Text

Oregon State Triga Reactor License Amendment for Irradiation of Fuel Bearing Targets for Production of Molybdenum-99
	ML12124A266
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Site:	Oregon State University
Issue date:	04/30/2012
From:	; Oregon State University
To:	; Office of Nuclear Reactor Regulation
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Radiation Center Oregon State University, 100 Radiation Center, Corvallis, Oregon 97331-5903 T 541-737-2341 I F 541-737-0480 I http://ne.oregonstate.edu/facilities/radiationcenter Oregon State UNIVERSITY January 8, 2013 Mr. Alexander Adams U. S. Nuclear Regulatory Commission Research and Test Reactors Branch A Office of Nuclear Reactor Regulation Mail Stop 012-G13 One White Flint North 11545 Rockville Pike Rockville, MD 20852-2738

Reference:

Oregon State University TRIGA Reactor (OSTR)

Docket No. 50-243, License No. R-106 Affidavit Letter dated April 13, 2012 License Amendment Letter dated April 13, 2012

Subject:

Replacement redacted version of License amendment application for the purpose of demonstrating 99Mo production capability in the OSTR Mr. Adams:

This letter serves as a request to completely replace the redacted version of the license amendment application submitted April 13, 2012, with the attached version. Some of the information in the licence amendment will be proprietary in nature and we requested that the information be withheld from public disclosure per 10 CFR 2.390(a)(4). An affidavit attesting to that affect was submitted and is still applicable to all the redacted information, with the exception of the second to last sentence of the first paragraph of section V.V of the amendment application. We request that the information in that specific sentence be withheld from public disclosure per 10 CFR 2.390(d)(1).

I hereby affirm, state, and declare under penalty of perjury that the foregoing is true and correct.

Executed on: i!

If you have any questions, please do not hesitate to contact me.

S~incere y, Stev eese Director cc: Document Control, NRC w/o attachements Rich Holdren, OSU Craig Bassett, NRC w/o attachements Andy Klein, OSU w/o attachements Rick Spinrad, OSU

Oregon State TRIGA Reactor License Amendment For Irradiation of Fuel Bearing Targets For Production of Molybdenum-99 Docket Number 50-243 License Number R-106 Submitted by the Oregon State University Radiation Center Oregon State University, Corvallis, OR 97333 April, 2012 1

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Table of Contents I S c o p e ................................................................................................... *................................................... 8 II Target Physical Description .......................................................................................................... 8 III Neutronic Analysis .............................................................................................................................. 11 IV Therm al Hydraulic Analysis ................................................................................................................. 15 V Accident Analysis ................................................................................................................................. 18 V.1 The Maxim um Hypothetical Accident (MHA) ........................................................................... 18 V.11 Target MHA SourceTerm ......................................................................................... 20 V.111 Calculation of Occupational and General Public Dose From Target MHA ................................ 26 V.IV Insertion of Excess Reactivity ..................................................................................................... 32 V.V Loss of Coolant Accident (LOCA) ................................................................................................ 33 V.VI Loss of Coolant Flow ....................................................................................................................... 33 V.VII Mishandling or Malfunction of Fuel ......................................................................................... 34 V.VIII Experim ent Malfunction ................................................................................................................ 34 V.IX Loss of Norm al Electrical Pow er ................................................................................................ 35 V.X External Events ............................................................................................................................... 35 V.XI Mishandling or Malfunction of Equipm ent .............................................................................. 35 VI Limiting Conditions for Operation with fuel bearing targets present in the OSTR Core ................ 35 VII Target Transportation and Storage ................................................................................................ 39 VIII Target Loading and Operation Plan ................................................................................................ 39 IX Conclusion ........................................................................................................................................... 40 Appendix A Steady State Therm al Hydraulic Analysis of the Targets ................................................. 41 1 INTRODUCTION ................................................................................................................................... 47 1.1 Purpose .......................................................................................................................................... 48 2 DESCRIPTIO N OF REACTO RAND SYSTEM ....................................................................................... 49 2.1 Fuel Description ............................................................................................................................. 55 2.1.1 Therm al Conductivity .................................................................................................... 57 2.1.2 Specific Heat ........................................................................................................................ 61 3 RELAP5-3D MODEL ................... 6.........................

64 3.1 Overview ........................................................................................................................................ 64 3.2 Model Description .......................................................................................................................... 65 3

3.3 AECL Groeneveld Look-up Tables ............................................................................................. 76 3.4 The Bernath Correlation ................................................................................................................. 77 4 RESULTS ............................................................................................................................................... 79 5 APPENDIX (AIR COOLING EVENT) ................................................................................................... 87 Appendix B Target Drawings ..................................................................................................................... 92 4

List of Tables Table 1, Summary of target design parameters ..................................................................................... 11 T able 2, T arget Reactivity W orth ............................................................................................................... 14 Table 3, Estimated Core Excess and Shutdown Margin with targets installed ..................................... 14 Table 4, Release Fraction Com ponents .................................................................................................. 23 Table 5, Airborne Radioactive Material Source Term ............................................................................ 24 Table 6, Lateral and Vertical Diffusion Coefficients and x/Q Values for Pasquill F and ...................... 27 Table 7, Occupational Radiation Doses in the Reactor Room Following a Single Target Failure at End of B om b ard m en t .............................................................................................................................................. 30 Table 8, Radiation Doses to Members of the General Public Following a Single Target Failure in Air at End of Bom bardm ent- Scenario A ............................................................................................. ....31 Table 9, Radiation Doses to Members of the General Public Following a Single Target Failure in Air at End of Bom bardm ent - Scenario B ....................................................................................................... 31 Table 10, Radiation Doses to Members of the General Public Following a Single Target Failure in Air at End of Bom bardm ent - Scenario C ....................................................................................................... 32 5

List of Figures Figure 1, Target Design ............................................................................................................................... 10 Figure 2, Power per elem ent and target ................................................................................................ 13 Figure 3, Target tem perature distribution vs. elevation ........................................................................ 16 Figure 4, Target DNBR vs. elevation ....................................................................................................... 16 Figure 5, Target M DNBR vs. target power ............................................................................................ 17 6

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I Scope This document contains documentation of the analysis performed in support of the irradiation of fuel bearing targets in the Oregon State TRIGA Reactor (OSTR). This amendment will allow simultaneous irradiation of up to three fuel bearing targets in G-ring positions 32/33/34 of the OSTR. A physical description of the target is given in section II. Neutronic analysis is discussed in section III and is used to determine the total power and power distribution within the highest power target. Thermal Hydraulic analysis is discussed in section IV and is used to determine the maximum temperature within the hottest target, and also the minimum departure from nucleate boiling ratio for the hottest target. Accident analysis is discussed in section V and the consequences of releasing the contents of the limiting target are shown to be acceptable. Required Technical Specification changes are listed in Section VI.

Target transportation and storage requirements are discussed in section VII. Guidelines for operation with targets in-core are given in section VIII.

II Target Physical Description Shown in Figure 1, a novel container has been designed that can be described (hereafter referred to as the "target").

uranium enrichment of no more than 19.75%. The The volume of the target that will contain the fuel is configured Dimensions of the actual height of the cladding. Based upon published values 8

used for both the neutronics and thermal hydraulic analysis. 1 Design quantities are summarized in Table 1.

Upper and lower end fittings are welded to the top and bottom of the using the standard TRIGA fuel handling I

9

Figure 1, Target Design 10

Table 1. Summarv of Target Design Parameters Parameter [Nominal Design Value Uranium Enrichment[%

Fuel Type Fueled volume [cmA3]

Fuel region height [cm]

Core centerline (reference) [cm] 0.000 Bottom of fueled region [cm]

Top of fueled region [cm]

Cladding material Aluminum Cladding thickness [cm] 0.320 III Neutronic Analysis Neutronic analysis was performed with MCNP5 2. The goal of the analysis was to calculate total power per element (fuel and targets) throughout the core for a typical core configuration, and then to calculate the power distribution throughout the hottest target. Target reactivity worth was also calculated. Global core parameters such as neutron lifetime, delayed neutron fraction, temperature coefficient of reactivity, etc. were not calculated. Since targets will comprise no more than a small fraction of the core (<3.5% by number for any configuration),

values for global parameters were assumed to remain unchanged from values determined in the OSU conversion safety analysis report (CSAR 3). Targets were modeled in MCNP5 as fuel and cladding, including the washers and weld regions of the cladding. Specific details of the target end pieces were not included in the MCNP5 models.

11

The analysis methodology used for the neutronic analysis was the same as that used in the CSAR. The MCNP5 input deck used was taken directly from that used for the beginning-of-life NORMAL core from the CSAR. The only changes made involved differences in G-ring positions to represent the current core configuration. These changes include:

" replacement of the graphite reflector elements in positions G32, G33, and G34 with the targets;

" replacement of the graphite reflector element in position GI15 with a fuel element;

" replacement of the graphite reflector element in position G14 with an air-filled aluminum tube representing the G-Ring In-Core Irradiation Tube;

" replacement of the graphite reflector element in G2 with an air-filled aluminum tube representing the pneumatic transfer system terminus; and,

" replacement of the graphite reflector elements in G1 and G3 with water.

Power in each element and target was determined by calculating the reaction rate and normalizing the total fission rate to a power of 1.1 MWth. The modeled core configuration for the target irradiation is shown in Figure 2. Maximum fuel element and target powers in this configuration were calculated to be

KW and

KW, respectively. It should be noted that the i KW is less than the maximum fuel element power of 18.52 KW evaluated in the CSAR. Adding targets to the periphery of the core has the effect of diluting power in other fuel elements, that is average and maximum power in a fuel element tends to decrease as targets are added to the core because some fraction of reactor power is generated in the targets themselves. The addition of targets will thus not lower any MDNBR values calculated in the CSAR. The thermal hydraulic :studies concentrated on the targets, as discussed in the next section.

The configuration shown in Figure 2 indicates that maximum target power is produced for a target in position G34 when three targets are present in the core. Preliminary analysis also showed that if the G-Ring were completely filled with target assemblies, the hottest target would still be located in the G34 position. Analysis of a core configuration containing only a single target in position G34 was also performed. The single target core had the same configuration as the core shown in Figure 2, except that positions G32 and G33 contained graphite reflector 12

elements. The total power in the target was calculated to be kW. These analyses show that irradiation of one, two or three targets in G-ring positions G32, G33 and G34 in any core configuration will be acceptable since fuel element power will be less than element power upper limits identified in the CSAR, and target power will be far below power levels that would result in target MDNBR less than 2.0.

Figure 2, Power per element and target The reactivity worth of each target was also calculated using MCNP5 by computing the value of k-effective with a target present and then re-computing k-effective with the target replaced by graphite. Calculated target reactivity worth is shown in Table 2. Adding a target in position G-32, G-33, or G-34 adds positive reactivity.

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i I Target Location G32 G33 G34 Total combined worth* 1 11

Combined worth is less than the sum of individual worths due to shadowing effects.

Based on prior experience with adding fuel elements, addition of targets to the G-ring will not significantly affect control rod worth. Control rod worth will be measured after installation of the targets. Core excess and NRC shutdown margin will be measured and verified to be within the limits allowed by technical specifications before full power target irradiation is allowed. The values of core excess and NRC shutdown margin for the three current core configurations are shown in Table 3. Estimated values for the same cores containing three targets each is estimated by subtracting M from the NRC shutdown margin values and adding ll to the core excess values. Addition of positive reactivity value targets decreases shutdown margin and increases core excess. All values fall within the technical specification limits for core excess and shutdown margin.

Table 3, Estimated Core Excess and Shutdown Margin with targets installed Core Total Rod NRC SDM : Core Excess NRC SDM Core Excess Configuration Worth (no targets) (no targets) (targets installed) (targets installed)

(no targets)

Normal $8.18 $1.62 $6.56 ICIT $7.41 $3.31 $4.10 CLICIT $8.03 $2.18 $5.85 MCNP5 was also used to perform criticality analysis of un-irradiated targets stored in the in-tank storage racks. Three targets were placed in water with a pitch of zero. The k-effective value of this configuration was Three targets placed in water with a pitch of 2.0 inches resulted in a k-effective value of . Both of these configurations are less than the technical specification limit of 0.9 and thus three targets can be safely stored in existing facilities.

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IV Thermal Hydraulic Analysis Thermal Hydraulic analysis was performed with RELAP5-3D. 4 Comprehensive details of the thermal hydraulic analysis are included in the document entitled Analysis of the Steady State Thermlal Hydraulic Behavior of the Oregon State TRIGA- Reactor Molybdenum Production Element that is presented in Appendix A. Results are summarized in this section.

md odel was created to analyze the targets The target analyzed was assumed to have the minimum sized that exists in the core. The target was assumed to have an inlet coolant temperature of 49'C and an outlet pressure equivalent to a water head of 14 feet above the core, consistent with technical specification limits. The target analyzed had a total power of kW with the power distribution as determined by prior MCNP analysis. All necessary model parameters are specified in the thermal hydraulic report.

Calculation results provided all parameters needed to determine coolant exit temperature, maximum fuel temperature, maximum cladding temperature and minimum-departure-from-nucleate-boiling-ratio (MDNBR). Critical heat flux used to determine MDNBR was calculated using both the Bemath and the Groeneveld 2006 correlations. These two correlations were used because Bernath has been used extensively in past analyses while Groeneveld provides more modem predictions. The maximum fuel temperature predicted in the hottest target is 137°C.

The maximum cladding temperature is 11 5°C and occurs slightly above core centerline on the inner cladding layer. Temperature distributions are shown in Figure 3. MDNBR values are also located slightly above core centerline. The lowest values of MDNBR for the ý target are 9.733 (, Bemath correlation) and 11.425 (. Bemath correlation). A plot of DNBR vs. elevation for the kW target is shown in Figure 4. Both MDNBR values are large and indicate a large margin before CHF conditions are reached.

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-1 Figure 3, Target temperature distribution vs. elevation Figure 4, Target DNBR vs. elevation 16

If target power is increased, temperature in the fuel, cladding and coolant would also increase causing a reduction in MDNBR. A plot of MDNBR vs. target power is shown in Figure

5. If target power is increased to 20 kW, the minimum value of MDNBR is still -5.1. All results of the thermal hydraulic analysis indicate that operating with targets in the G-ring will not result in unsafe temperature or MDNBR values.

31 2

Figure 5, Target MDNBR vs. target power 17

V Accident Analysis Nine credible accidents for research reactors were identified in NUREG-1537 5. These accidents are:

" The Maximum Hypothetical Accident (MHA)

Insertion of excess reactivity

" Loss of coolant accident (LOCA)

" Loss of coolant flow

Mishandling or malfunction of fuel

Experiment malfunction

" Loss of normal electrical power

" External events

Mishandling or malfunction of equipment The consequences of operating the OSTR with up to three targets present in the core are discussed in this section. Operating with targets present in the core does not increase the likelihood or consequences of any of these accidents beyond acceptable limits. Operating with targets present in the core also does not introduce the possibility of new accidents.

V.1 The Maximum Hypothetical Accident (MHA)

For the target, the proposed maximum hypothetical accident (MHA) is defined as the cladding rupture of one target irradiated continuously at full power for 365 days followed by the instantaneous release of volatile fission products outside the cladding and into the air or water.

Using the methodology of the CSAR, the limiting or bounding potential radiation doses to the reactor staff and to the general public in the unrestricted area were calculated, both with and without the presence of primary water, for three scenarios. These three scenarios were:

- Scenario A:

In this scenario, the entire north wall of the reactor room instantly vanishes.

No credible cause for this occurrence can be imagined. The noble gas and halogen fission products that have been released to the reactor room air are 18

assumed to mix instantly and uniformly with the room air. This reactor room air then moves out through the missing wall at the mean wind speed (1 m s-).

This is assumed to be a ground level release. It takes 8.52 seconds for the entire volume of the reactor room air to be evacuated. Thus, individuals outside the reactor room will be exposed to a radioactive cloud for a period of 8.52 seconds; Scenario B:

This scenario again assumes that the noble gas and halogen fission products instantly and uniformly mix with the reactor room air. The fission products that have been released to the reactor room air are then exhausted at the stack ventilation rate (4.39E6 cm 3 s-1). The path for this release is not specified. The air is assumed to be discharged at ground level, and no credit is taken for an elevated release. The time to evacuate the entire volume of the reactor room is 14.7 minutes, and this is, therefore, the exposure time for individuals outside the reactor room; and Scenario C:

This scenario also assumes that the noble gas and halogen fission products instantly and uniformly mix with the reactor room air. The reactor room air then leaks from the room at the leak rate of 1.69E4 cm3 sl as specified in the OSTR CSAR. The leakage from the room is through the walls brought about by a pressure differential between the room and outside. This pressure differential was assumed to arise through the unlikely combination of a drop in atmospheric pressure of 1.5" Hg and an increase in room temperature of 400 C in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Using the ideal gas law, PIV 1 /Tl= P2V2/T2, where temperature is in K, pressure is in inches of Hg, and volume is in cubic centimeters. The initial conditions of T 1=295.2 K (22.2 'C) and P 1=29.4" Hg in the RX bay of volume 3.74E4 cm 3 , the increase in temperature of 40 0 C would result in the pressure increasing to 33.4" Hg. The atmospheric drop of 1.5" Hg would mean an increase in room pressure to 34.9" Hg and assuming an outside temperature 19

of 8.7°C, the resulting bay volume would be 3.01E9 cm3. The leak rate is determined by the difference in the initial and final bay volumes which is 7.3E8 cm3 divided by 12 h or 43200 s to equal 1.69E4 cm 3 s-1. In this case, it would take 63.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months for the entire volume of the reactor room air to be evacuated, and this is the exposure time for individuals outside the reactor room. This is also assumed to be a ground level release.

V.11 Target MHA Source Term The source term for this calculation is identical to the methodology of the CSAR, except that it is driven by the target power of KW and that the release fraction of gases from the Mwas uniquely estimated. Just as in the CSAR, radioisotopes from two elements of the noble gases (krypton and xenon) and halogens (bromine and iodine) were selected, and the activities were calculated assuming a 365 day continuous operation. Data on each radioisotopes cumulative fission yield and half-live are the same as that used in the CSAR.

Once the fission products are released to the gap, this activity is available to be released when the cladding fails. If the release is in air (MHA), then this activity is released directly into the reactor room air. If the release occurs in the pool water, then the fission products must migrate through the water before being released to the reactor room air. Once released into the reactor room air, a further reduction of the halogen activity is expected to occur due to plate-out on the building surfaces.

Thus, the fraction (w) of the fission product inventory released from a single fuel element which reaches the reactor room air and, subsequently, the atmosphere in the unrestricted environment is:

w=efgh; where:

e = the fraction released from the fuel to the fuel-cladding gap; f-= the fraction released from the fuel-cladding gap to the reactor room air 20

(if no water is present), or to the pool water (if water is present);

g = the fraction released from the pool water to the reactor room air; and h = the fraction released from the reactor room air to the outside unrestricted environment, due to plate-out in the reactor room.

Because the form of the material is not TRIGA fuel, the release fraction from the target material to the target

material-cladding gap (the value for "e" above) will be different and unique for each isotope. Fission product release from the

was estimated by modeling the diffusion of gaseous fission products using the 21

Results for the isotope specific are given in Table 4.

Available information on release rates and plate-out vary significantly. Although it is understood that radioiodines will readily plate-out on surfaces, particularly wet surfaces, specific information related to research reactors is limited. NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, describes a LOCA at a spent fuel pool accident and recommends a release of 3% of the volatile fission products to the atmosphere but this is likely not applicable because of the differences in fuel form, temperature and bum-up. However, NUREG-1465 does estimate the scrubbing potential of water over the source term at 9

approximately 10%, or conversely that 90% of the available inventory is retained in the water9.

This is consistent with the values for median overlaying water pool scrubbing factors for reactor accidents described in NUREG/CR-5747, Estimate of Radionuclide Release CharacteristicsInto Containment Under Severe Accident Conditions, which ranged from 5 (20% or conversely 80%

retained) to 30 (3% or conversely 97% retained) although the values are likely higher compared to the conditions at the OSTR (expect higher retention factors for the OSTR) considering the energy and temperature of the scenarios.1 0 Regulatory Guide 3.33, Assumptions Usedfor Evaluating the PotentialRadiological Consequences ofAccidental Nuclear Criticality in a Fuel ReprocessingPlant, describes a criticality accident in solution and recommends a release rate of 100% for noble gases and 25% release of radioiodines but does not specifically address plate-out of iodine other than it should be handled on a case-by-case basis." IAEA Safety Series No. 53, Derivationof the Source Term andAnalysis of the Radiological Consequences of Research Reactor Accidents, describe release and transfer factors of halogens and noble gases for water-to-air that range from 0.5% to 0.05% and 0.5% to 100%, respectively. It also models the plate-out of radioiodines with deposition rate constants that are a function of material surfaces. Applying these values to the OSTR show that as little as 5% of the radioiodine remains airborne after only an hour. 12 Consistent with the CSAR, the values for these various release fractions are given in Table 1. For the accident in air, 100% of the noble gases are assumed to be available for release 22

I to the reactor bay air. For the halogens, there is a 50% plate-out assumed between the cladding gap and the air as well as a 50% plate-out on the reactor bay surfaces.

For the accident in water, it is assumed that most of the halogens released from the cladding gap remain in the water and are removed by the demineralizer. A conservative fraction, 5%, of the halogens are assumed to escape from the water to the reactor room air. Combining this with plate-out release of 50% from the gap to the water and 50% plate-out on the surfaces of the reactor bay, the result is that 2.5% of the halogens in the gap are capable of reaching the unrestricted environment. For the noble gases in water, 100% are assumed to be available for release to the unrestricted environment.

Table 4, Release Fraction Components

. : ** ' f a" "g 9*. . *4 *":..

Fission f . W product No pool water

[j "

With water No pool waterpool water Noble gas 1.0 1.0 1.0 1.0 1.0 Halogens 0.5 0.5 1.0 0.05 0.5 For the OSTR, the prevailing wind is from the south, blowing to the north. The minimum distance to the unrestricted environment (10 m), the minimum distance to the nearest occupied office building (100 m), and the minimum distance to the nearest permanent residence (267 m) are also to the north. For this accident, therefore, it was assumed that the wind is blowing from south to north.

For any atmospheric stability (Pasquill) class, a ground-level release always gives a higher concentration at any given distance than an elevated release. Thus, it was assumed that there were only ground level releases, which do not take credit for any release height.

Furthermore, the more stable the atmospheric class, the higher the concentration. Therefore, it was assumed that the most stable atmospheric class (Pasquill F) prevailed for all scenarios.

Also, the lower the wind speed, the higher the concentration. Thus, it was assumed that a low 23

wind speed of 1 m s- existed for all scenarios. This approach is identical to that found in the CSAR.

Once the fission products are released to the gap, this activity moves to either the primary water or the reactor bay air when the cladding fails. If the release is in air, then this activity is released directly into the reactor bay air. If the release occurs in the pool water, then the fission products must migrate through the water before being released to the reactor room air. Once released into the reactor room air, a further reduction of the halogen activity is expected to occur due to plate-out in the building. All of these values and methodology are identical to that used in the CSAR. A table summarizing the source term values can be found in Table 5.

Table 5, Airborne Radioactive Material Source Term Note Water.

Isotope Isotope D U-235 Wtr Reactor Target isotope Half-life ecay Fission Reactor B.ay Air IsoopeConstant Activity Product Bay Air Activity (S) (i/s) (mCi)

Yield Activity (mCi)

(mCi)

Br-82 127080 5.45E-06 6.10E-07 Br-83 8640 8.02E-05 5.38E-03 U Br-84m 360 1.93E-03 3.18E-04 U Reactor Water Bay NO Activity (mCi)

Water.

Reactor BayAir Br-84 1908 3.63E-04 1.OOE-02 U Br-85 172.2 4.03E-03 1.26E-02 U Br-86 55.5 1.25E-02 1.82E-02 U Br-87 55.9 1.24E-02 2.02E-02 U 1-131 692928 1.00E-06 2.88E-02 U 1-132 8208 8.44E-05 4.30E-02 U 1-133 74880 9.26E-06 6.70E-02

___4_

U 1-134 3156 2.20E-04 7.74E-02 U 1-135 23652 2.93E-05 6.29E-02 U U

1-136 83.4 8.31E-03 2.47E-02 Kr-83m 6696 1.04E-04 5.38E-03 U Kr-85m 16128 4.30E-05 1.26E-02 U Kr-85 3.39E+08 2.04E-09 2.74E-03 U Kr-87 4572 1.52E-04 2.51E-02 U Kr-88 10224 6.78E-05 3.57E-02 U Kr-89 189 3.67E-03 4.61E-02 U Xe-131m 1028160 6.74E-07 3.17E-04 U Xe-133m 189216 3.66E-06 1.95E-03 U -I 24

Xe-133 452736 1.53E-06 6.70E-02 Xe-135m 918 7.55E-04 1.21E-02 Xe-135 32760 2.12E-05 6.53E-02 Xe-137 229.2 3.02E-03 6.11E-02 Xe-138 846 8.19E-04 6.37E702 25

V.111 Calculation of Occupational and General Public Dose From Target MHA Calculation of the release from the building to the atmosphere and associated downwind dispersion was calculated using the same methodology as the CSAR. The committed dose equivalent (CDE) to the thyroid and the committed effective dose equivalent (CEDE) for members of the general public at a given distance downwind from the facility for all isotopes of concern were calculated using the methodology of the CSAR.

The methodology of NRC Regulatory Guide 1.145 was used to calculate the atmospheric dispersion factor as a function of distance, specifically Equation 3 from NRC Regulatory Guide 1.145:13 1

Q= WMUay~Z where:

X/Q = atmospheric dispersion factor (s m-3)

M = meandering correction factor (4, Figure 3 from Regulatory Guide 1.145)

Cy = lateral diffusion coefficient (in)

(z = vertical diffusion coefficient (m) g = mean wind speed (m s-1)

For distances greater than 100 m, the values for horizontal and vertical dispersion coefficients were also taken from this guide. For distances from 10 in to 100 m, which are not treated in NRC Regulatory Guide 1.145, data were obtained by extrapolating values to required distances. 14 The values for the lateral and vertical dispersion coefficients and X/Q are given in Table 6.

26

Table 6, Lateral and Vertical Diffusion Coefficients and z/Q 1 Values for Pasquill F and Mean Wind Speed of 1 m sec Distance Oy o, x/Q (M),. ' (M) (M)_.._'... (s m,3) 10 1.29 1.04 5.93 E-2 50 2.45 1.20 2.71 E-2 100 3.90 2.20 9.27 E-3 150 6.18 3.22 4.00 E-3 200 8.21 4.13 2.35 E-3 250 10.21 4.98 1.57 E-3 267 10.88 5.25 1.39 E-3 Additional parameters used in this accident were:

3

" reactor room ventilation exhaust rate: 4.39 E+6 cm sl; 3

reactor room leak rate: 1.69 E+4 cm s-;

reactor room volume: 3.88 E+9 cm3;

" area of north face of reactor building: 2.3 1E+2 in ;

" receptor breathing rate: 3.3 E-4 m 3 s 1 : (NRC "light work" rate); and

" dose conversion factors:

15 internal: based on DOE/EH-0071 ;

external: based on DOE/EH-0070.16 The committed dose equivalent (CDE) to the thyroid and the committed effective dose equivalent (CEDE):for members of the general public at a given distance downwind from the facility for all isotopes of concern may each be calculated by:

27

(CDE or CEDE)D = .(XD 0ft.BR DC A [e e-Ar 2IA]

where:

D (s m 3);

(X/Q)D = atmospheric dispersion factor at a given distance BR = breathing rate (in 3 s-I);

DCF,t,, = internal dose conversion factor for isotope i (mrem XCi-);

Ai = initial activity of isotope i released into the reactor room (ýi);

Rv = ventilation or leak rate of air from the reactor bay (in 3 s 1);

V reactor room volume (in3 );

X= ventilation constant = RVV (s-);

X= decay constant for isotope i (s-);

t= time when plume first arrives at the receptor point (s); and t2= time when plume has passed the receptor point (s).

The deep dose equivalent (DDE) to members of the general public at a given distance downwind from the facility for both the thyroid and the whole body may each be calculated by:

L DCF_IA2 le-Ar1 -e ýt (DDETIYrord or DDEI"B )D Q= A where:

DCFexti = external dose rate conversion factor for isotope i (mrem m3 [Ci-f s-).

For calculating dose to occupational workers in the reactor room, stay times of 2 and 5 minutes were used. Experience indicates that the reactor room can easily be evacuated in 2 minutes. The value of 5 minutes is thought to be a reasonable longer period of time assuming a 28 I

worker is performing some task (i.e., determining if a false alarm has occurred). The CDE and CEDE for personnel in the reactor room for a given stay-time may each be calculated by:

(CDEorCEDE)st = L DCFtiAjBR[1l-e1]

I 2

ieff V 1

IST where:

Xeff = Xi + Xv ; and tST = stay-time of personnel (s).

The DDE to personnel in the reactor room for a given stay-time for both the thyroid and the whole body may be calculated by:

SDCFet ,A 1- e-AefftST (DDETh/Wroid orDDEIB )ST = I L 2 ef The results of these calculations for all three scenarios are shown in Tables 7 through 10.

As seen from the tables, Scenario A gives the highest doses to the general public at any distance, as might be expected since the activity was released in a very short time leaving little time for radioactive decay. Scenario C gives the lowest doses at any given distance since the release has the lowest release rate. In all cases, doses for the general public and occupational workers are all below the annual dose limit specified by 10 CFR 20.

29

Table 7. Occupational Radiation Doses in the Reactor Room Following a Single Target Failure at End of Bombardment SScenario .I Release Environment Occupancy (minutes) 1 CDEThoid + DDEThyroid (mrem)

TEDE (mrem)

A Water 2 8 <1 A Water 5 8 <1 A Air 2 164 5 A Air 5 164 5 B Water 2 56 3 B Water 5 127 7 B Air 2 1099 34 B Air 5 2491 78 C Water 2 60 3 C Water 5 150 8 C Air 2 1175 37 C Air 5 2936 92 30

Table 8, Radiation Doses to Members of the General Public Following a Single Target Failure in Air at End of Bombardment - Scenario A Distance:,."' With PrimaryWater

With Primary .No Primary Water NoPrimaryý W

t *t r Water Water MI.. CDEThyroid + DDEThyroid TEDE CDEThyrold + DDEThyroid TEDE (mrem) (mrem (mrem) * ". (rnrem). . (mrem) 10 115 6 2254 71 50 53 3 1030 32 100 18 1 352 11 150 8 <1 152 5 200 5 <1 89 3 250 3 <1 60 2 267 3 <1 53 2 Table 9, Radiation Doses to Members of the General Public Following a Single Target Failure in Air at End of Bombardment - Scenario B With Primary Water With Primary No Primary Water No Primary Distance iCDEThyroid + DDEThyroid Water TEDE CDEThyroid + DDEThyroid Water TEDE .

(in) (mnrem) (mremr) (mrem) (rem) 10 115 6 2251 70 50 53 3 1029 32 100 18 1 352 11 150 8 <1 152 5 200 5 <1 89 3 250 3 <1 60 2 267 3 <1 53 2 31

Table 10. Radiation Doses to Members of the General Public Following a Single Target Failure in Air at End of Bombardment - Scenario C With Primary Wadter. With Primary No Primary Watiher No Primary Distance CDEThyroid + DDEThyroid Water TEDE CDEThyroid + DDEThyroid Water TEDE (m) (mrem) (mrem) (mrem) (mrem) 10 94 3 1873 56 50 43 1 856 25 100 15 <1 293 9 150 6 <1 126 4 200 4 <1 74 2 250 2 <1 50 1 267 2 <1 44 1 V.IV Insertion of Excess Reactivity The addition of up to three targets in the G-ring is not expected to significantly affect the worth of the control rods. Rod worth will be measured with targets installed prior to full power operation to confirm this assumption. Reactivity insertion accident calculations in the CSAR show that the maximum reactivity addition due to a continuous rod withdrawal accident will be

$0.96. This assumes initial reactor power is 100 watts and the accident is terminated by a high power SCRAM at 1.06 MW. Thermal calculations in previous sections assume continuous operation at 1.1 MW, so conditions during a transient event where power is raised quickly from 100 watts to 1.06 MW are bounded by the steady state analysis. The analysis also shows that since the maximum reactivity insertion is $0.96, a pulse will not occur.

The presence of targets in the core does not introduce any new mechanism that could cause a continuous rod withdrawal accident, nor does it increase the likelihood or severity of such an event. A continuous rod withdrawal accident will not initiate a reactor pulse since the reactor will SCRAM on high power before $1.00 of reactivity is inserted.

32

V.V Loss of Coolant Accident (LOCA)

An analysis was performed examining the temperature reached by the targets during a LOCA. The details of the analysis can be found in Appendix A. In summary, the same RELAP model used for the steady-state analysis was used to calculate the maximum temperature reached by the targets as a function of time for the LOCA to occur. The goal was to determine the shortest time for a LOCA to occur that would result in the maximum target temperature of 600'C (660'C is the melting point of aluminum). An inherent assumption is that the reactor operator is alerted to the loss of primary water from the low water annunciator and shuts down the reactor at which point it will then take a specified amount of time for the tank to drain. Once drained, all the targets in the core would be cooled by air only. The analysis shows that the maximum temperature reached by the targets was = for an instantaneous LOCA. The peak temperature is reached

after the reactor SCRAM and loss of coolant.

Therefore, it is improbable that the temperature of the targets would every reach the melting point of aluminum in the event of a LOCA.

Dose rate calculations due to shine from the uncovered core remain essentially unchanged. Dose rates are due to fission product inventory which is dependent on power history. The analysis assumes 365 days of continuous operation at full power. However, the intent will be to irradiate the targets for 6.5 continuous Effective Full Power Days (EFPD). Total core operating power will remain unchanged.

The presence of targets in the core does not introduce any new mechanism that could cause a loss of coolant accident, nor does it increase the likelihood or severity of such an event.

Handling of targets will result in increased cask work in the vicinity of the reactor, but targets will only be loaded and unloaded with the reactor shutdown, so increased frequency of cask operation does not increase the chance of a LOCA occurrence while operating at power.

V.VI Loss of Coolant Flow 33

Core flow is provided by natural circulation. Targets will be retained in the core lattice by the same mechanism (gravity) as the fuel, so it is not likely that targets could become mis-located and block core flow. The only credible mechanism for flow blockage is introduction of foreign objects in the core region. Multiple objects would be required to block all flow (one internal and multiple external flow paths) in any single target. The tank cover and careful operating practices prevents the introduction of foreign objects into the reactor tank.

Historically, this is a very rare event and if it were to occur, the reactor would be shut down.

The presence of targets in the core does not introduce any new mechanism that could cause a loss of coolant flow accident, nor does it increase the likelihood or severity of such an event.

V.VII Mishandling or Malfunction of Fuel The targets will be fabricated in a manner similar to the fuel. The targets will undergo a minimum of visual inspection and a helium leak test during post fabrication non-destructive testing. The targets will not likely be irradiated longer than 6.5 continuous EFPD. These constraints ensure that the likelihood of target malfunction is less than the likelihood of fuel malfunction. Targets will also be handled less than a typical fuel element during their respective lifetimes, so the likelihood of target mishandling is also less.

The presence of targets in the core does not introduce any new mechanism that could lead to mishandling or malfunction of fuel, nor does it increase the likelihood or severity of such an event.

V.VIII Experiment Malfunction The targets themselves can be considered an experiment. As discussed above, malfunction is less likely for a target than for a fuel element. Other than changing the flux in other experiment locations, the presence of targets in the core will not affect other experiments.

Target irradiation will likely be performed as a dedicated experiment with no other experiments being performed during the irradiation period, but the neutronic and thermal hydraulic analyses are not predicated on this assumption.

34

Each target is worth approximately reactivity. Removing a target from the core would add negative reactivity. In the unlikely event that the target were to fail is far less than a critical mass under any condition of moderation or geometry.

V.IX Loss of Normal Electrical Power The OSTR does not require an emergency backup power system to safely maintain core cooling. The presence of targets in the core does not introduce any new mechanism that could cause a loss of normal electric power, nor does it increase the likelihood or severity of such an event.

V.X External Events The presence of targets in the core does not increase the likelihood of tornadoes, hurricanes, seismic activity flooding or any other external events, nor does it increase the severity or consequences of such an event.

V.XI Mishandling or Malfunction of Equipment No credible accident initiating events were previously identified in the CSAR for this type of accident. The presence of targets in the core does not introduce any new mechanism that could lead to mishandling or malfunction of equipment, nor does it increase the likelihood or severity of such an event.

VI Limiting Conditions for Operation with fuel bearing targets present in the OSTR Core All Limiting Conditions for Operation (LCO's) established by the current Technical Specifications shall remain in effect. The existing LCO's will need to be augmented by the following additions which shall be in effect whenever the targets are present in one or more 35

OSTR in-core lattice positions. These changes are implemented to ensure assumptions made in this document exist in the OSTR at all times.

36

LCO TI, Permissible in-core Target lattice positions.

Applicability: Any time when targets are located in any core lattice position.

Obiective: To ensure assumptions made for the neutronic and thermal hydraulic analyses are not compromised.

Specification: Permissible target locations are core positions G32, G33, G34.

Targets shall not be placed in any other core lattice positions.

Basis: Analyzed target locations were G32, G33 and G34. Location G34 was found to produce the highest integrated power in a target. Thermal hydraulic analysis was based on power distribution in this hot target.

LCO T2, Pulse or square wave mode operation with targets located in any core lattice position.

Applicability: Any time when targets are located in any core lattice position.

Objective: To prevent all pulse activity while targets are present in any core lattice position.

Specification: The reactor shall not be operated in pulse mode or square wave mode while targets are present in any core lattice position.

Basis: Target performance has not been analyzed under rapid transient pulse conditions, therefore pulsing shall not be allowed when targets are present in the core. Pulse mode operation is prohibited. Square wave mode operation is also prohibited because it is possible to add more than $1.00 of reactivity in square wave mode. A rod withdrawal accident will not introduce sufficient reactivity to pulse the reactor.

LCO T3, Allowed Target Storage Locations.

Applicability: Any time targets are located in the reactor tank and not in transit or in an in-core lattice position.

37

Objective: To maintain k-effective of stored targets less than 0.9 under all conditions of moderation.

Specification: The targets shall be stored in the standard in-tank TRIGA storage racks. No other items shall be present in any storage rack containing targets.

Basis:

a. Storage racks are sufficiently far from the core such that the presence of targets in the core will not affect the criticality condition of targets in the storage racks. Criticality analysis assumes no other objects are present in the vicinity of the stored targets. The criticality analysis for the storage of the fuel assumed no other objects (i.e., fuel elements) were stored with the targets. The k-effective was calculated to be less than 0.9 when stored in the storage racks.

LCO T4, Target Fabrication Requirements.

Applicability: This specification applies to any target that will be placed in the reactor tank.

Objective: To assure that targets placed in the core may be used with a high degree of reliability with respect to their physical and nuclear properties.

Specification:

c. Cladding: aluminum, nominal thickness 0.32 cm.

Basis:

38

c. Cladding of this type provides adequate structural integrity while minimizing parasitic neutron absorption.

VII Target Transportation and Storage Transportation of un-irradiated and irradiated targets to or from the OSTR facility shall be performed in accordance with all regulations.

Un-irradiated targets will be placed in storage in an in-tank storage rack in a timely manner immediately after delivery. MCNP analysis shows that three targets with two inch spacing (square pitch) in water has a value of kff . Any in-tank storage rack containing one or more targets may not hold any other object of any type.

We also request a change in the license possession limit. We propose adding the following as license condition 2.b.2.f:

To receive, possess, use, but not separate, up to 0.45 kilograms of contained uranium-235 enriched to less than 20 percent VIII Target Loading and Operation Plan The loading of targets into the core and the operation of the reactor with targets in the core will be performed in accordance with the following guidelines. These guidelines specify the minimum required activities. Other additional activities may be performed in conjunction with target loading and subsequent reactor operation.

A reference core configuration shall be established prior to placing one or more targets in the core. *Excessreactivity and NRC shutdown margin shall be determined for the reference core. Control rod calibration and calibration of 39

nuclear instruments shall be performed for the reference core in accordance with existing OSTR Operating Procedures (OSTROPs) if needed. The reference core shall be established in such a manner that after target addition, the NRC shutdown margin and the core excess is predicted to be within technical specification limits.

" Reactivity worth of each target inserted individually into its intended operating location shall be measured.

Reactivity worth of all three targets inserted into their intended operating locations shall be measured.

A power calibration shall be performed before full power (1.0 MW) operation in accordance with existing OSTROPs. Power calibration will be performed in a timely manner as soon as reasonably possible after control rod calibration. Power calibration(s) shall be performed at intermediate power levels (50% and 90%, for example) prior to full power operation.

IX Conclusion Neutronic and thermal hydraulic analysis has shown that targets can be irradiated in the OSTR without adverse consequences. Analysis has also shown that in the event of the maximum hypothetical accident, the conservatively estimated doses received by occupational workers or the general public are below 10 CFR 20 limits. There is no known mechanism that could cause the MHA and there is no known mechanism that could cause simultaneous failures of more than one target.

40

Appendix A Steady-State Thermal Hydraulic Analysis of the Targets 41

ANALYSIS OF THE STEADY STATE THERMAL HYDRAULIC BEHAVIOR OF THE OREGON STATE TRIGA REACTOR

- MOLYBDENUM PRODUCTION ELEMENT -

Prepared by:

Department of Nuclear Engineering and Radiation Health Physics Oregon State University 116 Radiation Center Corvallis, OR 9733 1-5902 Prepared for:

Submittal to the Nuclear Regulatory Commission 42

TABLE OF CONTENTS Section Pa_ e I Sco p e ..................................................................................................................................................... 8 II Target Physical Description .......................................................................................................... 8 III Neutronic Analysis .............................................................................................................................. 11 IV Therm al Hydraulic Analysis ................................................................................................................. 15 V Accident Analysis ................................................................................................................................. 18 V.1 The M axim um Hypothetical Accident (M HA) ........................................................................... 18 V.11 Target M HA Source Term ........................................................................................ 20 V.111 Calculation of Occupational and General Public Dose From Target MHA ................................ 26 V.IV Insertion of Excess Reactivity ..................................................................................................... 32 V.V Loss of Coolant Accident (LOCA) .............................................................................................. 33 V.VI Loss of Coolant Flow ....................................................................................................................... 33 V.VII M ishandling or M alfunction of Fuel ......................................................................................... 34 V.VIII Experim ent M alfunction ................................................................................................................ 34 V.IX Loss of Norm al Electrical Power ................................................................................................ 35 V.X External Events ............................................................................................................................... 35 V.XI M ishandling or M alfunction of Equipm ent .............................................................................. 35 VI Limiting Conditions for Operation with fuel bearing targets present in the OSTR Core ............... 35 VII Target Transportation and Storage ................................................................................................ 39 VIII Target Loading and Operation Plan ................................................................................................ 39 IX Conclusion ........................................................................................................................................... 40 Appendix A .................................................................................................................................................. 41 1 INTRODUCTION ................................................................................................................................... 47 1.1 Purpose .......................................................................................................................................... 48 2 DESCRIPTION OF REACTOR AND SYSTEM ....................................................................................... 49 2.1 Fuel Description ............................................................................................................................. 55 2.1.1 Therm al Conductivity .................................................................................................... 57 2.1.2 Specific Heat ....................................................................................................................... 61 3 RELAP5-3D M ODEL .............................................................................................................................. 64 43

3.1 Overview .......................................................................................................................... ... .64 3.2 M odel Description .......................................................................................................................... 65 3.3 AECL Groeneveld Look-up Tables ............................................................................................ 76 3.4 The Bernath Correlation ................................................................................................................. 77 4 RES ULTS ............................................................................................................................................... 79 5 APPENDIX (AIR COOLING EVENT) ................................................................................................... 87 A p p e n d ix B .................................................................................................................................................. 92 44

LIST OF FIGURES Pa___ePa2e F ig ure 1, T arget D esign ............................................................................................................................... 10 Figure 2, Pow er per elem ent and target ................................................................................................ 13 Figure 3, Target tem perature distribution vs. elevation ....................................................................... 16 Figure 4, Target DN BR vs. elevation ...................................................................................................... 16 Figure 5, Target MDNBR vs. target pow er ............................................................................................. 17 Figure 1: Vertical sectional view of the OSTR [1] .................................................................................. 49 Figure 2: TRIGA fuel element design comprising the OSTR core ......................................................... 51 Figure 3: Mo-Element design considered for use in the OSTR core ..................................................... 52 Figure 4: LEU fuel element adjacent to a Mo-Element relative to the core grid plates ....................... 52 Figure 5: Core pow er per elem ent .............................................................................................................. 53 Figure 6: Axial power factor versus distance from Mo-Element axial centerline .................................. 54 Figure 7: Radial power factor versus distance from Mo-Element radial centerline ............................... 54 Figure 8: Tabulated values of helium thermal conductivity and fitted trend-line ................................. 59 Figure 9: Tabulated values of helium density and fitted trend-line ........................................................ 60 Figure 10: Thermal conductivity of .................... 61 Figure 11: Volumetric heat capacity of ............... 63 Figure 12:

RELAP5-3D Model Schem atic ......................................................................... 66 Figure 13: Hexagonal array axial average unit subchannel dimensions ................................................ 69 Figure 14:

RELAP5-3D Model Schematic ......................................................................... 72 Figure 15: Cross sectional view of Mo-Element composition ................................................................ 74 Figure 16: Radial nodal distribution in a M o-Element .......................................................................... 75 Figure 17: Mo-Element temperature distribution (a) color-plot and (b) isometric view ....................... 81 Figure 18: Axial tem perature distribution at = kW ....................................................................... 81 Figure 19: Radial temperature distribution at maximum axial fuel temperature at= kW ........... 82 Figure 20: Mo-Element Axial DNBR distribution at= kW ............................................................... 82 Figure 21: Mo-Elem ent channel properties ........................................................................................... 83 Figure 22: Mo-Elem ent m axim um tem peratures .................................................................................. 83 Figure 23: Mo-Elem ent MDNBR .................................................................................................................. 84 Figure 24: Decay power used for OSTR Air Cooling Event analysis ........................................................ 88 Figure 25: Maximum fuel temperature versus time for all Air Cooled Event Scenarios ........................ 90 Figure 26: Maximum fuel temperature in Mo-Element versus time delay between SCRAM and removal fro m co re ..................................................................................................................................................... 91 45

LIST OF TABLES Table Pa__e Table 1, Summary of Target Design Parameters ................................................................................... 11 T able 2, T arget R eactivity W orth ................................................................................................................ 14 Table 3, Estimated Core Excess and Shutdown Margin with targets installed ..................................... 14 Table 1: LEU 30/20 fuel design [1] ......................................................................................................... 50 Table 2: Hot Channel Molybdenum Element Power Summary .................................................................. 55 Table 3: Mo-Element Fuel Summary and Characteristics ..................................................................... 57 Table 4: RELAP5-3D Input for reactor and core geometry and heat transfer ............................................. 67 Table 5: Hydraulic flow parameters for the Mo-Element ........................... 68 Table 6: RELAP5-3D axial nodal lengths ................................................... 73 Table 7: Radial Mo-Element nodal locations (from Mo-Element center) ............................................ 76 Table 8: Steady State Results for Mo-Element (I kW) ................................................................. 79 Table 9: Decay pow er tim e table ................................................................................................................ 88 46

LIST OF ACRONYMS cSAR Conversion Safety Analysis Report HEU Highly Enriched Uranium INL Idaho National Laboratory LEU Low Enriched Uranium LOCA Loss of Coolant Accident OSTR Oregon State TRIGA Reactor OSU Oregon State University SAR Safety Analysis Report TRIGA Training Research Isotope General Atomics 1 INTRODUCTION Oregon State University (OSU) is currently investigating the safety bases regarding inserting a newly designed element into the Oregon State TRIGA Reactor (OSTR) Core for the purpose of producing Molybdenum. This newly designed element is referred to hereinafter as the "molybdenum element" (Mo-element).

RELAP5-3D version 2.4.2 is being used to perform the thermal/hydraulic safety analysis on this feasibility study. RELAP5-3D is a lumped parameter code that was originally developed by the Idaho National Laboratory (INL) for the purpose of performing integral nuclear reactor system thermal hydraulic analyses under normal operations and transient conditions. RELAP5-3D has been utilized to license the OSTR as a part of its recent core conversion from highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel [1]. Because RELAP5-3D is a lumped parameter code, geometric input parameters such as major and minor form losses must be inserted manually into the model by the user.

Two parameters that are of interest from a thermal hydraulic aspect during this feasibility study are the 47

geometric form losses associated with the inlet region and outlet region of the Mo-element as it sits in the OSTR core.

1.1 Purpose This study provides a presentation and discussion of numerical results obtained using RELAP5-3D version 2.4.2 that identify key thermal hydraulic parameters impacting the safety of the OSTR during normal operations while three Mo Elements are inserted in three discrete G-ring locations within the OSTR core.

48

2. DESCRIPTION OF REACTOR AND SYSTEM The OSTR is a Mark II TRIGA reactor licensed for steady state operation up to 1.1 MWth. The Mark II is an open pool design and operates at atmospheric pressure. The core is centered near the bottom of the reactor pool, seen in Figure 6. The top of the core is nominally located approximately sixteen feet below the surface of the reactor pool, with a minimum water level of fourteen feet driven by the OSTR's technical specifications. The OSTR pool is approximately six and a half feet in diameter and twenty feet deep [1]. The core is three and a half feet in diameter and two feet in height.

S iZIIII IIZI

-- ALUMINUM TANK FACILITYA HEAVY CONCRETE ON4TRACK DOO00R PO N RAD..U I . ON GRCHt ALMI .U 'ASN DOOR PLUG Figure 6: Vertical sectional view of the OSTR [1]

Seven rings make up the circular fuel element lattice configuration for thle OSTR identified as Rings A through G. Rings A and B contain triangular subchannels; moving outward radially on the grid plate the subchannel configurations become distorted and larger in rod-to-rod pitch. All fuel elements that comprise the OSTR core are congruent in material composition (shown in Table 4) and geometry (shown in Figure 7). During this study three Mo-Elements are inserted in positions G-32, G-33, and G-34. The 49

geometric configuration of the Mo-Elements to be inserted in said G ring positions is presented in Figure

8. Figure 9 provides a visual comparison of the geometric similarities between the Mo-Element and a LEU element currently comprising the OSTR core. Figure 9(a) and (b) present these elements near the center of the core for the purpose of placing them adjacent to each other, although the Mo-Elements will only be placed in the G Ring through all calculations in this study.

Table 4: LEU 30/20 fuel design [1]

ijel'Typv LUO30/20 Uranium content [mass %] 30 U-235 enrichment [mass % U] 19.75 Erbium content [mass %] 1.1 Fuel alloy inner diameter [mm] 6.35 Fuel alloy outer diameter [mm] 36.449 Fuel alloy length [mm] 381 Cladding material Type 304 SS Cladding thickness [mm] 0.508 Cladding outer diameter [mm] 37.465 50

STAINLESS STEEL TOP END OF FITTING, GRAPHITE V-...I i 87.38 mm STAINLESS STEEL TUBE CLADDING THICKNESS 0.508 mm ERBIUM-ZIRCONIUM HYDRIDE-8.5 WT%

URANIUM-..

673.1 mm 381 mm 37.34 mm K-i GRAPHITE I 88.138 mm STAINLESS STEEL BOTTOM END FI-TTI NG Figure 7: TRIGA fuel element design comprising the OSTR core 51

Figure 8: Mo-Element design considered for use in the OSTR core riguic 7.i--i LJZU eiZieitLUJdL~I1L RLU CICICH VU-M: 51211L ICIaLIVC LU LIM WIC pIULUt 52

(a) isometric view and (b) elevation view An MCNP5 model was developed to produce each element's explicit thermal power using the methods outlined in the Conversion Safety Analysis Report (cSAR) [1]. Of the three Mo-Elements, G-34 produces the highest thermal power (l kW,1 ) assuming an integral core power of 1.1 MWth; an axial and radial power distribution was then tabulated from G-34, producing Figure 11 and Figure 12. These power profiles and the explicit power tabulated for G-34 provide the neutronic inputs for this thermal hydraulic analysis.

Figure 10: Core power per element 53

I Figure 11: Axial power factor versus distance from Mo-Element axial centerline Figure 12: Radial power factor versus distance from Mo-Element radial centerline 54

Tabulating the power peaking factor for the most powerful (or Hot) Mo-Element may be done by taking the product of the "Hot Mo-n Element Peak Factor", "Hot Mo-Element Fuel Axial Peak Factor", and "Hot Mo-Element Fuel Radial Peak Factor" which are defined as:

Hot Channel Peak Factor = (maximum Mo-Element power)

/(core average element power)

Hot channel Fuel Axial Peak Factor (maximum axial power in the hot Mo-Element)

/(average axial power in the hot Mo-Element

" Hot Channel Fuel Radial Peak Factor (maximum radial power in the hot Mo-Element)

/(average radial power in the hot rod)

Utilizing these definitions yields Table 5. From Table 5 the effective Peak Factor for the hot Mo-Element is M, note that the current cSAR indicates that the hot rod power factor in all core configurations is 2.5 or greater [1].

Table 5: Hot Channel Molybdenum Element Power Summary

LDescri tion V . .......... Value Hot Mo-Element Location Hot Mo-Element Thermal Power [kW]*

Hot Mo-Element Peak Factor [Pmax/Pavg]

Hot Mo-Element Fuel Axial Peak Factor [Pmax/Pavg]

Hot Mo-Element Fuel Radial Peak Factor [Pmax/Pavg]

Effective Peak Factor

Hot Mo-Element thermal power corresponds to core power of 1. 1 MW,h 2.1 Fuel Description 235 The fuel core utilized within the Mo-Element enriched to 20 wt% U Previous studies have quantified numerous characteristics Laboratory experimentally identified 55

Figure 8 identifies an explicit value for the height The study's fuel core height of is tabulated by considering the dimensional values Taking a conservative approach, this study assumes the largest effective cross sectional loading area within U herefore identifying the maximum Mo-Element temperature within acceptable fabrication tolerances. The nominal The nominal

. Dividing the total occupied volume previously discussed by the maximum effective cross sectional loading area yields a height of .

Table 6 contains general information on the element, followed by several images which illustrate the element's geometry.

56

Table 6: Mo-Element Fuel Sulmary and Characteristics Property Description Fuel type Fill gas (assumed during study) Helium Enrichment [%] 20 Fuel density [g-cm 3 ]

2.1.1 Thermal Conductivity Thermo-physical properties of ,including thennal conductivity (k) and specific heat (Cp) have been theoretically and analytically studies by numerous investigators [2-4]. The thermal conductivity M for this study was collected from [5]. In this case thermal conductivity is defined by the relationship equation (2)

(2) where k 95 Thermal conductivity for 95% theoretical density (TD) fuel [W/m-K]

T Temperature [K]

Bu Bumn-ip [GWd/MTU]

57

f (Bu) Effect of fission products in crystal matrix (solution).=

g (Rt) Effect of irradiation defects =

Q Temperature dependence parameter = 6380 [K]

A Constant = M [m-K/W]

a Constant coefficient = M gad weight fraction of gadolinia B Constant = [m-K/W/K]

E Constant = [W-K/m]

F Constant = M [K]

h (T) temperature dependence of annealing on irradiation defects The model identified in equation (2) may be adjusted for as-fabricated fuel density (in fraction of TD) using equation ki =1.0789.k 95 1+ 0.5d(1 -d)(3 (3) where d is the density (fraction of TD).

The considered herein assumes a density fraction of the TD to be equal to unity, a fuel having no bumr-up, and no gadalinia present.

The fill gas comprising 58

thermal transport characteristics due to its increased insulating capability relative to nitrogen rich gas.

Tabulated values for thermal conductivity for helium were collected [6] and a trend-line was fit to these values as seen in Figure 13. The power fit equation fit to the data in Figure 13 for thermal conductivity of Helium as a function of temperature is shown in equation (4).

kH, = 0.0026065 x T0o7 133 9 [W/m-K] (4) 0.35 --

0.3.

I L

.. 0.25 4-U 0.2 39 K = 2.6065E-03 *T 7 .13 E-0I 0.15 -

1 200 300 400 500 600 700 800 Temperature [K]

Figure 13: Tabulated values of helium thermal conductivity and fitted trend-line Density of helium was obtained through the same method used to obtain helium thermal conductivity [6]

as shown in Figure 14. Equation (5) describes helium's density (PH,) as a function of temperature.

PH, = 48.526 x T-0 9 9 9 [kg/m3 ] (5) 59

I 0.2 0.18 t 0.16 0.14 46 5 p= .

".: 0.12 4 0.1 0.08 {

0.06 -F 26-() 999 -~-.---

0.04 0.02 200 300 400 500 600 700 800 Temperature [K]

Figure 14: Tabulated values of helium density and fitted trend-line An effective thermal conductivity of the fuel meat region in the Mo-Element is acquired through Therefore the thermal conductivity of Performing equation (6) yields the effective thermal conductivity shown in equation (7). A plot of thermal conductivity as a function of temperature for 60

Figure 15: Thermal conductivity of Helium 2.1.2 Specific Heat The specific heat 61

Ias recommended by previous authors [5, 7]. The volumetric heat capacity The specific heat of the helium fill gas (CPHe) is known to hold a constant value of 5193 J/kg-K [6]. The product of helium specific heat and temperature dependent density (represented in equation (5)) produces volumetric heat capacity of helium (pCp,He).

An effective volumetric heat capacity of the fuel meat region in the Mo-Element is acquired through averaged volumetric heat capacity 62

Performing equation (9) yields the effective volumetric heat capacity in equation (10). A plot of volumetric heat capacity as a function of temperature for Figure 16: Volumetric heat capacity of Helium 63

3 RELAP5-3D MODEL 3.1 Overview In this study, the predicted steady state thermal hydraulic performance of the OSU Mo-Element is determined for the reactor operating at 1.1 MWth with a water inlet temperature of 49°C and an effective outlet pressure equivalent to 14 feet of water head above the core. This analysis is conservative since the maximum license power of the OSTR is 1.1 MW,,,, but the reactor is operated at 1.0 MW with a high power SCRAM at 1.06 MWth. Per the Technical Specifications, the maximum pool temperature is 49 'C.

The maximum powered Mo-Element is used and assumed to be located in the smallest single subchannel location (equivalent to the B Ring position). The Mo-Element is different in geometry from all LEU flip elements in that it is RELAP5-3D version 2.4.2 was used to calculate the following for the G-34 Mo-Element assuming an integral 1.1 MWth:

subchannel flow rate

subchannel bulk coolant temperature distribution

subchannel departure from nucleate boiling ratio (DNBR) distribution using o Bernath Correlation [8]

o Groeneveld 2006 Look-up Tables [9]

Mo-Element cladding temperature distribution

" Mo-Element fuel core temperature distribution Power was then varied from the nominal I kW level that the G-34 Mo-Element produces at a steady integral core power level of 1.1 MWth to power levels below and above this nominal value and the following values were tabulated:

subchannel minimum DNBR (MDNBR) using o Bernath Correlation [8]

o Groeneveld 2006 Look-up Tables [9]

subchannel exit coolant temperature

Maximum Mo-Element cladding temperature 0 Maximum Mo-Element fuel core temperature 64

3.2 Model Description The analysis was perfonned using a two channel model divided into axial and radial segments (nodal distribution is described below). The RELAP5-3D model seen in Figure 17 consists of a Coolant Source, Cold Leg, Horizontal Connector, Lower Plenum, M Subchannel, M Subchannel, Upper Plenum, and Coolant Sink. This model is representative of the hydraulics associated with a single Mo-Element, assumed to be the hottest Mo-Element.

The Coolant Source is modeled as a time dependent volume in RELAP5-3D allowing for an inlet pressure and temperature boundary condition to be imposed on the system during the analysis. The Cold Leg is incorporated into the RELAP5-3D model in order to create a pressure differential between the cold coolant entering the and the heated coolant passing through the

. This pressure differential drives the natural circulation flow. The Horizontal Connector serves no physical purpose in the OSTR, but is rather a nonphysical connector between the Cold Leg and to allow communication between Volumes 102, 105 and 106 during the computational process. The (Volumes 105 and 106, respectively) are the volumes which are directly adjacent to the Mo-Element. It is assume in the RELAP5-3D model that the Mo-Element has the most conservative thennal hydraulic parameters found in the OSTR therefore, because the rod-to-rod pitch is the smallest in the B-Ring, the Mo-Element is assumed to be inserted in the B-Ring.

To simplified RELAP5-3D model, it was assumed that there is no cross flow between adjacent Sallowing a single element model to suffice. This assumption is conservative since higher values of temperature and lower margins to DNB have been predicted when cross flow between adjacent is ignored this was previously demonstrated during as a part of the cSAR development

[1]. Furthermore, other work has shown that a single element model provides critical heat flux results within -1.0% of those produced from two and eight channel models and that the single channel model produces the most conservative results relative to the two and eight chanmel models [10-12].

65

Horizontal Connector (103)

Figure 17: Two Channel RELAP5-3D Model Schematic The reactor geometric and hydraulic data for the RELAP5-3D input are given in Table 7. Flow channels in the OSTR are triangular, square, ore irregular, depending on the core location. The analysis assumes a triangular element lattice configuration to minimize the flow area and maintain a conservative safety basis.

66

Table 7: RELAP5-3D Input for reactor and core geometry and heat transfer Va u iatu

.P arai n etet, ... a::::::*

..(105)/ * .i * ":(.006) "J Inlet form loss coefficient Outlet form loss coefficient Inlet coolant temperature [°C] 49.0 Absolute pressure at the top of the core [Pa] 1.43E+05 A study conducted by General Atomics for the OSTR developed a methodology for calculating each effective subchannel form loss coefficient rather than local form losses within the core assuming LEU 30/20 element geometry [13], the same methodology was employed as a part of this study to tabulate the inlet and outlet form loss coefficients for the [14]. These coefficients are shown in Table 7. A constant pressure of 1.01 E+05 Pa is assumed to exist at the top of the reactor pool.

The OSTR Technical Specifications require a minimum water column height above the core to be 14 feet, equivalent to 4.20E+04 Pa. The sum of the atmospheric pressure and the minimum water column height was therefore imposed as a boundary condition at the top of the core (1.43E+05 Pa).

This analysis was performed on the maximum powered Mo-Element and was conducted assuming (conservatively) that the maximum powered Mo-Element was also located in the most restricted flow channel in the OSTR. The flow parameters for the most redistricted flow channel are given in Table 8.

67

Table 8: Hydraulic flow parameters for the Mo-Element Parameter "

Flow area [in 2]

rod-to-rod pitch [in]

Wetted perimeter [m]

Hydraulic diameter [m]

Heated diameter [m]

Mo-Element heated length [in]

Mo-Element heated surface area [in 2 ]

Mo-Element surface roughness [in]

The pitch for the B Ring subchannel is 4.064E-02 m [15]. A typical B Ring subchannel is graphically shown in Figure 18. The Mo-Element are (Figure 8). Equation (11) defines I and applies the formula for tabulated area of an element in a hexagonal array [16].

A= 2 p 2 D (11 4 )

From equation (11) the (12) 68

I

-Adjacent Element Hot Channel DI. Subchannel Hottest Element Figure 18: Hexagonal array axial average unit subchannel dimensions The Mo-Element heated surface areas in the RELAP5-3D model were calculated by referring to Figure 8. The axial length of the The total heated surface area of The wetted perimeter is defined perimeter is defined as P,,,etted = zrD. This equation produces a value of The hydraulic diameter is calculated per equation (13). With reference to the previously calculated wetted perimeter values and subchannel flow areas, the hydraulic diameters for the are tabulated as 69

D, 4A (13)

Piveited Figure 19 shows a direct comparison between a physical Mo-Element and the RELAP5-3D discretized subchannel volume. Node dimensions are given in Table 9. Nodes 01 and 3 1 represent the lower and upper grid plates, respectively. The lower grid plate is 1.91E-02 m thick. The upper grid plate is 1.59E-02 m thick. The bottom surface of the upper grid plate is 6.73 IE-01 m above the top surface of the lower grid plate [17]. The Mo-element axial nodal dimensions are shown in Figure 19; note that volumes 105 and 106 are congruent in the axial nodal values.

The bottom surface of the Mo-Element is m above the top surface of the lower grid plate (Figure 8). Nodes 02 through 04 extend from the bottom of the fuelled portion of the Mo-Element to the top of the lower grid plate and are evenly distributed in length From Figure 19 the fuel nodal lengths must be discretized accordingly; this done by use of equation (14) for Nodes 05 through 24:

L5 -

=24 (15) nlfiiel L05.24 refers to the nodal length for Nodes 05 through 24, n1i,,, is the number of nodes defined in the fuel region (i.e. 20 nodes).

M 70

(16) where t represents the height between the top of the lower grid plate and the bottom of the upper grid plate (6.9215E-01 m), LLAW is the height difference between 71

Figure 19: RELAP5-3D Model Schematic 72

Table 9: Table 9: ~~~REL AP5-3D P-Dail axial nodal oa lengths egh No.d*li

. De*cription Node Number' ><lN.dal Length [nm]

Upper Grid Plate 31 30

-29 28 27 26 25 24 23 22 21 20 19 18 17 16 15 14 13 12 11 10 09 08 07 06 05 04 03 02 73

Lower Grid Plate 01 A cross sectional view of a Mo-Element is shown in Figure 20. The radial nodal distribution is shown in Figure 21. The fuel portion of the Mo-Element consists of

ýto 20 wv%

The thermophysical properties employed in the RELAP5-3D model for the fuel meat composition include effective thenrinal conductivity found in equation (7) and volumetric heat capacity referred to in equation (10).

Figure 20: Cross sectional view of Mo-Element composition 74

Figure 21: Radial nodal distribution in a Mo-Element The mesh points within the fuel region used in the RELAP5-3D model correspond to one node for the

. The radial location of each node is identified in Table 10. RELAP5-3D input requires that radial mesh points be defined in order to specify all material properties and to calculate temperature gradients within the heat structure.

75

Table 10: Radial Mo-Element nodal locations (from Mo-Element center) 3.3 AECL Groeneveld Look-up Tables CHF = f (G, Xh geometry) (17)

By interpolating in the Groeneveld look-up tables given the mass flux (G), equilibrium quality (Xe) and absolute pressure (P,,h.,) produced in the RELAP5-3D model, a critical heat flux value for the system is 76

produced (CHFit). The appropriate correction factors, KI to K(, must then be multiplied by the interpolated CHF value.

CHF = K1 *K2 , K3

K 4

K 5 .K 6 *CHFi, (18)

Three correction factors affect this study: Kl, K/, and K4 . The other correction factors are assumed to have a value of one.

K_ (19) where the hydraulic diameter is measured in meters.

K, =Min0. 8,0.8 exp-O.5x j (20) where X, is the equilibrium quality. If X, is less than zero, it is set to zero.

K4 = exp.-He2a2, (21) where a = (22) and L [meters] is the distance from the start of the heated length to the middle of the node. If DH/L is greater than 0.2, it is set to 0.2. The quantities p, and pg [kg/nm3 ] are the liquid density and gas density respectively, while D. is the heated diameter [meters].

The appropriate critical heat flux values from the Groeneveld 2006 Look-up Tables were produced by applying these correction factors.

3.4 The Bemath Correlation CHF is defined in units of pound- centigrade per hr-ft2 per the following equations.

77

CHF = hBo (To"- Tb) (23) hso=

hBO=108901890/)h + DH) + Av Dj, (24) 48 ifD, < 0.1 ft (25)

D1,,

1l,+90 if D1, > 0.1 ft T"° =57In (Jbs)-54 PVb+1 4 (26)

Kabs ++/-15) 4 where:

hBo Limiting film coefficient [p.c.u./hr-ft2-°C]

Tb Fluid bulk temperature [°C]

T,,,BO Wall temperature at CHF [°C]

v Fluid velocity [fi/sec]

A "slope" Pabs Absolute pressure [psi]

DH Heated diameter [ft]

Dh Hydraulic diameter [ft]

78

4 RESULTS The driving force for flow in the OSTR core is supplied by the buoyancy of the heated water in the core.

Countering this force are the contraction and expansion losses at the entrance and exits to the channel, and friction losses due to coolant to element interfacial contact. A summary of the RELAP5-3D results for the OSTR Mo-Element at full power (M kW) is given in Figure 22 through Figure 25. A summary of results from Figure 22 through Figure 25 is given in Table 11.

Table 11: Steady State Results for Mo-Element ( kW)

Parameter Value-Flow rate for hottest rod [kg/s]

Maximum flow velocity [m/s]

Maximum wall heat flux [kW/m 2]

Maximum fuel centerline temperature [°C]

Maximum clad temperature [°C]

Exit clad temperature [°C]

Exit bulk coolant temperature [°C]

MDNBR [Groeneveld 2006, Bernath] 16.853. 11.425 16.465, 9.733 The steady state results shown above are for the Mo-Element located in the G-34 positing. The Mo-Element in the G-34 position has largest element power, and thus is the bounding Mo-Element steady state operation. The Mo-Element has an MDNBR of 17.171 in the and 14.762 in the at 1.1 MWIh steady state power using the Bernath Correlation. Figure 28 shows, MDNBR as a function of Mo-Element power. The highest Mo-Element power considered herein is at a level of 20 kW; the most conservative MDNBR obtained from this operating condition was found to be located in , evaluated by the Bernath Correlation with a value of 5.092. Using either the Bernath or the Groeneveld 2006 correlations, the Mo-Element operates at a power level well below that required for departure from nucleate boiling.

79

(a) 80

Figure 22: Mo-Element temperature distribution (a) color-plot and (b) isometric view Figure 23: Axial temperature distribution at M kW 81

Figure 24: Radial temperature distribution at maximum axial fuel temperature at = kW Figure 25: Mo-Element Axial DNBR distribution at kW 82

Figure 26: Mo-Element channel properties Figure 27: Mo-Element maximum temperatures 83

Figure 28: Mo-Element MDNBR 84

REFERENCES

1. Reese, S.R., et al., Safety Analysis Report for the Conversion of the Oregon State TRIGA Reactor Fronm HEUto LEU Fuel. 2007, Oregon State University.

2.

5. Luscher, W.G. and K.J. Geelhood, MaterialPropertv Correlations:Comparisonsbetween FRA4PCON-3.4. FRAPTRAN 1.4. andA 4TPRO. 2011, United States Nuclear Regulatory Commission.

6. Incropera, F.P., et al., Fundamentalsof Heat and Mass Transfer. 6 ed. 2006: John Wiley & Sons, Inc. 997.

7.

8. Bernath, L., A Theory of Local Boiling Burnout and Its Application to Existing Data. Chem. Eng.

Prog., Symp. Ser, 1960. 30(56): p.95-116.

9. Groeneveld, D.C., et al., The 2006 CHF look-up table. Nuclear Engineering and Design, 2007: p.

1-24.

10. Marcum, W.R., B.G. Woods, and S.R. Reese, Experimentaland Theoretical Comparisonof Fuel Temperature and Bulk Coolant Characteristicsin the Oregon State TRIGA Reactor During Steady State Operation.Nuclear Engineering and Design, 2010.240: p. 151-159.

11. Marcum, W.R., et al., Steady State Thermal Hydraulic Analysis of the Oregon State TRIGA Reactor Using RELAP5-3D. Nuclear Science and Engineering, 2009. 162: p. 261-274.

85

12. Marcum, W.R., Thermal Hydraulic Analysis of the Oregon State TRIGA Reactor Using RELAP5-3D, in Nuclear Engineering & RadiationHealth Physics Department.2008, Oregon State University: Corvallis. p. 166.

13. Bolin, J., TRIGA Reactor Thermal-HydraulicStudy. 2008, General Atoimcs: San Diego. p. 1-71.

14. Marcum, W.R., Determination of the Form Losses in the OSTR Molybdenum Element, in CalculationReport, OSU-MO-000 100-CALC-00 1, Editor. 2011, Oregon State University:

Corvallis.

15. Atomics, G., Top GridPlate. TRIGA Mark III, General Atomics. p. Dwg #: T13S21OJIO6s.

16. Todreas, N.E. and M.S. Kazimi, Nuclear systems II: Elements of Thermal Hydraulic Design. Vol.

2. 2001, New York: Taylor & Francis Group, LLC. 506.

17. Atomics, G., Reflector Assembly, Oregon: General Atomics. p. Dwg # T2D21OJ1 10-A.

18. Society, A.N., Decay Heat Power in Leight Water Reactors. 2005, American Nuclear Society: La Grange Park, Illinois.

86

5 APPENDIX (AIR COOLING EVENT)

The Mo-Element under discussion is considered a fueled experiment when utilized in the OSTR. As such, operators intend to remove the element after a period of irradiation time in order to examine the experiment. This fueled experiment is clad with aluminum. Because aluminum has a relatively low melting point, it is paramount to the safety of the experiment that the maximum temperature in the cladding does not approach aluminum's melting point when removed from the core region (liquid coolant) and exposed to air (non-condensable gas).

Conduction from the fueled region into the cladding region diffuses heat outward more prominently during decay heat mode than during full power operation. This results in the relative temperature difference between the maximum temperature in the fueled region and the clad region becoming much less than that during normal full power operations. Regardless of the operational mode, the fueled region continues to host the maximum temperature within the entire Mo-Element as this is the only region in which heat is being generated. Because this study is centered on a conservative safety analysis approach, the maximum temperature within the fueled region will be tracked during all scenarios within this event rather than the maximum clad temperature. Inductively, if the maximum temperature in the fueled region does not approach the melting temperature of aluminum, the clad region will never reach the melting temperature of aluminum.

This "Air Cooling Event" assumes that the Mo-Element has been operating for an effective infinite period of time allowing for thermal equilibrium to be reached. After thermal equilibrium takes place this event assumes the reactors SCRAMS (i.e. instantaneously shuts down). This assumption leads to only decay power within the Mo-element producing heat in a subcritical state. In all scenarios considered with reference to this specific event, decay power for the OSTR was calculated using the ANSI/ANS-5.4 2005 standard 5.1 "decay heat power in light water reactors". It is assumed that the Mo-Element proportionally follows this decay power profile based on its relative power ratio during full power operations. Using the decay power profile from the ANSI/ANS standard [18], and assuming that 100 % of the decay fission occurs in U2 35, Figure 29 is produced (and summarized in Table 12). It is important to note that while calculating the decay power using the ANSIANS standard it was assume that the total recovery energy associated with one fission of a nuclide was assumed to be 200 MeV/fission and that and the number of fissions per initial fissile atom was 1.0 (i.e. making the decay power profile as conservative as possible).

87

16 14 12

~10 4

2 0

10 100 1000 10000 Time After SCRAM [sec]

Figure 29: Decay power used for OSTR Air Cooling Event analysis Table 12: Decay power time table Timie [sec]- %~Pow~ej 1 13.3950 1.5 12.2325 2 11.3305 4 9.1550 6 7.9880 8 7.3750 10 6.8765 15 6.2432 20 5.8463 40 5.0289 60 4.6051 80 4.2775 100 4.0592 150 3.6588 88

200 3.4228 400 2.9040 600 2.6710 800 2.5322 1000 2.4302 Five scenarios are considered within this Air Cooling event and are summarized below:

Scenario I - The reactor has operated for an infinite period of time at full power, at t = 0 second the reactor is SCRAMed, simultaneously the Mo-Element is removed from the core and exposed to air cooling.

Scenario H - The reactor has operated for an infinite period of time at full power, at t = 0 seconds the reactor is SCRAMed. Five minutes (t = 300 seconds) after the reactor has SCRAMED the Mo-Element is removed from the core and exposed to air cooling.

Scenario II1- The reactor has operated for an infinite period of time at full power, at t = 0 seconds the reactor is SCRAMed. Ten minutes (t = 600 seconds) after the reactor has SCRAMED the Mo-Element is removed from the core and exposed to air cooling.

Scenario IV- The reactor has operated for an infinite period of time at full power, at t = 0 seconds the reactor is SCRAMed. Fifteen minutes (t = 900 seconds) after the reactor has SCRAMED the Mo-Element is removed from the core and exposed to air cooling.

Scenario V- The reactor has operated for an infinite period of time at full power, at t = 0 seconds the reactor is SCRAMed. Twenty minutes (t = 1200 seconds) after the reactor has SCRAMED the Mo-Element is removed from the core and exposed to air cooling.

The same RELAP5-3D model described in the body of this document was used to perform the simulated scenarios outlined above. For each scenario, RELAP5-3D assumes that the element is exposed to pure nitrogen (substitute for air) when removed from the core.

89

The maximum temperature within the fueled region of the Mo-Element was tracked over time for each scenario. This is presented in Figure 30. Note that because the aluminum conducts heat outward so quickly and the generally short overall conductive distance through the radial direction in the Mo-Element, heat is removed quickly from the fuel meat region. Figure 30 shows that it takes approximately 25 seconds for the maximum temperature to be reached in the Mo-Element under Scenario I, which produces the most conservative maximum fuel meat temperature of 159.97 'C. The maximum fuel temperature is also plotted against the time difference between SCRAM and exposure to air in Figure 31.

Notice that the decay power drops the maximum fuel meat temperature after allowing decay heat for approximately 200 seconds.

180 7-"

160 0 0 e

140 t

120 UI

U

a a U

< )

-I a

8

~0

SCENARIO I 4

SCENARIO I1 0

SCENARIO III SCENARIO IV

& SCENARIO V a.........._ t __..

?

50 100 150 200 25C Time After Removing From Core [see Figure 30: Maximum fuel temperature versus time for all Air Cooled Event Scenarios 90

180 160 .

5"140 o

=120 E 100

- 80 E

60 -

40T 20 .

0 200 400 600 800 1000 1200 Time After SCRAM when Removed From Core [sec]

Figure 31: Maximum fuel temperature in Mo-Element versus time delay between SCRAM and removal from core 91

Appendix B Target Drawings 92

93 References 2 "MCNP-A General Monte Carlo N-Particle Transport Code, Version 5," LA-CP-03--245, F. B.

brown, Ed., Los Alamos National Laboratory (2003).

3 "Safety Analysis Report for the Conversion of the Oregon State TRIGA Reactor from HEU to LEU Fuel," submitted November 2007.

4 RELAP5-3D Code Development Team, "Volume 1: code structure, system models, and solutions methods, in RELAP5-3D code manual," 2005, Idaho National Laboratory, Idaho Falls, Idaho. P. 600.

s NUREG-1537, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors," USNRC, February 1996.

9 NUREG-1465, Accident Source Termsfor Light-Water Nuclear Power Plants,U.S. Nuclear Regulatory Commission, February 1995.

10 NUREG/CR-5747, Estimate of Radionuclide Release CharacteristicsInto Containment Under Severe Accident Conditions, U.S. Nuclear Regulatory Commission, October 1993.

1' Regulatory Guide 3.33, Assumptions used for evaluating the potential radiological consequences of accidental nuclear criticality in a fuel reprocessing plant. U.S. Nuclear Regulatory Commission, April 1977.

12 IAEA Safety Reports Series No. 53, Derivation of the Source Term andAnalysis of the Radiological Consequences ofResearch Reactor Accidents, International Atomic Energy Agency, Vienna, Austria; 2008.

13 Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, August 1979.

94

14 "Calctilated Atmospheric Radioactivity from the OSU TRIGA Research Reactor Using the Gaussian Plume Diffusion Model," Bright M.K. Wong, Oregon State University Department of Nuclear Engineering Report 7903, August 1979.

15 "Internal Dose Conversion Factors for Calculation of Dose to the Public," DOE/EH-0071, U.S.

Department of Energy, Washington, D.C., 1988.

16 "External Dose Rate Conversion Factors for Calculation of Dose to the Public," DOE/EH-0070, U.S.

Department of Energy, Washington, D.C., 1988.

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