ML19221B602

North Carolina State University, Safety Analysis in Support of Fueled Experiments for the Ncsu Pulstar Reactor

ML19221B602
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Site:	North Carolina State University
Issue date:	06/21/2019
From:	North Carolina State University
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SAFETY ANALYSIS IN SUPPORT OF FUELED EXPERIMENTS FOR THE NCSU PULSTAR REACTOR Nuclear Reactor Program NORTH CAROLINA STATE UNIVERSITY RALEIGH, NORTH CAROLINA 27695 LICENSE NO. R-120 DOCKET NO. 50-297 21-JUNE-2019

INTRODUCTION Information is provided in this analysis as a supplement to the Final Safety Analysis Report (FSAR) in support of the license renewal for the NCSU PULSTAR nuclear reactor. Changes to Technical Specifications (TS) 1.2.9.e, 3.5, 3.8, and 4.4 regarding fueled experiments are also supported by this analysis. This analysis used fluence rates at 1 MW to determine saturation activities. At 2 MW, the fluence rates would be twice those at 1 MW and the mass needed to attain the same saturation activity is halved. Limits for fission rate and total number of fissions are calculated based on radiation dose. These experimental limits are the same for 1 MW or 2 MW at the same radiation dose. Ventilation rates for the reactor building are not affected by power level.

Based on experiment needs, the amounts and types of fissionable materials and a maximum fission rate were determined. The maximum fission rate was used in TS 3.8 to define the upper limit for fueled experiments.

Radiation doses from released fission gases and halogens are then calculated for occupied locations inside and outside the reactor building. Potential radiation dose is limited to three percent (3%) of the annual radiation dose limits given in 10 CFR Part 20; specifically, Total Effective Dose-Equivalent (TEDE) inside the reactor building is limited to 0.15 rem; Total Organ Dose-Equivalent (TODE) to the thyroid inside the reactor building is limited to 1.5 rem, and the TEDE in public areas outside the reactor building is limited to 0.003 rem. Calculations for the maximum fission rate were performed for accidental and planned releases of fission gases and halogens from the irradiation of U-235 and Pu-239. U-235 and Pu-239 were the limiting fissionable materials of those being requested for use in fueled experiments due to higher fission cross-sections and mass.

Threshold limits on fission rate and total number of fissions were established for experiments containing uranium which are based on limiting the radiation dose from a potential release to one percent (1%) of the annual public dose limit given in 10 CFR Part 20, i.e. a TEDE of 0.001 rem. The TEDE of 0.001 rem was calculated from exposure to a continuous release of fission gases and halogens with the reactor building in normal ventilation for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The resulting threshold limits are used in TS 1.2.9.e to define a fueled experiment using uranium. Experiments involving the neutron irradiation of uranium below these threshold limits are not classified as fueled experiments. Experiments involving the neutron irradiation of uranium in excess of these limits, the neutron irradiation of any amount of another fissionable material, or involving a planned vented release of fission gases or halogens are defined as fueled experiments.

The calculations performed result in a radiation dose per unit fission rate. The limiting fission rate is then determined using the criteria of not exceeding three percent of the 10 CFR 20 annual radiation dose limits.

The maximum duration of the experiment and the limiting fission rate were used to define the activity of longer-lived radionuclides produced. These longer-lived radionuclides are compared to Quantities of Concern given in 10 CFR Part 37.

Sections of the analysis are listed below:

1. Assumptions

2. Saturation activity

3. Exposure time

4. Filter retention

5. Released activity

6. Atmospheric dispersion

7. Time integrated exposure

8. Dose assessment

9. Experiment limits

10. Fueled experiment definition 1

11. Emergency Plan

12. Security, storage, and inventory

13. Possession limits

14. Calculation results

15. Conclusions

16. References Sections 1 through 13 provide information and equations used in the calculations.

Section 14 provides the calculation results and supporting data.

Section 15 provides conclusions used to support the requested changes to TS 3.8, TS3.5, TS 4.4, and TS 1.2.9. e.

SECTION 1: ASSUMPTIONS 1.1 Fueled Experiments 1.1A Conditions for a planned vented release from a fueled experiment:

1. Fission gases and halogens are released. Release of particulate, powder, liquid, and solid material is prevented by design of the experiment. .

2. A continuous, controlled release from the reactor building ventilation system during the experiment irradiation time is assumed. A minimum decay time of 30 minutes and maximum release rate of 3 liters per minute (lpm) is assumed.

3. Release occurs during the entire irradiation time.

4. Assumptions 7, 8, 9 and 16 below apply to vented experiments.

1.1B Conditions for accidental release from fueled experiments:

5. Radioactive materials are encapsulated until the time of failure.

6. Single-mode nonviolent failure of the encapsulation results in the release of radioactive noble gases and halogens into the minimum reactor building free air volume.

7. Neutron fluence rate is constant over the irradiation time and over the entire mass of the fissionable material present during the experiment irradiation time. No correction to the mass is made because of activation and fission reactions during the irradiation time. No correction is made to the fluence rate because of self-absorption by the mass of fissionable material or encapsulation materials.

8. Reactor ventilation system is in the normal mode until being activated by a radiation alarm from an abnormal release, which then places the ventilation system in confinement mode.

9. Radioactive noble gases and halogens are assumed to be present at the saturation activity from irradiation at the maximum fluence rate in the reactor experimental facilities.

10. Exposure times to personnel in the reactor building is a total of six minutes based on a radiation monitor response time of four minutes and an evacuation time of two minutes from the reactor building.

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11. Exposure times to the public are 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Evacuation of public areas occurs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. All released activity is removed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

12. The release is assumed to occur instantaneously and to be well mixed within the reactor building free air volume for accidental releases.

13. A correction factor of 0.1 is used for submersion dose within the reactor building for photons emitted by noble gases based on dimensions and geometry. A sphere rather than hemisphere is assumed.

14. The minimum reactor building free air volume is assumed to be 2.25x103 m3 based on reported and measured data and current design features given in TS 5.2a.

15. Confinement filter removal efficiency, or retention, is 99.97 percent for particulates and 99 percent for halogens. Retention for halogens is assumed to be 90 percent.

16. Atmospheric dispersion parameter, [X/Q] is calculated using established equations, data, and parameters given in the references. The Gaussian Plume Model (GPM) was used for all releases and release periods. Fumigation conditions were assumed to last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The GPM was modified for calm winds. Calm winds were assumed to last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1.2 Non-Fueled Experiments Conditions for irradiation of uranium at or below a TEDE of 0.001 rem:

17. The release of fission gases and halogens is assumed to be accidently released to the reactor building free air volume continuously for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using normal ventilation with no filtration.

SECTION 2: SATURATION ACTIVITY [Ref 12, 25, Section 14 Calculations 1, 2, 3, 4]

A sufficiently long production period is assumed to reach saturation activity of the fission gases and halogens.

These include isotopes of Kr, Xe, I, and Br. Saturation activity was calculated using experimental facility fluence rates, reported cross-sections, and reported cumulative fission yields.

Fission product inventory for the radionuclides available for release attains saturation activity with sufficient irradiation time. Saturation activity is estimated using Equation 2-1:

A() = kNY Eq. 2-1 where, A() is the saturation activity from thermal and non-thermal fission k is a group conversion constant to give activity k = (1x10-24 cm2/barn)(1 decay/atom)(1Ci /3.7x1010 dps) = 2.703x10-23 for Ci or k = 2.703x10-29 for µCi is the fission reaction cross section in barns is the neutron fluence rate in cm-2s-1 N is the number of atoms for the fissionable material present N = (Mass in grams, M)( 6.022x1023 atoms/mole)(1 mole / atomic mass number, A)

Y is the cumulative fission yield for a given radionuclide 3

Saturation activity is directly proportional to the fluence rate and mass of fissionable material. The fission rate, or production rate, for a given radionuclide is given by the product NY.

Fluence rate [Ref 5, 28]

Neutron fluence rates in experimental facilities are measured by the reactor staff following standard ASTM E261 Standard Methods for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques using NIST traceable materials. These measurements are made periodically, as experimental facilities change, or for specific experimental needs.

Release of halogens (I and Br fission products) is greater if no water is present since water partially retains halogens. A greater release would therefore occur for an irradiation facility outside the reactor pool, which are performed using the reactor beam tubes. Release of fission gases is not affected by the presence of water.

Therefore, the maximum fluence rates measured at 1 MW operation for a reactor beam tube experiment were used in this analysis to determine the saturation activity of the fission gases and halogens that are assumed to be released. At 2 MW, the fluence rates would be twice those at 1 MW and the time to reach saturation is halved.

The following fluence rates were used:

• Thermal neutron fluence rate of 1x1012 cm-2s-1

• Non-thermal fluence rate of 3x1011 cm-2s-1 Decay data [Ref 14]

Decay data, e.g. half-lives, were taken from data given in Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Joint Evaluated Fission and Fusion Project Report 20 (JEFF 3.1-3.1.1 Radioactive Decay Data and Fission Yield Sub-Library).

Cumulative fission yield data [Ref 15, 17]

Cumulative fission yields were taken from the following references:

• Evaluation and Compilation of Fission Product Yields, T.R. England and B.F. Rider, Los Alamos National Laboratory, October 1994, LA-UR 94-3106 ENDF 349

• Japan Atomic Energy Agency Nuclear Data Center Tables of Nuclear Data (JENDL data).

Fission cross-section data [Ref 16, 17, 18]

Fission cross-section data for thermal and non-thermal neutron energies used in this analysis:

• 0.025 eV for thermal neutron energy

• For non-thermal energies, the higher of the following was used

- Average from 1 x10-5 eV to 10 eV

- Resonance integral from 0.5 to 1 x105eV References for fission cross-section data used in this analysis:

• National Nuclear Data Center, Brookhaven National Laboratory, Evaluated Nuclear Data Files (ENDF libraries)

• OECD NEA Joint Evaluated Fission and Fusion Project Report 21

• Japan Atomic Energy Agency Nuclear Data Center Tables of Nuclear Data (JENDL data) 4

SECTION 3: EXPOSURE TIME [Ref 6,7, 22, 23]

Reactor building personnel Evacuation time measured from various locations inside the reactor building to the evacuation exit point for several individuals ranged from 10 seconds to 1 minute or less following initiation of the reactor building evacuation signal. Evacuation followed the facility emergency plan and procedures. Also, evacuation times were calculated for an average walking pace of 3 mph for the furthest distance in the reactor building to the assembly point outside the reactor building. Both the measured time and estimated walking time gave a time of 1 minute. A travel time of 2 minutes is assumed for conservatism.

Time for released activity to reach the ventilation system and be detected by the ventilation system radiation monitors is less than 2 minutes. This is based on a ventilation system flow rate of greater than 60 feet per minute and duct length of 100 feet and detector response time of 0.5 minutes. A response time of 4 minutes is assumed for conservatism.

For personnel in the reactor building, an exposure time of 6 minutes (360 s) is used based on the time needed for detector action to activate the building evacuation alarm (4 minutes) and for personnel to physically exit the reactor building (up to 2 minutes).

Public For the public, exposure times of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> were used for accidental releases:

• 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allows sufficient time for detection and response by facility personnel to determine affected public areas that need to be evacuated.

• 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is sufficient time for the entire released activity to be vented from the reactor building (more than 10 air changes). A public exposure time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is also associated with meeting emergency action levels given in the facility emergency plan, which would not be exceeded. After 24 h the reactor building has experienced over 10 air changes leaving a negligible fraction of the initial concentration, (e-10). After 24 h the accidental release is completely ventilated.

For planned vented releases, exposure times used for the public were 2, 24, 96, and 520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />. 520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> is the maximum irradiation time used to limit the production of long lived radionuclides.

The exposure is presumed to last during the irradiation time. Since irradiation times may be divided, full occupancy by the public during the irradiation time is assumed.

SECTION 4: AIRBORNE ACTIVITY CONTROLS AND FILTRATION [Ref 4, 11]

Airborne activity is controlled by sample e controls, radiation monitoring, and filtration.

Sample Controls Samples from fueled experiments that are not vented are controlled by encapsulation that meets TS 3.7.

Samples from vented fueled experiments are contained to prevent the release of sample material and allow the release fission gases and halogen vapors. Samples for vented experiments are contained using a sample holder, which is placed in the experimental facility. Sample holders as shown in Figure 4.1 may use a capsule or tubing with a restricted orifice (ring), mesh screen covering with smaller dimensions than the solid material, or filters to contain the sample.

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FIGURE 4-1: EXAMPLE VENTED FUELED EXPERIMENT SAMPLE HOLDERS Sample Capsule With Mesh Filter 6

Confinement System Initiation If an alarm setpoint of the airborne activity monitors in the radiation monitoring system or experiment exhaust from a vented fueled experiment is reached, then the confinement system is initiated. The setpoints for these airborne activity monitors are based on the released activity from a vented fueled experiment or an experiment accident.

High Efficiency Particulate Absorbers (HEPA) and charcoal beds are used in the confinement mode of ventilation.

Confinement filter testing Testing is performed per TS 4.5 on the ventilation system, including filter testing in accordance with TS 4.5.e every 2 years but not to exceed 30 months. Maintenance and surveillance procedures are in place for testing of the ventilation system. Testing methods follow ASME N510-1989 "Testing of Nuclear Air Treatment Systems. Testing is also required following major maintenance of the filters or housing. Testing and maintenance are documented in facility surveillance files as required by TS 6.4 and 6.8.

Acceptance criteria are retention of 0.9997 for HEPA tested with 0.3 micron aerosols and 0.99 for charcoal tested with Freon R-11. Charcoal filters are tested by the vendor prior to installation in the confinement system and have a reported retention of 0.99 for methyl iodine. A filter retention factor of 0.9 is used in this analysis for halogens.

Vented fueled experiment filters and certification The exhaust from a vented fueled experiment is filtered for halogens and particulates prior to entering the beam tube exhaust, which is connected to the reactor building ventilation system. A generic experiment arrangement is shown in Figure 4-2.

Commercially available HEPA filters for removal of particulates and carbon adsorber beds for iodine removal and delay of noble gases are available; e.g. activated carbon and coconut shell carbon.

HEPA and carbon filters for removal of particulates and halogens are located in the experiment exhaust prior to the delay volume. A sealed housing is used to contain the HEPA and carbon filters. HEPA filter and carbon adsorber shall have a tested retention of 0.95 or greater at flow rates of 3 lpm or less. Filter certification will be reviewed and retained. Carbon adsorbers for iodine removal are tested by the vendor following an applicable standard; e.g. ASTM D 3803 Standard Test Method for Nuclear-Grade Activated Carbon.

Certification of iodine adsorption by the supplier and adsorbent filter replacement has been added to TS 4.4 for the vented fueled experiment. The replacement time of 2 years, up to 30 months, is consistent with TS 4.5 and based on a shelf life of up to 5 years and noting that the exposure and operating characteristics to the confinement filters is similar. To meet LCOs for operation and surveillances, the confinement filters are continuously available for use and operated for a few minutes every week. The relative humidity and temperature of the vented experiment exhaust is similar to that for occupied areas in the reactor building.

Additionally, air sampling filters may be used in-line with the experiment exhaust for removal of particulates and halogens. Air sampling halogen filters are tested by the supplier following ASTM D 3803 with a reported retention factor greater than 95 percent for methyl iodine at flow rates of 3 lpm or less. Particulate air sampling filters shall have a reported retention of 95 percent or higher. The experiment air sampling filter housing will be sealed.

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FIGURE 4-2: GENERIC DIAGRAM OF A VENTED FUELED EXPERIMENT Vented Fueled Experiment Equipment Description of vented fueled experiment components:

• A pressurized inert gas is controlled by a regulator to provide a flow rate up to 3 lpm.

• The experiment container allows only fission gases and halogens to be released into the experiment exhaust.

• 30 minute delay volume. The delay volume may consist of a long length of coiling tubing, several small tanks in series, a charcoal bed for delaying noble gas release, or a well-mixed large tank.

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• Filters include particulate and charcoal adsorbers in a sealed housing located before and after the delay volume. Redundant filter trains may be used to allow for decay prior to changeout or maintenance and to provide a full capacity back up set of filters should a filter fail.

• Flow meter with a set point at 3 lpm monitors the experiment exhaust flow rate. TS 3.5 is revised to include the vented fueled experiment flow rate monitor.

• Experiment exhaust radiation monitor with a set point at the TS limit. Fission gases are monitored.

TS 3.5 is revised to include the fueled experiment exhaust monitor.

• Gas container with a known volume will be used with the experiment exhaust radiation monitor.

• Experiment exhaust is connected to the Beam Tube (BT) exhaust.

• The BT exhaust connection is located upstream of the confinement fan duct.

• Isolation valves are located at inert gas exit (experiment entry), experiment container exit (delay volume entry), and BT duct connection.

Due to the possibility of radioiodine being present in airborne effluent, initiation of the confinement system occurs for vented fueled experiments as follows:

• Vented fueled experiment exhaust gas radiation monitor or stack gas radiation monitor exceeding an alarm set point.

Alarm setpoints are based on the concentration of fission gases released from a vented fueled experiment at the TS 3.8 limiting fission rate of 9.6 E9 f/s. Monitoring of the vented fueled experiment exhaust fission gases provides early detection of abnormal airborne releases since the fission gases are at higher concentrations and not diluted.

• Stack particulate radiation monitoring channel exceeding an alarm set point.

The stack particulate monitor be equipped with a particulate filter to detect decay products of fission gases (Rb-88, Rb-89, Cs-138) and a radioiodine cartridge (e.g. TEDA charcoal or silver zeolite) to detect radioiodine (I-131, I-132, I-133, I-134, and I-135).

• If the activity released from a vented fueled experiment is high and the vented fueled experiment exhaust radiation monitor fails or a filter fails, then the radiation monitors sampling the stack exhaust (stack gas or stack particulate or stack exhaust) will alarm to initiate the confinement system.

• Upon receipt of a radiation alarm, the normal fans stop and one of the confinement fans starts. If the first confinement fan fails to start, the second confinement fan will start.

Leakage from the vented fueled experiment filters is unlikely due to the low flow rate of 3 lpm, in-line filter housing for the experiment exhaust, and sealing of the filter holder and connections. Leak testing is performed using facility procedures after new filters are installed and prior to initial use; e.g. by pressure testing. Surveillance for in-place testing on the vented fueled experiment filters is not provided. Should the experiment filters fail, radiation monitors would detect abnormal radioactivity, which then would isolate the experiment exhaust and initiate the confinement system.

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SECTION 5: RELEASED ACTIVITY [Ref 1, 3, Section 14 Calculations 1, 2, 3, 4]

The reactor building ventilation system is operated in the normal mode for fueled experiments since these experiments may last for an extended time. Releases are monitored by the radiation monitoring system (RMS). If abnormal levels are detected by the RMS, then the reactor building ventilation system initiates an evacuation alarm and switches to confinement mode. In confinement the exhaust is filtered using a charcoal bed and particulate absorber prior to release to the environment. Operating in normal mode maintains the confinement filters as an engineered safety feature.

Reactor building free air volume [Ref 4, 11]

Measurements of the reactor building experiment area were made and give a total volume of 3.43x103 m3. The free volume was measured to be 3.09x103 m3 by accounting for existing equipment and experiments. The measured free volumes are greater than the FSAR value of 2.4x103 m3 and TS value of 2.25x103 m3. Additional equipment, modifications, or experiments in the reactor building significantly affecting free air volume are not expected.

Therefore, the value of 2.25x103 m3 was used in this analysis. A lower free air volume is conservative since it increases the concentration and radiation dose.

Accidental releases For accidental releases, the radioactive material inventory of fission gases and halogens are assumed to be completely released and then instantaneously and uniformly distributed throughout the entire reactor building free air space. The released materials are then exhausted by the ventilation system and reactor stack to the environment. The concentration inside the reactor building, decay inside the reactor building, filter retention, exhaust ventilation rate, and atmospheric dispersion are considered in the analysis.

Concentration from an accidental release The sample is assumed to contain saturated activities of radioactive fission gases (Kr, Xe) and halogens (I, Br) at the time of encapsulation failure. The entire fission gas and halogen radioactivity is assumed to be instantaneously released and uniformly mixed into the minimum free reactor bay volume resulting in uniform airborne activity distribution throughout the entire reactor bay.

Initially, the release occurs for 2 minutes in normal ventilation with no filtration. Following this, the elevated concentrations of fission gases and halogens from an accidental release automatically initiate the evacuation and confinement system by the radiation monitoring system. The release is then filtered by the confinement filters for the remaining duration of the release.

The initial released concentration, C(0), in the reactor building is given by equation 5-1:

C(0) = A() / V Eq. 5-1 where, C(0) in Ci/m3 and A() in Ci V is the minimum reactor building experiment area free air volume = 2.25x103 m3 10

Average concentration for an accidental release inside the reactor building Over time, the initial concentration, C(0), is removed by decay and the ventilation system. Due to the high initial concentration, the ventilation system would be in confinement mode.

Average concentration, < C >, for exposure time T is given by:

< C > = C(0) e-ktdt = C(0) [ (1- e-kT) / (kT)] Eq. 5-2 where, k = + v in h -1 v = 0.453 h-1 at 600 cfm exhaust rate in confinement 0.453 h-1 = (28,317ml per cubic foot)(600 ml/ min) (60 min/h) / 2.25x109 ml v =1.41 h-1 at 1870 cfm in normal ventilation 1.41 h-1 = (28,317 ml per cubic foot)(1870 ml/ min) (60 min/h) / 2.25 x109 ml t is exposure time, with limits of integration from 0 to T, in hours Exposure times used: T is 0.066 h (4 minutes) for the exposure time in normal ventilation T is 0.033 h (2 minutes) for the evacuation time in confinement T is 2 h and 24 h for total public exposure time in confinement Release rate at the reactor stack from an accidental release The filtered release rate at the reactor stack, Q, is calculated as follows:

Q = C [1-R] F Eq. 5-3 where, Q is the release rate in Ci/s C is concentration in Ci/m3 , either <C> or C(0)

R is filter retention R = 0.9 for halogens and R = 0 for noble gases, R = 0 in normal ventilation F is the stack exhaust in m3/s F = 0.283 m3/s in confinement mode and 0.883 m3/s in normal mode Vented fueled experiments A continuous, controlled release during the experiment irradiation time occurs for vented fueled experiments.

(Refer to Figure 4-2).

Controls include:

• Filtration of particulates and halogens. Filters with rated retention greater than 95 percent for particulates and halogens are to be used. Retention for halogens is assumed to be 90 percent.

• Vented experiment has a minimum decay time of 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (30 minutes). The maximum experiment exhaust rate is 5.0x10-5 m3/s (3 liters per minute). At the maximum experiment exhaust rate a minimum experiment holdup volume of 0.09 m3 (90 liters) in needed to give a 30 minute decay time.

In addition, the experiment volume is designed so that activity is well mixed prior to release into the reactor building ventilation system; e.g. using a long tube or coil, a series of small air tanks, a low flow rate relative to the volume, well separated entry and exit flow ports, and baffles or diffusers within the experiment holdup volume.

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• Experiment exhaust flow is routed to the ventilation system using the beam tube exhaust and controlled by dedicated equipment with local flow rate indication. The experiment exhaust is capable of being isolated. The exhaust flow tubing from the experiment to the beam tube exhaust is sealed to prevent leakage into the reactor building free air space.

• Radiation monitoring and flow rate monitoring of the experiment exhaust prior to being routed to the reactor building ventilation system is required to identify and quantify the source of the release. The release is monitored for radioactivity with indication locally and in the control room. Local alarm annunciation and alarm indication in the control room is provided.

A sudden and significant release from a vented experiment, e.g. rupture of the holdup tank, would be an accidental release as previously described. Any release of exhausted air from the vented experiment into the reactor building would be diluted by the reactor building free air volume. Radiation monitoring of the reactor building air volume and exhausted air are performed continuously by the radiation monitoring system.

Concentration from a vented experiment at the experiment exhaust The saturation activity A() is assumed to be dispersed and held in the volume of a vented experiment, w, for decay giving the decayed, unfiltered saturation concentration, c():

c() = [A() / w] exp(-t) Eq. 5-4 where, c() in Ci/m3 A() is in Ci w is the experiment hold up volume; minimum value is 0.09 m3 (90 liters)

Decay time, t, is 30 minutes (30 minutes = 90 liters / 3 lpm)

Concentration from a vented experiment at the reactor stack c() is diluted by the reactor building ventilation system exhaust operating in the normal mode at the reactor stack. Concentration at the reactor stack, C, is given by:

C = c() [p / (p+F)] Eq. 5-5 substituting Eq. 5-4 into Eq. 5-5 gives, C = [A () / w] exp(-t)[p / (p+F)] Eq. 5-6 where, C and c() in Ci/m3 F is the normal ventilation flow rate of 0.883 m3/s (or 1870 cfm) p is the experiment exhaust rate and shall not exceed 5.0x10-5 m3/s (3 lpm)

[p/(p+F)] accounts for the dilution by the normal exhaust from the reactor building 12

Release rate for vented experiment The experiment exhaust is filtered and routed to the beam tube exhaust. The beam tube exhaust is connected to the reactor building ventilation system exhaust in the upper part of the reactor building.

The filtered release rate, q, from a vented experiment entering the beam tube exhaust is given by:

q = c() [1-R] p Eq. 5-7 substituting Eq. 5-4 into Eq. 5-7gives, q = [A () / w] exp(-t)[1-R] p; and q = Q Eq. 5-8 where, q is the decayed, filtered release rate in Ci/s that enters the beam tube exhaust p is the experiment exhaust rate in ml/s; maximum value is 5.0x10-5 m3/s (3 lpm) c() in Ci/m3 is the radioactive decay constant in 1/s Filter retention; R = 0.9 for halogens and R = 0 for noble gases Q is the decayed, filtered release at the reactor stack NOTES:

• Since the flow rate is constant into and out of the experiment delay volume, w, the net loss while in the holdup volume is due to radioactive decay.

• The decayed and filtered release rate at the reactor stack, Q, is the same as the decayed and filtered experiment release rate, q, since c() is diluted and exhausted by the same flow rate, F; i.e. Q = q.

Vented fueled experiment radiation and flow monitoring The vented fueled experiment exhaust gas is monitored for flow rate and radioactivity. Both have setpoints to keep public dose below 3 mrem. The experiment exhaust flow rate is limited to 3 lpm. Experiment exhaust gas radiation monitor setpoints are provided in Section 14 Calculation 10. Exceeding the setpoints of these monitors isolates the vented experiment exhaust and initiates the confinement system.

A vented fueled experiment exhaust gas radiation monitor is used for measuring radioactivity in the exhaust gas from a vented fueled experiment. Kr and Xe fission gases are readily released and therefore are monitored to provide an immediate assessment of the released activity. Setpoints of the experiment exhaust gas radiation monitor allow a planned, controlled release to occur without alarms If the flow rate or radiation levels are abnormal, the experiment exhaust is isolated and confinement system is initiated since the fission rate limit may be exceeded or an abnormal release may be occurring.

The stack gas and stack particulate radiation monitors are used to assess airborne effluent. The stack particulate monitor is equipped with a radioiodine cartridge during fueled experiments as required by the revised TS 3.8. The stack particulate monitor will detect abnormal radioiodine activity being released.

Setpoints for the stack gas and stack particulate monitors were adjusted so that allowed releases from a vented fueled experiment would not cause an alert (warn) or alarm signal. Released activity at the stack gas or stack particulate that are in excess of the fueled experiment limits or from an accident will cause an alarm and initiate the confinement system.

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It is noted that the radiation and flow rate monitor response is immediate relative to the exposure times used for calculating radiation dose to members of the public; e.g. 2 minutes for initiating the confinement system and stopping the experiment compared to an exposure time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> used in the dose calculations.

Therefore, the alarm setpoints occur well before a public dose of 3 mrem is reached.

The revised TS 4.4 includes calibration of the vented fueled experiment exhaust radiation monitor and flow rate meter. Annual calibration is required by TS and includes a channel calibration of the vented fueled experiment flow rate monitor to test isolation of the experiment exhaust, initiation of the confinement system, and Control Room annunciators.

Halogens are filtered and sampled at the experiment exhaust. Analysis of these samples is performed upon filter removal as required by facility health physics procedures.

SECTION 6: ATMOSPHERIC DISPERSION [Ref 1, 12, 19, 20, 27, Section 14 Calculation 5]

Atmospheric dispersion calculation methodology in the 2017 SAR and this amendment use the Gaussian Plume Model (GPM) at distances from 30 m to 5000 m for all exposure times. In addition, this analysis considered fumigation (i.e. trapping) conditions caused by an inversion and calm winds for periods up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Gaussian Plume Model (GPM)

The atmospheric dispersion parameter [X/Q] used in this amendment are for actual occupied locations.

The 2017 SAR used the maximum calculated atmospheric dispersion {X/Q] value at a given distance regardless of the ability for occupancy.

The GPM equation was used to calculate the atmospheric dispersion parameter [X/Q] for Pasquill-Gifford (PG) weather stability classes A through F:

y2 ( z h )2 ( z +h )2 1

2 y2

[

2 z2

] [

2 z2

]

[ X / Q] x , y , z = [e ][e +e ] Eq. 6-1 2 y z u where, [X/Q]x.y.z is the atmospheric dispersion parameter for downwind location (x,y,z) in s/m3

[X/Q] is the downwind concentration per unit release rate; X is in Ci/m3 and Q is in Ci/s x is the downwind distance from the stack to receptor in m y is the lateral distance from the plume centerline in m z is the receptor elevation in m y is the lateral dispersion parameter in m for PG weather stability classes z is the vertical dispersion parameter in m for PG weather stability classes h is the physical stack height in m, or 30 m is wind speed in m/s z and h are relative to the ground elevation of 0 m Dispersion parameters Pasquill-Gifford (PG) weather stability classes A through F are used for [X/Q] in the GPM and are characterized by the following:

• y is the lateral dispersion parameter in m

• z is the vertical dispersion parameter in m 14

Dispersion parameters y and z were calculated using fitting data from NUREG 1887 RASCAL 3.0.5:

Description of Models and Methods for downwind distances from 10 m to 5000 m. These calculated dispersion parameters for weather stability classes A through F were used in the [X/Q] equations.

Decay Corrections No decay corrections are made during transport by the atmosphere following the release or decay post-production prior to release since a failure may occur anytime during the experiment.

Stack Height ANSI/ANS-15.7 and US NRC Regulatory Guide 1.111 were used to calculate effective stack heights. From these calculations, the effective stack height was calculated to be only slightly greater than the physical stack height. For simplicity, the actual stack height of 30 m is used in atmospheric dispersion parameter [X/Q]

calculations.

Release time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or less For a release of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or less it is assumed that the weather stability class, wind speed, and wind direction remain constant. Assumptions made are as follows:

• The assumed wind speed () from is 1 m/s

• The most restrictive weather stability class for the given location is used

• The receptor is assumed to be on the plume centerline, i.e. y = 0 m With the noted assumptions, [X/Q] equation 6-1 becomes:

1

( z h )2 ( z + h )2

[ ] [ ]

[ X / Q ]x , y , z = [e 2 z2

+e 2 z2

] Eq. 6-2 2 y z The plume centerline equation above accounts for a receptor location at any elevation (z) relative to the ground level.

Release time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or longer Sector averaging applies if the wind direction deviates sufficiently across the sector width over time, i.e. a meandering plume over the lateral y dimension. Sector averaging is considered valid at downwind distances (x) if x/n > 2y and for periods greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

On inspection, for the reactor facility stack height where the relationship x/n> 2y is valid, the minimum distances are applicable for sector averaging for PG weather stability classes A through F as given in Table 6-1.

Table 6-1: Stability Classes Stability Class Minimum Distance (m)

A >50,000 B 25,000 C 2,500 D, E, F 100 15

The sector average model is as follows for any receptor elevation (z):

( z h )2 ( z +h )2 n f [ ] [ ]

[ X / Q]x , y , z = 2/ [e 2 z2

+e 2 z2

] Eq. 6-3 2x 2 z u where, the sector average [X/Q] is [ X / Q]

f is the frequency fraction for wind direction and wind speed Release time from 2 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The PG weather stability class frequency, wind direction frequency (f), and wind speed () remain constant.

The most restrictive PG weather stability class was used for a given downwind location (x,y,z). From ANSI/ANS-15.7, f is set at 1 and is 1 m/s.

If sector averaging is not valid, the [X/Q] equation for the GPM was used for all elevations (z):

1

( z h )2 ( z +h )2

[ ] [ ]

[ X / Q]x , y , z = [e 2 z2

+e 2 z2

] Eq. 6-4 2 y z If valid, the sector average [X/Q] equation was used for all elevations (z). Re-writing with the noted assumptions for f and gives equation 6-5:

( z h )2 ( z +h )2 2.032 [ ] [ ]

[ X / Q] x , y , z = [e 2 z2

+e 2 z2

] Eq. 6-5 2 z x where, 2.032 = (16 / 2) [2 / ]1/2 for n =16 The following simplifications to [X/Q] GPM calculations are made regarding releases from 2 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

• Stability classes A, B, and C were not sector averaged at any distance greater than 100 m for conservatism.

• Stability classes D, E, and F were sector averaged at distances greater than 100 m.

Fumigation (trapping during an inversion)

[X/Q] for fumigation conditions for the plume centerline (y = 0 m) were calculated at a wind speed of 1 m/s for periods up to 24 h using equation 6-6:

1

[ X / Q] = Eq. 6-6 2 hµ y where, h is the physical stack height of 30 m and replaces z The equation for fumigation was taken from Refs 19, 33, and 34. In fumigation conditions, the vertical dispersion is uniform from ground level to the stack height.

16

Inversions are associated with fumigation (i.e. trapping) conditions. Inversion frequency is given in 2017 SAR Table 2-18 as ranging from 32 to 43 percent in Greensboro, NC. Inversion duration is not given in the 2017 SAR. Fumigation in this analysis used a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a wind speed of 1 m/s [Ref 12,34].

Ground based inversions occur by rapid cooling of the ground on cloudless nights with light winds. With warming of the ground during the day by the sun, the inversion ends. It is noted that the reactor building is located in an area with significant paved surface area and that the reactor typically operates during daytime hours [Ref 12].

Inversions may also occur during periods of air stagnation. In North Carolina, approximately 15 air stagnation days per year are reported [Ref 35]. Air stagnation is defined as a mean wind speed of 4 m/s and a period of 4 days or more. This gives an annual frequency of less than five percent (15/365) for stagnant air.

It is noted that 4 m/s gives fumigation [X/Q] at one-fourth of those calculated in the 2017 SAR and that the duration of 4 days is 4 times of that used in this analysis. It is also noted that the average weather conditions used to calculate [X/Q] in the 2017 SAR Table 11-10 and Table 6-2 for periods up to 4 days were at wind speeds less than 4 m/s.

Maximum fumigation [X/Q] values are used to assess potential radiation dose in occupied locations near the reactor facility.

Calm winds Calm winds were assumed to exist for periods up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Calm winds have reported wind speeds less than 0.5 m/s. In calm winds the straight-line Gaussian plume model is not applicable and becomes undefined if the wind speed becomes zero. Ref 20 gives a model for calm winds that uses horizontal and vertical turbulence velocities (m/s) rather than normal dispersion parameters. For calm winds default turbulence velocities () of 0.13 m/s are used for the wind, cross wind, and vertical turbulence.

For the plume centerline (y = 0 m), [X/Q] for calm winds was calculated using equation 6-7:

1

[ X / Q] = Eq. 6-7 (2 ) ( x + h2) 3/ 2 2 A review of weather patterns is given in Section 2 of the 2017 SAR. Wind speed data for Jordan Hall at a height of 30 m on the university campus indicates that light winds (from 0 to2 m/s) occur approximately three percent of the time. Periods of calm winds (0 to 0.5 m/s) for Jordan Hall would occur less frequently.

Results for releases up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> From the above discussion, periods of calm winds and fumigation are considered to be infrequent and to exist for periods up to 24 h, In this amendment [X/Q] was calculated for calm winds at a wind speed of 0.5 m/s, fumigation at a wind speed of 1 m/s, and the GPM at a wind speed of 1 m/s for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Also, emergency action levels are associated with a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> regarding airborne radioactive effluent.

[X/Q] Dose calculations made for calm winds and fumigation for actual occupied locations are more conservative than those made using the GPM.

17

From Table 14-10 for periods of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less, the maximum [X/Q] values for occupied location from:

• Fumigation exceed those for the GPM for all distances

• Calm winds exceed those for the GPM for distances from 30 m to 50 m Atmospheric dispersion for actual occupied locations under calm winds and fumigation conditions as described in this analysis will be added to Section 11 of the 2017 SAR for license renewal.

Release times greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Release times greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are associated with vented experiments. Adjustments to the [X/Q]

calculations were made as given in ANSI/ANS-15.7 for times from 1 to 4 days and greater than 4 days for the PG stability class frequency (S), wind direction (f), and wind speed (u).

The product [ S f /u] is multiplied to the [X/Q] and [ X / Q] equations given above evaluated at a wind speed of 1 m/s and then summed for all PG stability classes to give the adjusted [X/Q] value. A summary of the

[X/Q] equations and adjustments are given in Table 6-2 below. Refer to Section 14 Calculation 5 in this analysis for the [X/Q] values that were calculated.

Table 6-2: Summary of the [X/Q] Equations and Adjustments PG Stability PG Stability Wind Wind speed Lateral Direction Duration Class Frequency, S Direction, f (m/s), u (y in m) 2h A through F 1 1 1 0, centerline 2 h to 24 h A, B, C 1 1 1 0, centerline 2 h to 24 h D, E, F 1 1 1 Sector Averaged 24 h Fumigation 1 1 0, centerline 24 h Calm wind 1 0.5 0, centerline 1 to 4 days D 0.4 1 3 Sector Averaged F 0.6 1 2 Sector Averaged

> 4 days C 0.333 0.15 3 0, centerline D 0.333 0.15 2 Sector Averaged F 0.333 0.15 2 Sector Averaged The following maximum [X/Q] values were used in this analysis to calculate time integrated exposures in occupied public areas at and beyond the site boundary:

• 8.54x10-3 s/m3 for a release times of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

• 7.79x10-4 s/m3 for a release time from 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />

• 9.15x10-5 s/m3 for a release time from 96 to 520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> In addition, maximum [X/Q] values for specific locations of interest were used to calculate the time integrated exposure.

[X/Q] using GPM equations given in Section 11 of the 2017 SAR and this analysis for occupied locations give the same values. For example:

• Table 14-10 and 2017 SAR Figure 11-2 for (x,z) location (200m, 10m vs 12m);

Talley and Reynolds for 2-24 hours: [X/Q] = 1.93x10-4 s/m3

• Table 14-10 and 2017 SAR Figure 11-3 for (x,z) location (350m, 20m);

North for 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />s: [X/Q] = 6.00x10-5 s/m3

• Table 14-10 and 2017 SAR Figure 11-5 for (x,z) location (150m, 30m);

DH Hill for greater than 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />s: [X/Q] = 9.15x10-5 s/m3 18

SECTION 7: TIME INTEGRATED EXPOSURE [Ref Section 14 Calculations 1, 2, 3, 4]

7.1 Fueled Experiments Time integrated exposures inside the reactor building from accidental release Time integrated exposures and Dose Conversion Factors (DCF) are used to calculate radiation dose. Time integrated exposures are given by the product of the average concentration over the exposure time and the exposure time.

Accidental releases initially occur with the reactor building in normal ventilation and then after 2 minutes the RMS or Reactor Operator activate the evacuation alarm and confinement ventilation.

The time-integrated exposure with removal by radioactive decay and ventilation system inside the reactor building from an accidental release was calculated as follows for exposure time, T:

r = <C>T Eq. 7-1 where, r is the time integrated exposure in µCi-h/ml

<C> in µCi/ml or Ci/m3; conversions are 1 µCi/ml = 1 µCi/cm3 = 1 Ci/m3 T is 0.066 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br /> inside the reactor building in normal ventilation T is the evacuation time of 0.033 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> inside the reactor building in confinement Calculated r for each ventilation mode and exposure time are then summed for the total r.

Time integrated exposures outside the reactor building from an accidental release Accidental releases initially occur with the reactor building in normal ventilation and switches to confinement after 0.066 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br /> (4 minutes).

Time-integrated exposure in public areas is reduced by removal of halogens and particulates by the confinement filters and by atmospheric dispersion. Time-integrated exposure outside the reactor building was calculated as follows for each exposure time, T:

p = <C>[1-R][X/Q]FT Eq. 7-2 Alternately, p = Q [X/Q] T Eq. 7-3 where, p is the time integrated exposure in µCi-h/ml for members of the public in µCi-h/ml

<C> in µCi/ml or Ci/m3 R = 0.9 for halogens and R = 0 for noble gases, R = 0 in normal ventilation F is the volumetric stack exhaust rate of 0.883 m3/s in normal ventilation and 0.283 m3/s in confinement

[X/Q] is the atmospheric dispersion parameter in s/m3 T is 0.066 h in normal ventilation T is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in confinement Q is Ci/s Conversions are 1 µCi/ml = 1 µCi/cm3 = 1 Ci/m3 Calculated p for each ventilation mode and exposure time are then summed for the total p.

19

Time integrated exposures outside the reactor building from vented experiments For vented experiments, the release activity is constant and continuous over the exposure time. The release is routed directly to the ventilation system, thereby not exposing occupants inside the reactor building to airborne activity. Time-integrated exposure to members of the public is given by the following equation:

p = q [X/Q] T = Q [X/Q] T Eq. 7-4 where, p is the time integrated exposure in µCi-h/ml for members of the public in µCi-h/ml Conversion constants: 1x10-6 Ci/µCi and 1x106 ml/m3 gives 1 Ci / m3 = 1 µCi / ml q and Q are the filtered release rate in Ci/s

[X/Q] is atmospheric dispersion parameter in s/m3 T is 2h, 24 h, 96 h, or 520 h for public exposure time outside the reactor building 7.2 Non-Fueled Experiment - Irradiation of Uranium Time integrated exposures for experiments with uranium For experiments utilizing small amounts of uranium below the limits for a fueled experiment, an accidental and continuous release is assumed to occur over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the activity dispersed into the reactor building in normal ventilation and no filtration. This case is similar to the accidental release except that the release is continuously made into the reactor building volume in normal ventilation with no filtration for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Airborne activity monitors would indicate abnormally high readings within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The time integrated exposure for an experiment using uranium is given by:

p = [A()/V][X/Q]FT Eq. 7-5 SECTION 8: DOSE ASSESSMENT Radiation monitoring system and air sampling Monitoring and air sampling of the reactor building exhaust and room air are continuously performed for radioactive particulates and gases as required by the reactor license and facility procedures and Radiation Protection Program. Air monitors provide indication in the control room and alarm at elevated levels. If the reactor building ventilation radiation monitors alarm, the evacuation alarm and confinement system initiate.

Setpoints are at low levels to allow for mitigation of any release and allow time for other actions to prevent activation of the emergency plan.

External dose For radiological control purposes, external dose rates are limited and controlled by the facility radiation protection program and facility procedures consistent with experimental limitations and conditions given in TS and 10 CFR Part 20 requirements, including ALARA (As Low As Reasonably Achievable) practices.

Appropriate access controls and radiation monitoring are used as required by the radiation protection program. The reactor radiation monitoring system and other radiation monitors as specified for the experiment are used to alert experimenters and reactor staff of abnormal radiation levels.

20

Radiation dose calculations and dose limits Radiation doses calculated include:

• Total Effective Dose-Equivalent (TEDE) for occupants inside and outside the reactor building.

• Total Organ Dose-Equivalent (TODE) to the thyroid for occupants inside the reactor building.

Dose from accidental release is limited as follows:

• TEDE to occupants inside the reactor building is limited to 0.15 rem.

• Thyroid TODE to occupants inside the reactor building is limited to 1.5 rem.

• TEDE for members of the public is limited to 0.003 rem.

Dose from a vented experiment is limited to a TEDE of 0.003 rem for occupational workers and members of the public.

Dose from experiments using uranium with a TEDE greater than 0.001 rem to members of the public or to personnel inside the reactor building are defined as fueled experiments.

External dose calculations [Ref 4, 33, Section 14 Calculations 7, 8]

External dose from exposure to the reactor building, overhead plume, reactor stack, and ventilation system ducts were calculated using average concentrations and exposure times.

Microshield was used to determine dose outside the reactor building from the following sources:

• Contaminated air present in the reactor building

• Overhead plume and Reactor stack The reactor building was modeled as a rectangular volume with a total air volume of 2.25x109 ml.

Dimensions were set at 50 feet high by 40 feet deep and 40 feet wide. The reactor walls are made of reinforced ordinary concrete with a density of 2.35 g/ml and a thickness of 30 cm.

The overhead plume, reactor stack, and ventilation ducts were modeled as line sources with no shielding.

Lengths and locations of interest for the line sources are as follows:

Overhead Plume:

• Horizontal line at a length of 100 meters at a height of 30 m.

• The highest dose point is at the line midpoint, i.e. x = 50 m and y = 0 m, at an elevation (z) of 12 m Reactor Stack:

• Vertical line at a length of 20 m. The exhaust duct enters the stack at a height of 10 m.

• The stack is 30 m high.

• Dose points are at the base of the entry point, i.e. z = 10 m at distances (x) from 5 m to 50 m.

Source terms for accidents are the initial concentration and average concentrations over 2 h and 24 h derived from the saturation activity dispersed within the reactor building volume in the confinement ventilation mode.

The highest dose point is opposite the midpoint of the line source, except for the stack. For the stack, occupied areas near the stack are at the bottom (or end) of the line. No correction for decay is made.

21

Based on the dimensions of the stack, the following relationship for activity per unit length was used for the stack line source:

Stack Activity, A(stack) = 3.93 C Eq. 8-1 where, Stack volume = 3.93 m3 for 20 m length and 0.5 m diameter C is the stack concentration in Ci/m3 At C = 1 Ci/m3, A(stack) = 3.93 Ci = (3.93 m3)(1 Ci /m3)

Under calm winds, the activity per unit length is at a maximum. The following relationship was used for the overhead line source:

Overhead Plume Activity, A(plume) = 56.6 C Eq. 8-2 where, C is stack concentration in Ci / m3 At C = 1 Ci/m3, A(plume) = 56.6 Ci = (1Ci/m3 )(0.283 m3/s) (100 m / 0.5 m/s) t, time in plume = 100 m / (0.5 m/s) = 200 s Ventilation ducts:

Activity in the beam tube ventilation ducts, A(d), is estimated from the decayed and filtered release rate, q, and time in the ventilation system.

Time in the ventilation system is estimated to be approximately 4 s based on linear distance of 150 feet of duct and the measured linear velocity of 40 feet per second. A(d) is given by:

A(d) = 4 q Eq. 8-3 where, q is the decayed and filtered release rate A(d) is distributed over multiple horizontal and vertical ducts. Maximum length of exhaust duct is 15 m. No correction for decay in the ventilation duct is made. Maximum dose rates from the ventilation exhaust ducts were calculated at the center of a line source using Microshield.

Dose from released activity [Ref Section 14 Calculations 1, 2, 3, 4, 5]

Radiation dose from the submersion and inhalation pathways for the radioactive materials released include the following as defined in 10 CFR Part 20:

• Deep dose-equivalent (DDE) from submersion

• TEDE from inhalation and submersion given by the sum of the DDE from submersion and the committed effective dose-equivalent (CEDE) from inhalation

• Thyroid TODE is given by the sum of the DDE from submersion and committed dose-equivalent (CDE) from inhalation 22

Dose to occupational workers and members of the public is determined as follows for each radioactive material released:

D = DCF f Eq. 8-4 where, D is dose, in rem is the Time Integrated Exposure (µCi-h/ml), either r or p r is taken from Eq 7-1 and p is taken from Eq 7-3, Eq 7-4, or Eq 7-5 DCF = Dose Conversion Factor in rem/h per µCi/ml f = 0.1 for submersion dose correction inside the reactor building, otherwise f = 1 Dose conversion factors (DCF) [Ref 21, 29, 30, 31]

Dose conversion factors (DCF) were taken from the following references:

• Inhalation DCF: Federal Guidance Report 11

• Submersion DCF: Federal Guidance Report 11 for noble gases Federal Guidance Report 12 for halogens Publication EPA400 for Xe-137 and Kr-89 Inhalation DCF were converted to rem per µCi-h/ml based on the adult breathing rate of 2.4x109 ml per 2000 h reported in 10 CFR Part 20 Appendix B.

Submersion dose correction [Ref 8, 9, 10, 12,26]

Reduction of submersion dose from photons emitted by released activity inside the reactor building is made based on room dimensions using the following:

f = fGk = en RGk Eq. 8-5 f = (4.92x10-5 / cm)(905 cm)(2)(1.1) = 9.8x10-2 ~ 0.1 Alternately, f = 2k[1-exp(-en r)] = 2(1.1)[1- exp(-4.92x10-5*905)] ~ 0.1 where, f is the submersion dose correction factor and has a value of ~ 0.1 or less and is applied to the submersion dose inside the reactor building.

f the ratio of dose from a finite cloud to dose from a semi-infinite cloud given by the product of uen r .

en = energy absorption coefficient in air for photons, for photons with an energy of 50 keV or more this value is < 4.92x10-5 per cm.

r = effective radius of 905 cm based on the reactor building volume of 3.0x109 ml.

G = geometry correction factor of 2 for a sphere (4 geometry) vs. hemisphere (2 geometry, semi-infinite cloud).

k = ratio of mass energy absorption coefficients for tissue to air to convert to tissue dose having a value of ~ 1.1 for photon energies from 50 keV to several MeV.

23

SECTION 9: EXPERIMENT LIMITS [Ref Section 14 Calculations 1, 2, 3, 4]

Fission rate The fission rate limit is used to control the radioactive material inventory that may be accidentally released or planned on being released during a vented experiment. The fission rate limit meets dose criteria for an accidental release or vented experiment.

For a given release, radiation dose and fission rate are calculated. To determine the fission rate, the radiation dose criterion to be met is compared to the radiation dose per unit fission rate.

Fission rate limits were calculated for U-235 and Pu-239. The lower fission rate limit is used for all fueled experiments in TS 3.8. Dose from released activity for individual fissionable materials would be the same or less than those at the fission rate limit. This allows any mixture of fissionable material to be used and met the TS dose criteria.

The fission rate limit is given by:

[f/s]Limit = Dose criterion / Calculated dose per unit fission rate Eq. 9-1 where, [f/s]Limit is the fission rate limit in fissions per second The dose criteria based on three percent of the annual limits given in 10 CFR Part 20 are:

• 0.15 rem TEDE to occupants inside the reactor building

• 1.5 rem TODE to the thyroid to occupants inside the reactor building

• 0.003 rem TEDE to members in public areas outside the reactor building Dose criteria for vented experiments are 0.003 rem TEDE for occupants inside the reactor building or to members of the public outside the reactor building.

Dose from experiments for experiments using uranium with a TEDE greater than 0.001 rem to members of the public or to personnel inside the reactor building are defined as fueled experiments.

The applicable dose (TEDE or Thyroid TODE) per unit f/s is used to calculate the fission rate limit.

Variance with fluence rate and time The mass of fissionable materials used in a fueled experiment is related to the number of atoms (N). N varies inversely with (fluence rate) to maintain the same limiting fission rate. From the limiting fission rate, the number of atoms is calculated for the fluence rate used in the fueled experiment:

N = [f/s]Limit / [] Eq. 9-2 where, N is converted to mass of the fissionable material The fission rate limit is based on the highest dose for a given exposure time. At other exposure times, the dose is lower. This is due to the assumption that saturation activities are always present and noting that the time integrated exposure is lower at other exposure times.

Therefore, no adjustment for irradiation time or exposure time is needed since it was accounted for in the limiting fission rate and total number of fissions.

24

Total number of fissions A limit on the total number of fissions is used to prevent accumulation of higher amounts of longer lived radionuclides, including those listed in Category 2 Quantities of Concern in 10 CFR Part 37.

For a given fissionable material used in a fueled experiment, a sample may be irradiated at different fluence rates at different times provided that the total number of fissions is not exceeded.

The total number of fissions is the sum of the product of the fission rate and irradiation time,T:

[fission]Limit = NT Eq. 9-3 The maximum exposure time is assumed to be the same as the maximum irradiation time of 520 h at the maximum fluence rate. 520 h was used in calculating the total number fissions allowed.

Experiment controls TS 3.8 limitations and conditions include those given previously and other TS requirements for experiments, reactivity, storage of fissionable materials and experiment reviews.

SECTION 10: FUELED EXPERIMENT DEFINITION [Ref Section 14 Calculation 9]

A fueled experiment is defined as an experiment involving any of the following:

• Neutron irradiation of uranium exceeding 1.9 x 106 fissions per second or 1.6 x 1011 fissions.

• Neutron irradiation of any amount of neptunium or plutonium

• A planned release of fission gases or halogens.

The definition of a fueled experiment is revised to allow irradiation of materials containing small amounts of uranium. Threshold limits on fission rate and total number of fissions are established for experiments containing uranium which are based on limiting the radiation dose from a potential release to one percent (1%) of the annual public dose limit given in 10 CFR Part 20, i.e. a TEDE of 0.001 rem. Experiments involving the neutron irradiation of uranium below these threshold limits are not classified as fueled experiments. Experiments involving the neutron irradiation of uranium in excess of these threshold limits, the neutron irradiation of any amount of another fissionable material, or involving a planned vented release of fission gases or halogens are defined as fueled experiments.

For experiments with samples containing uranium:

• The release of fission gases and halogens is assumed to be accidentally released to the reactor building free air volume continuously over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during the irradiation using normal ventilation with no filtration. Airborne activity monitors would indicate abnormally high readings within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• The fission rate limit used to define fueled experiments is based on a TEDE of 0.001 rem to members of the public or personnel inside the reactor building.

• The total number of fissions is based on an accidental 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release that may occur at any time during the 520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />.

25

Fueled experiments exclude fissionable material not subjected to neutron fluence, detectors containing fissionable material used in the operation of the reactor or used in an experiment, sealed sources, and fuel used in operation of the reactor.

• Source Material means: (1) uranium or thorium, or any combination thereof, in any physical or chemical form or (2) ores which contain by weight one-twentieth of one percent (0.05%) or more of:

(i) uranium, (ii) thorium or (iii) any combination thereof. Source material does not include special nuclear material.

• Special nuclear material means: (1) Plutonium, Uranium 233, Uranium enriched in the isotope 233 or in the isotope 235, and any other material which the Commission, pursuant to the provisions of section 51 of the Act, determines to be special nuclear material; or (2) any material artificially enriched.

• Sealed sources are defined as sources encased in a capsule designed to prevent leakage or escape of the material from the intended use of the source or potential minor mishaps. Manufactured detectors, sealed sources with registration certificates generated by the NRC and Agreement States, special form radioactive material as defined in 10 CFR Part 71, and NRC approved reactor fuel elements in cladding are examples of such excluded materials.

SECTION 11: EMERGENCY PLAN [Ref 4, 6, 7] - NO REVISION NEEDED TODE to the thyroid and TEDE radiation dose criteria for fueled experiments are below emergency action levels (EAL) given in the facility emergency plan. These radiation doses are also below those given for the fuel handling accident release scenario assumed in Section 13 of the FSAR.

No revision to the emergency plan is needed.

SECTION 12: SECURITY, STORAGE, and INVENTORY [Ref 11, 24, 32, Section 14 Calculations 6]

- NO REVISION NEEDED The possession limits are within 10 CFR Part 37 Category 2 limits for the fissionable materials requested and associated fission product inventory. Sr-90, Cs-137, and Pm-147 activities were calculated at the fission rate limit and maximum fluence rate and irradiation time. Other fission products listed in 10 CFR Part 37 are produced in insignificant quantities due to low cumulative fission yields. At the limit for the total number of fissions, the fraction of 10 CFR Part 37 Category 2 activity limits is approximately 6.0x10-7, which is well below the limiting value of 1.

TS 5.3 requirements for fueled experiments in storage shall be met, as applicable. Calculations and measurements made for reactor fuel are used for fueled experiment storage. These are documented using facility procedures to verify fueled experiments are stored in a configuration to keep keff no greater than 0.9.

Storage facilities are reviewed under TS 3.8, 10 CFR Part 50.59 for design changes, 10 CFR 50.54(p) for security, 10 CFR 50.54(q) for emergency planning, and 10 CFR Part 20 for radiation protection.

Fissionable materials used in fueled experiments are inventoried and accounted for as required by 10 CFR Part 70, the university broad scope license, and facility procedures.

With the limitations proposed, no revision of the security plan is needed.

26

SECTION 13: POSSESSION LIMITS Possession of up to 32 grams of Uranium-235, up to 1 g of Neptunium-237, and up to 5 g of Plutonium-239 for fueled experiments is requested based on experiment needs. Experiment needs include evaluation of fissionable materials used for reactor fuel and neutron detection.

Radionuclides initially present and those produced by activation of Uranium with subsequent decay include:

• Uranium: U-234, U-235, U-236, U-237, U-238, U-239

• Neptunium: Np-237, Np-238, Np-239

• Plutonium: Pu-238, Pu-239, Pu-240 U-234, U-235 and U-238 are present in natural abundances or uranium enriched in U-235.

Np-237 and Pu-239 are long-lived radionuclides.

SECTION 14: CALCULATION RESULTS Calculations 1 through 8 were performed for accidental and planned vented releases of fission gases and halogens from the irradiation of U-235 and Pu-239. U-235 and Pu-239 were determined to be the limiting fissionable materials of those being requested for use in fueled experiments due to higher fission cross-sections and mass.

From the calculations performed for inhalation and submersion doses, the limiting case for all fueled experiments was a planned release from a vented experiment using U-235 under the assumed conditions with a fission rate of 9.6x109 f/s. This limiting fission rate is used in TS 3.8.

For accidental releases, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> gave the higher time integrated exposures and radiation doses. This is due to all of the activity released being ventilated from the reactor building within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For vented experiments, A() is assumed to always be present and dose is directly proportional to the product [X/Q] T.

Conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> have the highest product of [X/Q] T and therefore the highest dose.

Calculation 9 was performed to define fueled experiments for experiments containing uranium. The fission rate calculated is used in the revised TS 1.2.9 e.

Calculation 10 was performed to determine radiation monitor set points for the revised TS 3.5.

27

CALCULATIONS 1 and 2 - Vented Experiments using U-235 and Pu-239 Tables 14-1 and 14-2 below give the data and results for TEDE from vented experiments in public areas outside the reactor building for fueled experiments using U-235 and Pu-239, respectively.

Fission rates of 9.6x109 f/s for U-235 and 1.4x1010 f/s for Pu-239 were calculated. At the fluence rates used in this analysis, a fission rate of 9.6x109 f/s is equivalent to 4.95x10-3 g of U-235 or 3.87x10-3 g of Pu-239.

Table 14-1: Calculation 1 - Vented Experiment using U-235 PARAMETER VALUES Parameter Value Units Parameter Value Units Nuclide U-235 Target atoms, N 1.27E+19 atoms Mass 4.95E-03 g Thermal fission rate 7.42E+09 f/s Mass Number, A 235 g/mol Non-thermal fission rate 2.17E+09 f/s Sigma thermal 585 b Total fission rate 9.59E+09 f/s Sigma non-thermal 571 b Reactor volume 2.25E+09 ml X/Q 8.54E-03 s/m3 F confinement 0.283 m3/s Thermal flux 1.00E+12 cm2/s v confinement 1.26E-04 1/s Non-thermal flux 3.00E+11 cm2/s F normal 0.883 m3/s Irradiation time 8.64E+04 sec v normal 3.92E-04 1/s Vented experiment exhaust 3 lpm Evacuation time in confine 120 s Vented experiment volume 90 liters Evacuation time in normal 240 s Public exposure 24 hours (1-R) halogens 0.1 ISOTOPIC DATA Half-Life Decay Cumulative Yield % Cumulative Yield % Eq. 2-1 Nuclide (sec) Constant (1/s) Thermal Fission Non-Thermal Fission Saturation Activity (µCi) 83mKr 6.70E+03 1.04E-04 5.36E-01 5.75E-01 1.41E+03 85mKr 1.61E+04 4.30E-05 1.29E+00 1.36E+00 3.38E+03 85Kr 3.39E+08 2.05E-09 2.83E-01 2.96E-01 7.41E+02 87Kr 4.57E+03 1.52E-04 2.56E+00 2.54E+00 6.63E+03 88Kr 1.02E+04 6.78E-05 3.55E+00 3.43E+00 9.14E+03 89Kr 1.89E+02 3.67E-03 4.51E+00 3.97E+00 1.14E+04 131mXe 1.03E+06 6.74E-07 4.05E-02 3.54E-02 1.02E+02 133mXe 1.89E+05 3.66E-06 1.89E-01 1.97E-01 4.95E+02 133Xe 4.53E+05 1.53E-06 6.70E+00 6.71E+00 1.74E+04 135mXe 9.18E+02 7.55E-04 1.10E+00 1.26E+00 2.95E+03 135Xe 3.28E+04 2.12E-05 6.54E+00 6.58E+00 1.70E+04 137Xe 2.29E+02 3.02E-03 6.13E+00 5.98E+00 1.58E+04 138Xe 8.46E+02 8.19E-04 6.30E+00 6.00E+00 1.62E+04 131I 6.93E+05 1.00E-06 2.89E+00 3.22E+00 7.69E+03 132I 8.26E+03 8.39E-05 4.31E+00 4.66E+00 1.14E+04 133I 7.49E+04 9.26E-06 6.70E+00 6.70E+00 1.74E+04 134I 3.16E+03 2.20E-04 7.83E+00 7.63E+00 2.02E+04 135I 2.37E+04 2.93E-05 6.28E+00 6.27E+00 1.63E+04 83Br 8.64E+03 8.02E-05 5.40E-01 5.76E-01 1.42E+03 84Br 1.91E+03 3.63E-04 9.67E-01 1.01E+00 2.53E+03 28

Table 14-1: Calculation 1 - continued VENTED RELEASE DOSE

SUMMARY

and RESULTS Eq. 5-8, Eq. 7-4 Eq. 8-4 Public Time Integrated DCF Public TEDE Nuclide Exposure (µCi-h/ml) (rem per µCi-h/ml) (rem) 83mKr 1.34E-07 1.52E-02 2.03E-09 85mKr 3.57E-07 1.10E+02 3.93E-05 85Kr 8.44E-08 1.74E+00 1.47E-07 87Kr 5.74E-07 5.25E+02 3.02E-04 88Kr 9.21E-07 1.33E+03 1.23E-03 89Kr 1.76E-09 1.20E+03 2.11E-06 131mXe 1.16E-08 5.48E+00 6.35E-08 133mXe 5.60E-08 1.99E+01 1.11E-06 133Xe 1.97E-06 2.25E+01 4.43E-05 135mXe 8.62E-08 2.79E+02 2.40E-05 135Xe 1.86E-06 1.73E+02 3.22E-04 Fission Rate [f/s]

137Xe 7.78E-09 1.10E+02 8.56E-07 9.60E+09 138Xe 4.21E-07 7.10E+02 2.99E-04 131I 8.74E-08 2.42E+02 2.12E-05 Fissions 132I 1.11E-07 1.49E+03 1.66E-04 1.80E+16 133I 1.95E-07 3.92E+02 7.62E-05 134I 1.55E-07 1.73E+03 2.68E-04 rem per [f/s]

135I 1.76E-07 1.06E+03 1.87E-04 3.12E-13 83Br 1.40E-08 5.09E+00 7.12E-08 84Br 1.50E-08 1.25E+03 1.88E-05 TEDE limit (rem)

Total Public TEDE (rem) = 3.00E-03 3.00E-03 Supporting Calculations for U-235:

The public dose from 9.6x109 f/s for U-235 is 3.0x10-3 rem.

Saturation activity (Reference Eq. 2-1): A() = kNY Kr-87: 6.63x103 µCi = (4.95x10-3g)(6.022x1023 / 235 g)(2.703x10-29) x [(585)(1x1012)(2.56/100)+(571)(3x1011)(2.54/100)] or 6.63x10-3 Ci Public Time Integrated Exposure (Reference Eq. 5-8 and Eq. 7-4):

p = [A () / w] exp (-t)[1-R] p [X/Q]T Kr87: 5.74x10-7µCi-h/ml = (6.63x103µCi/9x104ml)(50)e[-(1.52E-4)(9E4)/50](8.54x10-3s/m3)(1x10-6m3/ml)(24 h)

I133: 1.95x10-7µCi-h/ml = (1.74x104/9 x104)(50)e[-(1.53E-6)(9E4)/50](0.1)(8.54 x10-3)(1x10-6)(24 )

Public Dose for Xe-133 (Reference Eq. 8-4): D = DCF f TEDE is 4.43x10-5 rem = (1.97x10-6 µCi-h/ml)(22.5 rem per µCi-h/ml)

Fission rate limit from Eq. 9-1: 9.6x109 f/s = 1.0x10-3 rem / (3.13x10-13 rem per f/s)

Fission Limit from Eq. 9-3: 1.80x1016 fissions = (9.6x109 f/s) (1.87x106 s) 29

Table 14-2: Calculation 2 - Vented Experiment using Pu-239 PARAMETER VALUES Parameter Value Units Parameter Value Units Nuclide Pu239 Target atoms, N 1.43E+19 atoms Mass 5.67E-03 g Thermal fission rate 1.07E+10 f/s Mass Number, A 239 g/mol Non-thermal fission rate 3.38E+09 f/s Sigma thermal 748 b Total fission rate 1.41E+10 f/s Sigma non-thermal 789 b Reactor volume 2.25E+09 ml X/Q 8.54E-03 s/m3 F confinement 0.283 m3/s Thermal flux 1.00E+12 cm2/s v confinement 1.26E-04 1/s Non-thermal flux 3.00E+11 cm2/s F normal 0.883 m3/s Irradiation time 8.64E+04 sec v normal 3.92E-04 1/s Vented experiment exhaust 3 lpm Evacuation time in confine 120 s Vented experiment volume 90 liters Evacuation time in normal 240 s Public exposure 24 h (1-R) halogens 0.1 ISOTOPIC DATA Eq. 2-1 Half-Life Decay Cumulative Yield % Cumulative Yield % Saturation Nuclide (sec) Constant (1/s) Thermal Fission Non-Thermal Fission Activity (µCi) 83mKr 6.70E+03 1.04E-04 2.97E-01 3.15E-01 1.14E+03 85mKr 1.61E+04 4.30E-05 5.63E-01 5.94E-01 2.17E+03 85Kr 3.39E+08 2.05E-09 1.23E-01 1.38E-01 4.81E+02 87Kr 4.57E+03 1.52E-04 9.89E-01 1.04E+00 3.81E+03 88Kr 1.02E+04 6.78E-05 1.27E+00 1.29E+00 4.85E+03 89Kr 1.89E+02 3.67E-03 1.45E+00 1.45E+00 5.52E+03 131mXe 1.03E+06 6.74E-07 4.24E-02 4.27E-02 1.62E+02 133mXe 1.89E+05 3.66E-06 2.31E-01 2.45E-01 8.92E+02 133Xe 4.53E+05 1.53E-06 7.02E+00 6.97E+00 2.66E+04 135mXe 9.18E+02 7.55E-04 1.84E+00 2.08E+00 7.21E+03 135Xe 3.28E+04 2.12E-05 7.60E+00 7.54E+00 2.89E+04 137Xe 2.29E+02 3.02E-03 6.01E+00 5.58E+00 2.24E+04 138Xe 8.46E+02 8.19E-04 5.17E+00 4.71E+00 1.92E+04 131I 6.93E+05 1.00E-06 3.86E+00 3.88E+00 1.47E+04 132I 8.26E+03 8.39E-05 5.39E+00 5.32E+00 2.04E+04 133I 7.49E+04 9.26E-06 6.97E+00 6.91E+00 2.65E+04 134I 3.16E+03 2.20E-04 7.41E+00 7.11E+00 2.79E+04 135I 2.37E+04 2.93E-05 6.54E+00 6.08E+00 2.44E+04 83Br 8.64E+03 8.02E-05 2.97E-01 3.15E-01 1.14E+03 84Br 1.91E+03 3.63E-04 4.29E-01 4.63E-01 1.66E+03 30

Table 14-2: Calculation 2 - continued VENTED RELEASE DOSE

SUMMARY

and RESULTS Eq. 5-8, Eq. 7-4 Eq. 8-4 Public Time Integrated DCF Public Dose Nuclide Exposure (µCi-h/ml) (rem per µCi-h/ml) (rem) 83mKr 1.08E-07 1.52E-02 1.64E-09 85mKr 2.29E-07 1.10E+02 2.52E-05 85Kr 5.48E-08 1.74E+00 9.53E-08 87Kr 3.30E-07 5.25E+02 1.73E-04 88Kr 4.89E-07 1.33E+03 6.51E-04 89Kr 8.54E-10 1.20E+03 1.02E-06 131mXe 1.84E-08 5.48E+00 1.01E-07 133mXe 1.01E-07 1.99E+01 2.01E-06 133Xe 3.02E-06 2.25E+01 6.79E-05 135mXe 2.11E-07 2.79E+02 5.88E-05 135Xe 3.16E-06 1.73E+02 5.48E-04 Fission Rate [f/s]

137Xe 1.11E-08 1.10E+02 1.22E-06 1.41E+10 138Xe 5.01E-07 7.10E+02 3.56E-04 131I 1.67E-07 2.42E+02 4.05E-05 Fissions 132I 2.00E-07 1.49E+03 2.98E-04 2.63E+16 133I 2.96E-07 3.92E+02 1.16E-04 134I 2.14E-07 1.73E+03 3.70E-04 rem per [f/s]

135I 2.64E-07 1.06E+03 2.81E-04 2.13E-13 83Br 1.13E-08 5.09E+00 5.74E-08 84Br 9.84E-09 1.25E+03 1.23E-05 Dose limit, rem Total Public Dose (rem) = 3.00E-03 3.00E-03 Supporting Calculations for Pu-239:

3.87x10-3 g of Pu-239 gives a fission rate of 9.6x109 f/s:

3.87x10-3g = (5.67x10-3g)(9.6x109 f/s)/(1.41x1010 f/s)

The public dose at 9.6x109 f/s for Pu-239 is: 2.04x10-3 rem = (9.6x109 f/s)(2.13x10-13 rem per f/s)

CALCULATIONS 3 and 4: Accidental Release from Experiment using Pu-239 and U-235 Tables 14-3 through 14-6 below give the data and results for radiation dose from accidental releases from experiments using Pu-239.

Tables 14-7 through 14-9 give the data and results for radiation dose from accidental releases from experiments using U-235.

For accidental releases, the thyroid TODE dose criteria for fueled experiments of 1.5 rem is limiting for both U-235 and Pu-239. Fission rates of 1.90x1010 f/s for Pu-239 and 2.32x1010 f/s for U-235.

A fission rate of 1.90x1010 f/s is used to compare results from Pu-239 and U-235.

31

Table 14-3: Calculation 3 -Accidental release from experiment using Pu-239 PARAMETER VALUES Parameter Value Units Parameter Value Units Nuclide Pu239 Target atoms, N 1.93E+19 atoms Mass 7.65E-03 g Thermal fission rate 1.44E+10 f/s Mass Number, A 239 g/mol Non-thermal fission rate 4.56E+09 f/s Sigma thermal 748 b Total fission rate 1.90E+10 f/s Sigma non-thermal 789 b Reactor volume 2.25E+09 ml X/Q = 8.54E-03 s/m3 F confinement 0.283 m3/s Thermal flux 1.00E+12 cm2/s v confinement 1.26E-04 1/s Non-thermal flux 3.00E+11 cm2/s F normal 0.883 m3/s Irradiation time 8.64E+04 sec v normal 3.92E-04 1/s Vented experiment exhaust 3 lpm Evacuation time in confine 120 s Evacuation time in normal 240 s Vented experiment volume 90 Liters NG reactor correction 0.1 Public Hours 24 h (1-R) halogens 0.1 ISOTOPIC DATA Eq. 2-1 Half-Life Decay Cumulative Yield % Cumulative Yield % Saturation Nuclide (sec) Constant (1/s) Thermal Fission Non-Thermal Fission Activity (µCi) 83mKr 6.70E+03 1.04E-04 2.97E-01 3.15E-01 1.54E+03 85mKr 1.61E+04 4.30E-05 5.63E-01 5.94E-01 2.93E+03 85Kr 3.39E+08 2.05E-09 1.23E-01 1.38E-01 6.49E+02 87Kr 4.57E+03 1.52E-04 9.89E-01 1.04E+00 5.13E+03 88Kr 1.02E+04 6.78E-05 1.27E+00 1.29E+00 6.54E+03 89Kr 1.89E+02 3.67E-03 1.45E+00 1.45E+00 7.45E+03 131mXe 1.03E+06 6.74E-07 4.24E-02 4.27E-02 2.18E+02 133mXe 1.89E+05 3.66E-06 2.31E-01 2.45E-01 1.20E+03 133Xe 4.53E+05 1.53E-06 7.02E+00 6.97E+00 3.59E+04 135mXe 9.18E+02 7.55E-04 1.84E+00 2.08E+00 9.73E+03 135Xe 3.28E+04 2.12E-05 7.60E+00 7.54E+00 3.89E+04 137Xe 2.29E+02 3.02E-03 6.01E+00 5.58E+00 3.03E+04 138Xe 8.46E+02 8.19E-04 5.17E+00 4.71E+00 2.59E+04 131I 6.93E+05 1.00E-06 3.86E+00 3.88E+00 1.98E+04 132I 8.26E+03 8.39E-05 5.39E+00 5.32E+00 2.76E+04 133I 7.49E+04 9.26E-06 6.97E+00 6.91E+00 3.57E+04 134I 3.16E+03 2.20E-04 7.41E+00 7.11E+00 3.76E+04 135I 2.37E+04 2.93E-05 6.54E+00 6.08E+00 3.30E+04 83Br 8.64E+03 8.02E-05 2.97E-01 3.15E-01 1.54E+03 84Br 1.91E+03 3.63E-04 4.29E-01 4.63E-01 2.24E+03 32

Table 14-4: Calculation 3 - Time integrated exposures for Pu-239 for a public exposure time of 24 h Eq. 5-2, Eq. 7-1 Eq. 5-2, Eq. 7-1 Eq. 5-2, Eq. 7-2 Eq. 5-2, Eq. 7-2 Time Integrated Time Integrated Time Integrated Time Integrated Exposure - Confinement Exposure - Confinement Exposure- Normal Exposure-Normal Nuclide Reactor (µCi-h/ml) Public (µCi-h/ml) Reactor (µCi-h/ml) Public (µCi-h/ml) 83mKr 2.26E-08 2.01E-09 4.31E-08 3.25E-10 85mKr 4.29E-08 5.17E-09 8.23E-08 6.21E-10 85Kr 9.54E-09 1.54E-09 1.84E-08 1.38E-10 87Kr 7.48E-08 5.52E-09 1.43E-07 1.07E-09 88Kr 9.58E-08 1.01E-08 1.84E-07 1.38E-09 89Kr 8.86E-08 5.86E-10 1.41E-07 1.06E-09 131mXe 3.21E-09 5.15E-10 6.17E-09 4.65E-11 133mXe 1.77E-08 2.77E-09 3.40E-08 2.56E-10 133Xe 5.28E-07 8.42E-08 1.02E-06 7.66E-09 135mXe 1.37E-07 3.30E-09 2.52E-07 1.90E-09 135Xe 5.72E-07 7.91E-08 1.10E-06 8.28E-09 137Xe 3.74E-07 2.87E-09 6.12E-07 4.62E-09 138Xe 3.63E-07 8.20E-09 6.67E-07 5.03E-09 131I 2.91E-07 4.66E-09 5.60E-07 4.22E-09 132I 4.03E-07 3.92E-09 7.71E-07 5.81E-09 133I 5.24E-07 7.89E-09 1.01E-06 7.60E-09 134I 5.46E-07 3.25E-09 1.04E-06 7.82E-09 135I 4.84E-07 6.34E-09 9.29E-07 7.00E-09 83Br 2.26E-08 2.24E-10 4.33E-08 3.26E-10 84Br 3.23E-08 1.37E-10 6.08E-08 4.58E-10 Supporting Calculations:

Saturation Activity, A() (Reference Eq. 2-1): A() = k N Y I-131: 1.98x104 µCi = (7.65x10-3 g)(6.022x1023/239 g)(2.703x10-29) x [(748)(1x1012)(3.86/100)+(789)(3x1011)(3.88/100)] , or 1.98x10-2 Ci Time integrated exposure (Reference Eq. 5-2 and Eq. 7-1):

r = <C>T and < C > = C(0)[ (1- e-kT) / (kT)] or r = C(0)[ (1- e-kT) / k]

where:

r for each ventilation mode and exposure time is summed for the reactor building:

k = + v = 1.4176 per h in normal ventilation and 0.45858 per h in confinement T = 0.066 h in normal ventilation and 0.033 h in confinement Xe-133: r = (3.59x104 µCi / 2.25x109 ml) x {[1-exp((-1.4176)(0.066)) / 1.4176] + [1-exp((-0.45858)(0.033))/0.45858]}

= (5.28x10-7 + 1.02x10-6) µCi-h/ml

= 1.548x10-6 µCi-h/ml 33

Time integrated exposure (Reference Eq. 5-2 and Eq. 7-2):

p = <C>[1-R][X/Q]FT where:

p for each ventilation mode and exposure time is summed for the public:

k = + v = 1.4176 per h in normal ventilation and 0.45858 per h in confinement T = 0.066 h in normal ventilation and 24 h in confinement F = 0.883 m3/s in normal ventilation and 0.283 m3/s in confinement

[X/Q] = 8.54x10-3 s/m3 for T up to 24 h R=0 Xe-133: p = (3.59x104 µCi / 2.25E9 ml)(8.54x10-3) x {(0.883)[1-exp((-1.4176)(0.066)) / 1.4176]+ (0.283)[1-exp((-0.45858)(24))/0.45858]}

= (8.42x10-8 +7.66x10-9) µCi-h/ml = 9.186x10-8 µCi-h/ml Dose calculations (Reference Eq. 8-4): D = DCF f Xe-133 dose inside the reactor building:

3.47x10-6 rem = [(5.28x10-7 + 1.02x10-6) µCi-h/ml](22.5 rem per µCi-h/ml)(0.1)

= 1.19x10-6 + 2.28x10-6) rem Xe-133 dose outside the reactor building:

2.06x10-6 rem = [(8.42x10-8 +7.66x10-9) µCi-h/ml]( (22.5rem per µCi-h/ml)

= (1.89 x10-6 + 1.72x10-7) rem I-131 dose inside the reactor building:

TEDE is 3.36x10-2 rem = (2.91x10-7 + 5.60x10-7)(3.95x104+0.1(242))

= (1.15x10-2+ 2.21x10-2) rem Thyroid TODE is 1.10 rem = (2.91x10-7 + 5.60x10-7)(1.3 x106+0.1(242))

= (0.378 +0.726) rem I-131 dose outside the reactor building:

TEDE is 3.53x10-4 rem = (4.66 x10-9 + 4.22x10-9)(3.95 x104 + 242)

= (1.85x10-4 + 1.68x10-4) rem Fission rate (Reference Eq. 9-1):

The fission rate limit for an accidental release from Pu-239 irradiation is based on the thyroid dose inside the reactor building:

[f/s]Limit = Dose criterion / Calculated dose per unit fission rate 1.90x10-10 f/s = (1.5 rem) / (7.90x10-11 rem per f/s) 34

Table 14-5: Calculation 3 - Pu-239 Dose Calculation Results Eq. 8-4 Eq. 8-4 Eq. 8-4 Reactor Eq. 8-4 Eq. 8-4 Reactor Eq. 8-4 Effective Thyroid Reactor Vent. Public Reactor Normal Public Inhaltn. Inhaltn. Submer. Vent. Confine. Vent. Normal Vent. Normal DCF DCF DCF Confine. Thyroid Confine. Vent. Thyroid Vent.

(rem per (rem per (rem per TEDE Dose TEDE TEDE TODE TEDE Nuclide µCi-h/ml) µCi-h/ml) µCi-h/ml) (rem) (rem) (rem) (rem) (rem) (rem) 83mKr 1.52E-02 3.42E-11 3.42E-11 3.05E-11 6.55E-11 6.55E-11 4.93E-12 85mKr 1.10E+02 4.73E-07 4.73E-07 5.71E-07 9.08E-07 9.08E-07 6.84E-08 85Kr 1.74E+00 1.66E-09 1.66E-09 2.68E-09 3.19E-09 3.19E-09 2.41E-10 87Kr 5.25E+02 3.93E-06 3.93E-06 2.90E-06 7.49E-06 7.49E-06 5.65E-07 88Kr 1.33E+03 1.28E-05 1.28E-05 1.34E-05 2.45E-05 2.45E-05 1.84E-06 89Kr 1.20E+03 1.06E-05 1.06E-05 7.03E-07 1.69E-05 1.69E-05 1.28E-06 131mXe 5.48E+00 1.76E-09 1.76E-09 2.82E-09 3.38E-09 3.38E-09 2.54E-10 133mXe 1.99E+01 3.52E-08 3.52E-08 5.52E-08 6.77E-08 6.77E-08 5.10E-09 133Xe 2.25E+01 1.19E-06 1.19E-06 1.89E-06 2.28E-06 2.28E-06 1.72E-07 135mXe 2.79E+02 3.81E-06 3.81E-06 9.19E-07 7.02E-06 7.02E-06 5.29E-07 135Xe 1.73E+02 9.90E-06 9.90E-06 1.37E-05 1.90E-05 1.90E-05 1.43E-06 137Xe 1.10E+02 4.11E-06 4.11E-06 3.16E-07 6.74E-06 6.74E-06 5.08E-07 138Xe 7.10E+02 2.58E-05 2.58E-05 5.82E-06 4.74E-05 4.74E-05 3.57E-06 131I 3.95E+04 1.30E+06 2.42E+02 1.15E-02 3.78E-01 1.85E-04 2.21E-02 7.26E-01 1.68E-04 132I 4.57E+02 7.73E+03 1.49E+03 2.44E-04 3.17E-03 7.64E-06 4.68E-04 6.08E-03 1.13E-05 133I 7.02E+03 2.16E+05 3.92E+02 3.70E-03 1.13E-01 5.84E-05 7.11E-03 2.18E-01 5.63E-05 134I 1.58E+02 1.28E+03 1.73E+03 1.81E-04 7.93E-04 6.14E-06 3.43E-04 1.51E-03 1.48E-05 135I 1.47E+03 3.76E+04 1.06E+03 7.65E-04 1.82E-02 1.61E-05 1.47E-03 3.50E-02 1.78E-05 83Br 1.03E+02 5.09E+00 2.35E-06 1.15E-08 2.43E-08 4.50E-06 2.20E-08 3.54E-08 84Br 1.01E+02 1.25E+03 7.30E-06 4.04E-06 1.85E-07 1.37E-05 7.62E-06 6.20E-07 Total 1.65E-02 5.13E-01 3.14E-04 3.17E-02 9.86E-01 2.78E-04 Table 14-6: Calculation 3 - Fueled Experiment Summary for Pu-239 TEDE rem Thyroid TODE rem TEDE rem Parameter Reactor Building Reactor Building Public Areas Total dose in rem 4.81E-02 1.50E+00 5.92E-04 rem per f/s 2.54E-12 7.90E-11 3.12E-14 Dose limit in rem 0.15 1.5 3.00E-03 Fission rate limit: Eq. 9-1 5.91E+10 f/s 1.90E+10 f/s 9.61E+10 f/s Analysis Notes:

At 9.6x109 f/s, the doses are 0.5 times those listed in Table 14-6 Total dose in rem: TEDE for Public Areas = 5.92x10-4 rem = 3.14x10-4 + 2.78x10-4 rem Rem per f/s: TEDE for Reactor Building = 2.54x10-12 = 4.81x10-2 rem / 1.90x1010 f/s 35

Table 14-7: Calculation 4 - Accident release from experiment using U-235 PARAMETER VALUES Parameter Value Units Parameter Value Units Nuclide U-235 Target atoms, N 2.51E+19 atoms Mass 9.8E-03 g Thermal fission rate 1.47E+10 f/s Mass Number, A 235 g/mol Non-thermal fission rate 4.30E+09 f/s Sigma thermal 585 b Total fission rate 1.90E+10 f/s Sigma non-thermal 571 b Reactor volume 2.25E+09 ml X/Q = 8.54E-03 s/m3 F confinement 0.283 m3/s Thermal flux 1.00E+12 cm2/s v confinement 1.26E-04 1/s Non-thermal flux 3.00E+11 cm2/s F normal 0.883 m3/s Irradiation time 8.64E+04 sec v normal 3.92E-04 1/s Vented experiment exhaust 3 lpm Evacuation time 120 s Vented experiment volume 90 liters NG reactor correction 0.1 Public Hours 24 h (1-R) halogens 0.1 ISOTOPIC DATA Eq. 2-1 Half-Life Decay Cumulative Yield % Cumulative Yield % Saturation Nuclide (sec) Constant (1/s) Thermal Fission Non-Thermal Fission Activity (µCi) 83mKr 6.70E+03 1.04E-04 2.97E-01 3.15E-01 2.80E+03 85mKr 1.61E+04 4.30E-05 5.63E-01 5.94E-01 6.70E+03 85Kr 3.39E+08 2.05E-09 1.23E-01 1.38E-01 1.47E+03 87Kr 4.57E+03 1.52E-04 9.89E-01 1.04E+00 1.31E+04 88Kr 1.02E+04 6.78E-05 1.27E+00 1.29E+00 1.81E+04 89Kr 1.89E+02 3.67E-03 1.45E+00 1.45E+00 2.25E+04 131mXe 1.03E+06 6.74E-07 4.24E-02 4.27E-02 2.02E+02 133mXe 1.89E+05 3.66E-06 2.31E-01 2.45E-01 9.80E+02 133Xe 4.53E+05 1.53E-06 7.02E+00 6.97E+00 3.44E+04 135mXe 9.18E+02 7.55E-04 1.84E+00 2.08E+00 5.84E+03 135Xe 3.28E+04 2.12E-05 7.60E+00 7.54E+00 3.36E+04 137Xe 2.29E+02 3.02E-03 6.01E+00 5.58E+00 3.13E+04 138Xe 8.46E+02 8.19E-04 5.17E+00 4.71E+00 3.20E+04 131I 6.93E+05 1.00E-06 3.86E+00 3.88E+00 1.52E+04 132I 8.26E+03 8.39E-05 5.39E+00 5.32E+00 2.25E+04 133I 7.49E+04 9.26E-06 6.97E+00 6.91E+00 3.44E+04 134I 3.16E+03 2.20E-04 7.41E+00 7.11E+00 4.00E+04 135I 2.37E+04 2.93E-05 6.54E+00 6.08E+00 3.22E+04 83Br 8.64E+03 8.02E-05 2.97E-01 3.15E-01 2.81E+03 84Br 1.91E+03 3.63E-04 4.29E-01 4.63E-01 5.01E+03 36

Table 14-8: Calculation 4 - U-235 Accident release for a public exposure time of 24 h at 6.33x109 f/s Eq. 5-2, Eq. 7-1 Eq. 5-2, Eq. 7-1 Eq. 5-2, Eq. 7-2 Eq. 5-2, Eq. 7-2 Time Integrated Time Integrated Time Integrated Time Integrated Exposure - Confinement Exposure - Confinement Exposure- Normal Exposure-Normal Nuclide Reactor (µCi-h/ml) Public (µCi-h/ml) Reactor (µCi-h/ml) Public (µCi-h/ml) 83mKr 4.09E-08 3.64E-09 7.81E-08 5.89E-10 85mKr 9.83E-08 1.18E-08 1.89E-07 1.42E-09 85Kr 2.16E-08 3.48E-09 4.15E-08 3.13E-10 87Kr 1.91E-07 1.41E-08 3.64E-07 2.75E-09 88Kr 2.65E-07 2.79E-08 5.07E-07 3.82E-09 89Kr 2.68E-07 1.77E-09 4.26E-07 3.21E-09 131mXe 2.97E-09 4.77E-10 5.71E-09 4.30E-11 133mXe 1.44E-08 2.26E-09 2.77E-08 2.09E-10 133Xe 5.06E-07 8.06E-08 9.73E-07 7.33E-09 135mXe 8.21E-08 1.98E-09 1.51E-07 1.14E-09 135Xe 4.94E-07 6.83E-08 9.48E-07 7.15E-09 137Xe 3.86E-07 2.97E-09 6.33E-07 4.77E-09 138Xe 4.48E-07 1.01E-08 8.22E-07 6.20E-09 131I 2.24E-07 3.58E-09 4.30E-07 3.24E-09 132I 3.30E-07 3.21E-09 6.31E-07 4.75E-09 133I 5.05E-07 7.60E-09 9.72E-07 7.32E-09 134I 5.80E-07 3.45E-09 1.10E-06 8.30E-09 135I 4.73E-07 6.20E-09 9.08E-07 6.85E-09 83Br 4.11E-08 4.07E-10 7.88E-08 5.94E-10 84Br 7.21E-08 3.06E-10 1.36E-07 1.02E-09 Eq. 8-4 Eq. 8-4 Effective Thyroid Eq. 8-4 Reactor Eq. 8-4 Eq. 8-4 Reactor Eq. 8-4 Inhaltn. Inhaltn. Submer. Reactor Confine. Public Reactor Normal Public DCF DCF DCF Confine. Thyroid Confine. Normal Thyroid Normal (rem per (rem per (rem per TEDE Dose TEDE TEDE TODE TEDE Nuclide µCi-h/ml) µCi-h/ml) µCi-h/ml) (rem) (rem) (rem) (rem) (rem) (rem) 83mKr 1.52E-02 6.20E-11 6.20E-11 5.52E-11 1.19E-10 1.19E-10 8.93E-12 85mKr 1.10E+02 1.08E-06 1.08E-06 1.31E-06 2.08E-06 2.08E-06 1.57E-07 85Kr 1.74E+00 3.75E-09 3.75E-09 6.05E-09 7.22E-09 7.22E-09 5.44E-10 87Kr 5.25E+02 1.00E-05 1.00E-05 7.42E-06 1.91E-05 1.91E-05 1.44E-06 88Kr 1.33E+03 3.53E-05 3.53E-05 3.71E-05 6.76E-05 6.76E-05 5.09E-06 89Kr 1.20E+03 3.22E-05 3.22E-05 2.13E-06 5.12E-05 5.12E-05 3.86E-06 131mXe 5.48E+00 1.63E-09 1.63E-09 2.61E-09 3.13E-09 3.13E-09 2.36E-10 133mXe 1.99E+01 2.87E-08 2.87E-08 4.50E-08 5.51E-08 5.51E-08 4.16E-09 133Xe 2.25E+01 1.14E-06 1.14E-06 1.81E-06 2.18E-06 2.18E-06 1.65E-07 135mXe 2.79E+02 2.29E-06 2.29E-06 5.51E-07 4.21E-06 4.21E-06 3.17E-07 135Xe 1.73E+02 8.55E-06 8.55E-06 1.18E-05 1.64E-05 1.64E-05 1.24E-06 137Xe 1.10E+02 4.25E-06 4.25E-06 3.26E-07 6.96E-06 6.96E-06 5.25E-07 138Xe 7.10E+02 3.18E-05 3.18E-05 7.18E-06 5.84E-05 5.84E-05 4.40E-06 131I 3.95E+04 1.30E+06 2.42E+02 8.84E-03 2.90E-01 1.42E-04 1.70E-02 5.58E-01 1.29E-04 132I 4.57E+02 7.73E+03 1.49E+03 2.00E-04 2.60E-03 6.25E-06 3.83E-04 4.97E-03 9.27E-06 133I 7.02E+03 2.16E+05 3.92E+02 3.57E-03 1.09E-01 5.63E-05 6.85E-03 2.10E-01 5.42E-05 134I 1.58E+02 1.28E+03 1.73E+03 1.92E-04 8.42E-04 6.52E-06 3.64E-04 1.60E-03 1.57E-05 135I 1.47E+03 3.76E+04 1.06E+03 7.48E-04 1.78E-02 1.57E-05 1.44E-03 3.42E-02 1.74E-05 83Br 1.03E+02 5.09E+00 4.28E-06 2.09E-08 4.42E-08 8.19E-06 4.01E-08 6.44E-08 84Br 1.01E+02 1.25E+03 1.63E-05 9.04E-06 4.14E-07 3.07E-05 1.70E-05 1.39E-06 Total 1.37E-02 4.21E-01 2.97E-04 2.63E-02 8.09E-01 2.44E-04 37

Table 14-9: Calculation 4 - Fueled Experiment Summary for U-235 Effective rem Thyroid rem Effective rem Parameter Reactor Building Reactor Building Public Areas Total dose in rem 4.00E-02 1.23E+00 5.41E-04 rem per f/s 2.11E-12 6.47E-11 2.85E-14 Dose limit in rem 1.50E-01 1.5 3.00E-03 Fission rate limit (f/s): Eq. 9-1 7.13E+10 2.32E+10 1.05E+11 NOTE: At 9.6x109 f/s, the doses are 0.5 times those listed in Table 14-9 CALCULATION 5: Radiation Doses (TEDE) for Specific Public Locations of Interest Tables 14-10 lists [X/Q] values for specific public locations of interest that were calculated as described in Section 6. Maximum [X/Q] values were 8.54x10-3 s/m3 for periods up to 24 h, 7.79x10-4 s/m3 for 96 h, and 9.15x10-5 s/m3 for 520 h.

Tables 14-11 and 14-12 give the TEDE for public areas from U-235 and Pu-239 for the limiting fission rate of 9.6x109 f/s.

TEDE in publicly occupied areas outside the reactor building were calculated at a fission rate of 9.6 x109 f/s to be 3.0x10-4 rem or less from an accidental release and 3.0x10-3 rem or less from a vented experiment.

Table 14-10: Calculation 5 - [X/Q] values for specific public locations of interest Eq. 6-7 Eq. 6-6 [X/Q] Eq. 6-4 Eq. 6-4 Eq. 6-4

[X/Q] Calm Eq. 6-4 Eq. 6-5 Eq. 6-5 Eq. 6-5 Fumign. Wind [X/Q] [X/Q] [X/Q] [X/Q]

Distance Height Up to Up to GPM GPM GPM GPM Building or x z 24 h 24 h 2h 24 h 96 h 520 h Location (m) (m) (s/m3) (s/m3) (s/m3) (s/m3) (s/m3) (s/m3)

All 30 to 100 up to 12 8.54E-03 3.99E-04 2.31E-04 2.31E-04 1.39E-07 1.56E-06 All 100 to 150 up to 12 2.46E-04 4.73E-05 2.39E-04 2.39E-04 2.73E-06 3.72E-06 All 150 to 5000 up to 30 2.00E-03 4.89E-05 7.57E-03 2.15E-03 7.79E-04 9.15E-05 Withers, Mann 50 12 5.38E-03 1.73E-04 9.73E-05 9.73E-05 4.20E-14 6.80E-22 Broughton, Riddick 70 12 3.97E-03 9.36E-05 2.26E-04 2.26E-04 3.56E-08 1.03E-06 Patterson, Ricks 90 12 3.17E-03 5.80E-05 2.41E-04 2.41E-04 1.08E-06 2.86E-06 DH Hill 150 30 2.00E-03 2.17E-05 7.57E-03 2.15E-03 7.79E-04 9.15E-05 Cox 175 12 1.73E-03 1.58E-05 2.17E-04 2.17E-04 5.68E-06 4.49E-06 Dabney 200 24 1.54E-03 1.22E-05 1.49E-03 5.12E-04 1.85E-04 2.87E-05 Hillsborough St. 200 15 1.54E-03 1.22E-05 2.59E-04 2.59E-04 1.77E-05 7.60E-06 Talley, Reynolds 200 12 1.54E-03 1.21E-05 2.05E-04 2.05E-04 8.97E-06 5.10E-06 Carroll, Syme 325 12 9.93E-04 4.61E-06 1.75E-04 1.64E-04 1.51E-05 5.39E-06 North 350 20 8.23E-04 3.99E-06 4.90E-04 1.76E-04 6.00E-05 1.00E-05 MAXIMUM 8.54E-03 3.99E-04 7.57E-03 2.15E-03 7.79E-04 9.15E-05 38

[X/Q] value analysis notes:

• Site boundary is located approximately 30 m away from the exhaust stack.

• Closest buildings outside the site boundary are 50 m away (Withers, Mann).

• Closest residential areas are 200 m away (Hillsborough St)

• Student dormitories are 325 m away (Carroll, Syme, North).

• Most buildings are three stories in height.

• DH Hill library is the tallest building near the facility at 150 m away and 30 m high.

• Maximum [X/Q] values are associated with occupied locations that are elevated or closer to the release point from the 30 m reactor stack. Ground level [X/Q] have lower values.

• There are no occupied areas at distances, x, less than 150 m at a height, z, greater than 12 m.

Table 14-11: Calculation 5 - U-235 radiation doses (TEDE) for public areas at the limiting fission rate of 9.6x109 f/s Eq. 8-4 Eq. 8-4 Eq. 8-4 Eq. 8-4 Eq. 8-4 Eq. 8-4 Eq. 8-4 Fumign. Calm Wind GPM Fumign. Fumign. GPM GPM Dist. Height TEDE. TEDE. TEDE. TEDE TEDE TEDE TEDE Building or x z 24 h 24 h 24 h 2h 24 h 96 h 520 h Location (m) (m) (rem) (rem) (rem) (rem) (rem) (rem) (rem)

All 30 to 100 up to 12 2.73E-04 1.38E-04 7.39E-06 2.22E-04 3.00E-03 1.94E-07 1.19E-05 All 100 to 150 up to 12 7.86E-06 1.64E-05 7.65E-06 6.39E-06 8.64E-05 3.82E-06 2.83E-05 All 150 to 5000 up to 30 6.39E-05 1.69E-05 6.88E-05 5.20E-05 7.03E-04 1.09E-03 6.96E-04 Withers, Mann 50 12 1.72E-04 5.98E-05 3.11E-06 1.40E-04 1.89E-03 5.88E-14 5.17E-21 Broughton, Riddick 70 12 1.27E-04 3.24E-05 7.23E-06 1.03E-04 1.39E-03 4.98E-08 7.83E-06 Patterson, Ricks 90 12 1.01E-04 2.01E-05 7.71E-06 8.24E-05 1.11E-03 1.51E-06 2.18E-05 DH Hill 150 30 6.39E-05 7.51E-06 6.88E-05 5.20E-05 7.03E-04 1.09E-03 6.96E-04 Cox 175 12 5.53E-05 5.46E-06 6.94E-06 4.50E-05 6.08E-04 7.95E-06 3.42E-05 Dabney 200 24 4.92E-05 4.22E-06 1.64E-05 4.00E-05 5.41E-04 2.59E-04 2.18E-04 Hillsbrgh.St. 200 15 4.92E-05 4.22E-06 8.29E-06 4.00E-05 5.41E-04 2.48E-05 5.78E-05 Talley, Reynolds 200 12 4.92E-05 4.18E-06 6.56E-06 4.00E-05 5.41E-04 1.26E-05 3.88E-05 Carroll, Syme 325 12 3.17E-05 1.59E-06 5.25E-06 2.58E-05 3.49E-04 2.11E-05 4.10E-05 North 350 20 2.63E-05 1.38E-06 5.63E-06 2.14E-05 2.89E-04 8.40E-05 7.61E-05 MAXIMUM 2.73E-04 1.38E-04 6.88E-05 2.22E-04 3.00E-03 1.09E-03 6.96E-04 RELEASE Accident Accident Accident Accident Vented Vented Vented U-235 TEDE analysis notes:

• Eq. 8-4 for different locations at a given exposure time varies by the ratio of {X/Q] values; i.e. a different [X/Q]

is used in Eq. 7-3 or Eq. 7-4 to calculate the time integrated exposure, P.

• P is then used in Eq. 8-4 to calculate the public TEDE.

• To determine the TEDE at a specific location, the maximum TEDE may be multiplied by the ratio of the [X/Q]

used for a specific location under the listed weather conditions at a given exposure time to the maximum [X/Q]

used for the same weather conditions and exposure time.

Maximum rem at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for accidental release using GPM: 6.88x10-5 rem = 2.73x10-4 rem (2.15x10-3 / 8.54x10-3)

Talley, Reynolds at 520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> for vented experiment: 3.88x10-5 rem = 6.96x10-4 rem (5.10x10-6 / 9.15x10-5) 39

Table 14-12: Calculation 5 - Pu-239 radiation doses (TEDE) for other public areas Eq 8-4 Eq 8-4 Eq 8-4 Eq 8-4 Eq 8-4 Eq 8-4 Eq 8-4 Fumign. Calm Wind GPM Fumign. Fumign. GPM GPM Dist. Height TEDE TEDE TEDE TEDE TEDE TEDE TEDE Building or x z 24 h 24 h 24 h 2h 24 h 96 h 520 h Location (m) (m) (rem) (rem) (rem) (rem) (rem) (rem) (rem)

All 30 to 100 up to 12 3.00E-04 1.51E-04 8.10E-06 2.42E-04 2.05E-03 1.33E-07 8.12E-06 All 100 to 150 up to 12 8.64E-06 1.79E-05 8.38E-06 6.97E-06 5.91E-05 2.62E-06 1.94E-05 All 150 to 5000 up to 30 7.03E-05 1.85E-05 7.54E-05 5.67E-05 4.80E-04 7.48E-04 4.76E-04 Withers, Mann 50 12 1.89E-04 6.55E-05 3.41E-06 1.52E-04 1.29E-03 4.03E-14 3.54E-21 Broughton, Riddick 70 12 1.39E-04 3.54E-05 7.93E-06 1.12E-04 9.53E-04 3.42E-08 5.36E-06 Patterson, Ricks 90 12 1.11E-04 2.19E-05 8.45E-06 8.98E-05 7.61E-04 1.04E-06 1.49E-05 DH Hill 150 30 7.03E-05 8.21E-06 7.54E-05 5.67E-05 4.80E-04 7.48E-04 4.76E-04 Cox 175 12 6.08E-05 5.98E-06 7.61E-06 4.90E-05 4.15E-04 5.45E-06 2.34E-05 Dabney 200 24 5.41E-05 4.62E-06 1.80E-05 4.36E-05 3.70E-04 1.78E-04 1.49E-04 Hillsbrgh.St 200 15 5.41E-05 4.62E-06 9.08E-06 4.36E-05 3.70E-04 1.70E-05 3.95E-05 Talley, Reynolds 200 12 5.41E-05 4.58E-06 7.19E-06 4.36E-05 3.70E-04 8.61E-06 2.65E-05 Carroll, Syme 325 12 3.49E-05 1.74E-06 5.75E-06 2.81E-05 2.38E-04 1.45E-05 2.80E-05 North 350 20 2.89E-05 1.51E-06 6.17E-06 2.33E-05 1.98E-04 5.76E-05 5.20E-05 MAXIMUM 3.00E-04 1.51E-04 7.54E-05 2.42E-04 2.05E-03 7.48E-04 4.76E-04 RELEASE Accident Accident Accident Accident Vented Vented Vented CALCULATION 6: Activity for Radionuclides Listed in 10 CFR Part 37 as Quantities of Concern From Calculations 1 through 5, the limiting fission rate for a release from a fueled experiment was determined to be 9.6 x109 f/s. This gives a total number of fissions of 1.8x1016 for an irradiation time of 520 h.

Tables 14-13 and 14-14 below give the 10 CFR Part 37 Category 2 Limit Fractions for U-235 and Pu-239 irradiation at 9.6x109 f/s for 520 h.

At 1.8 x1016 total fissions, activities for radionuclides listed in 10 CFR Part 37 were calculated to be approximately 6.0 x10-7 times the activity limits for Catergory 2 Quantities of Concern. The total activity for Sr-90, Cs-137, and Pm-147 is approximately 123 µCi.

Table 14-13: Calculation 6 - 10 CFR Part 37 Category 2 Limit Fractions for U-235 irradiation at 9.6x109 f/s for 520 h Non- Eq. 2-1 10 CFR Decay Thermal Thermal Part 37 Half-life Constant Fission Fission Activity Category 2 Fraction of Nuclide (s) (1/s) Yield % Yield % (Ci) Limit (Ci) Limit Sr90 9.07E+08 7.65E-10 5.87 5.60 2.15E-05 3.70E+02 5.82E-08 Cs137 9.47E+08 7.32E-10 3.25 3.76 1.19E-05 2.70E+01 4.42E-07 Pm147 8.26E+07 8.39E-09 2.25 2.14 8.98E-05 1.08E+04 8.32E-09 Total Limit Fraction = 5.09E-07 40

Table 14-14: Calculation 6 - 10CFR Part 37 Category 2 Limit Fractions for Pu-239 irradiation at 9.6x109 f/s for 520 h Non- Eq. 2-1 10 CFR Decay Thermal Thermal Part 37 Half-life Constant Fission Fission Activity Category 2 Fraction of Nuclide (s) (1/s) Yield % Yield % (Ci) Limit (Ci) Limit Sr90 9.07E+08 7.65E-10 2.15 2.07 7.92E-06 3.70E+02 2.14E-08 Cs137 9.47E+08 7.32E-10 4.28 4.66 1.55E-05 2.70E+01 5.75E-07 Pm147 8.26E+07 8.39E-09 2.01 1.99 8.10E-05 1.08E+04 7.50E-09 Total Limit Fraction = 6.04E-07 CALCULATION 7: External Dose from the Reactor Building, Overhead Plume, and Reactor Stack Tables 14-15 and 14-16 below give the U-235 and Pu-239 Source Terms at a fission rate of 9.6x109 f/s.

Tables 14-17 through 14-19 give the external dose rates resulting from U-235 at a fission rate of 9.6x109 f/s.

Table 14-20 gives the combined external doses for Pu-239 at a fission rate of 9.6x109 f/s for comparison.

Figures 14-1 and 14-2 illustrate the Microshield model used for the overhead plume and reactor building.

The dose rates from U-235 were calculated to be higher than those from Pu-239 and therefore U-235 is limiting.

External dose from the reactor building, overhead plume, and reactor stack were calculated to give 2.4x10-5 rem or less to publicly occupied areas outside the reactor building. Most of the dose is associated with the reactor building.

Occupants inside Burlington labs would be evacuated within 15 minutes to areas outside the site boundary.

Evacuation time is based on the time for the reactor staff to exit the reactor building and verify personnel within the building have evacuated. This gives a dose of less than 1.4x10-5 rem based on a minimal distance of 1 m to 10 m occupied for 15 minutes.

University personnel notify reactor staff if roof top access is being made. Reactor facility procedures require the reactor staff to clear the roof top if an evacuation alarm occurs. Initial dose rate on the roof top was calculated to be 8.6x10-4 rem/h. Evacuation time is estimated as being less than 15 minutes based on the time for the reactor staff to exit the reactor building and notify personnel on the roof top. This gives a dose of approximately 2.2x10-4 rem or less to roof top occupants.

41

Table 14-15: Calculation 7 - U-235 Source Term at 9.6x109 f/s Eq 5-1 Eq 5-2 Eq 5-2 Eq 5-2 Eq 8-1 Eq 8-2 Eq 5-2 Eq 8-1 Eq 8-2 2h 24 h 2h 2h 24 h 24 h Average Average Average 2h Average Average 24 h Average Initial Reactor Reactor Stack Average Overhead Stack Average Overhead Conc. Conc. Conc. Conc. Stack Line Conc. Stack Line C(0) <C> <C> <C> Line Calm Wind <C> Line Calm Wind Nuclide (µCi/ml) (µCi/ml) (µCi/ml) (µCi/ml) (Ci) (Ci) (µCi/ml) (Ci) (Ci) 83mKr 6.28E-07 3.07E-07 3.17E-08 3.07E-07 1.21E-06 1.74E-05 3.17E-08 1.24E-07 1.79E-06 85mKr 1.50E-06 8.71E-07 1.03E-07 8.71E-07 3.42E-06 4.93E-05 1.03E-07 4.05E-07 5.84E-06 85Kr 3.30E-07 2.17E-07 3.03E-08 2.17E-07 8.51E-07 1.23E-05 3.03E-08 1.19E-07 1.72E-06 87Kr 2.94E-06 1.27E-06 1.23E-07 1.27E-06 5.00E-06 7.22E-05 1.23E-07 4.82E-07 6.96E-06 88Kr 4.06E-06 2.19E-06 2.43E-07 2.19E-06 8.60E-06 1.24E-04 2.43E-07 9.53E-07 1.37E-05 89Kr 5.06E-06 1.85E-07 1.54E-08 1.85E-07 7.27E-07 1.05E-05 1.54E-08 6.06E-08 8.74E-07 131mXe 4.53E-08 2.98E-08 4.15E-09 2.98E-08 1.17E-07 1.69E-06 4.15E-09 1.63E-08 2.35E-07 133mXe 2.20E-07 1.43E-07 1.97E-08 1.43E-07 5.61E-07 8.10E-06 1.97E-08 7.72E-08 1.11E-06 133Xe 7.72E-06 5.06E-06 7.02E-07 5.06E-06 1.98E-05 2.86E-04 7.02E-07 2.75E-06 3.97E-05 135mXe 1.31E-06 2.06E-07 1.72E-08 2.06E-07 8.09E-07 1.17E-05 1.72E-08 6.76E-08 9.75E-07 135Xe 7.55E-06 4.66E-06 5.94E-07 4.66E-06 1.83E-05 2.64E-04 5.94E-07 2.33E-06 3.36E-05 137Xe 7.02E-06 3.10E-07 2.58E-08 3.10E-07 1.22E-06 1.75E-05 2.58E-08 1.01E-07 1.46E-06 138Xe 7.18E-06 1.05E-06 8.79E-08 1.05E-06 4.14E-06 5.97E-05 8.79E-08 3.45E-07 4.98E-06 131I 3.42E-06 2.24E-06 3.12E-07 2.24E-07 8.79E-07 1.27E-05 3.12E-08 1.22E-07 1.77E-06 132I 5.06E-06 2.61E-06 2.79E-07 2.61E-07 1.02E-06 4.93E-06 2.79E-08 1.10E-07 1.58E-06 133I 7.72E-06 4.94E-06 6.61E-07 4.94E-07 1.94E-06 2.80E-05 6.61E-08 2.60E-07 3.75E-06 134I 8.97E-06 3.31E-06 3.01E-07 3.31E-07 1.30E-06 1.87E-05 3.01E-08 1.18E-07 1.70E-06 135I 7.24E-06 4.36E-06 5.40E-07 4.36E-07 1.71E-06 2.47E-05 5.40E-08 2.12E-07 3.06E-06 83Br 6.31E-07 3.29E-07 3.54E-08 3.29E-08 1.29E-07 1.86E-06 3.54E-09 1.39E-08 2.01E-07 84Br 1.13E-06 3.10E-07 2.66E-08 3.10E-08 1.22E-07 1.76E-06 2.66E-09 1.05E-08 1.51E-07 Supporting Calculations C(0) = [A()/V] where A() is taken from Eq. 2-1 as shown in Table 14-1 for U-235 Average Release Concentration (Reference Eq. 5-1 and 5-2):

< C > = C(0) e-ktdt = C(0) [ (1- e-kT) / (kT)]

For Kr-87:

C(0) = 6.63x103 µCi/ 2.25 x109 ml = 2.94x10-6 µCi/ml 2 h <C> = (2.94x10-6 µCi/ml) [(1-exp(-2.77 x10-4/s*7200s)) / (2.77 x10-4/s*7200s)]

= 1.27x10-6 µCi/ml Line Activity (Reference Eq. 8-1 and 8-2)

Stack Ci = 3.93 <C> at 24 h for Kr-85 = (3.93)(3.03x10-8) Ci = 1.19x10-7 Ci Overhead Line Ci = 56.6 <C> at 2 h for Xe-138 = (56.6)(1.05x10-6) Ci = 5.97x10-5 Ci 42

Table 14-16: Calculation 7 - Pu-239 Source Term at 9.6x109 f/s Eq 5-1 Eq 5-2 Eq 5-2 Eq 5-2 Eq 8-1 Eq 8-2 Eq 5-2 Eq 8-1 Eq 8-2 2h 24 h 2h 2h 24 h 24 h Average Average Average 2h Average Average 24 h Average Initial Reactor Reactor Stack Average Overhead Stack Average Overhead Conc. Conc. Conc. Conc. Stack Line Conc. Stack Line C(0) <C> <C> <C> Line Calm Wind <C> Line Calm Wind Nuclide (µCi/ml) (µCi/ml) (µCi/ml) (µCi/ml) (Ci) (Ci) (µCi/ml) (Ci) (Ci) 83mKr 3.47E-07 1.70E-07 1.75E-08 1.70E-07 6.67E-07 9.63E-06 1.75E-08 6.88E-08 9.92E-07 85mKr 6.58E-07 3.81E-07 4.51E-08 3.81E-07 1.49E-06 2.16E-05 4.51E-08 1.77E-07 2.55E-06 85Kr 1.46E-07 9.60E-08 1.34E-08 9.60E-08 3.77E-07 5.43E-06 1.34E-08 5.27E-08 7.60E-07 87Kr 1.15E-06 4.99E-07 4.81E-08 4.99E-07 1.96E-06 2.83E-05 4.81E-08 1.89E-07 2.73E-06 88Kr 1.47E-06 7.94E-07 8.79E-08 7.94E-07 3.11E-06 4.49E-05 8.79E-08 3.45E-07 4.98E-06 89Kr 1.67E-06 6.13E-08 5.11E-09 6.13E-08 2.41E-07 3.47E-06 5.11E-09 2.01E-08 2.89E-07 131mXe 4.90E-08 3.22E-08 4.48E-09 3.22E-08 1.26E-07 1.82E-06 4.48E-09 1.76E-08 2.54E-07 133mXe 2.71E-07 1.76E-07 2.42E-08 1.76E-07 6.91E-07 9.97E-06 2.42E-08 9.49E-08 1.37E-06 133Xe 8.08E-06 5.29E-06 7.34E-07 5.29E-06 2.08E-05 2.99E-04 7.34E-07 2.88E-06 4.16E-05 135mXe 2.19E-06 3.44E-07 2.87E-08 3.44E-07 1.35E-06 1.95E-05 2.87E-08 1.13E-07 1.63E-06 135Xe 8.75E-06 5.40E-06 6.89E-07 5.40E-06 2.12E-05 3.06E-04 6.89E-07 2.70E-06 3.90E-05 137Xe 6.81E-06 3.00E-07 2.50E-08 3.00E-07 1.18E-06 1.70E-05 2.50E-08 9.82E-08 1.42E-06 138Xe 5.83E-06 8.56E-07 7.14E-08 8.56E-07 3.36E-06 4.85E-05 7.14E-08 2.80E-07 4.05E-06 131I 4.45E-06 2.92E-06 4.06E-07 2.92E-07 1.15E-06 1.65E-05 4.06E-08 1.59E-07 2.30E-06 132I 6.19E-06 3.20E-06 3.42E-07 3.20E-07 1.25E-06 1.81E-05 3.42E-08 1.34E-07 1.94E-06 133I 8.02E-06 5.13E-06 6.87E-07 5.13E-07 2.01E-06 2.91E-05 6.87E-08 2.70E-07 3.89E-06 134I 8.46E-06 3.12E-06 2.83E-07 3.12E-07 1.22E-06 1.77E-05 2.83E-08 1.11E-07 1.61E-06 135I 7.41E-06 4.46E-06 5.53E-07 4.46E-07 1.75E-06 2.53E-05 5.53E-08 2.17E-07 3.13E-06 83Br 3.47E-07 1.81E-07 1.95E-08 1.81E-08 7.10E-08 1.03E-06 1.95E-09 7.66E-09 1.10E-07 84Br 5.04E-07 1.39E-07 1.19E-08 1.39E-08 5.45E-08 7.87E-07 1.19E-09 4.68E-09 6.76E-08 Table 14-17: Calculation 7 - External dose rates using Microshield and calculated dose from the reactor building for U-235 at 9.6x109 f/s Distance Initial 2 h TEDE 24 h TEDE 2 h TEDE 24 h TEDE m rem / h rem / h rem / h rem rem 1 5.58E-05 2.03E-05 2.19E-06 4.07E-05 5.25E-05 10 2.39E-05 8.82E-06 9.51E-07 1.76E-05 2.28E-05 20 9.96E-06 3.69E-06 3.96E-07 7.38E-06 9.50E-06 30 5.13E-06 1.91E-06 2.05E-07 3.81E-06 4.92E-06 40 3.06E-06 1.13E-06 1.22E-07 2.26E-06 2.92E-06 50 1.97E-06 7.29E-07 7.86E-08 1.46E-06 1.89E-06 Roof 8.55E-04 Notes:

• 24 h dose at 30 m = 4.92x10-6 rem = (24 h) (2.05x10-7 rem/h)

• The 1 m distance is associated with offices in Burlington labs. 10 m to 30 m are associated with distances to the site boundary. Nearby buildings are located at 30 m to 50 m.

43

Table 14-18: Calculation 7 - External dose rates and dose calculated using Microshield from overhead plume and reactor stack for U-235 at 9.6x109 f/s 2 h Average 2 h Average 24 h Average 24 h Average 2 h Average Overhead Stack and 24 h Average Overhead Stack and Location Stack Line Line Plume Stack Line Line Plume x,y,z TEDE Rate TEDE Rate TEDE TEDE Rate TEDE Rate TEDE (m) (rem / h) (rem / h) (rem) (rem / h) (rem / h) (rem) 30,0,0 1.59E-08 1.92E-07 4.16E-07 1.73E-09 2.09E-08 5.42E-07 40,0,0 1.01E-08 2.01E-07 4.22E-07 1.10E-09 2.18E-08 5.48E-07 50,0,0 6.84E-09 2.03E-07 4.20E-07 7.41E-10 2.21E-08 5.47E-07 10,0,12 1.34E-07 3.12E-07 8.92E-07 1.45E-08 3.39E-08 1.16E-06 20,0,12 4.38E-08 3.66E-07 8.20E-07 4.74E-09 3.96E-08 1.06E-06 30,0,12 2.10E-08 3.93E-07 8.28E-07 2.28E-09 4.26E-08 1.08E-06 40,0,12 1.21E-08 4.05E-07 8.34E-07 1.32E-09 4.41E-08 1.09E-06 50,0,12 7.74E-09 4.11E-07 8.37E-07 8.37E-10 4.44E-08 1.09E-06 Table 14-19: Calculation 7 - Combined external doses (sum of reactor building, overhead plume, and stack) for U-235 at 3.2x109 f/s Location x,y,z External Dose in 2 h External Dose in 24 h (m) TEDE (rem) TEDE (rem) 10,0,12 1.9E-05 2.4E-05 20,0,12 8.2E-06 1.1E-05 30,0,12 4.6E-06 6.0E-06 40,0,12 3.1E-06 4.0E-06 50,0,12 2.3E-06 3.0E-06 Note: 2 h dose at 50 m = 2.3x10-6 rem = (8.34x10-7 + 1.46x10-6) rem Table 14-20: Calculation 7 - Combined external doses (sum of reactor building, overhead plume, and stack) for Pu-239 at 3.2x109 f/s Location x,y,z External Dose in 2 h External Dose in 24 h (m) TEDE (rem) TEDE (rem) 10,0,12 1.6E-05 2.1E-05 20,0,12 6.9E-06 9.0E-06 30,0,12 3.9E-06 5.0E-06 40,0,12 2.5E-06 3.3E-06 50,0,12 1.8E-06 2.4E-06 44

Microshield models used for the overhead plume and reactor building are illustrated below.

Figure 14-1: Rectangular volume geometry (reactor building with shield is shown)

Figure 14-2: Line source geometry (overhead plume is shown) 45

CALCULATION 8: External dose rates from beam tube exhaust duct Tables 14-21 and 14-22 give the beam tube exhaust duct activity and total dose resulting from U-235 with a fission rate of 9.6x109 f/s with a 520 hour0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> exposure.

Tables 14-23 and 14-24 give the beam tube exhaust duct activity and total dose resulting from Pu-239 with a fission rate of 9.6x109 f/s with a 520 hour0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> exposure.

Results for U-235 were slightly higher than those for Pu-239. The total activity in the beam tube exhaust duct at a fission rate of 9.6x109 f/s for a vented experiment is 274 µCi for U-235 and 265 µCi for Pu-239.

External dose rates for occupied areas (3 m or greater) from the beam tube exhaust duct are 7.7x10-6 rem/h or less for U-235 and 7.3x10-6 rem/h or less for Pu-239. The total dose for 520 h exposure is 4.0x10-3 rem or less for U-235 and for Pu-239.

Table 14-21: Calculation 8 - Beam Tube Exhaust Duct Activity resulting from U-235 at 9.6x109 f/s Eq. 5-8 Eq. 8-3 Decayed, Filtered Vented Release Beam Tube Duct Nuclide Rate q, (µCi/s) A(d), (µCi) 83mKr 6.51E-01 2.61E+00 85mKr 1.74E+00 6.96E+00 85Kr 4.12E-01 1.65E+00 87Kr 2.80E+00 1.12E+01 88Kr 4.49E+00 1.80E+01 89Kr 8.59E-03 3.44E-02 131mXe 5.66E-02 2.26E-01 133mXe 2.73E-01 1.09E+00 133Xe 9.63E+00 3.85E+01 135mXe 4.21E-01 1.68E+00 135Xe 9.08E+00 3.63E+01 137Xe 3.80E-02 1.52E-01 138Xe 2.05E+00 8.22E+00 131I 4.26E-01 1.70E+00 132I 5.44E-01 2.17E+00 133I 9.49E-01 3.80E+00 134I 7.55E-01 3.02E+00 135I 8.58E-01 3.43E+00 83Br 6.83E-02 2.73E-01 84Br 7.31E-02 2.93E-01 3.53E+01 1.41E+02 Supporting Calculations:

Ventilation Exhaust Activity (Reference Eq. 5-8 and 8-3):

A(d) = 4 q = 4 [A() / w] p [exp(-w/p)] where t = w/p Kr-87: 2.8 µCi = (4s) (6.63x103 µCi / 9.0x104 ml)(50 ml/s)[exp(-1.52x10-4 x 9.0x104/50)], or 2.8x10-6 Ci 46

Using a line source geometry in Microshield with the source term [A(d)] from Table 14-21 gives the dose rates in Table 14-22:

Table 14-22: Calculation 8 - Beam Tube Exhaust Duct Dose Rates from Microshield and Calculated Dose resulting from U-235 at 9.6x109 f/s and 520 hour0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> exposure TEDE Distance Rate TEDE m (rem/h) (rem) 1 2.80E-05 1.5E-02 2 1.28E-05 6.6E-03 3 7.71E-06 4.0E-03 4 5.25E-06 2.7E-03 5 3.81E-06 2.0E-03 6 2.89E-06 1.5E-03 7 2.27E-06 1.2E-03 8 1.82E-06 9.5E-04 9 1.49E-06 7.7E-04 10 1.24E-06 6.4E-04 Table 14-23: Calculation 8 - Beam Tube Exhaust Duct Activity resulting from Pu-239 at 9.6x109 f/s Eq 5-8 Eq 8-3 Decayed, Filtered Beam Tube Vented Release Duct Nuclide Rate q, (µCi/s) A(d), (µCi) 83mKr 3.60E-01 1.44E+00 85mKr 7.61E-01 3.04E+00 85Kr 1.82E-01 7.30E-01 87Kr 1.10E+00 4.39E+00 88Kr 1.63E+00 6.51E+00 89Kr 2.84E-03 1.14E-02 131mXe 6.12E-02 2.45E-01 133mXe 3.36E-01 1.34E+00 133Xe 1.01E+01 4.03E+01 135mXe 7.02E-01 2.81E+00 135Xe 1.05E+01 4.21E+01 137Xe 3.68E-02 1.47E-01 138Xe 1.67E+00 6.68E+00 131I 5.56E-01 2.22E+00 132I 6.66E-01 2.66E+00 133I 9.86E-01 3.95E+00 134I 7.12E-01 2.85E+00 135I 8.79E-01 3.52E+00 83Br 3.76E-02 1.50E-01 84Br 3.28E-02 1.31E-01 Total 3.13E+01 1.25E+02 47

Table 14-24: Calculation 8 - Beam Tube Exhaust Duct Dose Rates from Microshield and Calculated Dose resulting from Pu-239 at 9.6x109 f/s and 520 hour0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> exposure TEDE Distance Rate TEDE (m) (rem/h) (rem) 1 2.65E-05 1.4E-02 2 1.21E-05 6.3E-03 3 7.30E-06 3.8E-03 4 4.98E-06 2.6E-03 5 3.60E-06 1.9E-03 6 2.74E-06 1.4E-03 7 2.15E-03 1.1E-03 8 1.73E-06 9.0E-04 9 1.41E-06 7.3E-04 10 1.18E-06 6.1E-04 CALCULATION 9: Irradiation of Uranium Table 14-25 gives the parameter values for experiments containing uranium.

Table 14-26 gives the time integrated exposures and radiation doses Table 14-27 gives the dose summary for experiments containing uranium.

At the fluence rates used in this analysis, a fission rate of 1.9x106 f/s is equivalent to 9.8x10-7 g of U-235.

Results indicate 1.9x106 f/s is limiting based on the dose criteria of one percent (1%) of the annual radiation dose limits given in 10 CFR Part 20 for experiments containing uranium.

For experiments containing uranium, a fission rate of 1.9x106 f/s has a TEDE of 1.0x10-3 rem or less to personnel within the reactor building and a TEDE of 1.0x10-5 rem or less to members of the public outside the reactor building. Experiments with uranium equal to or greater than 1.9x106 f/s or 1.6x1011 fissions are therefore defined as a fueled experiment. The total number of fissions is 1.6x1011 based on an accidental 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release that may occur at any time during an irradiation time of 520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />.

48

Table 14-25: Calculation 9 - Parameter values for experiments containing uranium Parameter Value Unit Parameter Value Units Nuclide U-235 Target atoms, N 2.53E+15 atoms Mass 9.86E-07 g Thermal fission rate 1.48E+06 f/s Mass Number, A 235 g/mol Non-thermal fission rate 4.33E+05 f/s Sigma thermal 585 b Total fission rate 1.91E+06 f/s Sigma non-thermal 571 b Reactor volume 2.25E+09 ml X/Q 8.54E-03 s/m3 F normal 0.883 m3/s Thermal flux 1.00E+12 cm2/s v normal 3.92E-04 1/s Non-thermal flux 3.00E+11 cm2/s NG reactor correction 0.1 Irradiation time 520 h Fissions 1.65E11 Public Hours 24 h Reactor exposure time 24 h Eq. 2-1 Half-Life Decay Cumulative Yield % Cumulative Yield % Saturation Nuclide (sec) Constant (1/s) Thermal Fission Non-Thermal Fission Activity (µCi) 83mKr 6.70E+03 1.04E-04 5.36E-01 5.75E-01 3.88E-02 85mKr 1.61E+04 4.30E-05 1.29E+00 1.36E+00 9.30E-02 85Kr 3.39E+08 2.05E-09 2.83E-01 2.96E-01 2.04E-02 87Kr 4.57E+03 1.52E-04 2.56E+00 2.54E+00 1.82E-01 88Kr 1.02E+04 6.78E-05 3.55E+00 3.43E+00 2.51E-01 89Kr 1.89E+02 3.67E-03 4.51E+00 3.97E+00 3.13E-01 131mXe 1.03E+06 6.74E-07 4.05E-02 3.54E-02 2.80E-03 133mXe 1.89E+05 3.66E-06 1.89E-01 1.97E-01 1.36E-02 133Xe 4.53E+05 1.53E-06 6.70E+00 6.71E+00 4.77E-01 135mXe 9.18E+02 7.55E-04 1.10E+00 1.26E+00 8.10E-02 135Xe 3.28E+04 2.12E-05 6.54E+00 6.58E+00 4.67E-01 137Xe 2.29E+02 3.02E-03 6.13E+00 5.98E+00 4.34E-01 138Xe 8.46E+02 8.19E-04 6.30E+00 6.00E+00 4.44E-01 131I 6.93E+05 1.00E-06 2.89E+00 3.22E+00 2.11E-01 132I 8.26E+03 8.39E-05 4.31E+00 4.66E+00 3.13E-01 133I 7.49E+04 9.26E-06 6.70E+00 6.70E+00 4.77E-01 134I 3.16E+03 2.20E-04 7.83E+00 7.63E+00 5.55E-01 135I 2.37E+04 2.93E-05 6.28E+00 6.27E+00 4.47E-01 83Br 8.64E+03 8.02E-05 5.40E-01 5.76E-01 3.90E-02 84Br 1.91E+03 3.63E-04 9.67E-01 1.01E+00 6.96E-02 49

Table 14-26: Calculation 9 - Time Integrated Exposures and Doses from Irradiation of uranium at 1.9 E6 f/s Eq 7-1* Eq 7-5 Eq 8-4 Eq 8-4 Time Time Eq 8-4 rem per - rem per rem per Reactor Public Integrated Integrated Reactor h/ml Ci-h/ml Ci-h/ml Normal Normal Exposure Exposure Normal Normal Normal Effective Thyroid Thyroid Reactor Public Inhalation Inhalation Submersion TEDE TODE TEDE Nuclide Ci-h/ml Ci-h/ml DCF DCF DCF rem rem rem 83mKr 4.14E-10 3.12E-12 1.52E-02 6.28E-13 6.28E-13 4.73E-14 85mKr 9.92E-10 7.48E-12 1.10E+02 1.09E-08 1.09E-08 8.24E-10 85Kr 2.17E-10 1.64E-12 1.74E+00 3.78E-11 3.78E-11 2.85E-12 87Kr 1.94E-09 1.46E-11 5.25E+02 1.02E-07 1.02E-07 7.69E-09 88Kr 2.68E-09 2.02E-11 1.33E+03 3.57E-07 3.57E-07 2.69E-08 89Kr 3.33E-09 2.51E-11 1.20E+03 4.00E-07 4.00E-07 3.02E-08 131mXe 2.99E-11 2.25E-13 5.48E+00 1.64E-11 1.64E-11 1.23E-12 133mXe 1.45E-10 1.09E-12 1.99E+01 2.89E-10 2.89E-10 2.18E-11 133Xe 5.09E-09 3.84E-11 2.25E+01 1.14E-08 1.14E-08 8.62E-10 135mXe 8.64E-10 6.51E-12 2.79E+02 2.41E-08 2.41E-08 1.81E-09 135Xe 4.98E-09 3.75E-11 1.73E+02 8.62E-08 8.62E-08 6.49E-09 137Xe 4.63E-09 3.49E-11 1.10E+02 5.10E-08 5.10E-08 3.84E-09 138Xe 4.73E-09 3.57E-11 7.10E+02 3.36E-07 3.36E-07 2.54E-08 131I 2.25E-09 1.70E-11 3.95E+04 1.30E+06 2.42E+02 8.90E-05 2.92E-03 6.74E-07 132I 3.34E-09 2.51E-11 4.57E+02 7.73E+03 1.49E+03 2.02E-06 2.63E-05 4.90E-08 133I 5.09E-09 3.84E-11 7.02E+03 2.16E+05 3.92E+02 3.59E-05 1.10E-03 2.84E-07 134I 5.91E-09 4.46E-11 1.58E+02 1.28E+03 1.73E+03 1.96E-06 8.59E-06 8.42E-08 135I 4.77E-09 3.60E-11 1.47E+03 3.76E+04 1.06E+03 7.54E-06 1.80E-04 9.12E-08 83Br 4.16E-10 3.14E-12 1.03E+02 5.09E+00 4.33E-08 2.12E-10 3.40E-10 84Br 7.42E-10 5.59E-12 1.01E+02 1.25E+03 1.68E-07 9.30E-08 7.57E-09

• Note: Eq 5-1 is used in Eq 7-1 for <C>, or <C> = C(0) giving = [A() / V] T Table 14-27: Calculation 9 - Dose Summary for experiments containing uranium at 1.9x106 f/s Reactor Building Public Areas Total TEDE, (rem) = 0.001 9.39E-06 TEDE Limit, (rem) = 0.001 0.001 rem dose per f/s= 5.24E-10 4.91E-12 Eq 9-1 Fission rate limit (f/s) = 1.91E+06 2.04E+08

• From Eq. 9-3, the total number of fissions is 1.6x1011 based on an accidental 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release that may occur at any time during an irradiation time of 520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />.

50

CALCULATION 10: Radiation Monitor Setpoint Calculations Setpoint bases Setpoints are based on released activity that exceeds TS 3.8 limits for vented fueled experiments and fueled experiment accidents. Controlled, planned releases from experiments should not cause alarms or annunciation from radiation monitors.

• The vented fueled experiment exhaust gas radiation monitor is used to monitor vented fueled experiments.

• The stack gas and stack particulate radiation monitors are used to monitor fueled experiments and vented fueled experiments.

Monitor response is based on the release concentration at the experiment exhaust and reactor stack.

Released concentration in the vented fueled experiment exhaust:

Using data from Tables 14-1 and 14-2 (Table 14-2 data was adjusted for a fission rate of 9.6x109 f/s), the vented fueled experiment exhaust concentrations were calculated and are given in Table 14-28:

Ci/ml = (Public TEDE in rem / DCF)

• 2.0x104 / [T(X/Q)] Eq 14-1 where, 2.0x104s/m3 = ( 28,317 ml / ft3 ) ( 1870 ft3 / min )

(0.883 m3/s)(3000 ml/min)

DCF, Time integrated exposure, and public dose is from Tables 14-1 and 14-2 e.g. Kr:88 for U-235 fission:8.98x10-2 Ci/ml =

[1.23x10-3 rem/1.33x103rem per Ci-h/ml]( 2 .0x104 s / m3 )

(24h)(8.54x10-4 m3/s)

Vented fueled experiment AEC:

Using data from Tables 14-28 and 10 CFR Part 20 Appendix B, the vented fueled experiment AEC were calculated for individual radionuclides:

AEC fraction = (Ci/ml) / AEC Eq 14-2 7 -2 -10 e.g. I-131 for Pu-239 fission: 5.55x10 AEC fraction = 1.11x10 Ci/ml / [2 x10 Ci/ml per AEC]

AEC Fraction in the vented fueled experiment exhaust:

From Table 14-28 below for vented fueled experiments, the average AEC fraction at the experiment exhaust 7

was calculated to be 1.4 .0x107 AEC for fission gases (average of 1.03x107 and 1.84x10 ).

51

Table 14-28: Calculation 10 - AEC Fractions and Experiment Exhaust Concentrations from Vented Fueled Experiments for Pu-239 and U-235 Eq. 5.8, Eq. 7.4 Eq. 8-4 Eq.8-4 Eq.14-1 Eq. 14-1 Eq.14-2 Eq.14-2 Pu-239 U-235 U-235 Pu-239 Pu-239 U-235 Pu-239 U-235 AEC DCF Public Time Public Public Experiment Experiment Experiment Experiment Integrated Exposures TEDE TEDE Exhaust Exhaust Exhaust Exhaust Nuclide Ci/ml rem per Ci-h/ml Ci-h/ml Ci-h/ml rem rem Ci/ml Ci/ml AEC fractions 83mKr 5.00E-05 1.52E-02 7.38E-08 1.34E-07 2.03E-09 1.12E-09 7.20E-03 1.30E-02 1.44E+02 2.60E+02 85mKr 1.00E-07 1.10E+02 1.56E-07 3.57E-07 3.93E-05 1.72E-05 1.52E-02 3.48E-02 1.52E+05 3.48E+05 85Kr 7.00E-07 1.74E+00 3.74E-08 8.44E-08 1.47E-07 6.50E-08 3.65E-03 8.23E-03 5.21E+03 1.18E+04 87Kr 2.00E-08 5.25E+02 2.25E-07 5.74E-07 3.02E-04 1.18E-04 2.20E-02 5.60E-02 1.10E+06 2.80E+06 88Kr 9.00E-09 1.33E+03 3.34E-07 9.21E-07 1.23E-03 4.44E-04 3.25E-02 8.98E-02 3.62E+06 9.98E+06 89Kr 9.50E-09 1.20E+03 5.83E-10 1.76E-09 2.11E-06 6.99E-07 5.68E-05 1.72E-04 5.98E+03 1.81E+04 131mXe 2.00E-06 5.48E+00 1.25E-08 1.16E-08 6.35E-08 6.87E-08 1.22E-03 1.13E-03 6.12E+02 5.66E+02 133mXe 6.00E-07 1.99E+01 6.89E-08 5.60E-08 1.11E-06 1.37E-06 6.72E-03 5.46E-03 1.12E+04 9.10E+03 133Xe 5.00E-07 2.25E+01 2.06E-06 1.97E-06 4.43E-05 4.64E-05 2.01E-01 1.92E-01 4.03E+05 3.85E+05 135mXe 4.00E-08 2.79E+02 1.44E-07 8.62E-08 2.40E-05 4.01E-05 1.40E-02 8.41E-03 3.51E+05 2.10E+05 135Xe 7.00E-08 1.73E+02 2.16E-06 1.86E-06 3.22E-04 3.74E-04 2.11E-01 1.82E-01 3.01E+06 2.59E+06 137Xe 1.00E-07 1.10E+02 7.54E-09 7.78E-09 8.56E-07 8.30E-07 7.36E-04 7.59E-04 7.36E+03 7.59E+03 138Xe 2.00E-08 7.10E+02 3.42E-07 4.21E-07 2.99E-04 2.43E-04 3.34E-02 4.11E-02 1.67E+06 2.05E+06 131I 2.00E-10 2.42E+02 1.14E-07 8.74E-08 2.12E-05 2.76E-05 1.11E-02 8.52E-03 5.55E+07 4.26E+07 132I 2.00E-08 1.49E+03 1.36E-07 1.11E-07 1.66E-04 2.04E-04 1.33E-02 1.09E-02 6.66E+05 5.43E+05 133I 1.00E-09 3.92E+02 2.02E-07 1.95E-07 7.62E-05 7.92E-05 1.97E-02 1.90E-02 1.97E+07 1.90E+07 134I 6.00E-08 1.73E+03 1.46E-07 1.55E-07 2.68E-04 2.53E-04 1.42E-02 1.51E-02 2.37E+05 2.52E+05 135I 6.00E-09 1.06E+03 1.80E-07 1.76E-07 1.87E-04 1.92E-04 1.76E-02 1.72E-02 2.93E+06 2.86E+06 83Br 9.00E-08 5.09E+00 7.70E-09 1.40E-08 7.12E-08 3.92E-08 7.51E-04 1.37E-03 8.35E+03 1.52E+04 84Br 8.00E-08 1.25E+03 6.72E-09 1.50E-08 1.88E-05 8.42E-06 6.55E-04 1.46E-03 8.19E+03 1.83E+04 Total 6.42E-06 7.24E-06 3.00E-03 2.05E-03 6.26E-01 7.06E-01 8.94E+07 8.37E+07 Gases 5.62E-06 6.49E-06 2.26E-03 1.29E-03 5.49E-01 6.33E-01 1.03E+07 1.84E+07 Halogens 7.93E-07 7.53E-07 7.37E-04 7.63E-04 7.74E-02 7.34E-02 7.91E+07 6.53E+07 Average AEC for vented fueled experiment The average AEC are approximately 4.37 x10-8Ci/ml for fission gases and 1.05x10-9 Ci/ml for halogens.

These are similar to the AEC given in 10 CFR Part 20 Appendix B for Xe-135m and I-133, respectively.

Pu-239 U-235 AEC (Ci/ml) AEC (Ci/ml) 5.31x10-8 3.44x10-8 for fission gases 9.78x10-10 1.13x10-9 for halogens e.g. for fission gases from U-235 fission: 0.633 Ci/ml / 1.84x107 AEC fractions = 3.44x10-8 Ci/ml Release Concentrations in Reactor Stack Vented experiment release concentration in the reactor stack was determined by diluting experiment exhaust concentration given in Table 14-28 with normal exhaust.

Stack Ci/ml = Exhaust Ci/ml ( 3000 ml / min ) Eq 14-3 (28,317 ml/ft3)(1870 ft3/min) 52

Initial accident concentrations for fueled experiments inside the reactor building, C(0), were taken from Tables 14-15 and 14-16.

Table 14-29: Calculation 10 - Initial Accident Concentrations and Concentrations from Vented Fueled Experiments in the Reactor Stack for Pu-239 and U-235 Vented Experiment Accident Release Release Concentration Initial Concentration Eq.14-3 Eq.14-3 C(0) Eq.5-1 C(0) Eq.5-1 Ci/ml Ci/ml Ci/ml Ci/ml Nuclide U-235 Pu-239 U-235 Pu-239 83mKr 7.39E-07 4.09E-07 6.28E-07 3.47E-07 85mKr 1.97E-06 8.63E-07 1.50E-06 6.58E-07 85Kr 4.67E-07 2.07E-07 3.30E-07 1.46E-07 87Kr 3.18E-06 1.25E-06 2.94E-06 1.15E-06 88Kr 5.09E-06 1.85E-06 4.06E-06 1.47E-06 89Kr 9.74E-09 3.22E-09 5.06E-06 1.67E-06 131mXe 6.42E-08 6.94E-08 4.53E-08 4.90E-08 133mXe 3.10E-07 3.81E-07 2.20E-07 2.71E-07 133Xe 1.09E-05 1.14E-05 7.72E-06 8.08E-06 135mXe 4.77E-07 7.96E-07 1.31E-06 2.19E-06 135Xe 1.03E-05 1.19E-05 7.55E-06 8.75E-06 137Xe 4.31E-08 4.17E-08 7.02E-06 6.81E-06 138Xe 2.33E-06 1.89E-06 7.18E-06 5.83E-06 131I 4.83E-07 6.30E-07 3.42E-06 4.45E-06 132I 6.17E-07 7.55E-07 5.06E-06 6.19E-06 133I 1.08E-06 1.12E-06 7.72E-06 8.02E-06 134I 8.56E-07 8.08E-07 8.97E-06 8.46E-06 135I 9.73E-07 9.97E-07 7.24E-06 7.41E-06 83Br 7.74E-08 4.26E-08 6.31E-07 3.47E-07 84Br 8.29E-08 3.72E-08 1.13E-06 5.04E-07 Gases 3.59E-05 3.11E-05 4.56E-05 3.74E-05 Halogens 4.17E-06 4.39E-06 3.42E-05 3.54E-05 Total 4.01E-05 3.55E-05 7.97E-05 7.28E-05 Vented experiment release concentration in the reactor stack:

e.g. Kr-88 for U-235; 5.09 x10-6 Ci/ml = (8.98x10-2 Ci/ml) ( 3000 ml / min )

(28,317 ml/ft )(1870 ft /min) 3 3 From Table 14-29, initial accident concentrations for fueled experiments exceed those for vented fueled experiments in the reactor stack. Therefore, stack gas monitor and stack particulate monitor setpoints based on the release from the vented fueled experiments at the proposed TS 3.8 limits are conservative.

Vented fueled experiment exhaust gas radiation monitor setpoints:

The average experiment exhaust gas concentration of 1.4x107 AEC, or 0.59 Ci/ml, is used as the alarm setpoint and the lower exhaust gas concentration of 1.0x107 AEC, or 0.55 Ci/ml, is used as the alert set point for the vented fueled experiment exhaust radiation monitor.

53

Stack Gas Monitor Setpoints The alarm setpoint for the stack gas radiation monitor given in the 2017 Safety Analysis Report (SAR) submitted for the license renewal is based on an Emergency Action Level of 50 AEC for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This gives approximately 15 mrem in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The alert setpoint is based on 10 mrem in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The limiting radionuclide used is Ar-41, which has an AEC of 1x10-8 Ci/ml.

Using the equations given in Section 11 of the 2017 SAR submitted for license renewal and the [X/Q] value of 8.54 x10-3 s/m3 for occupied areas gives:

50 AEC 6630 AEC = Eq 14-4

[(0.00854 sm3 )(0.883 m3 s)]

10 mrem 4420 AEC = 6631 AEC [ ] Eq 14-5 15 mrem For Ar-41, the setpoint concentrations are approximately 6.63x10-5 Ci/ml for the alarm and 4.42x10-5 Ci/ml for the alert. For vented fueled experiments, the average concentration of fission gases released is approximately3.35x10-5Ci/ml. Fission gas AEC is approximately 4x10-8 Ci/ml, giving an AEC fraction of 837. For vented fueled experiments, the lower concentration of fission gases released is approximately 3 x10-5 Ci/ml giving a AEC fraction of approximately 750. The fission gas release from a vented fueled experiment is lower than the Ar-41 release at the EAL.

Adjustment is made for the stack gas monitor setpoints in terms of Ar-41AEC and the vented fueled experiment AEC fraction:

Alarm setpoint based on Ar-41: 3300 AEC 837 AEC (4x10-8 Ci/ml /1x10-8 Ci/ml) Eq.14-6 Alert setpoint based on Ar-41: 3000 AEC 750 AEC (4x10-8 Ci/ml /1x10-8 Ci/ml) Eq.14-7 At the new alarm setpoint, off-site dose is approximately 1.7 mrem from fission gases:

1.7 mrem = (3.35x10-5 Ci/ml)(0.00854 s/m3)(0.883 m3/s)(24 h/8760 h)(100 mrem /4x10-8 Ci/ml)

Table 14-28 gives an average of 1.8 mrem from fission gases and 0.8 mrem from halogens.

At the new alarm setpoint, the associated Ar-41 radiation dose decreases from 15 mrem to approximately 8 mrem in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in occupied locations; 15 mrem (3300/6630) = 7.5 mrem. EAL dose is still met.

Stack Particulate Radiation Monitor Setpoints:

The alarm setpoint for the stack particulate radiation monitor given in the 2017 Safety Analysis Report (SAR) submitted for the license renewal is based on an Emergency Action Level of 100 AEC for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This gives approximately 15 mrem in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The alert setpoint is based on 10 mrem in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The limiting radionuclide used is Co-60, which has an AEC of 5x10-11 Ci/ml.

54

Using the equations given in Section 11 of the 2017 SAR submitted for license renewal and the [X/Q] value of 8.54 x10-3 s/m3 for occupied areas gives:

100 AEC 13,260 AEC = Eq 14-8

[(0.00854 sm3 )(0.883 m3 s)]

10 mrem 8840 AEC ~ 13,260 AEC[ ] Eq 14-9 15 mrem For Co-60, the setpoint concentrations are approximately 6.63x10-7 Ci/ml for the alarm and 4.42x10-7 Ci/ml for the alert. For vented fueled experiments, the average concentration of fission halogen released is approximately 4.28x10-6 Ci/ml. Fission halogen AEC is approximately 1x10-9 Ci/ml, giving an AEC fraction of 4280. For vented fueled experiments, the lower concentration of fission halogens released is approximately 4 x10-6 Ci/ml giving an AEC fraction of approximately 4000.

Co-60 release concentration is lower than that expected for halogens from a vented fueled experiment. At the concentration of 6.63x10-7Ci/ml for the alarm and 4.42x10-7Ci/ml for the alert, the fission halogen AEC fractions are 663 at the alarm setpoint and 442 at the alert setpoint. These are approximately 15 percent of those estimated based on halogens released from a vented fueled experiment.

A halogen release of 1 percent rather than 10 percent is used for the setpoint calculation assuming two filters are used in series with 90% removal efficiency each; this gives 0.01 = (0.1)(0.1). One filter is used for treatment and a second in-line filter is used for air sampling. Using 1 percent rather than 10 percent conservatively predicts a release of halogens from the vented fueled experiment.

Adjustment is made for the stack gas monitor setpoints in terms of Co-60 AEC and the vented fueled experiment AEC fraction:

Alarm setpoint based on Co-60 : 8560 AEC 4280 AEC (1x10-9Ci/ml /5x10-11Ci/ml)(0.01/0.1) Eq.14-10 Alert setpoint based on Co-60 : 8000 AEC 4000 AEC (1x10-9Ci/ml /5x10-11Ci/ml)(0.01/0.1) Eq.14-11 At the new alarm setpoint, the associated Co-60 radiation dose decreases from 15 mrem to approximately 10 mrem in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in occupied locations; 15 mrem (8560/13260) = 9.7 mrem. EAL criteria are still met.

Current Setpoints in TS:

AEC Monitor Nuclide (Ci/ml) Alarm Alert Units Stack Gas Ar-41 1x10-8 5000 1000 AEC Fraction

-11 Stack Particulate Co-60 5x10 5000 1000 AEC Fraction Alarm is based on EAL. Alert is based on detection of abnormal releases.

55

Setpoints using the revised [X/Q] of 8.54x10-3 s/m3 for occupied areas:

Eq.14-4,14-8 Eq.14-5,14-9 Monitor Nuclide AEC(Ci/ml) Alarm Alert Units

-8 Stack Gas Ar-41 1x10 6630 4420 AEC Fraction Stack Particulate Co-60 5x10-11 13,260 8840 AEC Fraction Alarm based on EAL. Alert based on 10 CFR Part 20 constraint dose.

Based on the setpoints calculated, TS 3.5 is revised for the following:

Eq.14-6,14-10 Eq.14-7, 14-11 Monitor Nuclide AEC (Ci/ml) Alarm Alert Units Vent set points Fission gas 4x10-8 1.4x107 1.0x107 AEC Fraction

-8 Stack Gas Ar-41 1x10 3300 3000 AEC Fraction

-11 Stack Particulate Co-60 5x10 8500 8000 AEC Fraction Alarm is based on vented fueled experiment release at TS 3.8 limits. Alert is based on lower release activity from a vented fueled experiment. New alarm setpoints are lower and therefore continue to meet EAL requirements.

SECTION 15: CONCLUSIONS Table 15-1 below provides a summary of calculated radiation doses inside and outside of the reactor building resulting from planned vented and accidental releases for fueled experiments performed at the limiting conditions given in the requested technical specifications.

Given the low limits requested for fission rates and total fissions, any doses arising from accidental or planned vented releases from fueled experiments will be more than an order of magnitude below 10 CFR 20 limits.

Table 15-1: Radiation Doses for Fueled Experiments for planned vented and accidental releases, as compared to 10 CFR Part 20 Limits Reactor Bldg. Reactor Bldg.

TEDE Thyroid Public TEDE Release Nuclide (rem) TODE (rem) (rem)

Vented U-235 0.004 0.003 Vented Pu-239 0.0038 0.002 Accidental U-235 0.02 0.62 0.00027 Accidental Pu-239 0.025 0.75 0.0003 Dose Limits for Fueled Experiments:: 0.15 1.5 0.003 10 CFR Part 20 Limits: 5.0 (1) 50 (2) 0.1 (3)

(1): TEDE Annual Limit for occupationally exposed radiation worker.

(2): Thyroid TODE for occupational radiation worker (3): TEDE Annual Limit for member of the public.

56

Fueled experiments, as analyzed, meet the following criteria:

• As given in Table 15-1 above, radiation dose does not exceed:

- TEDE of 0.15 rem to occupants inside the reactor building

- Thyroid TODE of 1.5 rem to occupants inside the reactor building

- TEDE of 0.003 rem in public areas outside the reactor building

• Emergency action levels as defined in the facility emergency plan are not exceeded.

• Limits for a reportable event as defined in TS 1.2.24 b are not exceeded.

• Fissionable materials are stored as required by TS 5.3 and the facility security plan and radiation protection program.

Based on the analysis for fueled experiments for the assumed conditions and radiation dose criteria, the following conclusions are made:

• The fission rate for fueled experiments is limited to 9.6x109 fissions per second.

• The total number of fissions for fueled experiments is limited to 1.8x1016 fissions.

• Individual or mixtures of fissionable material meeting the limiting fission rate and total number of fissions may be used in a fueled experiment.

• Fissionable material mass and experiment fluence rates may be adjusted such that the limiting fission rate is met.

• Uranium with a fission rate greater than 1.9x106 fissions per second or total number of fissions greater than 1.6x1011 fissions is defined as a fueled experiment.

• Materials containing Neptunium or Plutonium that undergo fission in an experiment are considered to be fueled experiments.

The following additional controls will be implemented for vented experiments:

• Filtration of particulates and halogens is required. Filters with rated retention greater than 95 percent for particulates and halogens are to be used. Retention for halogens was assumed to be 90 percent.

• A minimum decay time of 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (30 minutes) before entering the reactor building ventilation system.

• The maximum experiment exhaust rate is 3 liters per minute.

• The experiment exhaust is well mixed prior to release into the reactor building ventilation system.

• Experiment exhaust flow rate is routed to the ventilation system using the beam tube exhaust and controlled by dedicated equipment with local flow rate indication.

• The experiment exhaust is capable of being isolated.

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• The exhaust flow tubing from the experiment to the beam tube exhaust is sealed to prevent leakage into the reactor building free air space.

• Radiation monitoring and flow rate monitoring of the experiment exhaust prior to being routed to the reactor building ventilation system is required to identify the source of the release. The release is monitored for radioactivity with indication locally and in the control room. Local alarm annunciation and alarm indication in the control room is provided.

• Airborne activity monitors in the radiation monitoring system initiate the confinement system and isolate the vented fueled experiment exhaust if an alarm setpoint is reached.

SECTION 16: REFERENCES

1. ANSI/ANS 15.7, Research Reactor Site Evaluation

2. US NRC NUREG 1400, Air Sampling in the Workplace

3. US NRC Regulatory Guide 2.2, Development for Technical Specifications for Experiments in Research Reactors

4. North Carolina State PULSTAR Nuclear Reactor Final Safety Analysis Report

5. North Carolina State PULSTAR Nuclear Reactor Nuclear Services (fluence rate data)

6. North Carolina State PULSTAR Nuclear Reactor Emergency Plan

7. ANSI/ANS 15.16, Emergency Planning for Research Reactors

8. Health Physics Journal, 27, 153, G. Chabot et. al. 1974, A Simple Formula for Estimation of Surface Dose from Photons Emitted from a Finite Cloud

9. International Commission on Radiation Protection, Publication 30, Limits for Intakes of Radionuclides by Workers

10. NUREG 1572, Safety Evaluation Report related to the renewal of the operating license for the research reactor at North Carolina State University

11. North Carolina State PULSTAR Nuclear Reactor Technical Specifications

12. Radiological Assessment: Sources and Doses, American Nuclear Society, R. Faw, JK Shultis, 1999

13. Evaluation and Compilation of Fission Product Yields, T.R. England and B.F. Rider, Los Alamos National Laboratory, October, 1994, LA-UR 94-3106 ENDF 349

14. Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA)

Joint Evaluated Fission and Fusion Project Report 20 (JEFF 3.1-3.1.1 Radioactive Decay Data and Fission Yield Sub-Library)

15. National Nuclear Data Center, Brookhaven National Laboratory, Evaluated Nuclear Data Files (ENDF libraries)

16. OECD NEA Joint Evaluated Fission and Fusion Project Report 21 (JEFF 3.1-Nuclear Data Library)

17. Japan Atomic Energy Agency Nuclear Data Center Tables of Nuclear Data (JENDL data)

18. Nuclear Analysis 1.0 Users Manual, Vilece Consulting, 1996

19. Handbook of Health Physics and Radiological Health, Third edition, Shleien et al, Williams and Wilkins, 1998

20. US Nuclear Regulatory Commission, NUREG 1887 RASCAL 3.0.5: Description of Models and Methods.

21. 10 CFR Part 20, Standards for Protection Against Radiation

22. Aspelin, Karen, Establishing Pedestrian Walking Speeds, Portland State University, 2005

23. Study Compares Older and Younger Pedestrian Walking Speeds, TranSafety, Inc. 1997

24. 10 CFR Part 37, Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material

25. Introduction to Nuclear Engineering, J LaMarsh, Addison Wesley, Second edition, 1986 58

26. National Institute of Standards and Measurements, Tables of X-Ray Mass Attenuation Coefficients and Mass Energy-Absorption Coefficients from 1 keV to 20 MeV for Elements Z = 1 to 92 and 48 Additional Substances of Dosimetric Interest

27. US Nuclear Regulatory Commission Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water Cooled Reactors

28. ASTM E261, Standard Methods for Determining Neutron Fluence, Fluence Rate, and and Spectra by Radioactivation Techniques

29. Federal Guidance Report 11, EPA Report 520/1-88-020 Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Submersion, Inhalation, and Ingestion

30. Federal Guidance Report 12, EPA 402-R-93-081 External Exposure to Radionuclides in Air, Water, and Soil

31. EPA 400-R-92001 Manual of Protective Action Guides and Protective Actions for Nuclear Incidents

32. 10 CFR 50, Domestic Licensing of Production and Utilization Facilities

33. Microshield Manual, Grove Engineering

34. US NRC Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants

35. Meteorology and Atomic Energy 1968, Air Resources Laboratory, Environmental Research Laboratories, US Department of Commerce, D.H. Slade ed.

36. Air Stagnation Climatology for the United States (1948-1998), Julian X.L. Wang and James K. Angell, National Oceanic and Atmospheric Administration, Air Resources Laboratory, Environmental Research Laboratories, 1999 59