducing the dictece tai-the nearest site boundary and introdtning on activated charcoal flir.er into the reactcr rom is presented in the arcachment. It ray be m#ed
,that several effects are en work to reduce the risk to the public of a
. fissico product release and to furthar enhance confidence in de safety of the thermionic devices. Firs'c is the fact that none of'the thermionic devices out of the dozens tested ~ for tens of flionsands of y hours during the past two decades has leaked fission goducts even I>- into the secondary <;cntaiment. Second, is the fact that we propose a h.;i reduction in the mehrm power in each device to 5 W frm the 14.4 KW approved earlier for the R-100 Mark III therm!cnic tests. Third, and u very important, is the fact that copper bus bars in the primary containment will, with certainty, act very effectively to absorb iodine. In fact, it appears that much less than one percent of the iodine would be available for release frm the primary contauwent ccnpared to the 10 percent release assumed in this treatacnt ard the earlier analyses. Finally, the installation of an activate 3 charcoal filter in the Mark F reactor room will greatly reduce the thyroid dose frm any postulated.< release to persons at the alte boundary., In fact, it will reduce this dose below that already aIproved for thermionic operations with the 3% m site. boundary. GA, _ therefore, concludes that there is no increase in rin to the iblic health and safety as a j" result of the reductim of the clistance to the nearest site boundary u fra 350 m to about 115 m with attendant facility modifications.
Rather, the steps proposed herein actually result in a decreased risk.
Suoqested Wordino for_' f ecNical' Speci/ Mat. ions 6.,2 and 10.2 6 ,
Following is suggested wotting for dditions to Sections 6.2 and 10.2.6 of the subject ?.:echnical Specif cations. We are confident that these suggested chan;es, in addition to the conditions already'in effect, will asstre that adequate protecticn is afforded for the public in the exttmely unlikely evert of a fission procuut release from a themienic device.
Section 3.5 (a new section)
" Adequate resistance to leakage of air frm the reactor roca ,
shall be verified prnually by nmsurement when a charcoal' filter is required for reactor operatioans. The resistance to air leakage shall be at,least 1M , where 6 = 6.38 x 10-6 m sec/kg." '
3 m L- A
l Section 6.2.6 (a new section)
"When an activated charcoal trap is required for the testing of thermionic devices, its blower shall be verified to be operable at least once each week. A sample of the activated charcoal shall be tested semiannually (at intervals not to exceed 8 months) to confirm acceptably low levels of water / humidity content."
section 10.2.6 (i) (a new section)
"During tests when irradiated thermionic devices are present in the core, regardless of reactor power level, a high efficiency activated charcoal filter having a nminal flow rate capacity of 460 cfm shall be available in the reactor rom whenever the effective site boundary is less than 350 m from the reactor rom."
In view of the fact that- there is no increase in the likelihood or consequences of hazards as a result of the proposed reduction in distance to the nearest site boundary and the attendant facility changes, we look forward to favorable consideration of this appli-cation. Should you have any questions regarding this application, please contact me at (619) 455-2824 or Dr. W. L. Whittemore at (619) 455-3277. We thank you for your early attention to this request.
Enclosed is a $150 check for the administrative fee associated with this application.
Very truly yours, Keith E. Asmussen, Manager Licensing, Safety and Nuclear Cmpliance KEA/WLW/mk Attachments: 1. Safety Analysis for Decrease in Distance to Site Boundary
- 2. Check for $150 cc: Mr. John B. Martin, Regional Administrator U.S. Nuclear Regulatory Ctmnission, Region V
_ - - _ _ - _ _ -_________-__--_____J
l L .
STATE OF CALIFORNIA )
)ss COUNTY OF SAN DIEGO )
On this J.he 42nd day of hM , 1987, before me, PMA /6. bsH , the undersigned Notary Public, personally appeared K. E. Asmussen, Manager, Licensing, Safety and Nuclear Compliance, proved to me on the basis of satisfactory evidence to be the person whose name is subscribed to the within instrument, and acknowledged that he executed it. WITNESS my hand and official seal.
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^; o m e w. SEAL. .
BRENDA S. DAW 8ON 3
%- Notary,a Signature
); SAN DEGO COUNTY Wy Comm. Emp. asp. 7.1990
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4 License R-67 Docket 50-163 ATTACHMENT SAFETY ANALYSIS FOR DECREASE IN DISTANCE TO SITE BOUNDARY.
Introduction At some time in the future, the distance from GA Technologies' (GA's) TRIGA reactor facility to the nearest site boundary will be - reduced from 350 m to
)
about 115 m. This decrease in the distance to the nearest GA site boundary i
does not have an important effect on the health and safety of the public for ;
any experiments and tests except possibly those related to the irradiation of direct conversion (thermdonic) devices. All non-thermionic operations autho-rized under the R-67 license are. served adequately by the exclusion area having i
a radius of 120 feet (2 37 m) from the approximate center of the reactor facil-ityh building. In particular, the reactor itself is fueled with a matrix of ;
uranium-zirconium hydride which provides a very high degree of fission product l retention. The release fraction -- typically 10 4 for TRIGA reactor fuel --
l together with the demonstrated inherent safety of the reactor and its fuel as- i sure that the exclusion radius of 120 feet is entirely adequate. Experiments authorized under the applicable techr.ical specifications, except for thermion-ics tests, fit within an envelope that is also consistent with this same exclu-sion area. On the other hand, the thermionics tests involve larger amounts of nuclear fuel than other approved experiments. The safety analyses performed j
- i for ' the therudonic tests have evaluated the consequences for the public and i found that a larger separation is required between unrestricted areas that may !
1
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be occupied by members of the public and the TRIGA reactnr 'cility. The ear-lier approved analyses for much larger inventories of fission products showed that the radiological effects were within the limits of 10CIR20 for unrestrict-l ed areas starting at a distance of 350 meters from the TRIGA facility [1].
y For the present case considered herein, the distance from the TRIGA reactor l
facility to the nearest site boundary ud11 be reduced from 350 m to about 115 J
- m. When considering the effect of this change on the health and safety of the I I
i Page 1 )
License R-67 l
L Docket 50-163 public, it is L also important to note that several improvements and changes have been made in the present generation of thermionic devices, such as the reduced fission power. levels and further refinements in their construction. These changes result in reduced fission product inventory and increased barriers. to the release of fission products from a thermionic device, thus posing less risk to the public. -In addition, we propose to install an activated charcoal filter in the reactor room to trap the iodines in the unlikely event of a fission
. product release. The analysis to; follow will evaluate the combined effect of these changes on the radiological conditions at the site boundary in the very unlikely event of'a release of fission products from a thermionic device. A complete recalculation of the radiological doses and concentrations has been made.
In the following considerations, the new reduced distance to the site boundary will be taken as 100 meters, rather than the expected actual value of
'115 to 119 unters. 7te slightly shorter distance will allow for any small variations in the final site boundary. The actual change in' concentration at i distances in the range of 100 to 119 meters is relatively small because the ef-
.fect of the building wake becomes increasingly important, rather than disper-sion, for the shorter distances.
Basic Integrity of Thermionic Devices
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A basic consideration in assessing the risks to the public from possible release of fission products from a thermionic device is the fact that not one thermionic device tested under the GA Technologies program has ever leaked fis-sion products even into its secondary containment. Since dozens of thermionic devices have been irradiated for tens of thousands of hours, this result is particularly significant as regards confidence in safety.
The basic construction for two types of the current generation of thermi-onic devices is shown in Figs. I and 2. Figure 1 shows pertinent details of thermionic fuel emitters that have been under continuous testing since January Page 2
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License R-67 i
Docket 50-163 1985. The diameter of the in-core device is about 2.5 inches and is inserted into the space left by the removal of three TRIGA fuel elements. Figure 2 is a schematic representation of the newer type of device also to be tested in this program. Its in-core diameter is about 1.5 inches and it is inserted into the space left by the removal of one TRIGA fuel element. Apart from other differ-ences, the secondary containments are somewhat different. In the larger diame-ter devices (Fig. 1) the secondary containment is separated into a lower and upper region. The other type (Fig. 2) has a single secondary volune and is the same in this detail as the thermionic devices successfully tested in this pro-gram in the 1960s and 1970s. However, the important point to note is that each of these two types has a secondary containment that is surmounted by a very long, many segmented plastic seal. Both types of units have seals that are es-sentia11y the same as the one sketched in Fig. 2. The plastic plug is about 53 inches long and is poured in several steps to assure a void free plug. This plug contains segments to control thermal neutron upward leakage (boron frit) and to control gamma ray streaming (Pb0 glass). Since this plastic plug is 53 inches long and its top end is below the surface of the reactor pool, even if fission products were to enter the secondary containment, they would be well below the surface of the reactor pool and would be prevented by a robust barri-er from entry into the reactor room. It should be noted further that the sec-ondary containment is maintained at a slightly negative pressure so that any leakage would be inward. This assures that no fission products in the second-ary could leak out except by diffusion even if a path to the outside should appear, such as for example, a broken purge line.
Practical experience has demonstrated the integrity of the internal con-struction of the thermionic devices. Since no fission products have ever been found in the secondary, the reliability of the assembly and construction of its internals is well demonstrated. The extra long plastic plug now used above the secondary is considerably longer than that used in this program in the early 1970s. The combined barriers to the release of fission products is formidable and constitutes a major element in the protection of the facility workers and the public.
Page 3
License R-67 Docket 50-163
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Inventory of Fission Products Both the U-235 loading and the in-core location of the fueled device deter-mine the resulting equilibrium fission product inventory. All of the devices in.the present thermionic program are limited to a maximum of 100 grams of U-235. .As noted,above, two types of devicas will be used in , the present . pro-gram. Up until the present, only the larger diameter type typified by Fig. I have been used. Their larger, in-core diameter assures that only the special i three-element positions in the core can be occupied. The largest fission power from any one of these devices is 3.3 kW. The new, smaller diameter devices are presently under construction for testing in this program. The expected core locations for them will range from the B through E ring. The maximum power for any of these in the B-Ring udll be 3.75 kW. If by accident one of these were to be -installed and operated in the A-Ring core position, the peak fission power would be no more than 3.87 kW. In this amendment application, we will assume conservatively that the maximum fission power is 5 kW. This larger value will provide for alternative emitter designs (still limited to 100 grams U-235 per device) that may generate somewhat more power.
The fission product inventory for the halogens and noble gases for an equi-librium distribution from 5 kW operation is given in Table 1. While these data are derived independently, they correspond well with those presented in Appendix C of the earlier Safety Analysis Report (1) when properly scaled from the 14.4 kW operation that was approved earlier for thermionic tests with the Mark III TRIGA reactor.
For use in the evaluation of the doses and concentrations at the site boundary, Table 1 also gives the activities releasable from the reactor room.
It will be shown in Appendix I, attached hereto, that the sorbing of iodine by !
the copper present within the thermionic devices will surely reduce the inven- .
tory of iodine releasable from a device to much less than 1 percent. The !
sorbing effect of the internal copper is so great that this is an important ef-feet in reducing the amount of iodine that might be released in an accident. ,
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_ _____:__ _ _ - _ _ _ _ . . __ _ _ . _ ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .. J
License R-67 s
Docket 50-163 Table 1. Halogen and Noble Gas Equilibrium Activity for 5.0 kW Operation.*
Curies Releasable **
Isotope Decay Constant Activity within Device from Reactor Room Br-83 3.02-1 hr~1 2.04 + 1 Ci 2.04 + 1 B4-84 1.31+0 4.11 + 1 4.11 + 1 Br-85 1.39+1 4.29 + 1 4.29 + 1 I -131 3.58-3 1.19 + 2 5.95 + 0 I -132 3.07-1 1.77 + 2 8.85 + 0 I -133 3.34-2 2.87 + 2 1.44 + 1 I -134 7.93-1 3.22 + 2 1.61 + 1 I -135 1.04-1 2.68 + 2 1.34 + 1 I -136 2.90+1 2.59 + 2 1.30 + 1 Kr-83m 3.66-1 2.04 + 1 2.04 + 1 Kr-85m 1.59-1 4.29 + 1 4.29 + 1 Kr-85 7.67-6 3.01 + 0 3.01 + 0 Kr-87 5.35-1 1.08 + 2 1.08 + 2 Kr-88 -2.50-1 1.54 + 2 1.54 + 2 Kr-89 1.31+1 1.95 + 2 1.95 + 2 Kr-90 7.55+1 2.17 + 2 2.17 + 2 Kr-91 2.54+2 1.47 + 2 1.47 + 2 Xe-131m 2.41-3 1.47 + 2 9.47 - 1 Xe-133m 1.26-2 6.87 + 0 6.87 + 0 Xe-133 5.50-3 2.89 + 2 2.89 + 2 Xe-135m 2.67+0 8.07 + 1 8.07 + 1 Xe-135 7.62-2 1.01 + 2 1.01 + 2 Xe-137 1.07+1 2.51 + 2 2.51 + 2 Xe-138 2.45+0 2.34 + 2 2.34 + 2 Xe-139 6.08+1 2.35 + 2 2.35 + 2 Xe-140 1.56+2 1.61 + 2 1.61 + 2 l
l
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- 1.00+2 represents 1.00x10 8 l
- 5% release of I, 100% release of Br, Kr, Xe.
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License R-67 Docket 50-163 The thyroid dose at site boundary resulting from this trivial release of iodine would be vanishingly small.
Notwithstanding the credible sorbing of iodine by the internal copper com-ponents, we have assumed as before that 10 percent of the iodine inventory within a device could escape into the reactor room in a postulated accident (1) and that half of; this will plate out on the cold walls of the reactor room :
leaving 5 percent to decay within the room, be absorbed in the charcoal filter when one is used, or be pumped by the wind out of the building and dispersed across the site boundary. It umy be noted that the several review panels investigating the Three Mile Island accident have concluded that the plate out of iodine on the normally cool walls of the reactor room is very effective in the control of iodine. Their work justifies a plate out value much larger than the 50 percent assumed herein. In Table 1 we have listed 5 percent of the iodine nuclides and.100 percent of other halides and the noble gases as releas-able from the reactor room. These values were used in the calculations of the resulting thyroid dose, whole body gamma ray dose, and concentrations at the nearest site boundary (i.e., at a distance of 100 m). It may be noted that the 33 percent release of the noble gases assumed in the earlier R-100 license has been taken as 100 percent release in the present calculations because no mate-rials that can specifically sorb the fission product gases Xe and Kr are presently incorporated within the primary containment of the thermionic de-vices, as was the practice in their construction 15 years ago.
Evaluation of Doses and Concentrations at Site Boundary A complete analysis to determine the doses and concentrations of radioac-tive nuclides at the nearest (i.e. 100 m. ) site boundary has been performed.
In the event of an accidental release of fission products from a thermionic device it is assumed that the reactor operator will: (1) secure the reactor i room by deactivating the room ventilation system (which automatically closes l
l the ventilator duct damper); (2) energize the activated charcoal filter; and, (3) close the 2 x 4 ft8 air inlet to the reactor room. It may be noted that the reactor would be automatically scrammed by the license required Continuous Page 6 l l
l
License R-67 i
Docket 50-163 Air Monitor. To reach the site boundary, the gaseous fission products must be pumped by the wind out of the reactor building through leaks and dispersed across the site to the boundary.
To determine the leakage of the reactor room under the action of wind, three effects are evaluated. First, the leakage rate from the room is evaluat-ed by determining the relative ease with which air can flow out from the build-ing under the influence of a pressure difference. Second, the pressure due to wind is evaluated in terms of the drag force per unit area. Finally, the dilu-tion of the fission products by the wind is evaluated at the site boundary.
The flow resistance is determined by assuming that the building can be described as in the earlier Safety Analysis for the Mark III thermionic experi-ments (1); namely, p = g/pV (1) where q = ventilation rate p = pressure differences v = room volume and 1/p = resistance to air leakage.
Typical measurements on the current Mark F reactor room have shown that:
~
p = 6.38 x 10 8 m sec/kg when q = 300 cfm = 0.1416 c7/sec p = 0.225 inches of E go = 56.00 Kg. m/sec 2 v = 14000 cf = 396.4 uf Page 7
License R-67 ,
Docket-50-163 l
L i.
1- . . .
l The driving force, i.e., the negative pressure in the lee of the building,
-is calculated in terms of the drag per unit area of the projected vertical '
l surface of the building using an appropriate drag coefficient (2). This drag pressure is' equal to the average pressure differential between the stagnation side and lee side of the building. A value of one-half this pressure differen-tial was used to give a somewhat conservative value for the negative pressure at the lee side of the building against the pressure of the undisturbed air.
Finally, the fractional leakage rate (Ag) from the reactor building resulting from wind forces' acting on the building is then given as:
A EV 31 2 2
-where C = drag coe m eient = 1.61 D
p = air density = 1.286 kg/d
,u = wind velocity, m/sec For the Mark F reactor room, the fractional leakage rate is therefore:
Ag = 3.3 x 10 8 u2 sec ~1 (4)
The concentration of room radioactivity Qg (Ci/d ), for fission product 1, at the time of experiment failure assuming an equilibrium fission product in-ventory is:
S 10"'i10(1-e A
P)P/V (5) f g
= fraction of N isotope released to room, A
10 " "9" #'"" "" D7 ' * '" P* P*#
experiment power, Ci/kW.
Page 8 l I
t License R-67 Docket 50-163 Ag = decay constant of isotope i sec 1 t = operating time, 3600 x 40,000 sec 1, and P = experimental power, 5.0 kW The isotope - release fractions, f , used in the analysis were 0.05 for the f
iodines and 1.0 for all other gaseous isotopes; 1.e., bromines and noble gases.
A. discussion of the release. fraction for the iodines is presented in Appendix 1 I, attached. hereto,' which justifies the use of an even smaller value than 0.05.
Table 1 presents the equilibrium activities for a thermionic device operated at.
5 kW as well as the activities of these isotopes releasable from the reactor room.
It is assumed that the gaseous fission products are introduced instanta-neously into the reactor room. Subsequent to the release, the concentration within the reactor room will decrease under the influence of radioactive decay,. q reductionbythepuppingbywind,andreductionduetotrappingofthehalideh i on the activated charcoal filter. The time dependence of the fission product inventory for each isotope is given as:
q i
,q io . - (Ag+nAyf+A)t g ;
1 1
th -
Where A g = decay constant for the i g,,t,p,, ,,c 3 ,
l Af = charcoal filter flow, fraction of room volume per j unit time, see 1, l
ng = 1.0 if iodine; 0.if not,
~
A room fractional leakage rate, sec 1, given by I g = Equation 4, )
1 Page 9 I
4 Licenae R-67 Docket.50-163 t = time from beginning of event, sec.
' Thel fractional rate of flow through'a' charcoal filter with a 460 cfm throughput
- (which removes.the iodines with an efficiency of 99.9% in one pass) is:
i A =
g (460cfm/14000cf)/60
~
= 5.48 x 10 4 sec ~2 Outside the reactor. room the concentration of the i th isotope at the site boundary (at 100 m) is qf =.Agg Q V (y/q), Ci/m7 (7) where' y/q= atmospheric dilution factor, sec/mP.
The computation for the' dilution factors follows the procedure set forth fa Appendix G of the Safety. Analysis for the thermionic tests 'in the Mark III TRIGA reactor (1), except that the 100 m distance is close enough that the ef- ;
feet of building wake must be included. The dilution factor (y/q) is thus cal-culated from the expressions ;
i (y/q)=1/[rva +CA]u,'sec/m 8 (8)
- where a a = the appropriate Pasquill relationship for I
- Gaussian diffusion of the clouds (3),
C = 0.5 from Reference [3), ;
i A = minimum building cross sectional area normal to the wind l direction, 182 af, j l
l u = wind velocity, m/pec. !
l Integration of Equation 7 over time gives the total integrated concentration of j Page 10
License R-67 Docket 50-163 each gaseous radioactive isotope that an individual would experience at a point 100 meters from the source. The fraction F of MPC (F = E F ) can be evaluated by summing over the concentrations of the contributing isotopes. The whole body (D ) and thyroid (D ) doses for the various isotopes are calculated from these concentrations. Thus, ,
F =of" -- -- dt (9) mpe,i D = 900 of E qf dt (10)
D = 3600 f*B D,gg qg de (11)
E = gama energy, MeV B = breathing rate, d /s D,gg = effective thyroid dose in Rem per Ci inhaled, and 4
'mpc,i
= maximum concentration of i th isotope permitted in 10CFR20.
The values for p are those taken from 10CFR20. The values for E are those typical of a fission product release from TRIGA reactor. The values for B and D,fg are given below.
B = 3.47 x 10 4 d /s 0 ( t ( 8 hr
~
= 1.75 x 10 4 8(t ( 24 ,
= 2.32 x 10 4 24 ( t ( m D,ff = 1.486 x 108 Rem Ci I-131 Page 11
l_
t .
d l:
License R-67' p
Docket 50-163
= 5.288 x 10' I-132
= 3.951 x 105 I-133
= 2.538 x 104 I-134 1
= 1.231 x 10' I-135 ,j The expression for the fraction of MPC -(Eq. 9) depends upon the wind veloc-ity both through the room leakage:(Eq. 4)-and the dilution factor (Eq. 8). In- l greater detail, F is given as follows: j A
g Qg V (y/q) dt q F= ----------------
(12) i mpc, i
)
~
S 3.3 x 10 8 u io
~~~~~
1 1
" {ird~d~~~~CII.
y r: 1)~mpe,i 375"ic I6 f ~u f I~n~X~I If Xi Similar expressions ~ exist for D and D g . Each of these expressions l depends upon the wind velocity u. In each case there is a unique wind velocity
-: that will maximize the dose or concentration at the site boundary. This wind velocity is obtained by differentiating the appropriate expression (eg.,
. Equation. ' 12) and setting the results equal to zero. The velocities that i
maximize the three quantities of interest are l
U F
=
11.0 m/sec U = 13 m/see 1
U = 6.6 m/sec Page 12
License R-67 Docket 50-163 The wind velocities determined above are high compared to normal wind velocities (~ 1 m/s) assumed in evaluating site boundary doses. The reason for these large . velocities is that the udnd is in competition udth the charcoal
' trap. Thus to pump significant quantities of gaseous fission products to the site boundary, strong winds (~ 30 mph) are required since the charcoal filter removes most of the more dangerous isotopes with a 1/e period of about a half hour.- It seems more reasonable to assume a more typical wind velocity (such as 1.0 m/s) together with conservative Pasquill F conditions for calculating the
. concentrations and doses at the site boundary. Table 2 summarizes the two sets of concentrations _and doses and demonstrates that the risk to the public at the new, nearer site boundary is either comparable to or reduced compared to that Table 2. Comparison ~ of Doses and Concentrations for the Proposed 100 m.
Site Boundary ~ with Those Already Approved for the Present 350 m.
Site Bourdary >
Parameter Values Values Proposed Approved for 100 m. Site for 350 m. Boundary-Site Boundary (with charcoal filter) wind velocity 1.0 m/s to maximize wind velocity
- Max. U-235 (g) 600 100 100 per Device
- Max. Fission Power 14.4 5 5 (kW)
- Fraction of Integrated 0.712 0.84*(11 m/s) 0.55**(1.0 m/s)
HPC at Boundary 4
- Integrated Thyroid 3.2 2.ud*(13 m/s) 0.55**(1.0 m/s)
Dose D hyr (Rem)
- Integrated Whole Body -------
0.084*(6.6 m/s) 0.068**(1/0 m/s)
Gamma D wbg ** I "")
- Pasquill D for u)6 m/s ** Pasquill F for u = 1.0 m/s Page 13
License R-67
) Docket 50-163 l i for the presently approved 350 m site boundary. The thyroid dose of 0.55 Rem is far below the earlier guide lines in 10CFR20 and is also well below the 5.0 !
Rem thyroid dose above which the EPA guidelines require protective action for members of the public at the site boundary (4). The whole body gamma ray dose f
of 68 millirem is f ar below the 500 millirems annual dose limit established in 10CFR20.105 for an individual member of the public in unrestricted areas. The value of 0.55 MFC at the site boundary is consistent with the value approved earlier for the 350 m site boundary.
The activated charcoal filter is very important in its action to trap any iodine in the vapor phase. Because of the importance to the engineered safety system of both the activated charcoal filter and an appropriate degree of reac-tor room tightness, we have prepared additions to the Technical Specifications covering the use of such a filter and establishing appropriate surveillance requirements. Technical details of an activated charcoal filter can be found in Appendix 11 attached hereto.
REFERENCES i (1) GA Document GA 9622 (1970) submitted as part of the license application for thermionic tests in the R-100 Mark III TRIGA Reactor.
~
i (2] " Elementary Fluid Mechanics," (3rd Edition) J.K. Vennard, published by John Wiley & Company (1957), pp 352-353.
[3] " Meteorology and Atomic Energy, 1968," Ed., D. H. Slade, U.S. Atomic En-ergy Commission, July 1968, pp 101-103.
[4] " Manual of Protective Action Guides and Protection Actions for Nuclear Incidents," Environmental Protection Agency Report EPA-520/1-75-001, Sept.
1975 (Revised June 1980).
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4 1
APPENDIX I THE RELEASE OF IODINE FROM THE DIRECT CONVERSION DEVICES 1
The behavior of radiolodine is discussed at length in Refs. 1-4. It is 1
shown in these references (particularly Ref. 3) that copper will sorb large quantities of iodine. Silver and charcoal are mentioned as being better sorbants of iodine than copper. The primary containment in the thermionic devices always includes copper as bus bars, electrical leads, or other components . Therefore, it is reasonable to assume that a large fraction of the fission product iodine will be sorbed on the copper components and will not be available for release out of the primary containment.
In fact, Ref. 3 reports the results of a test which closely approximat-ed the conditions existing in the primary containment. Sections of copper mesh were exposed to iodine crystals a.t roor temperature in an air-filled
'dessicator for periods up to one year. These copper samples were periodi-cally weighed to determine their iodine content. "The results did not indicate any ultimate limit to the absorption of Jodine by copper metal. In run A which lasted for about a year, as much as 37% c,f the copper comprising a nesh section vas converted to the iodide. After allowing for the non-uniform rate of attack throughout the section, it is probable that the outermost wires of the section had very little metallic copper left in them, which was also suggested by their comparatively fragile nature."I3) The ;
copper samples had surface treatments which provided samples with very clean surfaces (etched in nitric acid) and highly oxidized surfaces (exposed to air for one hour at 500'C). The highly oxidized surface was the poorest ab-sorber but had sorbed ~ 10 mg/cd of iodine by the end of the one-year test.
The iodine concentration used in these tests was 2.8 g/d . A similar I')
test was performed with flat plates of copper at 285*C in vacuum with a vapor density of 0.83 g/d . The duration of the test was ~ 1/2 hr but the copper had aircady sorbed quantities of iodine varying from 0.3 - 1.2 mg/cd at its conclusion. In addition, no saturation effects were observed.
Page 1
A f
From the activities reported in Table 1 of the ATTACHMENT, one can calculate the mass of radioactive iodine nuclides in the inventory using the standard relationship.
N mass of iodine = 8.023 x 10ff * ^ '"U (E#"**}'
1r dN 1 where N = y [ g - J .
The mass of radioactive iodine at 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> is computed to be about 2.01 milligrams. Assuming that copper will sorb at least 10 mg iodine /cd , we find that less than 1 cm2 of surface area is needed to sorb all the iodine produced in the 40,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> test. Since the copper bus bars have dozens of cd of surface exposed in the primary containment, there is much more than sufficient copper to assure that all iodine is sorbed.
Therefore, it is reasonable to assume that essentially all of the iodine will be sorbed by copper components and that the only iodine avail-able for release will be that which remains in the vapor phase. The vapor 20'C.I ) The boiling point
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pressure of CuI is given as ~2.4 x 10 ' torr at of CuI is 1290*C. From these two points we can construct an approximate vapor pressure-temperature relationship of the form p=p 0*
if p is in torr and T in *K the constants are p0" * * * "
C = 9556*K Since the bus bars are water cooled they operate at ~ 30* - 50 C. If we assume a conservative bus bar temperature of 100*C we find that the vapor
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pressure is 3 x 10 8 torr.
Page 2
4
' The resulting vapor density is
$.*.}0 _ggff x -- "E"--- x --- 121.5 .___ . 2.31 x 10-12,j,,3 3
1 7600 torr 22,400 cm atm-Since the volume of the primary is less than 104 cm' , the maximum' mass of the available iodine in the vapor phase is 2.31 x 10 s g. Of the approxi-mately 2 milligram iodine inventory, the percent.ge available for release into the secondary ist b-f- - x 100 = 1.2 x 10 sx In- view of ' this minute percentage release, it is extremely - conservative (probably unnecessarily conservative) to assume ' that 10% of the iodine inventory in the primary containment is available for release into the secondary containment and hence into the reactor room.
REFERENCES
- 1. Morris, J.B., et al., " Removal of Iodine Vapour from Air by Metallic Copper," Chemical Engineering Division, Atomic Energy Research Estab-lishment (England) Report AERE-R4114, 1962.
- 2. Davis , R.J. , " Mechanisms of Sorption of Molecular Iodine," Oak Ridge National Laboratory Report ORNL-4126,-August 1967.
- 3. International Symposium on Fission Product Release and Transport Under Accident Conditions Proceedings, CONF-650407/ Volumes 1, 2, Oak Ridge, Tennessee, April 1965.
- 4. Shelton, R.A.J., Blairs, S. and Margrave, D., Nature 190, 1183 (1961).
Page 3
- _ _ _ _ _ _ _ _ _ _ . . _ _ __ _ b
APPENDIX II ACTIVATED CHARCOAL FILTER IN REACTOR ROOM As discussed in the ATTACHMENT, an activated charcoal filter will sig-nificantly reduce the thyroid dose at the site boundary attendant upon the postulated release of fission products from a thermionic device. In the following, a specific filter unit available from Barneby-Cheney (Cleve-land, Ohio) will be used as an example of a suitable filter. The through-put for each unit is a nominal 460 cfm with a transit time through the charcoal filter of 0.25 sec. The manufacturer claims trapping efficiency of 99.9% for this type of operation. The industrial quality motor and blower is capable of continuous duty even though its use is normally intermittent in the present application.. Figure 1 illustrates the approximate size of the filter unit and the arrangement of the activated charcoal in its holder.
Several additional features are required for satisfactory operation.
l A HEPA filter is included downstream of the charcoal filter to assure that
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no activated charcoal dust particles (fines) will be distributed through-I l out the reactor room when the filter system is in operation. The major
" poison" for activated charcoal is water from the humidity normally l ~
L present within the reactor room. To prevent deleterious action of this moisture, air tight seals are provided for the inlet and outlet of the ac-tivated charcoal system when it is not in operation. To verify that the charcoal remains unaffected by humidity, sample coupons of activated char-coal will be inserted within the sealed compartment. Periodically, a cou-pon will be removed; sealed tightly against ambient air; sent to the .
, 1 manufacturer for analysis and assurance that water content is not a prob-lem; and replaced with a new activated charcoal specimen. It is the plan to start operation with three (3) removable coupons. The present recom-mendation of the manufacturer is for semiannual coupon analysis for mois-ture content.
PAGE 1
1 ,
In the proposed ' operation, the activated charcoal filter system will remain sealed until needed in the event of an accidental release of fis-sion products , or until a' run is made to check the system operability.
Periodic weekly checks on operability of the motor will assure that the bearings in the motor and' air blower remain free and operable. On advice from the manufacturer, a five minute operation weekly, even in ~100% hu-midity condition, cannot significantly " poison" the activated charcoal with: water.- ,0nce every two or three years of such operation (approximately 5 minutes operation each week) it is recommended that the activation of the specimen coupon be checked in addition to the water content (which is to be tested approximately semiannually).
The filter unit will be situated within the reactor room. This will take air-from the reactor room and return the filtered air back into the same room. 'The flow rate-of 460 cfm from its blower will adequately cir-culate and distribute the processed air. In one pass of air-through the activated charcoal, ~ the efficiency for removal of iodine isotopes is 99.9%. With a filter flow rate' of 460 cfm and a room volume of about-14000 cf, the fractional rate of removal of iodine isotopes will be:
- x 60 = 1.97 hr~1 PAGE 2
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