ML113530010
| ML113530010 | |
| Person / Time | |
|---|---|
| Site: | University of California - Irvine |
| Issue date: | 12/02/2011 |
| From: | Geoffrey Miller University of California - Irvine |
| To: | Meyer W Division of Policy and Rulemaking |
| Meyer W, NRR/DPR/PRLB, 301-415-0897 | |
| References | |
| Download: ML113530010 (13) | |
Text
UNIVERSITY OF CALIFORNIA-IRVINE RESEARCH REACTOR LICENSE NO. R-116 DOCKET NO. 50-326 TECHNICAL RAI RESPONSES (DATED 12/2/2011)
REDACTED VERSION*
SECURITY-RELATED INFORMATION REMOVED
- REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS
T UNIVERSITY OF CALIFORNIA, IRVINE BERKELEY
- DAVIS
- IRVINE LOS ANGELES - RIVERSIDE
° SAN DIEGO SAN FRANCISCO (JI!*
SANTA BARBARA SANTA CRUZ George E. Miller Senior Lecturer Emeritus Department of Chemistry and Director, Nuclear Reactor Facility Faculty A dvisor for Science, UCI Center for Education Partnerships US Nuclear Regulatory Commission Document Control Desk Washington DC 20555 Attention: Walter Meyer, Senior Project Manager IRVINE, CA 92697-2025 (949) 824-6649 or 824-6082 FAX: (949) 824-8571 email : gemiller@uci.edu December 2, 2011 Re: Docket 50-326 Relicense
Dear Mr Meyer:
Please find attached two items in completion of our response to requests for additional information.
I declare under penalty of perjury that the foregoing and the attached are true and correct to my knowledge.
Executed on December 2nd, 2011 Dr. George E. Miller--/KA ý-
", '-) o
13.2 Maximum Hypothetical Accident (MHA) - Rupture of Single Fuel Element in Air.
13.2.1 Summary The MHA fbr reactors with TRIGA hydride fuel has been adopted (NUREG-1537) to be the release of fission products from a single fuel element whose cladding has been stripped. This is analyzed both, under water and in air. The assumption for the MHA is made that maximum fission product inventory has been achieved in the most active fuel element, i.e., that the reactor has been operating for an infinite time period prior to the release. This is obviously a highly conservative assumption for the UCI reactor which has operated on a very intermittent schedule since it was commissioned. The actual likelihood for such a complete rupture event is very small, but is not further considered here. This section examines the consequences of the postulated MHA. The most significant activity released from the perspective of personal exposure risk is that of iodine isotopes since this delivers exposure to the thyroid gland. Direct radiation is also considered for risk to any occupants of adjacent areas not subject to risk from the volatile materials.
The released radioactivity is the product of the fission product inventory and the release fraction.
Consideration is then given to the exposure risk to personnel within the facility, those in surrounding spaces including laboratories or offices, and release beyond the building via the building ventilation system. Factors included are the facility volume, the ventilation purge rate and an estimated plating factor for iodine radio-nuclides on walls and facility structures which will reduce inhalation exposures.
The total number of fission product nuclei present from the fuel was estimated belowI and the release fraction was used from NUREG 1282. The modest levels predicted in the MHA come about as a result of the very low value of fission product release fraction established for TRIGA fuel elements that have stripped their cladding.
13.2.2 Fission Product Inventory The source term for the reactor inventory was calculated by General Atomic using SCALE 6.1 code assuming 1000 effective full power days (EFPD) and separated into halogen and noble gas groups. The first group comprises bromine and iodine isotopes that will dissolve if water is present, and be airborne if it is not. The second group comprises the insoluble volatiles: krypton and xenon isotopes. The second are the major source of radioactivity in the room (and outside) if the unclad element were to be under water. SCALE 6.1 uses the ORIGEN code for isotope generation and depletion calculations.
Details of the calculation methodologies and validation references are contained in the original GA report cited. The fuel composition used was based on the number densities calculated 2 earlier for this core. The calculation showed the reactor turning subcritical within about 100 EFPD's mostly due to build-up of 35 Xe, but this was ignored to determine the "saturation" fission product inventory, which occurred at about 125 EFPD. Exceptions are the very long-lived nuclides such as 85Kr and i21, and short-lived activation daughter products which are separately identified (by *) in Table 13-1.
GA 911209, "ORIGEN and Dose Calculations for the UCI TRIGA Reactor. ", Revision A., September 2011.
2 GA 911196, "Nuclear Analysis of the University of California TRIGA Reactor", Rev I July 2011.
Page 1 of 11 UCINRF Section 13.2, revision 3E November 10, 2011
TABLE 13-1 GASEOUS FISSION PRODUCTS IN SINGLE UCI TRIGA MAXIMUM POWER FUEL ELEMENT Group I Group 11 Nuclide Br 82 Br 82m*
Br 83 Br 84 Br 84m Br 85 Br 86 Br 87 1-128*
1 129*
!-130*
1-130m*
1131 1-132 1133 1134 I-I34m 1135 1136 I-I 36m 1-137 Kr 83m Kr 85*
Kr 85m Kr 87 Kr 88 Kr 89 Kr 90 Kr91 Xe 131m Xe 133 Xe 133m Xe 135 Xe 135m Xe 137 Xe 138 Xe 139 Xe 140 Half-life 35.28 h 6.13 m 2.4 h 31.8 m
- 6. m 2.9 m 55.1 s 55.65 s 24.99 m 1.57 x 107 y 12.36 h 8.84 m 8.02 d 2.295 h 20.8 h 52.5 m 3.52 m 6.57 h 83.4 s 46.9 s 24.13 s Total lodines Total Group I 1.83 h 10.776 y 4.48 h 76.3 m 2.84 h 3.15 m 32.32 s 8.57 s 11.93 d 5.25 d 2.19d 9.14 h 15.29 m 3.818 m 14.08 m 39.68 s 13.6 s Total Group 11 Inventory (curies) 9 2
2 Page 2 of 11 UCINRF Section 13.2, revision 3E November 10, 2011
13.2.3 Fission Product Release Fraction NUREG-1282 quotes a release fraction of I x 105 that has been measured for 8.5% uranium TRIGA fuel. Some indications in experiments are that the release slowly increases with fuel temperature.
However, tacilities have adopted a release fraction = 1.5 x 10-5 as an appropriate value for hypothetical use for fuel temperatures below 4000C.
Multiplying this release fraction by the inventory of gaseous fission products produced in the maximum power fuel element, as given in Table 13-1, gives the total activity that would be released should the integrity of a fuel element cladding be compromised. These values are shown in Table 13-2 Table 13-2 Releases from Maximum Power Fuel Element Unclad in Air Total Total Released Curies (Curies)
Noble Gases Iodine Halogens 13.2.4 Pool Water Activity if Rupture is Under Water.
We review briefly the course of events if the rupture did occur in water, in which case the soluble (Group 1) halogen products will remain in the water, totaling or
(
Since the volume of water in the reactor pool is 8.7 x i0 7 cm 3, the activity concentration is
ý
(
) In 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the activity would decrease by more than a factor of 10 to as all but Br-82, 1-131 and 1-133 nuclides have half-lives shorter than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, and the activity contribution of 1-129 is negligible. The demineralizer can be used to safely concentrate these for subsequent disposal. 3 millicuries of activity on the resin is a readily handled quantity by trained personnel. Thus the personnel exposure consequences of such an incident are relatively minor.
13.2.5 Predicted Exposures to Personnel Inside the Reactor Room if Rupture in Air.
The cited General Atomic report' computed the external dose rates and thyroid TEDE rates to personnel in the reactor room as a result of the described MHA: fuel element unclad in air, using a fission product release fraction of 1.5 x 105. The concentrations in Tables 13-3 and 13-4 are based on the fission product release uniformly dispersed within the 6.6 x 108 cm 3 volume of the reactor room.
External dose conversion factors for submersion in air were obtained from Federal Guidance Report No. 12-3 except for Br-85, 1-136, Kr-89, Kr-90, and Xe-137. Scaling factors for these radionuclides were derived from a DOE tabulation of external dose conversion factorsa.The Group I external dose rate to an individual in the reactor room is 31 mrem/hr and the Group II external dose rate is almost 19 mrem/hr for a total external dose rate of almost 51 mrem/hr. Some of the radionuclides from the ORIGEN calculation do not have external dose conversion factors. These radionuclides either have very weak gamma emissions or very short half-lives so their contribution to the total dose is negligible.
3 "Federal Guidance Report No. 12: External Exposure to Radionuclides in Air, Water, and Soil," US Environmental Protection Agency, EPA-402-R-93-08 1, September 1993.
' "External Dose-Rate Conversion Factors for Calculation of Dose to the Public," US Dept. of Energy, DOE/EH-0070, July 1988.
Page 3 of 11 UCINRF Section 13.2, revision 3E November 10, 2011
Table 13-3 Group I Fission Product Release and External Dose DCF External Dose Nuclide Inventory (Ci)
Release (Ci)
Concentration (Ci/m
- 3)
(Sv per Bq-s/m 3)
(mrem/hr)
Br-82 1.30E-13 6.97E-04 Br-82m Br-83 3.82E-16 2.39E-03 Br-84 9.41E-14 1.04E+00 Br-84m Br-85 3.56E-15 5.70E-02 Br-86 Br-87 1-128 4.16E-15 4.77E-05 1-129 3.80E-16 2.92E-10 1-130 1.04E-13 5.23E-03 1-130m 1-131 1.82E-14 6.06E-01 1-132 1.12E-13 5.66E+00 1-133 2.94E-14 2.27E+00 1-134 1.30E-13 1.18E+01 1-134m 1-135 7.98E-14 5.80E+00 1-136 1.35E-13 4.17E+00 1-136m 1-137 Total Group I 31.4 Table 13-4 Group II Fission Product Release and External Dose Inventory Concentration DCF (Sv per External Dose Nuclide (Ci)
Release (Ci)
(Ci/m
- 3)
Bq-s/m 3 )
(mrem/hr)
Kr-83m 1.50E-18 9.26E-06 Kr-85 1.19E-16 7.89E-05 Kr-85m 7.48E-15 1.15E-01 Kr-87 4.12E-14 1.23E+00 Kr-88 1.02E-13 4.11E+00 Kr-89 9.21E-14 4.77E+00 Kr-90 6.07E-14 3.38E+00 Kr-91 Xe-131m 3.89E-16 1.52E-04 Xe-133 1.56E-15 1.21E-01 Xe-133m 1.37E-15 1.14E-03 Xe-135 1.19E-14 6.48E-01 Xe-135m 2.04E-14 1.90E-01 Xe-137 9.06E-15 6.42E-01 Xe-138 5.77E-14 4.19E+00 Xe-139 Xe-140 Page 4 of 11 UCINRF Section 13.2, revision 3E November 10, 2011
1Total Group 1 1
I I
1 29.4 The thyroid committed dose rate and whole body committed effective dose rate from inhalation to an individual within the reactor room are presented in Table 13-5 based on the inventory and releases presented in Table 13-4. The concentration in the room is multiplied by the breathing rate of 3.47 x 10-4 m3/s to give the inhaled amount of iodine. Committed effective dose equivalent (CEDE) and thyroid dose conversion factors (DCFs) for inhalation were obtained from Federal Guidance Report No.
- 15. Some of the radionuclides from the ORIGEN calculation did not have inhalation dose conversion factors. These radionuclides have very short half-lifes such that their contribution to the total dose is negligible. The total thyroid dose rate is 1.37 mrem/sec or 4.9 rem/hr. The CEDE whole body dose rate is 0.046 mrem/sec or 0.17 rem/hr. Adding the external dose rate gives a Total Effective Dose Equivalent (TEDE) of 5.1 rem/hr.
Table 13-5 Thyroid and CEDE Dose Rates from Reactor Room Inhalation Reactor Room Thryoid Concentration DCF Thyroid Dose CEDE DCF CEDE Dose Nuclide (Ci/m
- 3)
(Sv per Bq)
(mrem/sec)
(Sv per Bq)
(mrem/sec)
Br-82 2.38E-10 1.23E-07 4.13E-10 2.13E-07 Br-82m Br-83 3.29E-12 1.99E-06 2.41E-11 1.46E-05 Br-84 3.12E-12 3.33E-06 2.61E-11 2.79E-05 Br-84m Br-85 Br-86 Br-87 1-128 5.34E-11 5.91E-08 1.28E-11 1.42E-08 1-129 1.56E-06 1.16E-07 4.69E-08 3.48E-09 1-130 1.99E-08 9.64E-05 7.14E-10 3.46E-06 1-130m 1-131 2.92E-07 9.37E-01 8.89E-09 2.85E-02 1-132 1.74E-09 8.48E-03 1.03E-10 5.02E-04 1-133 4.86E-08 3.62E-01 1.58E-09 1.18E-02 1-134 2.88E-10 2.51E-03 3.55E-10 3.10E-03 1-134m 1-135 8.46E-09 5.92E-02 3.32E-10 2.33E-03 1-136 1-136m 1-137 Total 1.37 0.046 On detection of release of radioactive fission products into the reactor room, the continuous air monitor will close the normal exhaust and start the emergency purge. The room air will then be exhausted
' "Federal Guidance Report No. 11: Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors lbr Inhalation, Submersion, and Ingestion," US Environmental Protection Agency, EPA-520/l-88-020, September 1988.
Page 5 of 11 UCINRF Section 13.2, revision 3E November 10, 2011
through a filter at 240 cfm (0.113 m3/sec) which will not remove the noble gases and may not remove all the iodine isotopes released. However, credit may also be taken for plate-out on surfaces of the iodine isotopes. In Oregon State University's SAR, they adopted a method which suggests halogen release plates out to reduce the airborne concentrations to less than 25% of the initial release.
With this assumption, the revised dose rate expected to the thyroid in the unlikely event of a fuel element cladding rupture in air, is 0.25 x 5.1 rem/hr = 1.3 rem/hr. If the water in the pooi is assumed to be present there is a high likelihood that this will rapidly absorb the majority of halogens from the air and reduce the source for thyroid exposure to an individual in the facility to even less than 25% of that released.
Given a 5 minute evacuation time for personnel from the facility, the revised maximum TEDE estimate is 0.1 rem to the thyroid. Without plating credit, the value is 0.4 rem.
Given that 50 rem to the thyroid is the limiting annual exposure6 a person will have a longer time than 5 minutes to evacuate. This is a very conservative estimate since in addition to the comments made above that would reduce exposure, 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> operation at full power prior to the accident was assumed as was instant cladding rupture. Nor was credit taken for operation of the room emergency
,purge exhaust which will also act to reduce the dose received.
13.2.6 Predicted Exposure to Persons Outside the Building The ventilation system discharges the emergency purge exhaust into the main building exhaust which has three exhaust fans designed to create a flow of 3 x 80,000 cfm, further diluted to 3 x 128,000 cfm at a height of 120 feet above the building roof by means of Axijet structures. From that point, natural dispersion, aided by the plume jet, will further dilute the effluents.
From the above, it is reasonable to assume a conservative dilution of the exhaust effluent by a factor of 384,000/240 = 1600. Using this, and assuming no climatic or other dilution outside the building exhaust flow, the maximum dose to an individual may be calculated.
The calculation assumes:
(1) Complete mixing in the reactor room at all times (2) The person is immersed in the effluent from the building exhaust flow in which the concentration of radioactivity Xt at any time t is equal to Ct/1600 where Ct is the concentration in the exhaust and in the room and Co is the initial concentration in the room.
(3) Inhaled activity At =X, multiplied by the breathing rate, R, of 3.47 x I0-4 m3/s.
(4) Activity is removed from the room by leakage (AL = 0.113/660 see -1) and radioactive decay, AD-The integrated concentration of radionuclides that an individual is exposed to up to time t is
=
C° 1 [1 - exp(-(A, + AD) t) 1600(A/ + AD)
Similarly, the inhaled activity up to time t is At RC.
[1 - exp(-(A,. + Al) t)]
1600(A/. + AD) 10CFR 20.1201 (l)(i)
Page 6 of 11 UCINRF Section 13.2, revision 3E November 10, 2011
For infinite exposure, the quantity in the square brackets equals one. The Group I and Group 1I external doses to an individual exposed to the rooftop exhaust for infinite exposure are presented in Tables 13-6 and 13-7. The thyroid and CEDE whole body inhalation doses for infinite exposure are presented in Table 13-8. The TEDE dose is the sum of the external and CEDE dose which equals 0.18 torem. No credit is taken for plate-out, filtering or downwind dispersion.
Atmospheric dispersion is commonly expressed in terms of the X/Q ratio which is the ratio of concentration (Ci/m 3) over source rate (Ci/sec). At the rooftop exhaust, the concentration is 1600 The source rate is Q
0.113 [mc xC° The effective dispersion at the rooftop exhaust is 5.531 x 10-3 sec/m 3. Atmospheric dispersion is calculated using the following equation X
1_I Q
- raau where u is the wind speed in m/sec, and o-y and oa are the standard deviations of the Gaussian plume in the y and z directions. The Gaussian plume dispersion parameters as a function of downwind distance x in meters are calculated using the following relationships from the MACCS computer code description in NUREG/CR-4691, Vol. 27.
=
axb d
-z c X Using conservative weather conditions of Stability Class F, the Gaussian plume parameters are:
a=0.0722, b=0.903 1, c=0.2, and d=0.6020. Assuming a conservative wind speed of 1 m/sec (2.2 mph),
a virtual source distance of 247 m is calculated using the methodology of the MACCS computer code.
The associated atmospheric dispersion parameters are o'y=0.5 m and or=5.5 m. This virtual source distance is the distance from a point source which reproduces the effective atmospheric dispersion at the rooftop exhaust. The nearest residence to Rowland Hall is Campus Village Housing which is at least 200 m to the west. The atmospheric dispersion parameters at a distance of 447 m are ory=l 7.9 m and u= 7.9 m which gives an atmospheric dispension of 2.261 x 10-3 sec/m 3. This atmospheric dispersion results in a further dilution of the rooftop exhaust by a factor of 2.45. The resulting TEDE dose for infinite time exposure is 0.073 mrem.
However there is a very small possibility that at the time of the MHA, all the exhaust fans are not operational. If it is assumed that only one of the three fans is functioning, then (since all ducting is in parallel) the concentration in the air released will be three times that calculated above, resulting in doses three times higher.
Thus under the extremely hypothetical conditions, the calculated maximum MHA doses are as follows:
TEDE for infinite time/cloud at point of release: 0.18 mrem x 3 = 0.54 mrem 7 "MELCOR Accident Consequence Code System (MACCS) - Model Description," U.S., Nuclear Regulatory Commission, NUREG/CR-4691, Vol. 2, February 1990.
Page 7 of 1I UCINRF Section 13.2, revision 3E November 10, 2011
1 TEDE for infinite time/cloud at virtual distance of 447 m: 0.073 x 3 = 0.22 mrem Page 8 of 11 UCINRF Section 13.2, revision 3E November 10, 2011
Table 13-6 Group I External Dose From Rooftop Exhaust Integrated DCF Concentration (Sv per External Dose Nuclide (Ci-s/m 3)
Bq-s/m 3)
(mrem)
Br-82 1.30E-13 6.85E-07 Br-82m Br-83 3.82E-16 1.65E-06 Br-84 9.41E-14 3.39E-04 Br-84m Br-85 3.56E-15 2.38E-06 Br-86 Br-87 1-128 4.16E-15 1.31E-08 1-129 3.80E-16 2.96E-13 1-130 1.04E-13 4.86E-06 1-130m 1-131 1.82E-14 6.11E-04 1-132 1.12E-13 3.85E-03 1-133 2.94E-14 2.18E-03 1-134 1.30E-13 5.22E-03 1-134m 1-135 7.98E-14 5.02E-03 1-136 1.35E-13 8.53E-05 1-136m 1-137 Total Group I 0.017 Table 13-7 Group II External Dose From Rooftop Exhaust Integrated DCF Concentration (Sv per External Dose Nuclide (Ci-s/m 3)
Bq-s/m 3)
(mrem)
Kr-83m 1.50E-18 5.82E-09 Kr-85 1.19E-16 8.OOE-08 Kr-85m 7.48E-15 9.36E-05 Kr-87 4.12E-14 6.62E-04 Kr-88 1.02E-13 2.98E-03 Kr-89 9.21E-14 2.16E-04 Kr-90 6.07E-14 2.72E-05 Kr-91 Xe-131m 3.89E-16 1.53E-07 Xe-133 1.56E-15 1.22E-04 Xe-133m 1.37E-15 1.13E-06 Xe-135 1.19E-14 5.85E-04 Xe-135m 2.04E-14 3.55E-05 Xe-137 9.06E-15 3.49E-05 Xe-138 5.77E-14 7.34E-04 Xe-139 Xe-140 Total Group Ii 1
1 1
0.005 Page 9 of I 1 UCINRF Section 13.2, revision 3E November 10, 2011
Table 13-8 Thyroid and CEDE Inhalation Doses from Rooftop Exhaust Total Activity Thyroid DCF Thyroid CEDE DCF CEDE Dose Nuclide Inhaled (Ci)
(Sv per Bq)
Dose (mrem)
(Sv per Bq)
(mrem)
Br-82 2.38E-10 4.35E-07 4.13E-10 7.55E-07 Br-82m Br-83 3.29E-12 4.94E-06 2.41E-11 3.62E-05 Br-84 3.12E-12 3.90E-06 2.61E-11 3.26E-05 Br-84m Br-85 Br-86 Br-87 1-128 5.34E-11 5.83E-08 1.28E-11 1.40E-08 1-129 1.56E-06 4.22E-07 4.69E-08 1.27E-08 1-130 1.99E-08 3.23E-04 7.14E-10 1.16E-05 1-130m 1-131 2.92E-07 3.40E+00 8.89E-09 1.04E-01 1-132 1.74E-09 2.08E-02 1.03E-10 1.23E-03 1-133 4.86E-08 1.25E+00 1.58E-09 4.07E-02 1-134 2.88E-10 4.01E-03 3.55E-10 4.95E-03 1-134m 1-135 8.46E-09 1.85E-01 3.32E-10 7.25E-03 1-136 1-136m 1-137 Total 4.86 0.158 The high variability of local climatic conditions actually prevalent when an accident occurs make it relatively ineffective to try to predict accident exposures with any greater degree of certainty. It seems very likely that the present predictions will be conservative, and that accidents of this hypothetically very serious nature at this facility will not result in unacceptably hazardous exposures either to personnel in the facility, nor to the general public. Since the probability of such an event could increase during fuel handling operations, in order to further guard against the likelihood of public exposures, operational regulations at the reactor preclude the presence of any inessential or untrained personnel within the facility whenever operations such as fuel handling are in progress. Events involving handling of irradiated fuel elements in air are subject to even greater restrictions and personnel training, and include procedures to exclude the public from all areas adjacent to the facility.
This was the practice followed during the transfer of one cask load of irradiated fuel elements into the facility in 1974. The protocol for that procedure is on file for further use and inspection.
13.2.7 Predicted Exposure to Rowland Hall Occupants Adjacent to the Facility The calculations presented in section 13.2.5 can be applied to assessment of potential exposure in an MHA to individuals located in adjacent areas outside the reactor facility.
There is no pathway for airborne activities to reach the adjacent areas of the building, so that the deep dose rates established in that section for individuals within the facility may be utilized. Such Page 10 of 11 UCINRF Section 13.2, revision 3E November 10, 2011
individuals will not be immersed in an infinite cloud of radioactive volatiles, so the facility may be considered as a single source. Further only gamma-ray exposures would be experienced since the walls and windows are thick enough to absorb all beta radiation. There are no doors directly opening from the facility to adjacent building areas, so no leakage to interior building areas is anticipated.
Section 13.2.5 established a maximum dose rate to an individual within the facility of about 50 mrem/hr. Since this is an immersion calculation, it is reasonable to assume a distance factor before estimating the dose rate through a wall. The closest a person can stand outside the facility is in the adjacent hallway at the "window" wall. Assuming a whole body distance on each side of the wall, and including the wall thickness, it is reasonable to credit a distance factor of 5 feet. Making a further assumption that the source is an effective point source inside the facility, this could provide a distance factor of 52 or 25. Thus the external dose rate estimate is reduced by 50/25 or to 2 mrem/hr. making no allowances for attenuation by air or wall materials, This is at the acceptable rate8 for public individuals over short time periods. In such an accident, the building will be cleared within less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the accident so that extensive exposure at this rate (which would need to exceed 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to attain the 50 mrem annual limit 9) would not be realized. Since there is no internal exposure, this exposure is the TEDE estimate. As it is so low, eveh if there are large uncertainties in the estimate, this appears to be entirely acceptable. Anyone at a further distance, such as in nearby laboratories or offices, would receive less as a result of additional distance factors and absorption reductions. It is important to emphasize that the building ventilation design is such that no mechanism exists for mixing of air between the reactor area and adjacent rooms, so no internal exposure can result from the MHA except to persons in the facility..
S10CFR20.1302(2)(ii) 9I 0CFR20.13 02(2)(i i)
Page It of 11 UCINRF Section 13.2, revision 3E November 10, 2011