ML11269A030
| ML11269A030 | |
| Person / Time | |
|---|---|
| Site: | Armed Forces Radiobiology Research Institute |
| Issue date: | 09/06/2011 |
| From: | Melanson M US Dept of Defense, Armed Forces Radiobiology Research Institute |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME1587 | |
| Download: ML11269A030 (9) | |
Text
ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE (AFRRI)
RESEARCH REACTOR LICENSE NO. R-84 DOCKET NO. 50-170 RESPONSE TO RAI#12 DATED SEPTEMBER 6,2011 REDACTED VERSION*
SECURITY-RELATED INFORMATION REMOVED
- REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS
la6'cF/e 5 a- 1.
ARMED FORCES RADIOBIOLOGY RESEARCH INS'ItUTE.
8901 W84NflNIN AVIENU BsrTHOoA, MAIPLANP 206809-O3
. . . . . . .. Se.3. ,
. ::. * ,;,,;.,,.. : ,, September6, 2011 SI...
Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE APPLICATION FOR LICENSE RENEWAL (TAC NO. ME1587)
Sir:
By letter dated July 19, 2010, the Nuclear Regulatory Commission requested additional information necessary to allowprocessing of our research reactor license renewal applicadon(LiAcense R-.84, Docket 50-170).
In subsequent conversations with Mr. Walter Myer, we. were grapted aip extension until.
September 9,2011 to provide an, answer for question 12. Our r'espnse to that questioih is enclosed. Remaining to be answered are questions 3, 5, and 6bfor which we have requested an additionalextension. - " '. "
If you need further information, please. contact lr! Steve Miller '*" at 301-295-92453 or " ** " . .1 -
" " "- "1.".'*** 1 3 * "k;.,.
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I declare under penalty of perjury that the foregoing and all enclosed information is true and correct to the best of my knowledge. Execqd on September 6, 2011.
Enclo sure: 'eM kI.A A MELANSON I ~ ~)I s j~l C),:.' S, USA, 7..'
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- 12. NUREG-1537,Part1, Section 13.1.6 provides guidance-forthe licensee,to discuss events.that could resultfrom experiment malfunction. The licensee is requestedto justify its assumption that the releaseof irradiationArgon accident scenariois the worst conceivable casefor radiologicalconsequencesfrom an experiment. -The licensee should present a range of experimentalmalfunction accidents considered. The Argon activation assumptionsand calculationsshould be presented In more detall All-expediments performed'as part-of tfieTRIGA'reactor oper-tioU airedir-eviewed by thie Reactor and Radiation Facilities Safety Subcommittee and supervised by trained, licensed, supervisory personnel. The Technical Specifications contain requirements that must be met before performing:experiments using the AFRRI-TRIGA reactor. Although improbable, an experiment could fail and therefore the consequences of a variety of experimental acddents.have been considered.
The most common experiments performed at AFRRI involve the irradiation of biological samples. The radiological consequences from the failure of an experiment of this type are very minimal, as failure would not pose any risk to the reactor structure itself or result in a release of significant quantities of radioactive material to the staff or public.
The consequences of experiment malfunction of non-biological samples are described below:
According to the Technical Specifications, the irradiation of explosive materials in quantities greater than 25 mg is prohibited. Smaller quantities may be Irradiated assuming they are housed in a container capable of withstanding a pressure burst greater than twice the pressure resulting from detonation of the sample. The calculations demonstrating the ability of the container to withstand the pressure burst are to be reviewed by the Reactor and Radiation Facilities Safety Subcommittee and approved by the Reactor Facility Director. Failure of an explosive experiment therefore does not classify as a worst conceivable event.
Samples containing corrosive material must be doubly encapsulated according to the Technical Specifications and therefore the consequences of a failure are limited. Failure of a corrosive experiment in the reactor pool would be diluted by the primary coolant and, while resulting.in the need for a cleanup and Inspection of reactor fuel and instrumentation, would not present a worst case radiological event. ....
It is also notable that the sum of all experiments will not-exceed $3.00. The AFRRI-TRIGA reactor is capable of pulsing up to $3.00'and has proven that changes in reactivity of this magnitude are not damaging to the reactor. Therefore, simultaneous failure of all experiments would not result in reactor conditions that are less conservative than a standard pulse operation.
Another experiment Identified in the SAR as the irradiation of a 20 liter container of argon gas in ERI for one hour at 1 MW. This results in a total argon-41 Inventory of 5.6 a. Release of this quantity from the reactor stack yields an effective dose to the closest member of the public of 0.2 mrem and is not considered the worst case radiological evint resulting from an experiment malfunction.
The irradiation-of fueled experiments at AFRRI are limited so-that the total inventory of Iodine Isotopes 131 through 135 is not greater than' 1.3 C0 and the maximum strontium-90 inventory is. not greater than 5 mbi. Given the fission prodiict yields reported in NUREG/CR-2387,the 1.3 Ci limit of radioiodines. 131-through -1351is reached before 5 mrCof -
strontium-90 In fueled experiments. The release of these quantities-directly fromthe:AFRRI stack has been-analyzed and determined to represent-the worst case scenario of a release of radioactive material from.,experimental failure..; nthis.
accident scenario, aifueled experiment is irradiated untl-1.98 total Ci of radioiodines 131-135 are-present.. This total activity, is more-than.150% of the limit-specified in the Technical Specifications, and thus provides, a conservative,;
estimate for the radiological consequences-of this accident scenario. - . . .
For this experiment, It is assumed that 1 g of 19.75%enriched LEU.is irradiated:in-the AFRRI-core for 72.5 minutes at 1 MW. The assumed thermal neutron flux at this power level and sample location is Ux10. n/cm 2/s. The source term for this experiment was generated using ORIGEN, with radioisotope activities ofinterest-shown in Table 1.
TABLE 1. ORIGEN source term for fueled experiment.
isotope .1Half Life
"{ . .. .
Ac"tvi of Total r
I, AdhAtv Reiea~ed
--... t~ the P
.Br-82 .1.47d, I . ' 1 to *.c '..a)*
Br-83, 2.4h .
Br-84m. 6;0 m Br-84 31.8 m. .' , _ _ _ ._. .I Br-85 2.87 m Br-86 -;55.5. s _ _,'
Br-87 55.9s .
1-131 8-02 d _
1-132 2.28 h _
- ;" '....,., .I.'
1-133 20.8 h 1-134 52.6 m 1-135 6.57 h 1-136 1.39 m Kr-83m 1.86 h _____'__
Kr-85m '448h Kr-85 10.76 1 : '..
Kr-87 1.27 h Kr-88 2.4 h Kr-89 3.15 m .. . ...
Xe-131m 11.9dd. ____,
Xe-133m 2.19 d , .1.
Xe-133 5.24 d .... _ _
Xe-135m 15.3 m _
Xe-135 9.1 h Xe-137 3.82 m Xe-138 14.1 m * ., .k ~ .
It is very conservatively assumed that 25% of the halogens released from the sample into the fueled experiment sample holder are eventually available for inhalation by a radiation worker in the reactor room or a member of the public in the unrestricted environment. This value is based on historical usage and recommendations (Ref. 1-9), where Ref. 1 recommends a 50% release fraction for the halogens from the gap of a fuel element to the air. For the purpose of evaluating the consequences of a failed fueled experiment, the release fraction from the gap of a fuel element is assumed to be equal to the release fraction from the sample holder. Ref. 2 and Ref. 3 apply a natural reduction factor of 50% due to plateout in the reactor building., The 25% total halogens released results.from combining the 50% release from the sample holder with the.50% plateout.. However, this.25%yva!ue appears to-be quite conservative, as Ref. 6 and Ref. 7 quote a 1.7% release fraction from the gap of a fuel element rather than a 50% release fraction from the gap. The experienceat TMI-2;,along with recnt'experiments, also Indicates that the 50% halogen release fraction from the gap is much too large and reports that possibly as little as 0.06% of the Iodine reaching the dadding gap may be released into
the reactor room, due in part to a large amount of the elemental iodine reacting with cesium to form Csl, a compound much less volatile and more water soluble than elemental iodine (Ref. 7). It is reasonable to assume that this same reaction occurs in the fueled experiment sample holder, therefore reducing the amount of radioiodine released from the fueled experiment sample holder to the reactor room. It is assumed that 100% of the noble gases are available for release to the unrestricted environment.
This accident analysis assumes that 100% of the fission products present in the sample are released to the sample holder, with no restrictions on release from the sample. matrix.itself (unlike the MHA where an additional fission product reduction is attributed to the design of the TRIGA fuel matrix and its ability to restrict the release of fission products).
Because of this conservative assumption, the physical state of the fueled experiment does not need to be specified. For example, it is understood that a larger fraction of fission products release from a liquid sample than a solid sample. By assuming 100% release, the accident analysis provides a true worst case scenario regardless of t'h'ephysical characteristics of the fueled experiment sample. In reality, both liquid and solid fueled experiments would restrict the release of fission fragments to the sample holder to some extent, thus reducing the doses to radiation workers and members of the public.
The minimum distance to the unrestricted environment, as well as the minimum distance to the nearest occupied building, are assumed to be in the same direction as the prevailing wind. These assumptions will result in the highest possible radiation doses to members of the public.
For any atmospheric stability (Pasquill) class, a pround-level release always leads to a higher effluent concentration at any given distance than an elevated release. Accordingly, It is assumed for this accident analysis that only pround level effluent releases occur, and no credit is taken for either release heights or building wake effects. Furthermore, atmospheric modeling indicates that the more stable the atmospheric class and the lower the wind speed, the higher the effluent concentration. Therefore,,this analysis assumed both the most stable atmospheric class (Pasquill F)and a low wind speed (1 m/s) were present. The time that a receptor is exposed to the plume is determined by calculating the time required to exhaust the reactor room at the standard ventilation exhaust rate. For this analysis, the time is 9.1 min.
The methodology for atmospheric diffusion models presented in NRC Regulatory Guide 1.145 was used (Ref. 10) in the accident analysis. For distances greater than 100 m, the values for horizontal and vertical dispersion coefficients were also taken from Regulatorý Gulde 1.145. For distances from 10 m to 100 m, not addressed in Regulatory Guide 1.145, data from the OSTR SAR was used (Ref. 11). The values for the dispersion coefficients and x/Q are given in Table 2.
TABLE 2. Atmospheric Dispersion Coefficients and x/Q Values for Pasquill F and Mean Wind Speed of 1 m/s.
Distance(m) ay(m) o z(m),: x/Q(s/m)).
10 1.29 1.04 5.93E-02 50 2.45 1.2 2.71E-02 100 3.9 2.2 9.27E-03 150 6.18 3122. 4.00E;03. ' ..
200 8.21 4.13 2.35E-03 250 10.21 4.98 1.57E-03 Furthermore, it was assumed that all-of the fission products were released to the iunrestricted area by a single reactor room air change; which would maximize"the dose rate to persons exposed to the*l1urie during the accident.
Additional parameters used in this accident *were:
0 Reactor room ventilation exhaust rate: 1.68 m3/s
- Reactor room volume: 917 m3
" Receptor breathing rate: 3.3x10 4 m3/s (NRC "light Work" rate).-
" Dose conversion factors:
Internal based on DOE/EH-0071 (Ref. 12)
External based on DOE/EH-0070 (Ref. 13)
The committed dose equivalent (CDE) to the'thyroid and the committed effective dose equivalent (CEDE) for members of the general public at a given distance downwind from the facility for all isotopes of concern are calculated by:
- . , * ". . L. ",
." ;; . .I.. . , . ,1 (x/I)D - atmospheric dispersion factor at a given distahce D (s/ma)
BR = breat hing rate (m 3/s)
DCF 1* = internal dose conversion factor for isotope i (mremfi/pC) *.*. "
Aq = initial activity of.isotope(ý I ( .... *.
R, = ventilation of air from the reactor room (m3A /s)" *' .:".. .. " * ' .:
............................................. . rf *-.,.
Vira rromv~olume V reacto'rroomvoume (M) zm* .. t ... ... I.;.
- ~... ,' *.,: .~! "
A, = ventilation constant = R,/V (s-1 )
= decay constant for isotope i (s")
t= time when the plume first arrives at the receptor point (s) t= time when plume has passed the receptor point (s)
The deep dose equivalent (DDE) to members of the general public at a given distance downwind from the facility for both the thyroid and whole body are each calculated by:
... 'D.* DCF.,j A.(e-Atf - e-A)J DCF, 1 exterhal dose rate conversion factor for isotope i (mirem m3*/Ci s)
For calculating the dose to occupational workers in the reactor room, a stay time of 5 minutes was used. Experience indicates that the reactor room can easily be evacuated in less than 2 minutes however; the value of 5 minutes is used to account for any time the worker may be delayed performing a task. The CDE and CEDE for personnel in the reactor room for a given stay-time may each be calculated by:
[DCFigt, A, BR (1 - e-BR(i e1ts) J (CDE or CEDE)r =
-. D3 E or 1 tsT - stay timeof personnel The DDE to personnelin the reactor room for a given stay time for both the-thyroid and the whol bbidy are calculated by:
.t , .
- ,(~DE~Ihy,.Qg OrDDEwB)sT_ ZiCVA(~.....~iS)
The results of these calculations are shown in Tables 3-5. In all cases, doses for the general public and occupational workers were below the annual dose limits specified by 10 CFR 20.
There were two different scenarios analyzed in this accident scenario. In scenario #1, the isolation dampers fail following the release of radioactive material into the reactor room. As a result, the radioactive material is vented from the AFRRI stack to the unrestricted public. In scenario #2, the Isolation dampers operate as designed and limit the radioactive material release from the reactor room. This latter scenario results In a higher exposure to the reactor staff member in the reactor room. As the radioactive materials disperse In the reactor room, the room becomes a source term for external exposure to staff members within the building, as well as to members of the public outside in the vicinity of the AFRRI facility. Although the reactor room does not completely seal when the dampers are dosed, the slow leakage of radioactive material results in a lower dose to the public than the instantaneous release analyzed in scenario #1. Therefore, the release through room leakage as an Internal exposure is not detailed in this analysis.
TABLE 3. Radiation Doses to Members of the Public for Scenario #1.
Distance (m) CDEmw, + DDE*m (mrem) TEDE (toero,)
10 249 95 50 111 41 100 37 14 150 416 ' 6 200 9 .. 3 250 6 . 2 TABLE 4. Occupational Radiation Doses In the Reactor Room for Scenario #1.
Reactor: Room Occupancy (min) CDEn W + DDEm. (mmrero) TEDE (mrem) 5 1145-' "" 482 "
~~~~~~~. . . . . . .,. .:,... ,:..,. .,- .,....., ., - .-.
~~~~~~.
. . , , , .. .",,'.; .",: ".,.; ,.., .i: . ".. .,
.3 . . ~, 3
TABLE 5. Occupational Radiation Doses in the Reactor.Room for Scenario.#2.,
Reactor Room Occupancy (min) CDE~b + DDETI,, (mrem)n I , TEDE (mrem)
- 5. 1473 . . 613 Direct external exposures to Individuals outside of the: reactor room originating from airborne radioactive material inside the reactorroom were calculated assuming the source term -to bi'thd entire reactor roomovolume. Theseexposure rates encompassed three distinct locations, and were calculated using MicroShield TOV8.02. Receptor A was located 3 ft. from any reactor wall; but not within the reactor room. Receptor Bwas located 20 ft. from any reactor wall, with an additional concreteblock wall between receptor B and the reactor wall. ;Receitbr C was located 100 ft. from any reactor wall, with an additional concrete block wall between receptor C~and-lhe reactor wall.'
Receptor A,represents the staff member in closest proximity to the reactor, typically able to evacuate the area in less than 2 minutes. JToincorporate further, conservatism, the evac ation time for Receptor Awas set-at 5 minutes.
Receptor B represents the closest proximity to the reactor's Controlled Access Area within the AFRRI complex. Receptor B's location represents the highest exposure rate to a staff member who is outside of the Controlled Access Area. All other staff locations throughout AFRRI are a greater distance from the reactor room, and have significantly more shielding. From past emergency drill experiences, it is estimated that the entire AFRRI complex can be evacuated in less than 20 minutes.
Receptor C represents the closest location of an emergency evacuation assemblage point. For the purposes of this calculation, it was assumed that a member of the public could stay at this assemblage point for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the accident. In reality, personnel would be evacuated to a more distant location in this type of accident. The exposures for each receptor are presented in Table 6.
TABLE 6. Radiation Exposures Outside of the Reactor Room in Scenario #2.
Receptor Exposure Rate (mR/hr) Evacuation Time (min) Exposure (mR)
A 92.87 5 7.74 B 15.88 20 5.29 C 1.61 120 3.22 It is important to note that these dose rates are at the time of the failure of the fueled experiment and do not include decay corrections for the duration of any of the evacuation times. This adds a significant conservatism into the estimated exposures. The results presented indicate the contribution of exposure from the source term inside the reactor room to anyone outside the reactor room is well within the 10 CFR 20 limits.
REFERENCES
- 1. "The Calculations of Distance Factors for Power and Test Reactor Sites" DiMunno, JJ. et al., TID-14844, U.S.
Atomic Energy Commission, March 1962.
- 2. Regulatory Guide 3.33 "Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Fuel Reprocessing Plant" U.S. Nuclear Regulatory Commission, April 1977.
- 3. Regulatory Guide 3.34 "Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Uranium Fuel Fabrication Plant" U.S. Nuclear Regulatory Commission, July 1979.
- 4. Regulatory Guide 1.5 "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors" U.S. Nuclear Regulatory Commission, June 1974.
- 5. "A Guide to Radiological Accident Considerations for Siting and Design of DOE Nonreactor Nuclear Facilities" Elder, JC. et al., LA-10294-MS, Los Alamos National Laboratory, January 1986.
- 6. Nuclear Power Reactor Safety Lewis, EE., John Wiley and Sons, 1977, p.521.
- 7. Nuclear Engineering. Theory. and Technology of Commercial Nuclear PowerNuclear Engineering. Theory, and Technology of Commercial Nuclear Knief, RA., Hemisphere Publishing, 1992, pp.353,431.
- 8. "Fuel Elements for Pulsed TRIGA Research Reactors" Simnad, MT. et al., Nuc. Tech. 28, January 1976.
- 9. "The U-ZrH, Alloy: Its Properties and Use in TRIGA Fuer" Simnad, MT. General Atomic Report E-117-883, February 1980.
- 10. Regulatory Guide 1.145 "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" U.S. Nuclear Regulatory Commission, August 1979.
- 11. "Calculated Atmospheric Radioactivity from the OSU TRIGA Research Reactor Using the Gaussian Plume Diffusion Model" Bright, MK. et al., Oregon State University Department of Nuclear Engineering Report 7903, August 1979.
- 12. "Internal Dose Conversion Factors for Calculation of Dose to the Public" DOE/EH-0071, U.S. Department of Energy, Washington DC, 1988.
- 13. "External Dose Conversion Factors for Calculation of Dose to the Public" DOE/EH-0070, U.S. Department of Energy, Washington DC, 1988.
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