ML20083C136
| ML20083C136 | |
| Person / Time | |
|---|---|
| Site: | University of Maryland |
| Issue date: | 12/19/1983 |
| From: | Belcher R MARYLAND, UNIV. OF, COLLEGE PARK, MD |
| To: | Thomas C Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20083C140 | List: |
| References | |
| NUDOCS 8312220138 | |
| Download: ML20083C136 (18) | |
Text
P University of Marytand Department of Chemical and Nuclear Engineering COLLEGE PARK, MARYLAND 20742 December 19, 1983 Cecil O.
Thomas, Chief Standardization and Special Projects Branch Division of Licensing U.S.
Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Thomas:
Attached are the answers to Questions 17, 18, and 35 of the 39 Questions concerning the MUTR Safety Analysis Report requested by your staff.
The answers to the other 36 questions were submitted on December 12, 1983.
Also attached are the plans for the grid plate and grid plate support which were inadvertently omitted from the answer to Question 5.
Sincerely, 4
Ral L.
Belcher Director MUTR RLB:bg Enclosures l
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PDR Nuclear Engineering Program Tel. 301-454 2430/6 j
I
- e 0-17.
Identify the assumptions used in estimating the average 41 annual release of Ar from the MUTR to be on the order of 100 mci.
What operational data do you have for verification?
What is your best estimate of annual release if you operated on a maximum schedule consistent with your license limits?
Ans-17.
The assumptions used to estimate thesannual release of Argon 41 from the MUTR include the following:
12 2
1.62x10 n/cm 7,,c average thermal flux
=
=
th 3
17,763 cm Volume of water in core
=
3 2.26x10 cm Volume of water in pool
=
6870 seconds Cycle time in pool
=
AT across core
= 18*C Average time of H O in core = 5.4 seconds 2
3 Flow rate in core
= 3289 cm /sec 15 3
Concentration of A-40 in pool = 8x10 aton/cm The equilibrium concentration in the pool was found to 4
be = 8.21x10 atoms of A-41/cm.
Following the procedure l
used in the University of Texas FSAR the source of A-41 7
into the reactor room was found to be 5.05x10 atoms /sec.
Room volume = 1.7x10' cm The exhaust fans are assumed off l
l The equilibrium activity is 8.02x10-pCi/cm 20.1 m
m y.
-w-e-.-,----.m
.--,s----------m- - -.., - -
This compares very favorably with the limit established
-6 3
in 10CFR20 which is equal to 2x10 pCi/cm,
The emission of 5.05x10 atoms /sec at 250 KW operation at equilibrium results in 12 atoms 1.82x10 2.5MwH The history of operation of the MUTR is such that the maximum recorded burnup in a one year period was 32 MwH.
Thus the total Argon-41 activity emitted into the room in that year was 67 millicuries.
If we operated a maximum schedule permitted by license (250KW, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> / day, 200 days) we would generate 400 MwH (greatly exceeding present practice) and 834 milli-curies.
References University of Texas FSAR 1982 J.R.
Lamarsh, Introduction to Nuclear Engineering, 2nd Ed., Addison Wesley, 1983, p. 522.
~
l 20.2
--. _--,,____l
O 0-18.
What calibration procedures are used for the radiation monitors in the reactor room?
What is the function and purpose of the monitor near the exhaust vent at the " stack"?
At what exposure rate and room concentration of airborne radioactivity does it initiate alarms and/or protection action?
How does this compare with 10 CFR 20?
Ans-18.
The function of the monitor near the exhaust vent is to measure radiation levels in the room.
It is redundant with the " bridge" monitor.
It is set to alarm at an exposure rate of 10 mrem /hr.
From calculations made to respond to Question 35, we may make the following estimate.
Assume that one fuel' element clad fails and that the rare gases all escape into the room.
Assume the pool to be full of water and the iodines are all dissolved.
Then 38 millicuries of mixed fission rare gases escape into the room.
Using a room volume of
-5 1.7x10 cm, this results in a concentration of 2.24x10 "I
From the results of calculations done for Question.
3 cm 35, a 10 minute immersion in the fission gas atmosphere would result in a dose rate of 3 mrem /10 min or approximately 18 mrem per hour.
Thus this " exhaust" monitor would alarm and perform its function in this scenario. 10 mren/ hour be this analysis would correspond to a concentration
-5 pC1 of mixed fission gases of about 1.2x10 cm3,,
20.3
O I
-6 2.5 mrem / hour would. correspond to a level of 3.1x10 pCi/cm.
By way of example 10CFR20 limits Xe-135 to
-6
-6 3
4x10 Ci/cm, and Kr-87 and 88 to lx10 pCi/cm.
This monitor is calibrated by placing a gannaa source of known strength a known distance from the detector.
The response of the monitor is then adjusted to correspond to the known field.
References:
See Question 35.
e 20.4
f Q-35.
Provide more details on the fuel element cladding failure accident described on pp. 11-7 to 11-10 of your SAR.
- a. What fraction of the total inventory was assumed present in the centrally located fuel element?
- b. Which specific isotopes were included in the source terms and what were their activities?
-4 3
c.
Show how the water activity value of 6.68 x 10 pCi/cm was obtained.
- d. Explain the method and provide the equations used for the calculation of atmospheric dilution and radiation exposure inside and outside the building.
- e. What meteorological conditions, release height, building leakage rate, and dose conversion factors were assumed?
- f. Where you refer to General Atomic research, give specific references.
-5
- g. Your indication of a release fraction of 1.5 x 10 at 600*C is not consistent with the latest reference work available to us.
See, for example: Simand, et.
al.,
" Fuel Elements for Pulsed TRIGA Research Reactors,"
Nucl. Tech., 28, 31-56, Jan. 1976.
- h. Please consider the loss of cladding integrity of one fuel rod in air.
(See Columbia University hearing on TRIGA Reactor).
Ans-35.
Please see next page.
9 45
Ans-35.
The analysis reported on pp. 11-7 to 11-10 of the MUTR SAR is based on calculations performed circa 1970-1971.
That analysis was for a scenario of the cladding failure of a single element.
The details of that analysis has been lost.
Therefore a new, more conservative analysis has been performed.
This scenario involves the cladding failure of a four-element bundle (a standard MUTR fuel bundle) in air.
The bundle is assumed to have twice the average burnup.
The burnup considered is for a case of the MUTR operating at 250 kilowatts (full licensed power) for ten hours per day and for a time period of 200 days.
This greatly exceeds the operating history of the MUTR.
The specific isotopes considered and there activities in the four element bundle are given in the accompanying Table 35-1.
They are derived from the estimate given for a TRIGA which has undergone one megawatt-year of operation which is reported in NUREG/CR-2387 (PNL-4028).
The release fraction used for this analysis is lx10 this is a conservative assumption as Sinnad reports a
-5 value of 1.5,x10 for fuel temperatures below 400*C.
With the assumption of a peak power to average power equal to two, the maximum burnup element would be operating at less than 400*C.
With the assumption that the elements are failed in air the scenario would be one of off-loading fuel so that the reactor is not considered to be operating 45.1 l
when failure occurs.
Thus the temperature would be considerably less than 400*C.
The activity of each isotope released is also presented in Table 35.1.
From Table 35.1 the total quantity of volatile fission products present in this maximum burnup four-element bundle is:
3 Rare gases 1.52x10 curies Iodines 2.15x10 curies Using a release fraction of lx10", the total release would be Rare gases 152 millicuries Iodines 215 millicuries If we consider a rupture of the bundle in the pool, with 7
3 a total pool volume of 2.26x10 cm and assuming only the iodines go into solution, the activity in the pool
-3 3
water would be 9.5x10 microcuries/cm.
Twenty-four hours later only the I-131 and I-l33 would contribute
-3 3
significantly and the activity would be 2.14x10 microcuries/cm,
In order to estimate the dose to individuals in the reactor building in a conservative manner, it is assumed that the four element bundle fails in air.
The released fraction of the activity all becomes airborne and disperses uniformly throughout the reactor building.
The building exhaust is assumed not to be operating and the free volume of the reactor building is taken.as 1.7x10' cm.
This results in the concentrations given as C in a le 35.1.
External 0
radiation doses and thyroid doses were calculated using these concentrations.
The assumption is made that the t
45.2
immersion. time of individuals in the reactor building is ten minutes.
The simplicity of the MUTR building makes this a reasonable assumption as the MUTR can easily be evacuated in that time.
The thyroid dose due to inhalation of iodine inside the building is determined from the following equation.
BC D ( 1-e-a) 35-1 D
=
t
-12 10 y
where B = Breathing rate for the standard man =
3
-4 m
3.47x10 sec CO = Air concentration (Ci/m )
D
= d se to thyroid (area) t t, = inumersion time (seconds)
A
= decay constant g = dose conversion factor.
D The dose conversion factors are taken from Table E-7 of Regulatory Guide 1.109.
The external dose to individuals inside the building is calculated using the following equation derived following immarsh.
l
( ~*
}
A, CO x
D, =
g 45.3
l 1
t where x
= Beta or gamma designations 0.229 A
=
g Ay = 0.262 I
average gn==a energy emitted (MeV)
E
=
averrage effective beta emitted (MeV)
E
=
g A = decay constant t,= immersion time (600 seconds)
D = dose in rema.
The doses derived from these calculations are given in Table 35.2.
The second case considered is for that of individuals exposed outside the building in the event of failure of the exhaust system to function properly..
This would r
result in an exhaust rate of the building contents through-3 two exhaust vents with a total capacity of 2.86 [h Approximately one-half of this exhaust is directed toward opposite sides of the building.
Thus the vent exhaust 3
on one side of the building is given as V=1.43,",.
Atmospheric dilution, X/Q, is calculated from the following equation (Lamarsh, p. 566)
~
X=
1 Q
CA where V = wind speed taken as one meter /second C = shape factor taken as 0.5 45.4
A = cross sectional area of the building estimated to be 270 m
-3 Thus X/Q = 4.12x10 The release height in the above equation is taken to be zero and the receptor point is in the lee of the building.
3 With an exhaust rate of 1.43 "sec, the concentrations outside of the building are given by (g)V CO X
=
o where 3
,"e c V = exhaust rate C C = concentration in the building (Ci/m3) 0
-3
= 5.89x10 C
or XO 0
The values of X are tabulated in Table 35.1.
O The 50 year committment thyroid dose to individuals outside the building was calculated using equation 35-1 (above) with X in P ace of C and an effective decay constant l
0 O
and a A which accounts for the diminution of the source c
term with time as we exhaust from the building.
(See Lamarsh, p. 578)
Exhaust rate c " A
- Building volume
-1 (A + 1.68.10- ) sec A
=
c The dose conversion factors were again taken from Reg.
Guide 1.109, Table E-7.
Infinite immersion was assumed.
These 50 year thyroid doses are given in Table 35-2.
45.6
=
The following.two equations were used for calculating the external dose to individuals outside the building, immersed in the F ume.
Again infinite isnersion was l
assumed.
4 3.16x10 gDg D
=
Y 3c 0.229 X E0 D
=
8 Ac where D
= ganuma dose in millirem g = beta dose in rem D
Dy = dose conversion factor from Reg. Guide 1.109, Table B-1.
1
= concentration (G M )
. ?
XO 4
3.16x10 = conversi h factor, see Reg. Guide 1.109 A
= 1.68x10-3' + A c
0.229 =, conversion facton (See T amarsh)
E = effective beta energy per decay (MeV)(Table 35-1).
g The results of these calculations are presented in Table 35-2.
~
e e
e 45.7
l
~
T:blo 35-1 C
X0 O
Activity Air Concen-Air Concen in four tration inside tration outside Element Building Building Bundle E
E Y
B Nuclide t
MeV MeV Ci Ci/m Ci/m g
-6
-9 Kr83m 1.9h
.0008 28.3 1.66x10 9.78x10
-6
-8 Kr85m 4.4h
.16
.223 65.5 3.85x10 2.27x10
-8
-10 Kr85 10.8y
.004
.213 1.1 6.27x10 3.69x10
-6
-8 Kr87 1.3h 1.07 1.05 125.9 7.4x10 4.36x10
-8 Kr88 2.8h 2.05
.341 179.9 1.06x10-6.25x10
-5
-8 Kr89 3.2 min 2.40 1.5 221.1 1.3x10 7.66x10
-I Xal33m 2.3d
.037
.155 8.9 5.21x10-3.07x10
-5
-7 Xn133 5.3d
.029
.146 517.8 3.05x10 1.80x10
-6
-8 Xel35m 0.3h
.46
.097 136.3 8.0x10 4.71x10
-8 Xa-135 9.lh
.25
.322 233.7 1.37x10" 8.07x10
-5
-8 I-131 8.1d
.40
.197 246.3 1.45x10 8.54x10 I-132 2.3h 1.96
.448 378.9 2.23x10-1.31x10" I-133 20.3h 0.56
.423 440.5 2.59x10-1.53x10-
-5
-7 I-134 0.9h 3.02
.455 580.
3.41x10 2.0lx10 I-135 6.7h 1.77
.308 504.5 2.97x10-1.75x10-l l
l
[
45.8 r
,.s 4
Table 35-2 50 yr.
External Location Thyroid Dose 8,y dose Inside Building 8.3 ren 52 millirem outside Building 47 millires 0.052 millirem I
l 45.9 l' -
References:
J.R.
Lamarsh, Introduction to Nuclear Engineering, 2nd Edition, Addison Wesley, 1983 M.T.
Simnad, et.al., "Puel Elements for Pulsed TRIGA Research Reactors, Nucl. Tech., 28, Jan. 1976.
Regulatory Guide 1.109, Calculations of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CRF50, Appendix I.
S.C.
Hawley, et.al., " Credible Accident Analyses for TRIGA and TRIGA-Pueled Reactors", NUREG/CR-2387, PNL-4028.
l 45.10
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