ML13182A084

Armed Forces Radiobiology Research Institute - Response to Request for Additional Information Regarding the Application for License Renewal (TAC ME1587)

ML13182A084
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	06/28/2013
From:	Huff L A US Dept of Defense, Armed Forces Radiobiology Research Institute
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME1587
Download: ML13182A084 (18)

Text

) 0 cF9-5'0 ."ýARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE8901 WISCONSIN AVENUEBETHESDA, MARYLAND 20889-5603June 28, 2013Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001SUBJECT:REQUEST FOR ADDITIONAL INFORMATION REGARDING THEAPPLICATION FOR LICENSE RENEWAL (TAC NO. ME1587)Sir:By letter dated May 3, 2013, the Nuclear Regulatory Commission requested additionalinformation necessary to allow processing of our research reactor license renewalapplication (License R-84, Docket 50-170). Answers to those questions are enclosed. Inaddition, following a discussion with our license renewal Project Manager at the AFRRIfacility on June 27, 2013, a correction to Table 4-19 from Chapter 4 of the proposedSafety Analysis Report submitted on March 4, 2010 is enclosed.If you need further information, please contact Mr. Steve Miller at 301-295-9245 orstephen .miller@ usuhs.edu.I declare under penalty of perjury that the foregoing is true and correct to the best of myknowledge. Executed on June 28, 2013.L. Andrew HuffCOL, USAFDirector Financial Qualifications1. Pursuant to 10 CFR 50.33(f) (2), "[t]he applicant shall submit estimates for total annualoperating costs for each of the first five years of operations of the facility." Since theinformation included in the previous correspondence was for the period of fiscal years (FYs)2012 through 2016, please provide the following additional information:(a) Projected operating costs of the AFRRI facility for each of the FY2013 thru FY2018(the first five year period after the projected license renewal). If the cost estimateshave not changed since the previous submittal for the period of FY2013 throughFY2016, please so state.The cost estimates for FY 2013 through FY 2016 remain the same.Estimated operating costs ($ millions) to support AFRRI for FY 2017 and 2018 are:FY AFRRI TRIGA reactor2017 $15.7 $1.32018 $16.0 $1.4(b) Has the source(s) of funding to cover the operating costs for the above FYs changedsince the August 13, 2010, submittal?No

2. By letter dated August 13, 2010, you provided an updated decommissioning costestimate for the facility that was developed using NUREG/CR- 1756, "Technology, Safetyand Costs of Decommissioning Reference Nuclear Research and Test Reactors." Thedecommissioning cost estimate was $14.831 million in 2011 dollars. The cost estimatesummarized costs by labor, radioactive wastes disposal, energy, and a 25-percentcontingency factor.(a) Please indicate if the basis for how the cost estimate was developed haschanged. If NUREG/CR- 1756 is still the basis, please so state.NUREG/CR-1756 remains the basis.(b) Please indicate if there are any changes to the means of adjusting the costestimate and associated funding level periodically over the life of the facility.No changes.

3. AFRRI provided a Statement of Intent (SOl), dated August 11, 2010, stating that"[f]unding will be sought from the [U.S.] Department of Defense in accordance withestablished programming and budgeting procedures, " per 10 CFR 50. 75(e)(1)(iv).(a) Please indicate if there have been any changes to the SOl and if decommissioningfunding obligations of the AFRRI facility continue to be backed by the full faith andcredit of the U.S. Government.There are no changes to the SOl and decommissioning funding obligations of the AFRRI facilityremain backed by the full faith and credit of the U.S. Government.

Technical Specifications1. Technical Specification (TS) 4.1 (a) states, "the reactivity worth of each control rod andthe shutdown margin shall be determined annually but at intervals not to exceed 15months or following any significant core configuration changes". The term "significantcore configuration changes" is not defined in your submissions. Please submit adefinition for significant core configuration changes or justify why it is not necessary.The following definition has been added to TS 1.0:CORE CONFIGURATION: The core configuration includes the number, type, orarrangement of fuel elements and standard control rods/transient rod occupyingthe core grid.TS 4.1 (a) has been changed and now reads:The reactivity worth of each control rod and the shutdown margin shall bedetermined annually but at intervals not to exceed 15 months or following anysignificant (>$0.25) core configuration changes.

2. TS 4.1 (e) states, "The core excess reactivity shall be measured at the beginning of eachday of operation involving the movement of control rods, or prior to each continuousoperation extending more than a day. During extended shutdown periods, the coreexcess reactivity shall be measured at least annually, not to exceed 15 months." ThisTS conforms to the recommendation in American Nuclear Society/American NationalStandards Institute, Inc. (ANS/ANSI) 15. 1 except that ANS/ANSI 15. 1 states "andfollowing significant core configuration and/or control rod changes". Please justify why itis not necessary to measure core excess reactivity following significant coreconfiguration changes and/or control rod changes or submit a change to your TSs.TS 4.1 (e) has been changed and now reads:The core excess reactivity shall be measured at the beginning of each day ofoperation involving the movement of control rods, or prior to each continuousoperation extending more than a day, and following any significant (>$0.25) coreconfiguration changes. During extended reactor shutdown periods, the coreexcess reactivity shall be measured at least annually, not to exceed 15 months.

3. A description of measurements of the fuel elements and fuel follower control rods isincluded in TS 4.2.5 and 5.2.2(e).(a) TS 4.2.5 states, in part, "Fuel elements and fuel follower control rods indicatingan elongation greater than 0. 100 inch, a lateral bending greater than 0.0625 inch,or significant visible damage shall be considered damaged, and shall not be usedin the reactor core." This statement appears to be a limiting condition foroperation (LCO). Please justify why this TS is a surveillance and not an LCO orsubmit a change to the TSs.TS 3.9 has been changed and now reads:3.9. FUEL PARAMETERSApplicabilityThis specification applies to all TRIGA fuel elements and fuel follower control rods.ObiectiveThe objective is to maintain integrity of the fuel element cladding.SpecificationThe reactor shall not operate with damaged fuel elements or fuel follower control rods, exceptfor the purpose of locating damaged fuel elements or fuel follower control rods. A fuel elementor fuel follower control rod shall be considered damaged and must be removed from the core if:a. The transverse bend exceeds 0.0625 inches over the length of the cladding;b. The length exceeds its original length by 0.100 inches;c. A cladding defect exists as indicated by the release of fission products; ord. Visual inspection identifies bulges, gross pitting, or corrosion.BasisGross failure or obvious visual deterioration of the fuel is sufficient to warrant declaration of thefuel as damaged. The elongation and bend limits are the values found acceptable to theUSNRC (NUREG-1537).

TS 4.2.5 has been changed and now reads:4.2.5. REACTOR FUEL ELEMENTS AND FUEL FOLLOWER CONTROL RODSAppl~icabilityThis specification applies to the surveillance requirements for the fuel elements and fuel followercontrol rods.ObiectiveThe objective is to verify that the specifications for fuel elements and fuel follower control rodconditions are met.SpecificationFuel elements and fuel follower control rods shall be inspected visually for damage ordeterioration and measured for length and bend in accordance with the following:a. Before being placed in the core for the first time or following long-term storage;b. Every two years (not to exceed 30 months), or at intervals not to exceed 500 pulses ofinsertion greater than $2.00, whichever comes first, for elements in the B, C, and D ringsand for fuel follower control rods;c. Every four years (not to exceed 54 months) for elements in the E and F rings;d. If damage, deterioration, or unacceptable length or bend measurements are found inone or more fuel elements or FFCRs, all fuel elements and FFCRs in the core shall beinspected for damage or deterioration and measured for length and bend.BasisThe frequency of inspection and measurement is based on the parameters most likely to affectthe fuel cladding of a pulse reactor. Inspecting fuel elements in rings with higher power factorswill provide early indication of fuel damage while significantly reducing the amount of fuelmovement required.(b) TS 5.2.2(e), Reactor Core, states that "fuel elements indicating an elongationgreater than 0. 100 inch, a lateral bending greater than 0.0625 inch, or significantvisible damage shall be considered damaged, and shall not be used in thereactor core." Please justify why this is a design feature and not an LCO.See response to RAI 3 (a).

4. The safety analysis report states that the calculation of the maximum temperature duringthe loss of coolant accident (LOCA) vs. fuel rod power density during operation wasbased on decay heating after operation at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per week for 40 years. What is thebasis for this assumption-is there a limit which caps the number of operating hours perweek? Please explain why a Technical Specifications constraint of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is notnecessary or propose a Technical Specification.A previous RAI was submitted to address this issue. In summary, there are twoanalyses that address the LOCA at AFRRI.The bounding analysis is based on General Atomics Report No. E-1 17-196. This reportidentifies the ability of natural convection to maintain fuel cladding temperature below900'C following infinite operation and instantaneous coolant loss, as long as a powerdensity of 21 kW per element is not exceeded. Operation of the AFRRI TRIGA at a fullpower of 1.1 MW yields a peak power density of 19.4 kW in the B04 fuel elementposition. Given that the B04 location represents the highest power density within theAFRRI core, this scenario is bounded by the GA analysis.However, infinite operation of the AFRRI TRIGA reactor is a purely hypotheticalscenario. Fission product build up within the AFRRI core limits full power operation toless than 24 continuous hours, at which point the negative reactivity from fissionproducts prevent operation at full power. Historically, the AFRRI TRIGA has operatedan average of approximately 500 kW-hrs per week. In order to provide a more accurateand useful analysis of a LOCA at the AFRRI TRIGA reactor, a comparison was made toa LOCA following 1 MW operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per week for 40 years. Although thisanalysis is still substantially higher than the run history at AFRRI, it further indicates themargin of safety when considering a LOCA. In addition to this, a comparison to theOregon State TRIGA Reactor (OSTR) SAR calculations for a power history of 70 MW-hrs per week for 40 years was also discussed.Previous RAI response:As described in 13.2.1 of the AFRRI SAR, the reactor fuel elements rely on ambient airnatural convection through the core to cool the reactor fuel in the event of a loss-of-coolant accident (LOCA). A buoyancy force to drive this natural convection isdeveloped by a hot air column within the core and a cooler column of air outside of thecore. This accident has been analyzed for an instantaneous loss of water and a loss ofwater occurring over a 15 minute period. In addition, this analysis includes scenarios forinfinite full power operation and 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per week operation over a 40 year period.If all reactor coolant was suddenly lost, the primary concern would be the integrity of thefuel cladding. Maintaining a fuel cladding temperature below the NUREG-1282specification of 9500C ensures that the cladding maintains sufficient strength to preventfailure under the pressure of hydrogen gas buildup within the element. Therefore,operating conditions must be such that in the event of a LOCA the fuel claddingtemperature will not exceed 9500C.General Atomics Report No. E-1 17-196 provides a detailed analysis showing that airnatural convection cooling is adequate to maintain fuel cladding below 9000C assumingpower levels no higher than 21 kW per element are achieved. This calculation assumesan instantaneous loss of all coolant and infinite operation at 1 MW. If a 15 minute delay between reactor scram and total coolant loss is assumed, fuel cladding will remain below9000C up to a power level of 22 kW per element. Analysis provided in the Oregon StateTRIGA Reactor SAR shows that if reactor operating history is assumed to be 70 MW-hrsper week over a 40 year period, these limits for maximum power level per elementincrease to 25.2 and 26.4 kW, respectively.As described in 13.2.1.3 of the AFRRI SAR, assuming an operational history of 72 MW-hrs per week for 40 years, maximum fuel cladding temperatures following LOCA are5480C assuming instantaneous coolant loss and 4770C assuming a 15 minute delaytime. These calculations assume operation at 1 MW.The current maximum licensed steady state power of the AFRRI reactor is 1.1 MW.Section 4.5.8 of the AFRRI SAR discusses the power peaking within the 85-3 core anddetermines the highest power factor of 1.552 to occur in position B04. Assuming 1.1MW with 88 fuel elements, the B04 element has a power of 19.4 kW. This is within the21 kW per element limit set for the worst-case LOCA, and substantially less than themore representative 26.4 kW per element limit. In reality, AFRRI's operating history isfar below 70 MW-hours per week, with a historical average closer to 500 kW-hours perweek.

5. TS 1.10, the definition of Initial Reactor Startup is, "The first reactor startup following fuelelement relocation within the core." This definition does not conform to therecommendations of NUREG- 1537 which suggests that initial reactor startup is definedas "the first startup after the reactor is secured." Please justify why your definition isacceptable or submit a change to the TSs.The definition of Initial Reactor Startup (TS 1.10) and Reactor Startup (TS 1.26) have beenremoved.

6. TS 1.24, the definition of Reactor Secured is, "Either sufficient fuel is removed to ensurea $1.00 (or greater) shutdown margin ... "does not conform to the recommendations ofANSI/ANS 15. 1 which state "Either there is insufficient moderator available in the reactorto attain criticality... "Please justify why this is acceptable or submit a change to theTSs.TS 1.24 (a) has been changed and now reads:a. Either there is insufficient moderator available in the reactor to attain criticality orthere is insufficient fissile material present in the reactor to attain criticality underoptimum available conditions of moderation and reflection;

7. TS 3.1.3(b), Reactivity Limitations states, "The shutdown margin provided by theremaining control rods with the most reactive control rod fully withdrawn shall be $3.50... "However, the definition for shutdown margin uses the term "the most reactiveposition." Is the term "the control rod fully withdrawn" equivalent to "the most reactiveposition" and do the two terms refer to the same position? Please justify why differentterms were used, and why this is acceptable or submit a change to the TSs.TS 3.1.3(b) has been changed and now reads:The shutdown margin provided by the remaining control rods with the most reactive rodin the most reactive position shall be greater than ...NOTE: The initial NRC question refers to TS 3.1.3(b). It is our understanding that there is atypo in the question and that $3.50 should actually be $0.50. The response assumesthat this question refers to a $0.50 shutdown margin.

8. TS 3.5.2, Effluents, Argon-41 Discharge Limit, and TS 6.6(b), Operating Reports,Gaseous Waste, indicate that Argon-41 is the only effluent measured. Please justify whyother effluents are not measured and indicate how samples are obtained as well as themethod used to determine that the samples are statistically representative.During normal reactor operations, Ar-41 and N-16 are the only airborne radioisotopesproduced. Given the short half-life of N-16 (-7 seconds), the dose to members of thepublic from N-16 production in the AFRRI reactor pool is insignificant.A description of the AFRRI Stack Gas Monitoring System (SGM) is found in Section7.7.3.2 of the AFRRI SAR. Briefly, a sample of air exhausting from the reactor room andexposure rooms is directed through a Nal scintillation detection system locateddownstream of the reactor absolute air filters. Air samples are collected through a seriesof tubes within the stack. These tubes traverse the entire cross section of the stack andpull air from numerous locations within the same planar space. The alarms on thedetector have local and remote readouts to alert reactor personnel if higher than normalradioactive effluent levels are present in the exhausting air.

9. TS 3.6(b), Limitations on Experiments states, "Each fueled experiment shall be limitedso that the total invehtory of iodine isotopes 131 through 135 in the experiment is notgreater than 1.0 curie and the maximum strontium-90 inventory is not greater than 5millicuries." The basis states that these limits assure that the dose to members of thepublic will not exceed the limits of 1 0 CFR Part 20. Please explain why this does notapply to workers or submit a change to the TS bases.The basis for TS 3.6(b) has been changed and will now read:The 1.0 curie limitation on iodine-131 through 135 assures that, in the event of amalfunction of a fueled experiment leading to total release of radioactive materialincluding fission products, the dose to any individual will not exceed the limits specifiedin 10 CFR 20.

10. TS 3.4, Ventilation states, "The reactor shall not been operated unless the facilityventilation system fan is operating, except for periods of time during which the dampersshall be closed. In the event of a release of airborne radioactivity in the reactor roomabove both routine reactor operation and normal background values, the ventilationsystem to the reactor room shall be secured via closure dampers automatically by asignal from the reactor deck air particulate monitor." Please explain what is meant by"except for periods of time during which the dampers shall be closed." Please describeanalyses that have been completed with regard to releases/exposures under both openand closed positions of the dampers.TS 3.4 has changed and now reads:3.4. VENTILATION SYSTEMApplicabilityThis specification applies to the operation of the facility ventilation system.ObiectiveThe objective is to assure that the ventilation system is operable to mitigate theconsequences of possible releases of radioactive materials resulting from reactoroperation.SpecificationThe reactor shall not be operated unless the facility ventilation system is operating,except for periods of time not to exceed two (2) hours to permit repair, maintenanceor testing of the ventilation system. In the event of a release of airborneradioactivity in the reactor room above both routine reactor operation and normalbackground values, the ventilation system to the reactor room shall beautomatically secured via closure dampers by a signal from the reactor deck airparticulate monitor.BasisDuring normal operation of the ventilation system, the concentration of argon-41 inunrestricted areas is below the limits allowed by 10 CFR 20. In the event of a fuelcladding rupture resulting in a substantial release of airborne particulateradioactivity, the ventilation system shall be shut down, thereby isolating the reactorroom automatically by spring-loaded, positive sealing dampers. Therefore,operation of the reactor with the ventilation system shut down for short periods oftime ensures the same degree of control of release of radioactive materials.Moreover, radiation monitors within the building independent of those in theventilation system will give warning of high levels of. radiation that might occurduring operation with the ventilation system secured.Analyses comparing doses to staff and members of the public in the open and closed damperpositions are found in Chapter 13 of the proposed Safety Analysis Report. Specifically, the revised response to RAI #12 submitted to the NRC on January 17, 2012 discusses thebounding MHA for each damper position.

The following table will replace the existing Table 4-19 of the proposed SAR.Table 4-19AFRRI TRIGA RELAP5 Thermal Results Summary for Core Operating at1.1 MW.Parameter Initial CoreAxial peaking factor -average element 1.316Axial peaking factor -hot element 1.343Hot element power factor 1.560Inlet coolant temperature 480C, 1180FCoolant saturation temperature at core inlet 110.30C, 230.50FExit coolant temperature -average element 67.110C, 152.80FExit coolant temperature -hot element 82.510C, 180.50FAverage temperature in pool above core 60.20C, 140.40FCoolant mass flow 13.60 kg/sec, 107,900 lb/hrAverage flow velocity 29.48 cm/sec, 0.967 ft/secCore average fuel temperature 247.10C, 476.70FPeak fuel temperature in average fuel element 360.00C, 679.90FMaximum wall temperature in hot element 149.2-C, 300.60FPeak fuel temperature in hot fuel element 440.70C, 825.30FAverage heat flux 27.87 W/cm2, 88,362 BTU/hr-ft2Maximum heat flux in hot element 58.40 W/cm2, 185,125 BTU/hr-ft2Minimum DNB ratio of 1.0 1.99 MW