ML12272A303

Request for Additional Information Regarding the Application for License Renewal

ML12272A303
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	09/21/2012
From:	Huff L US Dept of Defense, Armed Forces Radiobiology Research Institute
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME1587
Download: ML12272A303 (6)

Text

mARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE 8901 WISCONSIN AVENUE BETHESDA, MARYLAND 20889-5603 September 21, 2012 0)

Nuclear Regulatory Commission ATFN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE APPLICATION FOR LICENSE RENEWAL (TAC NO. ME1587)

Sir:

The Nuclear Regulatory Commission previously requested additional information necessary to allow processing of our research reactor license renewal application (License R-84, Docket 50-170). We are now submitting revised answers to three of those questions. The enclosed revised answers replace the original submission as indicated:

1. Response to Question 6 of first RAI set. Originally submitted on April 20,2012.

2. Response to Question 18 of first RAI set. Last submitted on October 20, 2011.

3. Response to Question 3b of second RAI set. Originally submitted on October 21, 2010.

If you need further information, please contact Mr. Steve Miller at 301-295-9245 or I declare under penalty of perjury that the foregoing and all enclosed information is true and correct to the best of my knowledge. Executed n September 21, 2012.

Enclosures:

L REW HUFF as COL, SAF Acting Director

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6. NUREG-1537 Part1,Section 13 provides guidance to the licensee to discuss potential occidentscenarios. Section 13.1.5 of the SAR presents the results of an analysis of a reactivity insertionof $0.51. Justify the magnitudeof this assumed reactivityinsertionIn comparison with the maximum reactivityInsertionassociated with any single experiment.

The following analysis will replace the reference to a reactivity insertion of $0.51 of the SAR:

The failure of an experiment or experiments could result in Instantaneous insertion of reactivity. The worst possible case would be the prompt addition of $3.00 (2.1% Ak/k) within the reactor core. The Technical Specifications establish that the sum of the absolute reactivity worths of all experiments in the reactor and in the associated experimental facilities shall not exceed $3.00 (2.1% Ak/k). The instantaneous insertion of $3.00 (2.1% Ak/k) to the reactor core as a result of a worst case reactivity insertion is bounded by the analysis of the $3.50 (2.45%

Ak/k) pulse limit and would not result in any adverse safety conditions within the AFRRI TRIGA core.

In addition to this analysis, the loss-of-coolant (LOCA) scenario has been updated to reflect the 1.1 MW licensed power limit. The following analysis will be added to SAR Section 13.2.1.3.

As described in 13.2.1 of the AFRRI SAR, the reactor fuel elements rely on ambient air natural 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 is developed by a hot air column within the core and a cooler column of air outside of the core. This accident has been analyzed for an instantaneous loss of water and a loss of water occurring over a 15 minute period. In addition, this analysis indudes scenarios for infinite 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 the fuel cladding. Maintaining a fuel cladding temperature below the NUREG-1282 specification of 9500 C ensures that the cladding maintains sufficient strength to prevent failure 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 cladding temperature will not exceed 9500 C.

General Atomics Report No. E-1 17-196 provides a detailed analysis showing that air natural convection cooling Is adequate to maintain fuel cladding below 900*C assuming power levels no higher than 21 kW per element are achieved. This calculation assumes an 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 below 9000 C up to a power level of 22 kW per element. Analysis provided In the Oregon State TRIGA Reactor SAR shows that if

reactor operating history Isassumed to be 70 MW-hrs per week over a 40 year period, these limits for maximum power level per element increase 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 are 5480 C assuming instantaneous coolant loss and 4770 C assuming a 15 minute delay time. 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 and determines the highest power factor of 1.552 to occur in position B04. Assuming 1.1 MW with 88 fuel elements, the B04 element has a power of 19.4 kW. This is within the 21 kW per element limit set for the worst-case LOCA, and substantially less than the more representative 26.4 kW per element limit. In reality, AFRRI's operating history is far below 70 MW-hours per week, with a historical average closer to 500 kW-hours per week.

(Revised September 21, 2012)

18. TS 3.1.2: ANSI/ANS-15.1-2007, Section 3.1(3) provides guidancefor the LCOfor pulse limits.

In TS 3.1.2, the LCO for pulse mode operatiOnsspecifies the maximum step insertionof reactivity shall be $4.00 in the pulse mode. Anaolyses performned by the licensee indicate that this magnitude pulse may achieve a peakfuel temperature that exceeds the fuel vendor's recently recommendedpeak temperatureof 830 degrees C duringpulse mode operations. Please analyze and discuss how TS 3.2.1 should be revised to meet the fuel vendor recommendation.

TS 3.1.2 has been modified to change the maximum allowable pulse size to $3.50. This smaller size pulse will not affect AFRRI operations since experimental pulses over $3.50 have not been required in more than 20 years and are not expected to be needed in the future. The TS basis has been revised to indicate the current temperature analysis from the new Safety Analysis Report Chapter 4 submitted on March 4, 2010.

The Limiting Safety System Setting from the TS will remain at 6000 C. Maintaining a fuel cladding temperature below the NUREG-1282 specification of 9500C (when the cladding equals the fuel temperature) ensures that the cladding maintains sufficient strength to prevent failure under the pressure of hydrogen gas buildup within the element. This 950 0C temperature limit does not account for the cooling provided by the primary water and thus yields a conservative value. When fuel cladding is maintained below 500tC, a peak fuel temperature limit of 11500C is recommended. The AFRRI TRIGA Limiting Safety System Setting specified in the Technical Specifications Is 600f. This provides a significant safety margin to ensure the Safety Limit Is not exceeded during reactor operations.

There are two fuel temperature monitoring channels within the reactor core (one in the B ring and one In the C ring). The highest power density occurs in these two rings, and therefore provides temperature monitoring in the hottest locations of the reactor core. Table 4-14 of the AFRRI SAR identifies the rod power factors for each fuel location in the reactor core. Within the B ring, the highest and lowest power factors are 1.552 and 1.525, respectively. Assuming the instrumented fuel element (IFE) is located in the lowest power density position (801), a temperature indication of 6000C would yield a peak temperature at the highest power density location (B04) of 6110C. This value is well within the conservative NUREG-1282 limit of 950WC.

Within the C ring, the highest and lowest power factors are 1.438 and 1.374, respectively.

Assuming the instrumented fuel element (IFE) is located in the lowest power density position (C12), a temperature indication of 6000C would yield a peak temperature at the highest power density location (C09) of 6280 C Similarly, this value is well within the conservative NUREG-1282 limit of 9500C.

(Revised September 21, 2012)

3b. Where Is the person located? If this dose Is ftom immersion in the Ar-41 plume when it reachesground level, confirm that a higherdose is not possiblefrom radiationshinefrom the plume passingover a person closer to the focility than the point at which the plume reaches groundlevel orfrom a person exposed to directradiationshinefrom the Ar-41 source before releasefrom the AFRRI researchreoctorfocility.

In order to provide a thorough evaluation of the consequences from Ar-41 release, the following will address the estimated doses to members of the public outside of the AFRRI facility and radiation workers within the reactor room.

Theoretical Dose to Members of the Public from Ar-41 Release The following analysis illustrates that even under operating conditions that are orders of magnitude above those found at AFRRI, no member of the public will exceed the 10 CFR 20 limit of 100 mrem as a result of Ar-41 release during reactor operations.

The maximally exposed member of the public is located at the Zachary and Elizabeth Fisher House, 91 meters from the AFRRI research reactor stack. This location represents the closest "full occupancy" area, e.g. an area occupied full-time by an individual or permanent residence position from the AFRRI research reactor stack. Other features located less than 91 meters from the AFRRI research reactor stack are roads, sidewalks, parking lots, parking garages, and the AFRRI front or rear patios. None of the aforementioned outdoor areas could be considered either a permanent residence or a full occupancy area.

In order to calculate the theoretical dose to a person exposed to direct radiation shine from an Ar-41 source released from the AFRRI research reactor facility, simulations were conducted utilizing MicroSkyshineg. If we assume continuous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation at 1 MW with an Ar-41 production rate of 5.1 ptCi/sec, we obtain a value of 0.44 Ci of Ar-41. If,for the sake of conservatism, 0.44 Ci is rounded to 0.5 Ci, and we presume a member of the public was standing ten feet from the AFRRI building exterior, the exposure rate from direct radiation shine would be approximately 0.023 mR/hr (0.00023 mSv/hr).

In radiation protection, the occupancy factor for an area is defined as the average fraction of time that the maximally exposed individual is present while radiation is being produced. NCRP Report No. 151 "Structural Shielding Design and Evaluation from Megavoltage X-and Gamma-ray Radiotherapy Fadlities" recommends an occupancy factor of 1/40 for outdoor areas with only transient pedestrian or vehicular traffic, unattended parking lots, vehicular drop off areas (unattended), stairways, and unattended elevators. Applying this 1/40w occupancy factor to an 8,760-hour year results in an occupancy time of 219 hours0.00253 days <br />0.0608 hours <br />3.621032e-4 weeks <br />8.33295e-5 months <br />, resulting in a dose of 5 mR in a year, well below 10 CFR 20 limits.

If we apply the NCRP-recommended occupancy factor of 1/20 for outdoor areas with seating (the AFRRI patios) to an 8,760-hour year, the result is an occupancy time of 438 hours0.00507 days <br />0.122 hours <br />7.242063e-4 weeks <br />1.66659e-4 months <br />, resulting in a dose of 10 mR in a year. If, for the sake of conservatism, we assume the occupancy time was increased to a standard 2,000-hour working year with an occupancy factor of 1 (100%), the resulting 46 mR dose to a member of the public remains well below 10 CFR 20 limits. In practice, the perimeter of AFRRI is continually monitored by AFRRI security personnel to discourage loitering; hence the occupancy of these outdoor areas is limited to well below the occupancy times recommended in NCRP Report No. 151.

Theoretical Dose to Radiation Worker within the Reactor Room from Ar-41 The following analysis illustrates that even under operating conditions that are orders of magnitude above those found at AFRRI, no radiation worker will exceed the 10 CFR 20 limit of 5 rem as a result of Ar-41 release during reactor operations.

The reactor room is posted as a high radiation area and a Reactor Controlled Area. This means that all personnel entering the room must be equipped with dosimetry per AFRRI Instruction 6055.8F "Radiation Protection Program." All dosimeters for AFRRI radiation workers are analyzed approximately quarterly and flagged if they exceed AFRRI ALARA limits, which are established below 10 CFR 20 limits. The AFRRI personnel monitoring program ensures that no personnel will exceed 10 CFR 20 limits as a result of Ar-41 release into the reactor room during operations.

As described in the AFRRI SAR, the equilibrium Ar-41 concentration in the reactor room during operation at 1 MW is 3.2 x 10U jLCi/cm 3. Although this concentration exceeds the 10 CFR 20 DAC of 3 x1 4 i LCi/cm 3 , in an absolute worst case scenario a radiation worker remaining in the reactor room for an entire working quarter (520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />) while the reactor was continuously operated at 1 MW would receive an approximate dose of 1.3 rem. This dose exceeds the AFRRI ALARA limit and would trigger an investigation into the radiation worker's dose and a limit on their exposures for the remainder of the year, thus ensuring that the worker would never exceed 10 CRF 20 limits. For comparison, to achieve the 5 rem dose described in 10 CFR 20, the reactor must operate at 1 MW during the entire working year (2,000 MW-hrs) while the historical operating records at AFRRI indicate an average annual operation of 20 MW-hrs. A radiation worker remaining in the reactor room for all 20 MW-hrs of operation in a single year would receive an approximate dose of 53 mrem, -10% of the 10 CFR 20 limit.

(Revised September 21, 2012)