ML12032A054

Response to Request for Additional Information Regarding the Application for License Renewal (TAC No. ME1587)

ML12032A054
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	01/17/2012
From:	Melanson M A US Dept of Defense, Armed Forces Radiobiology Research Institute
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME1587
Download: ML12032A054 (9)

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/ 1ýr) C F R 3_e' 94:ýARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE8901 WISCONSIN AVENUEBETHESDA, MARYLAND 20889-5603January 17, 2012Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001SUBJECT: REQUEST FOR ADDITIONAL INFORMATION REGARDING THEAPPLICATION FOR LICENSE RENEWAL (TAC NO. ME1587)(License R-84, Docket 50-170)Sir:After conversation with Mr. Walter Meyer, we are submitting a complete revision of ourresponse to Question 12 of your initial RAI. Our November 28, 2011 response to Question 12should be withdrawn and replaced by this version. The only changes from the November 28,2011 version are indicated by a vertical line in the left margin on page six. Because this itemcontains security-related information, we request that all eight pages be withheld from publicdisclosure under 10 CFR 2.390.If you need further information, please contact Mr. Steve Miller at 301-295-9245 ormillers @afrri.usuhs.mil.I declare under penalty of perjury that the foregoing and all enclosed information is true andcorrect to the best of my knowledge. Executed, on January 17, 2012.Enclosure:asMARK A. MELANSONCOL, MS, USADirector-.1 .*I I

12. NUREG-1537, Part 1, Section 13.1.6 provides guidance for the licensee to discuss events that couldresult from experiment malfunction. The licensee is requested to justify its assumption that the releaseof irradiation Argon accident scenario is the worst conceivable case for radiological consequences froman experiment. The licensee should present a range of experimental malfunction accidents considered.The Argon activation assumptions and calculations should be presented in more detail.All experiments performed as part of the TRIGA reactor operations are reviewed by the Reactor andRadiation Facilities Safety Subcommittee and supervised by trained, licensed, supervisory personnel.The Technical Specifications contain requirements that must be met before performing experimentsusing the AFRRI-TRIGA reactor, including the requirement that the failure of one experiment cannotcontribute to the failure of any other experiment. Although improbable, an experiment could fail andtherefore the consequences of a variety of experimental accidents have been considered.The most common experiments performed at AFRRI involve the irradiation of biological samples. Theradiological consequences from the failure of an experiment of this type are very minimal, as failurewould not pose any risk to the reactor structure itself or result in a release of significant quantities ofradioactive 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 greaterthan 25 mg is prohibited. Smaller quantities may be irradiated assuming they are housed in a containercapable of withstanding a pressure burst greater than twice the pressure resulting from detonation ofthe sample. The calculations demonstrating the ability of the container to withstand the pressure burstare to be reviewed by the Reactor and Radiation Facilities Safety Subcommittee and approved by theReactor Facility Director. Failure of an explosive experiment therefore does not classify as a worstconceivable event.Samples containing corrosive material must be doubly encapsulated according to the TechnicalSpecifications and therefore the consequences of a failure are limited. Failure of a corrosive experimentin the reactor pool would be diluted by the primary coolant and, while resulting in the need for acleanup and inspection of reactor fuel and instrumentation, would not present a worst case radiologicalevent.It is also notable that the sum of all experiments will not exceed $3.00. The AFRRI-TRIGA reactor iscapable of pulsing up to $3.00 and has proven that changes in reactivity of this magnitude are notdamaging to the reactor. Therefore, simultaneous failure of all experiments would result in reactorconditions that are more conservative than a standard pulse operation.Another experiment identified in the SAR as the irradiation of a 20 liter container of argon gas in ER1 forone hour at 1 MW. This results in a total argon-41 inventory of 5.6 Ci. Release of this quantity from thereactor stack yields an effective dose to the closest member of the public of 0.2 mrem and is notconsidered the worst case radiological event resulting from an experiment malfunction.Revised January 17, 2012 The irradiation of fueled experiments at AFRRI are limited so that the total inventory of iodine isotopes131 through 135 in the experiment is not greater than 1.0 Ci and the maximum strontium-90 inventoryis not greater than 5 mCi. Given the fission product yields reported in NUREG/CR-2387, the 1.0 Ci limitof radioiodines 131 through 135 is reached before 5 mCi of strontium-90 in fueled experiments. Therelease of these quantities directly from the AFRRI stack has been analyzed and determined to representthe worst case scenario of a release of radioactive material from experimental failure. In this accidentscenario, a fueled experiment is irradiated until 1.01 Ci of radioiodines 131-135 are present.For this experiment, it is assumed that 1 g of 19.75% enriched LEU is irradiated in the AFRRI core for 42minutes.at 1 MW. The assumed thermal neutron flux at this power level and sample location is lx1013n/cm2_s. The source term for this experiment was generated using ORIGEN, with radioisotope activitiesof interest shown in Table 1.TABLE 1. ORIGEN source term for fueled experiment.Isotope Half Life Activity in Experiment (mCi) Activity Released to Environment (mCi)Br-82 35.3 h 4.OOE-04 1.OOE-04Br-83 2.4 h 4.15E+01 1.04E+01Br-84m 6.0 m 9.55E+00 2.39E+00Br-84 31.8 m 3.38E+02 8.46E+01Br-85 2.87 m 6.95E+02 1.74E+02Br-86 55.5 s 9.86E+02 2.46E+02Br-87 55.9 s 1.18E+03 2.95E+021-131 8.02 d 5.35E-01 1.34E-011-132 2.28 h 2.46E+00 6.16E-011-133 20.8 h 3.48E+01 8.70E+001-134 52.6 m 7.17E+02 1.79E+021-135 6.57 h 2.56E+02 6.40E+011-136 83.4 s 1.50E+03 3.75E+02Kr-83m 1.86 h 4.69E400 4.69E+00Kr-85m 4.48 h 6.72E+01 6.72E+01Kr-85 10.76 y 1.16E-04 1.16E-04Kr-87 76.2 m 4.61E+02 4.61E+02Kr-88 2.84 h 3.17E+02 3.17E+02Kr-89 3.15 m 2.59E+03 2.59E+03Xe-131m 11.9 d 3.20E-06 3.20E-06Xe-133m 2.19 d 1.34E-02 1.34E-02Xe-133 5.24 d 4.96E-02 4.96E-02Xe-135m 15.3 m 1.10E+02 1.10E+02Xe-135 9.1 h 1.12E+01 1.12E+01Xe-137 3.82 m 3.47E+03 3.47E+03Xe-138 14.1 m 3.16E+03 3.16E+03Revised January 17, 2012 It is very conservatively assumed that 25% of the halogens released from the sample into the fueledexperiment sample holder are eventually available for inhalation by a radiation worker in the reactorroom or a member of the public in the unrestricted environment. This value is based on historical usageand recommendations (Ref. 1-9), where Ref. 1 recommends a 50% release fraction for the halogensfrom the gap of a fuel element to the air. For the purpose of evaluating the consequences of a failedfueled experiment, the release fraction due to plateout from the gap of a fuel element is assumed to beequal to the release fraction due to plateout from the sample holder. Ref. 2 and Ref. 3 apply a naturalreduction factor of 50% due to plateout in the reactor building. The 25% total halogens released resultsfrom combining the 50% release from the sample holder with the 50% plateout. However, this 25%value appears to be quite conservative, as Ref. 6 and Ref. 7 quote a 1.7% release fraction from the gapof a fuel element rather than a 50% release fraction from the gap. The experience at TMI-2, along withrecent experiments, also indicates that the 50% halogen release fraction from the gap is much too largeand reports that possibly as little as 0.06% of the iodine reaching the cladding gap may be released intothe reactor room, due in part to a large amount of the elemental iodine reacting with cesium to formCsl, a compound much less volatile and more water soluble than elemental iodine (Ref. 7). It isreasonable to assume that this same reaction occurs in the fueled experiment sample holder, thereforereducing the amount of radioiodine released from the fueled experiment sample holder to the reactorroom. It is assumed that 100% of the noble gases are available for release to the unrestrictedenvironment.This accident analysis assumes that 100% of the fission products present in the sample are released tothe sample holder, with no restrictions on release from the sample matrix itself (unlike the fuel elementgap release where an additional fission product reduction is attributed to the design of the TRIGA fuelmatrix and its ability to restrict the release of fission products). Because of this conservativeassumption, the physical state of the fueled experiment does not need to be specified. For example, it isunderstood that a larger fraction of fission products release from a liquid sample than a solid sample. Byassuming 100% release, the accident analysis provides a true worst case scenario regardless of thephysical properties of the fueled experiment sample. In reality, both liquid and solid fueled experimentswould restrict the release of fission fragments to the sample holder to some extent, thus reducing thedoses to radiation workers and members of the public.The minimum distance to the unrestricted environment, as well as the minimum distance to the nearestoccupied building, are assumed to be in the same direction as the prevailing wind. These assumptionsresult in the highest possible radiation doses to members of the public.For any atmospheric stability (Pasquill) class, a ground-level release always leads to a higher effluentconcentration at any given distance than an elevated release. Accordingly, it is assumed for thisaccident analysis that only ground level effluent releases occur, and no credit is taken for either releaseheights or building wake effects. In reality, the release of radioactive material would emit from theAFRRI stack at a height of 13 meters; resulting in lower doses to the public than those reported in Table3. A COMPLY calculation estimates the dose at 10 meters from the AFRRI stack to be reduced by afactor of 10 when accounting for the elevated release. Furthermore, atmospheric modeling indicatesthat the more stable the atmospheric class and the lower the wind speed, the higher the effluentRevised January 17, 2012 concentration. Therefore, this analysis assumes both the most stable atmospheric class (Pasquill F) anda low wind speed (1 m/s) are present. The time that a receptor is exposed to the plume is determinedby 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 are used(Ref. 10) in the accident analysis. For distances greater than 100 m, the values for horizontal andvertical dispersion coefficients are also taken from Regulatory Guide 1.145. For distances from 10 m to100 m, not addressed in Regulatory Guide 1.145, data from the OSTR SAR are used (Ref. 11). The valuesfor 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 1m/s.Distance (m) G y (m) a z (m) x/Q (s/m3)10 1.29 1.04 5.93E-0250 2.45 1.2 2.71E-02100 3.9 2.2 9.27E-03150 6.18 3.22 4.00E-03200 8.21 4.13 2.35E-03250 10.21 4.98 1.57E-03Furthermore, it is assumed that all of the fission products are released to the unrestricted area by asingle reactor room air change, which would maximize the dose rate to persons exposed to the plumeduring the accident.Additional parameters used in this accident are:* Reactor room ventilation exhaust rate: 1.68 m3/s* Reactor room volume: 917 M3* Receptor breathing rate: 3.3x10' 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 isotopesof concern are calculated by:Revised January 17, 2012 (x/Q)D = atmospheric dispersion factor at a given distance D (s/m3)BR = breathing rate (m3/s)DCFiti = internal dose conversion factor for isotope i (mrem/gCi)A, = initial activity of isotope I (gCi)Rv = ventilation of air from the reactor room (m3/s)V = reactor room volume (M3),= ventilation constant = Rv/V (s-').= decay constant for isotope i (s1)t= time when the plume first arrives at the receptor point (s)t2= 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 fromthe facility for both the thyroid and whole body are each calculated by:(DDEthyroid or DDEWB)D = WDCFext't A ,A(e-a'tL1 -e-Lt21 = external dose rate conversion factor for isotope i (mrem m3/pCi s)For calculating the dose to occupational workers in the reactor room, a stay time of 5 minutes is used.Experience indicates that the reactor room can easily be evacuated in less than 2 minutes however; thevalue of 5 minutes is used to account for any time the worker may be delayed performing a task. TheCDE and CEDE for personnel in the reactor room for a given stay-time may each be calculated by:(CDE or CEDE)s > [DCFinti AlBR(- 2 e ff V ]AeffA'f = V=+tsr = stay time of personnelThe DDE to personnel in the reactor room for a given stay time for both the thyroid and the whole bodyare calculated by:Revised January 17, 2012 (DDEth3roid or DDEWB)ST = [ [DCFextji Aj(1- e-effts)]1[ 'Aeff VThe results of these calculations are shown in Tables 3-5. In all cases, doses for the general public andoccupational workers are below the annual dose limits specified by 10 CFR 20.There are two different scenarios analyzed in this accident. In scenario #1, the isolation dampers failfollowing the release of radioactive material into the reactor room. As a result, the radioactive materialis vented from the AFRRI stack to the unrestricted public. In scenario #2, the isolation dampers operateas designed and limit the radioactive material release from the reactor room. This latter scenario resultsin a higher exposure to the reactor staff member in the reactor room. As the radioactive materialsdisperse in the reactor room, the room becomes a source term for external exposure to staff memberswithin 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 closed, the slow leakage ofradioactive material results in a lower dose to the public than the instantaneous release analyzed inscenario #1. Therefore, the release through room leakage as an internal exposure is not detailed in thisanalysis.TABLE 3. Radiation Doses to Members of the Public for Scenario #1.Distance (m) TEDE (mrem)10 7650 33100 11150 5200 3250 2IITABLE 4. Occupational Radiation Doses in the Reactor Room for Scenario #1.Reactor Room Occupancy (min) TEDE (mtero)5 401TABLE 5. Occupational Radiation Doses in the Reactor Room for Scenario #2.Reactor Room Occupancy (min) TEDE (mrem)5 508Direct external exposures to individuals outside of the reactor room originating from airborneradioactive material inside the reactor room are calculated assuming the source term to be the entirereactor room volume. These exposure rates encompass three distinct locations, and are calculatedIIRevised January 17, 2012 using MicroShield TM V8.02. Receptor A is located 3 ft. from any reactor wall, but not within the reactorroom. Receptor B is located 20 ft. from any reactor wall, with an additional concrete block wall betweenreceptor B and the reactor wall. Receptor C is located 100 ft. from any reactor wall, with an additionalconcrete block wall between receptor C and the reactor wall.Receptor A represents the staff member in closest proximity to the reactor, typically able to evacuatethe area in less than 2 minutes. To incorporate further conservatism, the evacuation time for ReceptorA is set at 5 minutes.Receptor B represents the closest proximity to the reactor's Controlled Access Area within the AFRRIcomplex. Receptor B's location represents the highest exposure rate to a staff member who is outsideof the Controlled Access Area. All other staff locations throughout AFRRI are a greater distance from thereactor room, and have significantly more shielding. From past emergency drill experiences, it isestimated 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 thepurposes of this calculation, it was assumed that a member of the public could stay at this assemblagepoint 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 distantlocation 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 lime (min) Exposure (mR)A 118 5 9.9B 21 20 7C 2 120 4It is important to note that these dose rates are at the time of the failure of the fueled experiment anddo not include decay corrections for the duration of any of the evacuation times. This adds a significantconservatism into the estimated exposures. The results presented indicate the contribution of exposurefrom the source term inside the reactor room to anyone outside the reactor room is well within the 10CFR 20 limits.REFERENCES1. "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 RadiologicalConsequences of Accidental Nuclear Criticality in a Fuel Reprocessing Plant" U.S. NuclearRegulatory Commission, April 1977.3. Regulatory Guide 3.34 "Assumptions Used for Evaluating the Potential RadiologicalConsequences of Accidental Nuclear Criticality in a Uranium Fuel Fabrication Plant" U.S. NuclearRegulatory Commission, July 1979.Revised January 17, 2012

4. Regulatory Guide 1.5 "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Loss of Coolant Accident for Pressurized Water Reactors" U.S. Nuclear RegulatoryCommission, June 1974.5. "A Guide to Radiological Accident Considerations for Siting and Design of DOE NonreactorNuclear 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 Power _Knief, RA.,Hemisphere Publishing, 1992, pp.353, 431.8. "Fuel Elements for Pulsed TRIGA Research Reactors" Simnad, MT. et al., Nuc. Tech. 28, January1976.9. "The U-ZrHx Alloy: Its Properties and Use in TRIGA Fuel" Simnad, MT. General Atomic Report E-117-883, February 1980.10. Regulatory Guide 1.145 "Atmospheric Dispersion Models for Potential Accident ConsequenceAssessments at Nuclear Power Plants" U.S. Nuclear Regulatory Commission, August 1979.11. "Calculated Atmospheric Radioactivity from the OSU TRIGA Research Reactor Using theGaussian Plume Diffusion Model" Bright, MK. et al., Oregon State University Department ofNuclear 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.Revised January 17, 2012