ML113410168

Dow Chemical Company - *Redacted* Responses to NRC Request for Additional Information Technical

ML113410168
Person / Time
Site:	Dow Chemical Company
Issue date:	11/10/2011
From:	O'Connor P Dow Chemical Co
To:	Geoffrey Wertz Research and Test Reactors Licensing Branch
Meyer W, NRR/DPR/PRLB, 301-415-0897
References
TAC ME1595
Download: ML113410168 (69)

Text

DOW CHEMICAL COMPANY RESEARCH REACTOR LICENSE NO. R-108 DOCKET NO. 50-264 TECHNICAL RAI RESPONSES (DATED 11/10/2011)

REDACTED VERSION*

SECURITY-RELATED INFORMATION REMOVED

• REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS

The Dow Chemical Company Midland, Michigan 48667 Security-related information - withhold from public disclosure under 10 CFR 2.390 Mr. Geoffrey Wertz Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

Subject:

The Dow Chemical Company-License No. R-108; Docket No. 50-264 Enclosed are the DTRR Revised responses to RAI questions 11, 14, 15-1, 15-2, 15-3, 16, 17-1, 17-2, 17-3, 18, 19, 41, 52, 53-1, 53-2, 54, 56 in support of the license renewal.

Should you have any questions or need additional information, please contact the Facility Director, Paul O'Connor, at 989-638-6185.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November 10, 2011 Paul O'Connor, Ph.D.

Director Dow TRIGA Research Reactor Subscribed and sworn to before me this )0 day of November, 2011 Notary Pub c Z

-NORY PUBLIC ounty, M ichigan

."couN M y C o m m is s io n E x p ire s :

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• Ju ne 2 8. 20 17 004111N,\\ *,it cc: Wayde Konze, R&D Director - Analytical Sciences Paul O'Connor, Director Siaka Yusuf, Reactor Supervisor

DTRR Revised response to questions: 11, 14, 15-1, 15-2, 15-3, 16, 17-1, 17-2, 17-3, 18,19, 41, 52, 53-1, 53-2, 54, 56 in support of the license renewal.

November 2011 1

11. NUREG-1537, Part 1, Section 4.2.5, "Core Support Structure" requests the applicant to provide design information pertaining to the core support structure. DTRR SAR, Chapter D, does not provide sufficient information. Please provide figures depicting the upper and lower core plates and provide the dimensions and locations of all penetrations that allow coolant to flow through them.

DTRR response:

Two aluminum grid plates fix the position of fuel elements, dummy elements and neutron source. Figure 4 is the schematic drawing of the upper grid plate with the position of the control rods, pneumatic transfer location and source indicated. The upper plate is 3/4 inch aluminum and 1.5 inch diameter holes to position the fuel, dummy elements, control rods, etc. The bottom plate is 3/4 inch aluminum and has holes to receive the end pins of the fuel and dummy elements, Thirty-six holes for the natural convection cooling are found on the lower grid plate. The water passes through the upper grid plate by means of the gap between the triflute section of the fuel and the upper grid plate. The penetrations on the lower grid plate are in concentric rings with 7, 12, and 17 holes, respectively. Figure 5 is a photograph of the lower grid plate. This photograph substantiates the as built drawings given by General Atomics, and previous Safety Analysis Reports submitted by The Dow Chemical Company.

In addition to these holes, the fuels are raised above the bottom grid plate by the fuel pin (5/8" diameter by 2" long). The gap between the fuel and the lower grid plate therefore allows the bottom coolant to flow, in an "open" configuration through the core from the bottom.

Visual inspections of the core support structure indicate there are no corrosion, no deformation, and no cracking. We believe that the structures can support the core for the next 20 and more years.

2

• F4-Rabbit Terminus

• S-AmBe Source

• D-Dummy element

• AL-Aluminum Fuel element oS1-ShimI Control Rod

• S2-Shim2 Control Rod oReg-Regulating Control Rod Figure 4. Upper grid plate 3

Figure 5. Lower grid plate looking from the bottom.

14.

NUREG-1 537, Part 1, Section 4.5, "Nuclear Designk requests the applicant to provide a detailed description of analytical methods used in the nuclear design, including computer codes used to characterize technical parameters pertaining to its reactor. DTRR SAR, Chapter D, does not provide sufficient information. Please provide descriptions of the DTRR nuclear design analyses, including the methods and the computer codes used for the analyses.

DTRR Response The DTRR nuclear design analysis was carried out using MCNP5, a general purpose Monte-Carlo codes which is capable of high fidelity modeling of a nuclear reactor. The details of the analysis and the results obtained are included in a separate report titled "Analysis of the neutronic behavior of the DTRR". The analysis was completed in October of 2011 and submitted with this response.

15. NUREG-1537, Part 1, Section 4.5.1, "Normal Operating Conditions" requests the applicant to provide a description of the limiting core configuration (LCC), the core configuration that would yield the highest power density using the fuel specified for the reactor. All other core configurations utilized by the applicant should be encompassed by the safety analysis of this configuration. The description should indicate the number, types, and locations of all core components on the grid plate including fuel, control rods, neutron reflectors, and 4

moderators.

15.1 DTRR SAR, Section D.5.5, provides a list of reactivity worths but control rod worths are not included. Please provide control rod worths specific to the LCC at the requested power level.

15.2 DTRR SAR, Section A.3, describes the original fuel configuration as having 75 stainless steel (SS)-clad elements and one Aluminum (Al)-clad element. The DTRR SAR does not provide information relating to the DTRR fuel element and control rod layout for the requested power level. Please provide a complete description of the LCC for the requested power level and provide a core diagram showing all components.

15.3 The limit on excess reactivity is established in DTRR SAR Table 4. However, the actual excess reactivity of the DTRR LCC is not identified in the DTRR SAR. Please provide the calculated excess reactivity for the LCC at the requested power level.

DTRR response:

15.1 At 300 kW the excess reactivity is limited to $3.00, and shutdown margin is $0.50 based on a cold xenon negligible condition (<$0.30). There are three control rods, namely, Shiml, Shim2 and the Regulating rod. The control rod worth for the Shiml and Shim2 is approximately $3.00 each. The Regulating rod is worth, approximately, $1.00. The reactivity worth of the control rods are Shiml $2.68, Shim2 $2.73, Regulating rod $1.01 as measured on January 11 th 2011.

The LCC is described in section 3.6 of the report titled "Analysis of the neutronic behavior of the DTRR", which was completed in October of 2011 and submitted with this response. The parameters are comparable to the measured parameters listed here.

15.2 The core configuration is found in Figure 1. The core is loaded with 79 Stainless Steel clad fuel elements, 1 Aluminum clad fuel element and 5 graphite dummies. The LCC, described in section 3.6 of the report titled "Analysis of the neutronic behavior of the DTRR" which was completed in October of 2011 and submitted with this response. It contains 76 Stainless Steel clad fuel elements and 9 graphite dummies. The other difference is that all the elements in B-ring were fresh, un-depleted new fuel elements. This LCC provides the highest power peaking factor since the most reactive fuel rods are in the most reactive positions of the reactor core and the least reactive positions contains all the dummies.

15.3. The excess reactivity of the DTRR as currently configured is $2.28 as measured at 5 Watts on January 11 th, 2011. The LCC is described in section 3.6 of the report titled "Analysis of the neutronic behavior of the DTRR". As expected the core excess of the LCC is only $2.24 which less than the DTRR TS core excess limit.

5

oF4-Rabbit Terminus

• S-AmBe Source

" - -*D-Dummy element

• AL-Aluminum Fuel element oS1-Shiml Control Rod

• S2-Shim2 Control Rod F°

-Reg-Regulating Control Rod ot-)

C

((Du Figure 1. Upper grid plate

16.

NUREG-1537, Part 1, Section 4.5.2, "Reactor Core Physics Parameters" requests the applicant to provide a description of the full set of core physics parameters for the LCC that are used in their safety analyses and the methods used to determine them. DTRR SAR, Table 4, provides some of the values cited (i.e., P3d, prompt-neutron-lifetime, fuel temperature and the void coefficient). However, it is unclear if these are generic values or if they are applicable to the LCC of the DTRR and to the safety analyses in Chapter M. Please provide a description of the full set of core physics parameters for the LCC that are used in the DTRR safety analyses and the methods used to determine them.

DTRR Response The Limiting Core Configuration (LCC) has now been developed and modeled using MCNP5 and RELAP codes. The details of the analysis and the results are described in two separate reports, 1) "Analysis of the neutronic behavior of the DTRR" and 2) "Analysis of the thermal hydraulic and reactivity insertion behavior of the DTRR. These reports are submitted with this response. The values and the sources of the parameters, peff, prompt-neutron-lifetime and fuel temperature coefficients which were used in these models are included in the report. The void coefficient is negligible was not considered in the analysis.

6 I

17. NUREG-1537, Part 1, Section 4.5.3, "Operating Limits" requests the applicant to provide information regarding the operating limits applicable to the LCC of its reactor.

DTRR SAR, Section D does not provide sufficient information.

17.1 Please describe any limits or conditions on the evaluation of excess reactivity contributors, such as those due to temperature variations and poisons (e.g.,

xenon and samarium). Please describe algebraically how DTRR determines excess reactivity showing all components.

17.2 Please describe any limits or conditions on the evaluation of shutdown margin, including a discussion of uncertainties.

17.3 Safety Limit (SL) is based on fuel temperature, and the Limiting Safety System Setting (LSSS) is based on core power (DTRR TS 2.1 and DTRR TS 2.2). Please describe the relationship between these parameters and how the DTRR operation using the LCC at the new requested power level will result in fuel temperatures that are bounded by the SL.

DTRR response:

17.1 The Limiting Core Configuration (LCC) has now been developed and modeled using MCNP5 and RELAP codes. The details of the analysis and the results are described in two separate reports, 1) "Analysis of the neutronic behavior of the DTRR" and 2) "Analysis of the thermal hydraulic and reactivity insertion behavior of the DTRR". The core excess measurements and models are done for clean cold core.

17.2 The Limiting Core Configuration (LCC) has now been developed and modeled using MCNP5 and RELAP codes. The details of the analysis and the results are described in two separate reports, 1) "Analysis of the neutronic behavior of the DTRR" and 2) "Analysis of the thermal hydraulic and reactivity insertion behavior of the DTRR". The shutdown margin measurements and models are done for clean cold core.

17.3 The DTRR has since withdrawn the request for a new power level other than 300kW.

However, the relationship between the SL, Temperature, LSSS, 300kW reactor power are described the Neutronics and Thermal-hydraulic models using MNCP5 and RELAP codes, which are submitted as separate reports with this response. The analysis show that setting the LSSS at 300kW is adequate to prevent violation of the SL of 500'C

18.

NUREG-1 537, Part 1, Section 4.6, "Thermal-Hydraulic Design" requests the applicant to provide information and analyses of thermal-hydraulic conditions in its reactor demonstrating that sufficient cooling capacity exists for steady-state operations at the maximum licensed power level. DTRR SAR, Chapter D, does not provide sufficient information. Please provide information pertaining to the minimum DNBR for the DTRR using the LCC at the new requested power level. Please describe the analytical methods used to determine the DNBR, including the core inlet and exit conditions assumed and other assumptions and correlations employed.

7

DTRR Response The Limiting Core Configuration (LCC) has now been developed and modeled using MCNP5 and RELAP codes. The details of the analysis and the results are described in two separate reports, 1) "Analysis of the neutronic behavior of the DTRR" and 2) "Analysis of the thermal hydraulic and reactivity insertion behavior of the DTRR". The DNBR for both the current core and the LCC have been modeled. The analysis show that operating at 300kW (the limiting TS) and with l5ft of water above the core, natural circulation is sufficient to maintain a DNBR of 6.76.

19.

NUREG-1 537, Part 1, Section 5.2, "Primary Coolant System" requests the applicant to provide a description of the primary coolant system, including information to substantiate the removal of heat from the fuel during maximum licensed power operation and decay heat when the reactor is shutdown. DTRR SAR, Sections E.1 and E.3, do not provide information demonstrating the adequacy of the primary system to perform this task.

Please provide information showing the adequacy of the primary system to cool the reactor under all anticipated conditions of operation at the new requested power level.

DTRR Response DTRR has since withdrawn the request for a new power level other than 300kW. The Limiting Core Configuration (LCC) has now been developed and modeled using MCNP5 and RELAP codes for 300kW. The details of the analysis and the results are described in two separate reports,

1) "Analysis of the neutronic behavior of the DTRR" and 2) "Analysis of the thermal hydraulic and reactivity insertion behavior of the DTRR". The maximum fuel temperature of the LCC is 246.7°C and is well below the technical specification of 5000C. This analysis also demonstrates that natural circulation within the reactor tank is sufficient to remove heat at the rated power of 300kW.

41.

NUREG-1537, Part 1, Section 10.2, "Experimental Facilities" requests the applicant to provide a description of the radiological considerations associated with the design and the use of the experimental facilities, generation of radioactive gases, release of fission products or other radioactive contaminants, and exposure of personnel to neutron and gamma beams. DTRR SAR, Section J.2, does not provide this information. Please provide this information for operation at the new requested power level.

DTRR response The DTRR is currently operated at varying power levels up to the licensed power of 300 kW. At this time the DTRR is withdrawing its request to up rate in power. The DTRR will continue to operate at 300 kW. All routine experiments are reviewed prior to irradiation. Dose rate on removal is estimated using power, irradiation time, decay time and composition of the samples, These estimates are noted on the TRIGA activation request form. Dose rate to the experimenter is controlled using ALARA. Samples are unloaded from the rotary specimen rack using a "fishing pole". Samples may be returned to the rotary specimen rack or a lead cave in the reactor room if the dose rate exceeds expectation. Samples are unloaded from the secondary capsules using long handled tongs in side of the fume hood which is vented HEPA filter. Operation of the 8

pneumatic system is not allowed when individuals are on the roof. Personnel are not typically in the reactor room during operation and therefore not as risk for direct exposure to neutron or gamma beams. Bends are located in both the pneumatic system and rotary specimen rack system to minimize dose rates in the reactor room. The ARM, CAM and Geiger tube continuously provide an indication of the condition in the reactor room. The schematics of the rabbit system and the directional flow of air during operation are shown in Figures 41-l a and 41*

lb.

Inserting a Rabbit Figure 41-1 a: The Pneumatic system showing air flow direction during insertion of a rabbit capsule 9

Withdrawing a Rabbit Figure 41-1b: The Pneumatic system showing air flow direction during withdrawal of a rabbit capsule Ar-41 is produced when Ar-40 in the air (P-tube, sample capsules and rabbit terminus) absorbs a neutron and is activated to Ar-41. The rabbit system discharges into a hood exhaust duct which flows at 1100 CFM (34,000 Lin). The volume of Rabbit Terminus is 301.5 cc (1.25" dia. x 15" long). The weight of Ar in this volume at 1 atm. is 3.64 mg (0.001293 g/cm 3 x 0.934% x 301.5 cm 3). Per Erdtmann NAA tables, production of 41-Ar at 300 KW is 300 dpm/microgram per irradiation minute.

3.64?ngx300 dpm x 1000 iglq A =

n x-rg f m g = 0.

2.2 x 1 0 6 dpni min gLCi The rabbit system is typically only operated approximately 15 minutes per day. In a worst case scenario, if the reactor was operated continuously during working hours (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day) with the rabbit system in operation, the total generation rate of Ar-41 would be approximately 235.2

.tCi/day.

10

The rabbit system discharges into a hood exhaust duct which flows at 1100 CFM (31,000 L/min). Assuming a 95% usage fraction of the hood, the daily exhaust rate of the hood is 4.26 x 107 Liday Using these numbers, the daily average concentration of Ar-41 being exhausted from the 1602 Building is:

235.2 C

day

-7 = 5.52 x 10.9, ICi/mL. (2) 4.2 6 xjO7._LXj1000 day L

The 10 CFR 20, Appendix B allowable effluent release concentration of Ar-41 through the air pathway is 1 x 10-8 tCi/ml. The Ar-41 releases from the reactor are less than the allowable release concentration, which corresponds to a Total Effective Dose Equivalent of approximately 28 n-rem/yr, assuming somebody was continuously present at the location of release. This is below regulatory limits for releases of radioactive material in 10 CFR Part 20. Note that actual operation of the rabbit system occurs for only about 15 minutes per day, which keeps routine releases well below the 10 CFR 20 ALARA goal of 10 mrem/yr.

If this material was released directly into the reactor room, it would be dispersed throughout the reactor room and be ventilated out of the area via the room ventilation.

Ar-41 concentrations in the reactor room can be calculated using a steady-state well-mixed box model (AIHA, 2000) by:

C = G/Q

Where, C = room concentration (ytCi/m3)

G = generation rate (IiCi/min) (0.49 pCi/min)

Q = room ventilation rate (m3/min) (50 m3/min)

An Ar-41 concentration of 9.8 x 10-9 uCi/mL could be generated. If a worker was continuously present in an environment of this concentration, during reactor operations, their Total Effective Dose Equivalent would only be 17 mrem/yr.

11

52.

NUREG-1537, Part 1, Section 13.1.1, "Maximum Hypothetical Accident" requests the applicant to provide a maximum hypothetical accident (MHA) and demonstrate that it bounds all potential credible accidents at the facility. The MHA for TRIGA reactors is typically the failure of one fuel element in the air with the release of gaseous fission products. DTRR SAR M. 1.3, analyzes a fuel failure in the pool, but it does not meet the expectation of being a bounding accident analysis. Please provide an analysis of the MHA for the DTRR that bounds all other accident analysis. Please include all assumptions, sequence of events and the potential radiological consequences.

DTRR response The rupture of a fuel element outside of the pool water is the DTRR "Maximum Hypothetical Accident" because it has the potential to result in the release of fission products and cause maximum exposure to personnel. Fuel elements are rarely, if ever, removed from the pool water, but, in order to bound the consequence analysis, it is assumed that the damaged fuel element is outside the reactor pool at the time of the accident. This type of event has been analyzed by F. C. Foushee and R. H, Peters, "Summary of TRIGA Fuel Fission Product Release Experiments", Gulf Energy and Environmental Systems Report A-10801, 1971. Similar conclusions are reported by S. C. Hawley and R. L. Kathren, "Credible Accident Analyses for TRIGA and TRIGA-fueled Reactors", NUREG/CR-2387, PNL-4028 (1982). The Dow TRIGA Reactor contains fuel with similar characteristics as the Oregon State LEU fuel (Oregon State, 2007). The Oregon State SAR calculates fission product inventory in a fuel element after being run for one year at a power level of 1.1 MW at a peak power density of 18.52 kW per fuel element with zero decay time. Correcting this initial fission product inventory for the Dow TRIGA reactor's lower peak power density of 6.08 kW per fuel element (from the analysis of a limiting core configuration, to be reported), the fission product inventory predicted for the worst-case single fuel element is presented in Tables 52-I and 52-2,. The use of the Dow TRIGA research reactor over the past 40 years indicates that the reactor has not been continuously operated for a sufficient period of time to achieve saturation of the fission product inventory and is not likely to be so operated, therefore the inventory is conservatively estimated and the doses estimated are for emergency planning only.

The fraction of gaseous fission products that will be released during a failure of a TRIGA reactor fuel element at a temperature of less than 3500 C is 1.5 x 10-5 (Foushee and Peters, 1971; Stahl, 1982). The volume of the reactor room is about 130 cubic meters, with a turnover rate of about 50 cubic meters per minute. However, no credit was taken in the consequence calculations for worker exposures in the reactor room for reductions in air concentrations due to the ventilation of the room. The effective dose equivalent for worker exposure was calculated assuming perfect mixing in the room and instantaneous release to the reactor room.

Restricted area The effective dose equivalent to an individual in the restricted area enveloped in the radioactive cloud of released halogens and noble gases was calculated as the sum of the inhalation dose of the halogens and the submersion dose of the noble gases. Dose conversion factors for inhalation were taken from 10 CFR 20, Appendix B, column 2 for the calculation of effective dose equivalent to the whole body and for the dose calculations to the thyroid for the iodine compounds. For the thyroid dose from the 12

bromine compounds and the submersion dose, dose conversion factors from FGR No. 11 (Environmental Protection Agency, 1988) were used.

Dose rates from inhalation and submersion doses in the reactor room are calculated to be 1.76 rem

/hour. The committed dose equivalent to the thyroid is calculated as 52 rem/hour. Calculations of the dose rate by isotope are summarized in Tables 52-1 and 52-2. Assuming that it takes no more than 30 minutes for the operators to bring the reactor to a safe condition and evacuate the area, if necessary, the total effective dose equivalent to reactor operators from this incident would be 0.88 rem, and the committed dose equivalent to the thyroid would be 26 rem.

Table 52-1. Dose Inside the Reactor Room from Halogens from Fuel Element Rupture Accident Scenario Halogens Activity in Concentration Inhalation Dose Dose Fuel Rod in Reactor Conversion Contribution (Ci)

Room (Ci/m3)

Factor (Sv/Bq)

(rem/hr)

Br-82 3.51E-10 2.30E-06 Br-83 2.41E-11 3.26E-04 Br-84 2.61E-11 6.45E-04 Br-84M 2.61E-l1 2.41E-05 Br-85 2.61E-1 I 8.16E-04 Br-86 2.61 E-11 1.14E-03 Br-87 2.61E-11 1.28E-03 1-131 8.89E-09 6.91E-01 1-132 1.03E-10 1.24E-02 1-133 1.58E-09 2.87E-01 1-134 3.55E-11 7.34E-03 1-135 3.32E-10 5.64E-02 1-136 8.89E-09 5.67E-01 Total Halogens 1.62 13

Table 52-2. Dose Inside the Reactor Room from Noble Gases from Fuel Element Rupture Accident Scenario Noble Activity Concentration in Submersion Dose Gases (Ci)

Reactor Room Dose Contribution (Ci/m3)

Conversion (rem/hr)

Factor (Sv/Bq)

Kr-83M 2.98E-11 3.36E-04 Kr-85 4.70E-13 2.65E-06 Kr-85M 2.98E-11 7.73E-04 Kr-87 1.42E-10 7.28E-03 Kr-88 3.60E-10 2.60E-02 Kr-89 3.60E-10 3.35E-02 Xe-131M 1.48E-12 1.17E-06 Xe-133M 5.38E-12 2.43E-05 Xe-133 6.07E-12 9.19E-04 Xe-135M 7.53E-11 2.17E-03 Xe-135 4.68E-11 7.02E-03 Xe-137 1.92E-10 2.63E-02 Xe-138 1.92E-10 2.67E-02 Kr-83M 2.98E-11 3.36E-04 Kr-85 4.70E-13 2.65E-06 Kr-85M 2.98E-1 1 7.73E-04 Kr-87 1.42E-10 7.28E-03 Total Noble Gases:

0.131 Unrestricted area The only occupied space that the reactor room is connected to in the building is the reactor control room. If the reactor room ventilation system is operational, the reactor room is at negative pressure as compared to surrounding facilities, and there will be no potential for release of the radionuclides to other areas in the building. If the reactor room ventilation system was not functional, air within the reactor control room could leak around the door to the reactor room into the control room. However, in this circumstance, the building would be evacuated of non-radiation workers to prevent additional personnel exposure. Exposures in other locations within the building would be bounded by dose estimates from air leakage into the control room.

Air leakage through a narrow opening can be predicted by:

Q = CdA[2Ap]5 P

14

Where, Q = Flow Rate (m3/sec)

Cd = Discharge coefficient (0.61 for a flat plate orifice)

A = Area of opening (m2) (0.0095 in2)

Ap = pressure differential between rooms (Pa) p = density of air (kg/m 3) (1.225 kg/m3)

Using the characteristics of the reactor room door and a conservative pressure differential between the rooms of 20 Pa, a flow rate of 0.023 m3/sec is calculated. This air flow rate is assumed to flow into the control room for 15 minutes before the building is evacuated. It is conservatively assumed that no air is released from the control room. The control room has a volume of 66.3 m3, and no credit is taken for decay of radionuclides as they permeate into the control room. Average concentrations of radionuclides generated in the control room from this scenario and calculations of the dose rate for members of the public located inside the building are summarized in Tables 52-3 and 52-4.

The total effective dose equivalent rate from inhalation and submersion in the control room are calculated to be 186 mrem /hour. The committed dose equivalent to the thyroid is calculated as 5.21 rem/hour. Assuming that it takes no more than 15 minutes for the building to be evacuated, the total effective dose equivalent to members of the public in the reactor building from this incident would be less than 47 mrem, and the committed dose equivalent to the thyroid would be less than 1.3 rem.

Exposures to operators in the control room would be bounded by the exposure estimates inside the reactor room, above.

Table 52-3. Dose Inside the Control Room from Halogens from Fuel Element Rupture Accident Scenario Halogens Release Rate 15-Minute Average Inhalation Dose into Control Concentration in Dose Contribution Room Control Room Conversion (rem/hr)

(Ci/sec)

(Ci/m3)

Factor (Sv/Bq)

Br-82 3.51E-10 2.47E-08 Br-83 2.41E-1 I 5.09E-05 Br-84 2.61E-11 L.O1E-04 Br-84M 2.61E-11 3.77E-06 Br-85 2.61E-ll 1.28E-04 Br-86 2.61E-11 1.78E-04 Br-87 2.61E-11 2.OOE-04 1-131 8.89E-09 1.08E-01 1-132 1.03E-10 1.93E-03 1-133 1.58E-09 4.48E-02 1-134 3.55E-1 I 1.15E-03 1-135 3.32E-10 8.81E-03 1-136 8.89E-09 3.31E-03 Total Halogens 0.165_

15

Table 52-4. Dose Inside the Control Room from Noble Gases from Fuel Element Rupture Accident Scenario Noble Gases Release Rate 15-Minute Average Submersion Dose into Control Concentration in Dose Contributi Room Control Room Conversion on (rem/hr)

(Ci/sec)

(Ci/m3)

Factor (Sv/Bq)

Kr-83M 2.98E-1I 5.25E-05 Kr-85 4.70E-13 4.15E-07 Kr-85M 2.98E-11 1.21E-04 Kr-87 1.42E-10 1.14E-03 Kr-88 3.60E-10 4.07E-03 Kr-89 3.60E-10 5.23E-03 Xe-131M 1.48E-12 1.82E-07 Xe-133M 5.38E-12 3.79E-06 Xe-133 6.07E-12 1.44E-04 Xe-135M 7.53E-11 3.40E-04 Xe-135 4.68E-11 1.1OE-03 Xe-137 1.92E-10 4.11E-03 Xe-138 1.92E-10 4.17E-03 Kr-83M 2.98E-1 1 5.25E-05 Kr-85 4.70E-13 4.15E-07 Kr-85M 2.98E-11 1.21E-04 Kr-87 1.42E-10 1.14E-03 Total Noble Gases:

0.0205 Unrestricted area ' offsite Doses to unrestricted areas are calculated using a normal ventilation rate for the reactor room of 50 m 3/min. Based on this ventilation rate and the reactor room volume of 130 M 3, it will take 2.6 minutes for the released radionuclides to be released to the environment when the ventilation system is running. Using the Pasquill categories and a Gaussian approach to dispersion, the maximum concentration of fission products to a member of the public is at the fence-line, 23 m from the reactor building. The short range is due to the low height of the ventilation exit, 2.73 meters (8 ft) above the building. The wind is assumed to be blowing in the direction of the fence at the time of the accident, and conservative atmospheric stability conditions (category F) are assumed. Transport calculations are performed following the guidance in Regulatory Guide 1.145 (U.S. Nuclear Regulatory Commission, 1982). To be conservative, the downwind conditions are calculated assuming a fumigation condition and equation (2) in Section 1.3.1 of the Regulatory Guide is used, as it is higher than the result from equation (1) for this scenario.

The downwind concentrations of radionuclides are calculated by multiplying the release rate by the Atmospheric Relative Concentration Value, calculated by:

16

2' 1

Q 3,ztu 10oo.

Where, X/Q = Atmospheric Relative Concentration Value (sec/m3)

U10 = Wind speed at a height of 10 m (2 m/sec)

Gy = Dispersion factor in the y-dimension at 23 m downwind (1 m)

(7, = Dispersion factor in the z-dimension at 23 m downwind (0.5 m)

Radioactive decay is conservatively not considered during transport calculations. The total effective dose equivalent to a member of the public at the unrestricted location is 6.5 mrem and the committed dose equivalent to the thyroid at this location is 127 mrem. Calculations of the dose rate by isotope for offsite exposures are summarized in Tables 52-5 and 52-6.

Table 52-5. Downwind Dose from Halogens from Fuel Element Rupture Accident Scenario Halogens Release Inh DCF Downwind Dose Rate Rate (Sv/Bq)

Concentration (rern/hr)

(Ci/min)

(Ci/m3)

Br-82 3.51E-10 2.04E-07 Br-83 2.41E-1l 2.88E-05 Br-84 2.61E-ll 5.71E-O05 Br-84M 2.61E-l1 2.14C-06 Br-85 2.61E-11 7.22E-05 Br-86 2.61E-11 1.O1E-04 Br-87 2.61E-11 9.79E-05 1-131 8.89E-09 6.11E-02 1-132 1.03E-10 1.1OE-03 1-133 1.58E-09 2.54E-02 1-134 3.55E-11 6.50E-04 1-135 3.32E-10 4.99L- 03 1-136 8.89E-09 4.56E-02 Total:

0.139 17

Table 52-6. Downwind Dose from Noble Gases from Fuel Element Rupture Accident Scenario Noble Gases Release Submersion Dose Downwind Dose Rate Rate Conversion Concentration (rem/hr)

(_i/min)

Factor (Sv/Bq)

(Ci/m3)

Kr-83M 3.36E-04 2.97E-05 Kr-85 2.65E-06 2.35E-07 Kr-85M 7.73E-04 6.84E-05 Kr-87 7.28E-03 6.44E-04 Kr-88 2.60E-02 2.30E-03 Kr-89 3.35E-02 2.84E-03 Xe-131M 1.17E-06 1.03E-07 Xe-133M 2.43E-05 2.15E&06 Xe-133 9.19E-04 8.13E-05 Xe-135M 2.17E-03 1.92E-04 Xe-135 7.02E-03 6.21&E04 Xe-137 2.63E-02 2.25E-03 Xe-138 2.67E-02 2.36E-03 Totals:

__1.14E-02 References F.C. Foushee and R. H. Peters, Summary of TRIGA Fuel Fission Product Release Experiments, Vol. 11, General Atomic Company Report Gulf EES-A10801 1; and S. Langer and N. L.

Baldwin, Fission Product Release Experiments on Uranium-Zirconium Hydride Fuels, Vol. I, General Atomic Company Report Gulf GA-A10781 (1971).

D. Stahol. Fuels for Research and Test Reactors, Status Review: July 1982. Argonne National Laboratory report ANL-83-5 (1982).

18

53. NUREG-1537, Part 1, Section 13.1.2, "Insertion of Excess Reactivity" requests the applicant to provide an analysis of reactivity insertion events. Similarly, NUREG-1537, Part 1, Section 4.5.3,"Operating Limits," requests that the applicant provide an analysis of the uncontrolled withdrawal of the highest reactivity control rod. DTRR SAR, Section M. 1.2, does not provide sufficient information regarding reactivity insertion events.

53.1 Please provide an analysis of possible reactivity insertion events for the DTRR.

53.2 Please provide an analysis of the uncontrolled rod withdrawal event for DTRR using the highest reactivity control rod.

DTRR response:

53.1 The analysis of possible reactivity insertion events for the DTRR has now been completed.

The Limiting Core Configuration (LCC) has now been developed and modeled using MCNP5 and RELAP codes. The details of the analysis and the results are described in two separate reports, 1) "the neutronic behavior of the DTRR" and 2) "Analysis of the thermal hydraulic and reactivity insertion behavior of the DTRR". The reports are included with this response.

53.2 The analysis of uncontrolled rod withdrawal events for the DTRR using the highest reactive control rod has now been completed. The Limiting Core Configuration (LCC) has now been developed and modeled using MCNP5 and RELAP codes. The details of the analysis and the results are described in two separate reports, 1) "Analysis of the neutronic behavior of the DTRR" and 2) "Analysis of the thermal hydraulic and reactivity insertion behavior of the DTRR". The reports show that the SL was to challenged as a result of these reactivity insertions.

54.

NUREG-1537, Part 1, Section 13.1.3, "Loss of Coolant" requests the applicant to provide analysis that assures that doses to the public that could result from a loss of coolant accident do not exceed 10 CFR Part 20 limits. DTRR SAR, Section M. 1.1, Table 7 presents exposures resulting from a loss of coolant accident. There is no statement regarding occupational or public dose limits and whether they are met. Please explain this accident analysis in further detail and in terms of meeting the regulatory limits.

DTRR response:

The water level in the surrounding area is above the core height and therefore tank breach will not result in a total loss of coolant. However, a site-specific analysis was completed for an uncovered core after several hours of operation at 300 kW and reported in the SER for DTRR, U.S. Nuclear Regulatory Commission, 1989). The results from this analysis are in the following table:

19

Time after complete Direct radiation - 18 ft Indirect Radiation loss of coolant directly above core shield top edge of the (R/hr) tank (R/hr) (Position 1 or 4 in Figure 54-1) 10 seconds 3000 0.78 1 day 360 0.090 1 week 130 0.042 1 month 35 0.012 Exposures inside the Reactor Room The elevated radiation fields generated from this hypothetical accident will be highly collimated above the reactor pool. The core sits inside a 17" diameter opening in the reflector. The top of the reflector is 16' below the top of the 76" diameter reactor pool. The reactor room roof is 12' above the top of the reactor pool. Based on the maximum scattering angle that direct radiation may be emitted from the reactor (from the far left side of the core to the far right side of the reactor pool, and vice-versa), the direct radiation beam from the core will only have a diameter of 12.3 feet at the roof of the reactor room, which is still much smaller than the entire room.

Therefore, workers and members of the public located outside of the reactor room will not be exposed to the direct radiation from the reactor core. Responders to the incident will also avoid the area of the room directly above the core in order to avoid exposure to the direct radiation from the reactor core.

Exposures outside the Reactor Building To estimate potential radiation exposure levels from scattered radiation outside of the reactor room (indicated as position 2 in Figure 54-1, which is outside of the reactor building) measured radiation scatter data from dose rates during the operation of the neutron beam tube will be used to estimate potential dose rates generated around the reactor building. Specifically, during an operation of the central beam tube in 1991, measurement surveys were made and reported to the Radiation Safety Committee, (The Dow Chemical Company, 1991). This report documented measured gamma and neutron doses during the operation of a neutron beam tube that consisted of a 1.5" streaming pathway that allowed collimated neutrons to travel through a helium filled aluminum pipe without being shielded. Figure 54-1 shows the layout of the reactor pool and the areas surrounding the reactor room.

The highest total gamma and neutron dose rate measured during this survey was 4.2 mrem/hr and was located on the east side of the building (marked as location #2 in the drawing). During this survey, a measurement of 5.9 mrem/hr was made also on the east side of the reactor but outside of the intense direct radiation field (marked as location #1 in the drawing). This is proportional to the 1991 predicted radiation field of 780 mrem/hr (a factor of 132), immediately after an incident. Therefore, using this factor, the 4.2 mrem/hr measured outside of the reactor building will be equivalent to 554 mrem/hr. This number represents the highest expected dose rate, outside the reactor room, from a totally exposed core. Note that this takes no credit for the decay 20

of the fission products during the time that it would take to drain the reactor pool, which would lower the calculated dose rates significantly.

Reactor room radiation alarms are monitored by Dow Security. In the event of an alarm, Dow Security would immediately respond and clear the area around the reactor of personnel. This response would occur within 30 minutes. Therefore, the highest potential dose to a member of the public from this incident would be a Total Effective Dose Equivalent of 227 mrem, assuming an individual was located immediately outside the emergency door on the east side of the reactor for the entire duration of the incident until they were cleared from the area by security.

D (mrem) = 554 mrem/hr

• 0.5 hr =227 mrem This is the maximum conservative estimate of the public dose during a hypothetical loss of coolant accident.

Exposures in the Control Room To estimate potential radiation exposure levels from scattered radiation inside the reactor control room (position 3 in Figure 54-1), a similar method will be used to predict scattered radiation dose rates. The dose rate measured during the beam tube experiment in the control room reactor console was 0.55 mrem/hr and the dose rate on the north side of the reactor as 8.5 mrem/hr (marked as location #4 in Figure 54-1). Scaling the dose rates using the factor 91 (i.e. 780/8.5),

as before, a control room dose rate of 50 mrem/hr from scattered radiation is the calculated expected dose rate immediately following such an incident.

The Dow Chemical Company operates an on-site fire department, which would respond to this incident and be able to refill the reactor pool within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the incident occurring. Assuming that this response requires an individual to be located in the reactor room for 30 minutes and in the control room for the remaining 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, a total employee Total Effective Dose Equivalent of less than 565 mrem would be received by an employee due to this incident.

D (mrem) = (780 mrem/hr

• 0.5 hr) + (50 mrem/hr

• 3.5 hr) = 565 mrem This is the maximum conservative estimate of the occupational dose during a hypothetical loss of coolant accident.

It is important to note that location 3 shown in figure 54-1 and whose dose rate is addressed above, is the only occupied location just outside of the reactor room.

21

0r 0~

0 07 C-i

'0 N I Note: Drawing not to scale I

56.

NUREG-1537, Part 1, Section 13.1.6, "Experiment Malfunction" requests the applicant to provide analysis of an experiment malfunction event. DTRR SAR, Section M. 1.4, does not include analysis of an experiment failure with release of radioactivity. Please provide an analysis and consequences of an experiment malfunction for the experiment with the highest potential release of radioactivity.

DTRR response:

All experiments are reviewed before insertion and all experiments are separated from the fuel cladding by at least one barrier for example, the pneumatic tube and central thimble. All experiments that could damage components of the reactor are required by technical specification to be double encapsulated. Samples are typically under 8 grams, with a majority of the samples irradiated consisting of carbon, hydrogen and oxygen (plastic and organics).

The dose consequences of the release of 10 microCi of 1-131-1-135 are calculated for workers assuming that 100% of the material is released into the reactor room, and the ventilation system is shut off, causing the material to be trapped within the rector room. It is assumed that the worker spends 60 minutes within the reactor room to resolve the incident. The release of the iodine would generate a concentration of 7.69x10-8 Ci/m 3 inside the reactor room.

Worker doses are calculated using Dose Conversion Factors for effective dose and dose to the thyroid for 1-131 (conservative for iodine radionuclides) from Federal Guidance Report #11 (Eckerman, et al. 1988). This calculation will bound the dose to any member of the public who is located within the building and any exposure estimates to workers located within laboratories adjacent to the reactor room.

The total effective dose to the worker is calculated to be 3.04 mrem, and the thyroid dose is calculated to be 99.7 mrem.

Note that from the exposure scenarios for a ruptured fuel element, in response to RAI 52, shows that 1-131 contributed 42% of the total dose for the scenario. Since this accident scenario would have a similar mix of radionuclides, it is not anticipated that contributions from additional radionuclides would increase this dose estimate by more than a factor of 3, which would keep exposure estimates from this scenario well below the dose estimates for the fuel rupture accident scenario and 10 CFR Part 20 exposure limits.

For exposures to members of the public, it is assumed that the ventilation system is operational and vents the released iodine outside the reactor building. Based on the ventilation rate and volume of the reactor room, it would take 2.6 minutes to have one full air change of the reactor room and release all of the iodine, which results in a release rate of 6.41 x 10-8 Ci/sec from the facility. Downwind air concentrations at the plant fence-line located 23 m to 23

the west of the reactor building are determined following guidance in Regulatory Guide 1.145 (U.S. Nuclear Regulatory Commission, 1982) to be 6.80 x 10-9 Ci/m 3. Doses to members of the public are calculated using Dose Conversion Factors for effective dose and dose to the thyroid for 1-131 (conservative for iodine radionuclides) from Federal Guidance Report #11 (Eckerman, et al. 1988).

The total effective dose equivalent to the maximally exposed offsite member of the public is calculated to be 1.4 mrem and the thyroid dose is calculated to be 47.1 mrem.

References U.S. Environmental Protection Agency. 1988. Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion. Federal Guidance Report No. 11. Washington, D.C.: U.S. Environmental Protection Agency.

The Dow Chemical Company. 1991. Radiation Dose and Exposure Rate Evaluations During Neutron Radiographic Operation of the Dow TRIGA* Research Reactor at 100 and 240 Kilowatts, Special Analysis, Michigan Division Analytical Laboratory, 1602 Building, November 19, 1991. HEH RAD14(8).

U.S. Nuclear Regulatory Commission. 1983. Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. Regulatory Guide 1.145.

Washington, DC: U. S. Nuclear Regulatory Commission.

U.S. Nuclear Regulatory Commission. 1989. Safety Evaluation Report related to the renewal of the facility license for the research reactor at the Dow Chemical Company. NUREG-1312.

Washington, DC: U.S. Nuclear Regulatory Commission.

U.S. Nuclear Regulatory Commission. Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. Nuclear Regulatory Commission, Washington, DC.

24

ANALYSIS OF THE NEUTRONIC BEHAVIOR OF THE DOW TRIGA RESEARCH REACTOR Submitted to the NRC in support of the DTRR License Renewal Prepared by:

Michael R Hartman 11 October 2011

TABLE OF CONTENTS LIST OF TABLES.................................................................................................................................

ii LIST OF FIGURES..............................................................................................................................

iii

1.

Introduction..............................................................................................................................

1

2.

Core M odel...............................................................................................................................

1

3.

M odel Results...........................................................................................................................

9 3.1 Effective Delayed Neutron Fraction............................................................................

9 3.2 Control Rod W orth.....................................................................................................

10 3.3 Shutdow n M argin.....................................................................................................

12 3.4 Fuel Prom pt-Tem perature Coefficient.....................................................................

12 3.5 Core Power Distribution............................................................................................

13 3.6 Lim iting Core Configuration.......................................................................................

14

4.

Sum m ary.................................................................................................................................

15 APPENDIX A..................................................................................................................................

A-1

LIST OF TABLES Table 1. Burnup history for various DTRR core configurations.................................................

6 Table 2. Material composition for core components within the MCNP5 model of the D T R R co re..........................................................................................................................

8 Table 3. Summary of integral control rod worths....................................................................

12 Table 4. Summary of MCNP5 prompt-temperature coefficient calculations...........................

13 ii

LIST OF FIGURES Figure 1.

Vertical cross section of the MCNP5 model developed for the DTRR.......................

1 Figure 2.

Horizontal cross section of the MCNP5 model developed for the DTRR taken at the mid-plane of the active fuel region......................................................

2 Figure 3.

Horizontal cross section of the MCNP5 model developed for the DTRR taken at the mid-plane of the active fuel region, demonstrating the details of the model within the reactor core region. Grid locations for select locations are shown in parentheses. (2011 core configuration)...............................

2 Figure 4.

DTRR core configuration for the 1967 critical core. (72 fuel elements)................... 4 Figure 5.

DTRR core configuration in 1987. This core arrangement served as the reference core for the period 1967-1991. (76 fuel elements).................

4 Figure 6.

DTRR core configuration following the addition of fuel to grid locations F15 and F17 in 1991. (78 fuel elem ents)..........................................................................

5 Figure 7.

DTRR core configuration originally established in 1997. This core arrangement was retained following a fuel shuffle in 2001. (78 fuel e le m e n ts)........................................................................................................................

5 Figure 8.

DTRR core arrangement in 2005 following the addition of fuel to grid locations F25 and F29. This core configuration represents the current core configuration. (80 fuel elem ents)..............................................................................

6 Figure 9.

Schematic illustration of a standard TRIGA fuel element (left). The cross sectional view (right) displays the internal structure of the fuel element.

Note that each of the individual fuel slugs was divided into three radial segments within the MCNP5 model to permit the determination of intra-rod power distributions and uranium depletion......................................................

7 Figure 10. Integral rod worth curve for the SHIM 1 rod based on rod positions from the 2011 DTRR rod calibration procedure...............................................................

10 Figure 11. Integral rod worth curve for the SHIM 2 rod based on rod positions from the 2011 DTRR rod calibration procedure...............................................................

11 Figure 12. Integral rod worth curve for the REGULATING rod based on rod positions from the 2011 DTRR rod calibration procedure......................................................

11 Figure 13. Results of the MCNP5 calculation of the prompt-temperature coefficient for the DTRR fuel. The magnitude of the prompt-temperature coefficient is shown as a function of temperature with the average value over a temperature region shown above that region........................................................

13 Figure 14. Map of the DTRR core showing the power per element at the maximum allowable operating power of 300 kW for the 2011 core configuration. The hot rod is found to be in location B-5 with an element power of 5.91 kW.

(80 fuel elem ents)..................................................................................................

14 iii

Figure 15. Map of the DTRR core showing the power per element at the maximum allowable operating power of 300 kW for the limiting core configuration.

The hot rod is found to be in location C-6 with an element power of 6.08 kW. (76 fuel elem ents)............................................................................................

15 iv

1. Introduction The following report summarizes an investigation into the neutronic behavior of the DOW TRIGA Research Reactor (DTRR). The DTRR is a TRIGA Mark-I reactor which is licensed to operate at powers up to 300 kW. The reactor consists of a series of six concentric rings in which are located fuel elements, graphite reflectors, and various experimental facilities.

The core is surrounded by a cylindrical annulus of graphite which acts as a neutron reflector. The core and reflector are located in a below-grade aluminum tank filled with high-purity water. The water acts as a neutron moderator, a coolant, and as a radiation shield.

The purpose of this report is to provide a modern analysis of the neutronic characteristics of the DTRR in support of a 20-year license renewal through the U.S. Nuclear Regulatory Commission.

Modeling of the DTRR was done using MCNP5 1, a general purpose Monte-Carlo code which is capable of high-fidelity modeling of the DTRR. The results of the MCNP5 modeling provide a basis to evaluate the neutronic performance of the DTRR in support of the license renewal.

2. Core Model The MCNP5 model was constructed using facility drawings and technical data acquired from General Atomics (GA). Representative views of the MCNP5 model are shown in Figures 1,2, and 3.

REGULATING Lazy Susa Reactor cor Central thimble (

ular graphite ctor assembly ter surrounding ular graphite lector assembly Figure 1. Vertical cross section of the MCNP5 model developed for the DTRR.

1. "MCNP - A General Monte Carlo N-Particle Code, Version 5," LA-CP-03-0245, F. B. Brown, Ed., Los Alamos National Laboratory (2003).

1

Water surrounding annular graphite reflector assembly Annular graphite reflector assembly Reactor core Figure 2. Horizontal cross section of the MCNP5 model developed for the DTRR taken at the mid-plane of the active fuel region.

Graphite reflector element SHIM 1 (C-11)

Central thimble (A-1) 0 a P-tube terminus (F-4)

Fuel element SHIM 2 (C-3)

, Neutron source (F-9)

REGULATING rod (E-13)

Figure 3. Horizontal cross section of the MCNP5 model developed for the DTRR taken at the mid-plane of the active fuel region, demonstrating the details of the model within the reactor core region. Grid locations for select locations are shown in parentheses. (2011 core configuration) 2

Historically, the arrangement of fuel and experimental facilities within the DTRR has consisted of essentially six different configurations. The detailed location-by-location arrangement of these configurations is provided in Appendix A. The first of these arrangements, shown in Figure 4, corresponds to the initial core configuration which was attained when the DTRR was first taken critical on 06 July 1967. The fuel for this initial core loading consisted of stainless steel-clad used fuel elements which were purchased from GA. Following the initial criticality additional fuel, including a new aluminum-clad element, was added to increase the core excess reactivity.

Following the addition of this additional fuel, the DTRR core assumed a configuration similar to that which existed during the 1987 fuel inspection, shown in Figure 5.

In 1991, two new stainless-steel clad fuel elements were introduced into the DTRR in grid locations F15 and F17, replacing the graphite reflector elements which had previously occupied these locations in the 1987 core configuration. The 1991 core arrangement is shown in Figure 6.

In 1997, several elements within the B-ring and C-ring were interchanged, and additionally, the graphite elements in grid locations El and E19 were interchanged with the fuel elements in F1 and F29, respectively, resulting in the arrangement shown in Figure 7. In 2001, the new fuel elements which were introduced into grid location F15 and F17 in 1991 were exchanged with the fuel elements in B5 and B4, respectively. This fuel movement resulted in a core with the same fuel arrangement as that shown in Figure 7 for the 1997 core. In 2005, graphite reflector elements in grid locations F25 and F29 were replaced with new stainless-steel clad fuel elements. The 2005 core configuration, shown in Figure 8, has remained unchanged and represents the current arrangement of the DTRR core.

The DTRR is fueled with Standard TRIGA fuel comprised of a matrix of ZrHx (x - 1.6) containing 8,5 wt. % of uranium with a nominal enrichment of 20 wt. % U-235. The arrangement of a typical fuel element is shown in Figure 9.

All but one of the fuel elements in the DTRR core have stainless steel cladding, while the remaining element (F-28) is clad with aluminum. One of the most important aspects of performing an accurate analysis of the neutronic performance of the DTRR is the determination of the fuel composition throughout the lifetime of the core. While the "as manufactured" data was available for the fuel received from GA for the DTRR, the fuel was used and detailed records of the burnup history were unavailable. To overcome this limitation, a MCNP5 model of the 1967 critical core was established using the as manufactured data for the initial fuel composition.

The model was used to establish the average intra-rod power distribution for the nine fuel segments within each element. Using this intra-rod distribution, the U-235 of each fuel element was reduced until good agreement was achieved between the model and the as-measured core excess reactivity for the 1967 critical core configuration.

It was determined that a U-235 depletion of 10.55% for each element resulted in a MCNP5 model with a core excess of $0.21 which is to be compared with the 1967 critical core which had a measured core excess of $0.13. After establishing the initial U-235 content of each fuel element in the 1967 core, the various core arrangements and the recorded energy produced in these arrangements, 3

Figure 4. DTRR core configuration for the 1967 critical core. (72 fuel elements)

Figure 5. DTRR core configuration in 1987. This core arrangement served as the reference core for the period 1967-1991. (76 fuel elements) 4

Figure 6. DTRR core configuration following the addition of fuel to grid locations F15 and F17 in 1991. (78 fuel elements)

Figure 7. DTRR core configuration originally established in 1997. This core arrangement was retained following a fuel shuffle in 2001. (78 fuel elements) 5

Figure 8. DTRR core arrangement in 2005 following the addition of fuel to grid locations F25 and F29. This core configuration represents the current core configuration. (80 fuel elements) as detailed in Table 1,was used to deplete the U-235 content of each fuel segment within each element and determine the current value for use in the 2011 MCNP5 model. The material compositions used for other components within the DTRR MCNP5 model are listed in Table 2.

Table 1. Burnup history for various DTRR core configurations.

1967 to 1991 9.60 1991 to 1997 8.51 1997 to 2001 6.59 2001 to 2005 6.94 2005 to 2011 8.70

Fuel Ha TriFlute Fuel Cl aphite el Slug el Slug Pin el Slug aphite Lower Fitting Note: The arrangement shown is for a stainless steel clad element with a cladding thickness of 0.020". For an aluminum clad element, there is no zirconium pin and the cladding thickness is 0.030".

Figure 9. Schematic illustration of a standard TRIGA fuel element (left). The cross sectional view (right) displays the internal structure of the fuel element. Note that each of the individual fuel slugs was divided into three radial segments within the MCNP5 model to permit the determination of intra-rod power distributions and uranium depletion.

7

Table 2. Material composition for core components within the MCNP5 model of the DTRR core.

Physical:

Material Density.

Element Mass Fraction

[g/cm31 C

3.500E-04 Mn 1.OOOE-02 Si 5.OOOE-03 Type 304 P

2.500E-04 Stainless Steel S

1.OOOE-04 Cr 1.850E-01 Ni 9.250E-02 Fe 7.068E-01 Graphite Reflector 1.560 C

1.000 in fuel element Graphite in Graphite Reflector 1.620 C

1.000 Elements Graphite in Annular Reflector Ase by1.75 C

1.000 Assembly Zirconium Fuel Pin 6.506 Zr 1.000 Al 1.000 Aluminum 1100 2.700 B

1.OOOE-06 Si 6.OOOE-03 Fe 3.500E-03 Cu 2.800E-03 Mn 7.OOOE-04 Aluminum 6061-T6 2.700 Mg 1.OOOE-02 Cr 2.OOOE-03 Zn 1.291E-03 B

1.OOOE-06 Al 9.7373E-01 H

1.119E-01 Water 1.000 0

8.881E-01 N

0.800 Air 1.029E-03 N

0.200 0

0.200 C

3.105E-04 Mn 8.871E-03 Stainless Steel + Water Mix Si 4.435E-03 P

2.218E-00 for Triflute Region (all S

8.871E-05 elements except B4, B5, F25, 4.429 C

8.871E-05 and F29 in the 2011 Core Arrangement)

Ni 8.206E-02 Fe 6.270E-01 H

1.263E-02 0

1.003E-01 C

2.781E-04 Mn 7.947E-03 Si 3.973E-03 Stainless Steel + Water Mix P

1.987E-04 for Triflute Region (elements 3.262 1.470E-01 B4, B5, F25, and F29 in the Cr 2011 Core Arrangement)

Ni 7.351E-02 Fe 5.617E-01 H

2.297E-02 0

1.824E-01 8

Physical Material Density Element MassmFraction.

[g/cm3]

C 3.496E-04 Mn 9.988E-03 Si 4.994E-03 P

2.497E-04 Stainless Steel + Air S

9.988E-05 7.867E-01 in Lazy Susan Cr 1.848E-01 Ni 9.239E-02 Fe 7.060E-01 N

9.418E-04 O

2.355E-04 Si 5.979E-03 Fe 3.488E-03 Cu 2.790E-03 Mn 6.976E-04 Aluminum + Air Mg 9.966E-03 2.709E-01 Cr 1.993E-03 in Lazy Susan Zn 1.287E-03 B

9.966E-07 Al 9.704E-01 N

2.735E-03 O

6.837E-04 Boron Carbide B

7.826E-01 in Control Rods 1.764 C

2.174E-01 Si 4.378E-03 Fe 2.554E-03 Cu 2.043E-03 Mn 5.108E-04 Aluminum + Water for Mg 7.297E-03 Triflute Region of Graphite 1.850 Cr 1.459E-03 Reflector Elements and

.8 C.422E-04 Aluminum-Clad Fuel Zn 9.422E-04 B

7.297E-07 Al 7.105E-01 H

3.024E-02 O

2.400E-01

3. Model Results 3.1 Effective Delayed Neutron Fraction The effective delayed neutron fraction for the MCNP5 model representing the 2011 core configuration was determined using the expression kP fleff = 1-where k, is the system eigenvalue assuming all neutrons from fission are born with the energy distribution appropriate for prompt neutrons and kP÷d is the system eigenvalue where fission neutrons are born with an appropriately weighted energy distribution for both prompt and delayed neutrons.

The effective delayed neutron fraction was determined to by 0.0070 +/-

0.0003. This is identical to the value given in the DTRR Safety Analysis Report.

9

3.2 Control Rod Worth The SHIM 1, SHIM 2, and REGULATING rods were calibrated using the rod positions from the actual rod calibration performed in January of 2011. Calibrating the rods in this manner is beneficial in that it permits the rod worths calculated using MCNP5 to be directly compared with experimental data, and in addition, it provides a number of critical rod configurations for the DTRR which can be utilized to assess the validity of the DTRR MCNP5 model for the 2011 core configuration.

MCNP5 simulations were performed to simulate the DTRR rod calibration procedure. The fourteen critical core configurations were used to determine the bias of the MCNP5 model for the 2011 core, and a bias of $0.035 + $0.053 was found. This implies that the MCNP5 model for the 2011 DTRR core provides a very good representation of the actual 2011 core. The rod pull simulations were used to determine the reactivity associated with each rod pull, and these were summed to determine the integral rod worth for each curve. Note that due to lower worth of the REGULATING rod, it is only possible to determine a portion of the integral rod worth curve for the SHIM 1 and SHIM 2 rods, and the remainder of the curve is determined by fitting a curve to the data and extrapolating it over the uncalibrated portion of the rod. The results of the rod calibrations are shown in Figures 10, 11, and 12, and the integral rod worths are compared with the experimentally determined values in Table 3. The MCNP5 results are noted to be systematically high relative to the measured values. This deviation may be due to the choice of the prompt-neutron lifetime utilized in determining the measured data.

4 3.5 3

• Measured 2.5 2

U* Calculated 01.

0 1.

u0.5 0

M 0

200 400 600 800 1000 Position [ steps]

Figure 10. Integral rod worth curve for the SHIM 1 rod based on rod positions from the 2011 DTRR rod calibration procedure.

10

4 c-t0

-o 0

W to C

Ecc 3.5 3

2.5 2

1.5 1

0.5 0

0 200 400 600 800 Position [ steps]

1000 Figure 11. Integral rod worth curve for the SHIM 2 rod based on rod positions from the 2011 DTRR rod calibration procedure.

1.5 1.3 1.1 0.9

• Measured t

U Calculated o0.7 S0.5 0

00 0.3

• ~0.1

¢* -0.1 0

200 400 600 Position [ steps]

800 1000 Figure 12. Integral rod worth curve for the REGULATING rod based on rod positions from the 2011 DTRR rod calibration procedure 11

Table 3. Summary of integral control rod worths.

Rod Measured Integral MCNP5 Calculated Rod Worth [$1 Integral Rod Worth [$]

REGULATING 1.01 1.40+0.11 SHIM 1 2.60 3.85 + 0.20 SHIM 2 2.72 3.50+0.19 Sum of all rods 6.33 8.75 + 0.29 3.3 Shutdown Margin The shutdown margin for the current DTRR core was determined using the MCNP5 calculated integral rod worths, listed in Table 3, in conjunction with the MCNP5 calculated core excess reactivity of $2.81 + 0.04. The calculated shutdown margin was found using the relationship, Shutdown _ Total worth Worth of most Core excess Margin of all rods reactive rod reactivity This is the same methodology utilized by the DTRR to calculate shutdown margin, based upon measured control rod worths. For the current core, the MCNP5 based shutdown margin was determined assuming that the SHIM 1 rod was fully withdrawn from the core, yielding a shutdown margin of $2.09 + 0.36. The calculated shutdown margin exceeded the minimum shutdown margin of $0.50 specified within the DTRR Technical Specifications for all core conditions.

3.4 Fuel Prompt-Temperature Coefficient The prompt-temperature coefficient for the DTRR, cxf, was calculated by varying the fuel temperature within the MCNP5 model along with the thermal neutron libraries, S(xc,3), for the hydrogen and zirconium within the zirconium hydride matrix which contains the fuel. Due to the limited cross section libraries available within MCNP5, simulations could only be performed at a limited number of temperatures. Calculations were performed at 293.6 K, 600 K, 900 K, and 1200 K. There are no S(cx,p) libraries for zirconium and hydrogen at 900 K, so the 900 K data point was obtained by simulating the fuel temperature at 900 K through the proper choice of cross section libraries for the isotopes found within the fuel, and then performing two simulations using S((x,p) data at 800 K and 1000 K. The MCNP5 simulation resutls corresponding to S(c3,p) at 800 K and 1000 K were then arithmetically averaged to determine an appropriate value at 900 K. The results of this calculation are shown in Figure 13 and summarized in Table 4 below.

w0-

3-1/b.

CU-F- 0 4-ý

-CL 0.030 0.025 0.020 0.015 0.010 0.005 0.000 0.0198 0

200 400 600 800 1000 Temperature [°C]

Figure 13. Results of the MCNP5 calculation of the prompt-temperature coefficient for the DTRR fuel. The magnitude of the prompt-temperature coefficient is shown as a function of temperature with the average value over a temperature region shown above that region.

Table 4. Summary of MCNP5 prompt-temperature coefficient calculations.

327 to 627

-0.0198 627 to 927

-0.0171 3.5 Core Power Distribution Using the MCNP5 model, the core power distribution for the 2011 core configuration was found for an operating power of 300 kW with all rods fully withdrawn from the core. The power produced in each element is shown in Figure 14. The average element power was determined to be 3.75 kW while the element which produced the maximum power was found to be in grid location B-5, producing 5.91 kW.

13

Figure 14. Map of the DTRR core showing the power per element at the maximum allowable operating power of 300 kW for the 2011 core configuration. The hot rod is found to be in location B-5 with an element power of 5.91 kW. (80 fuel elements) 3.6 Limiting Core Configuration In an effort to determine the maximum power per element which could be expected for the DTRR core, a hypothetical core, denoted as the limiting core configuration, was modeled using MCNP5.

In the limiting core configuration, the fuel in the B-ring was replaced with fresh fuel which had the nominal composition of the fuel added to the core in 2005. This fuel was chosen to be representative of fuel which would be manufactured under current fuel specifications and modern fabrication techniques. In addition to the replacement of the burnt fuel in the B-ring with fresh fuel, four low-power elements on the periphery of the core were replaced with graphite reflector elements. The model was again run with all rods completely withdrawn from the core. The calculated core excess for the limiting core configuration was $2.24 + 0.06. The average element power was determined to be 3.95 kW while the element which produced the maximum power was found to be in grid location C-6, producing 6.07 kW. The limiting core configuration represents an extreme modification of the current 2011 core configuration, and while the hot rod was found to produce more power in the limiting core configuration, it was only -3% higher than in the current 2011 core configuration. The limiting core configuration and the power per element at a core power of 300 kW are shown in Figure 15.

14

Figure 15. Map of the DTRR core showing the power per element at the maximum allowable operating power of 300 kW for the limiting core configuration. The hot rod is found to be in location C-6 with an element power of 6.08 kW. (76 fuel elements)

4. Summary The DTRR was simulated using MCNP5 to determine select core physics parameters such as beta-effective (Peff),

integral control rod worth, shutdown margin, and the fuel prompt-temperature reactivity coefficient (cu).

The results of this analysis demonstrate that the licensing basis established in the DTRR Technical Specifications is valid, and furthermore that the reactor physics parameters used in various safety analyses within the DTRR Safety Analysis Report are in reasonable agreement with the values calculated herein.

15

APPENDIX A

SUMMARY

OF DTRR CORE LOADING CONFIGURATIONS UTILIZED IN THE ANALYSIS OF THE NETURON BEHAVIOR A-i

The table below summarizes the core loading for the core configurations utilized within this analysis.

Grid locations shaded in green represent a change in the loading of that particular location relative to the prior core configuration. The yellow shading for grid location F28 indicates the presence of an aluminum clad element in that location.

Element Identification Year Grid Location 1967 1987 1991 1997 2001 2005 2011 Al B1 4099 4099 4099 2402 2402 2402 B2 4100 4100 4100 4076 4076 4076 B3 4115 4115 4115 4115 4115 4115 4115 B4 4093 4093 4093 10220 10220 B5 4116 4116 4116 10219 10219 B6 4059 4059 4059 2437 2437 2437 Cl 4076 4076 4076 4093 4093 4093 C2 2426 2426 2426 2426 2426 2426 2426 C3 SHIM 2 SHIM 2 SHIM 2 SHIM 2 SHIM 2 SHIM 2 SHIM 2 C4 2363 2363 2363 2363 2363 2363 2363 CS 2433 2433 2433 2433 2433 2433 2483 C6 2424 2424 2424 2424 2424 2424 2424 C7 2402 2402 2402 4099 4099 4099 CS 2437 2437 2437 2442 2442 2442 C9 2454 2454 2454 2454 2454 2454 2454 Clo 2364 2364 2364 1

4116 4116 4116 Cll SHIM 1 SHIM 1 SHIM 1 SHIM 1 SHIM 1 SHIM 1 SHIM 1 C12 2442 2442 2442 2364 2364 2364 D1 2398 2398 2398 2398 2398 2398 2398 D2 2384 2384 2384 2384 2384 2384 2384 D3 2395 2395 2395 2395 2395 2395 2395 D4 2378 2378 2378 2378 2378 2378 2378 D5 2381 2381 2381 2381 2381 2381 2381 D6 2396 2396 2396 2396 2396 2396 2396 D7 2370 2370 2370 2370 2370 2370 2370 D8 2440 2440 2440 2440 2440 2440 2440 D9 2432 2432 2432 2432 2432 2432 2412 DO0 2434 2434 2434 2434 2434 2434 2434 D11 2455 2455 2455 2455 2455 2455 2455 D12 2453 2453 2453 2453 2453 2453 2453 D13 2438 2438 2438 2438 2438 2438 2438 D14 2421 2421 2421 2421 2421 2421 2421 D15 2429 2429 2429 2429 2429 2429 2429 D16 2365 2365 2365 2365 2365 2365 2365 A-2

D17 D18 El E2 E3 E4 ES E6 E7 E8 E9 El0 Ell E12 E13 E14 E15 E16 E17 E18 E19 E20 E21 E22 E23 E24 F1 F2 F3 F4 F5 F6 F7 F8 F9 F10 Fll F12 F13 F14 F15 F16 F17 F18 F19 2408 2449 2428 2367 2399 2415 3398 2416 2403 2392 2423 2431 2412 3021 REGULATING 2390 2422 3015 2427 2417 3509 4056 4124 3823 3824 3822 4092 4073 4075 RABBIT 4112 4101 4052 4067 2379 2372 2444 2441 2420 WATER WATER WATER WATER WATER WATER 2408 2499 2367 2399 2415 3398 2416 2392 2423 2439 2412 3021 REGULATING 2390 2422 2427 2417 4056 4124 2428 3824 3822 4092 4073 4075 RABBIT 4112 4101 4052 4067 2372 2403 2408 2499 GRAPHITE 2367 2399 2415 3398 2416 2379 2392 2423 2439 2412 3021 REGULATING 2390 2422 2386 2427 2417 GRAPHITE 4056 4124 2428 3824 3822 4092 4073 4075 RABBIT 4112 4101 4052 4067 SOURCE 2372 2403 GRAPHITE 2408 2449 2367 2399 2415 3398 2416 2379 2392 2423 2439 2412 3021 REGULATING 2390 2422 2386 2427 2417 4056 4124 2428 3824 3822 4073 4075 RABBIT 4112 4101 4052 4067 SOURCE 2372 2403 GRAPHITE 2420 2441 2408 2449 4092 2367 2399 2415 3398 2416 2379 2392 2423 2439 2412 3021 REGULATING 2390 2422 2386 2427 2417 2380 4056 4124 2428 3824 3822 GRAPHITE 4073 4075 RABBIT 4112 4101 4052 4067 SOURCE 2372 2403 GRAPHITE 2420 2441 2431 3015 GRAPHITE 2408 2449 4092 2367 2399 2415 3398 2416 2379 2392 2423 2439 2412 3021 REGULATING 2390 2422 2386 2427 2417 2380 4056 4124 2428 3824 3822 GRAPHITE 4073 4075 RABBIT 4112 4101 4052 4067 SOURCE 2372 2403 GRAPHITE 2420 2441 4059 2431 4100 3015 GRAPHITE 2408 2449 4092 2367 2399 2415 3398 2416 2379 2392 2423 2439 2412 3021 REGULATING 2390 2422 2386 2427 2417 2380 4056 4124 2428 3824 3822 GRAPHITE 4073 4075 RABBIT 4112 4101 4052 4067 SOURCE 2372 2403 GRAPHITE 2420 2441 4059 2431 4100 3015 GRAPHITE A-3 10219 2431 10220 3015 3015 GRAPHITE GRAPHITE

F20 F21 F22 F23 F24 F25 F26 F27 F28 F29 F30 WATER 2393 WATER 2444 WATER GRAPHITE WATER 2376 WATER 3509 WATER GRAPHITE WATER 3823 WATER GRAPHITE 2397 4969 2380 2380 2380 2409 2409 2409 2393 2444 GRAPHITE 2376 3509 GRAPHITE 3823 GRAPHITE 2409 2393 2444 GRAPHITE 2376 3509 GRAPHITE 3823 (RAPi-ITF 2393 2444 GRAPHITE 2376 3509 3823 (RADWITF 2393 2444 GRAPHITE 23"6 3509 11359 3821 rPAPDWITF 4969 4969 4969 GRAPHITE 11358 2409 2409 2409 I

I A-4

ANALYSIS OF THE THERMAL HYDRAULIC AND REACTIVITY INSERTION BEHAVIOR OF THE DOW TRIGA RESEARCH REACTOR Submitted to the NRC in support of the DTRR License Renewal Prepared by:

Michael R Hartman 08 November 2011

TABLE OF CONTENTS LIST OF TABLES.................................................................................................................................

ii LIST OF FIGURES..............................................................................................................................

iii

1.

Introduction..............................................................................................................................

1

2.

Therm al Hydraulic Analysis................................................................................................

1

3.

Reactivity Insertions...............................................................................................................

12

4.

Sum m ary.................................................................................................................................

15

LIST OF TABLES Table 1. Physical data utilized in the RELAP5-3D model of the DTRR........................................

2 Table 2. Loss coefficients utilized in the RELAP5-3D model of the DTRR....................................

3 Table 3. Summary of conditions for the three thermal hydraulic cases considered in the analysis of the thermal hydraulic performance of the DTRR.................................

7 Table 4. Summary of thermal hydraulic analysis for Case #1...................................................

9 Table 5. Summary of thermal hydraulic analysis for Case #2...................................................

10 Table 6. Summary of thermal hydraulic analysis for Case #3...................................................

11 ii

LIST OF FIGURES Figure 1. Schematic illustration of the thermal hydraulic sub-channel utilized in the analysis of the thermal hydraulic performance of the DTRR......................................

2 Figure 2. Schematic illustration of the RELAP5-3D hot channel model utilized in the therm al hydraulic analysis of the DTRR.....................................................................

4 Figure 3. A schematic illustration of the profile of a DTRR fuel element (right) along with a cross sectional view, showing the internal structure of the fuel element (left). The axial discretization used to establish the nodes in the RELAP5-3D model are shown adjacent to the cross sectional view...........................

5 Figure 4. Schematic illustration of the radial discretization utilized within the RELAP5-3D model of the DTRR. The nodes are show below the radial cross section............. 6 Figure 5. Normalized axial power profile for the hot rod (element B-5) in the 2011 core configuration of the DTRR...................................................................................

7 Figure 6. Normalized radial power profile for the hot rod (element B-5) in the 2011 core configuration of the DTRR...................................................................................

8 Figure 7. Normalized axial power profile for the hot rod (element C-6) in the LCC core configuration of the DTRR..........................................................................................

8 Figure 8. Normalized axial power profile for the hot rod (element C-6) in the LCC core configuration of the DTRR..........................................................................................

9 Figure 9. Fuel centerline, outer cladding, and bulk coolant temperature for the hot-channel in Case #1....................................................................................................

10 Figure 10. Fuel centerline, outer cladding, and bulk coolant temperature for the hot-channel in Case #2....................................................................................................

11 Figure 11. Fuel centerline, outer cladding, and bulk coolant temperature for the hot-channel in Case #3....................................................................................................

12 iii

1. Introduction The following report summarizes an investigation into the thermal hydraulic behavior and the response to reactivity insertions of the DOW TRIGA Research Reactor (DTRR).

The DTRR is a TRIGA Mark-I reactor which is licensed to operate at powers up to 300 kW. The reactor consists of a series of six concentric rings in which are located fuel elements, graphite reflectors, and various experimental facilities. The core is surrounded by a cylindrical annulus of graphite which acts as a neutron reflector. The core and reflector are located in a below-grade aluminum tank filled with high-purity water. The water acts as a neutron moderator, a coolant, and as a radiation shield. Heat produced within the DTRR is rejected from the fuel to the water in the tank via natural convection cooling.

The purpose of this report is to provide a modern analysis of the thermal hydraulic characteristics of the DTRR in support of a 20-year license renewal through the U.S. Nuclear Regulatory Commission. Thermal hydraulic modeling of the DTRR was done using RELAP5-3Db.

The reactivity accidents were analyzed using a point-kinetics model of the DTRR in conjunction with data from the recent neutronic and thermal hydraulic analyses of the DTRR. The results of these analyses provide a basis to evaluate the thermal hydraulic and reactivity accident performance of the DTRR in support of the license renewal.

2. Thermal Hydraulic Analysis To analyze the thermal hydraulic performance of the DTRR a "hot-channel" methodology was employed whereby the most limiting thermal hydraulic channel in the DTRR is coupled to the fuel element with the highest power production. It should be noted that in reality the highest power element is not necessarily associated with the most limiting thermal hydraulic channel.

However, performing the analysis in this manner creates a bounding analysis for all possible combinations of hot rods and thermal hydraulic channels in the DTRR. The limiting thermal hydraulic channel in the DTRR exists in the B-ring where the fuel elements have a hexagonal arrangement with an element-to-element pitch, P, of 0.04054 m. The limiting thermal hydrauiic channel in the DTRR is shown schematically in Figure 1.

The outer diameter of the fuel elements, D, for the DTRR fuel is Using these, the flow area of the limiting thermal hydraulic channel was found, using Equation (1),

2 2

p2 A

2~

=-

4 DFuel, l

to be 3.884E-4 m2. The hydraulic diameter, Dh, is calculated using Equation (2),

4(AFlow)

(2)

Dh = (Wetted perimeter)

1. RELAP5-3D Code Development Team, "Volume 1: code structure, system models, and solution methods, in RELAP5-3D code manual" Idaho National Laboratory, Idaho Falls, ID (2005).

1

resulting in a hydraulic diameter of 1.327E-2 m for the limiting thermal hydraulic channel in the DTRR.

Using this data, a model of the hot-channel was constructed in RELAP5-3D.

The model conservatively assumes that there is no cross flow amongst adjacent thermal hydraulic channels.

In reality, there will be mixing between adjacent thermal hydraulic channels and as such the analysis reported herein has an inherent level of conservatism.

The geometry and loss coefficients utilized within the RELAP5-3D model are summarized in Table 1 and Table 2, respectively.

Fuel Element

/I Thermal Hydraulic Sub-Channel Fuel Element Pitch, P Figure 1. Schematic illustration of the thermal hydraulic sub-channel utilized in the analysis of the thermal hydraulic performance of the DTRR.

Table 1. Physical data utilized in the RELAP5-3D model of the DTRR.

Flow area [M2]

3.88E-04 Fuel Element Pitch [m]

0.04054 Wetted Perimeter [m]

0.117 Hydraulic diameter [m]

1.301E-02 Diameter of heated region [m]

Fuel element heated length [m]

Fuel element surface area [M2]

Fuel element surface roughness [m]

2.134E-06 2

Table 2. Loss coefficients utilized in the RELAP5-3D model of the DTRR.

MHe.seription Coeffen Inlet pressure loss coefficient 2.26 Exit pressure loss coefficient 0.63 The RELAP5-3D model used to analyze the thermal hydraulic performance of the DTRR is shown schematically in Figure 2. The model consists of a coolant source (volume 100), a cold leg (volume 101), a horizontal connector (volume 102), a hot channel (volume 103), and a coolant sink (volume 104). The coolant source permits the application of the time-dependent pressure and temperature boundary conditions for the thermal hydraulic analysis. The cold leg is used to establish the pressure differential across the thermal hydraulic sub channel and is utilized to establish the natural convection flow through the channel. The horizontal connector only serves to provide a physical connection between the cold leg and the hot channel. The hot channel contains the fuel element with the highest power and the limiting thermal hydraulic channel (corresponding in this case to the thermal hydraulic channels in the B-ring). The fuel element volume within the RELAP5-3D model is discretized both axially and radially, as shown in Figures 3 and 4. The discretization corresponds to that applied within the neutronics calculation and permits the power densities determined in the neutronics calculations to be directly applied to the heat source within the RELAP5-3D hot channel model. For the axial discretization, the upper and lower grid plates (nodes 01 and 24) have a length of 0.01905 m, the lower graphite region (node 02) has a length of 0.14643 m, the nodes in the fueled region (nodes 03 through 22) have a length of and the upper graphite region (node 23) has a length of 0.14567 m. For the radial discretization, node 01 is located on the centerline of the fuel, the zirconium pin (node

02) has a radius of 0.00318 m, each of the fuel nodes (nodes 03 to 22) are separated by a radial distance of the outer gap has a thickness of 1E-05 m, and the cladding thickness is 0.00087 m.

3

Coolant Source (100)

Coolant Sink Cold Leg (101)

Hot Channel (103)

Horizontal Connector (102)

Figure 2. Schematic illustration of the RELAP5-3D hot channel model utilized in the thermal hydraulic analysis of the DTRR.

4

Fuel Ha TriFlute Fuel Cl raphite el Slug uel Slug Pin el Slug raphite Lower E Fitting Figure 3. A schematic illustration of the profile of a DTRR fuel element (right) along with a cross sectional view, showing the internal structure of the fuel element (left). The axial discretization used to establish the nodes in the RELAP5-3D model are shown adjacent to the cross sectional view.

5

-1.~.

Lirconlum rin Gap Fuel Clad 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16 17 18 19 20 21 22t 24 23 Figure 4. Schematic illustration of the radial discretization utilized within the RELAP5-3D model of the DTRR. The nodes are show below the radial cross section The Bernath correlation 2 was used in the analysis of the thermal hydraulic performance of the DTRR. The choice of Bernath was predicated on the following factors: 1) the correlation has been extensively used for the analysis of research reactors, and 2) the DNBR predictions from the Bernath correlation have been shown to be conservative with respect to other potential choices for the correlation.

Three separate cases were considered for the thermal hydraulic analysis of the DTRR. Case #1 corresponds to the nominal operating characteristics of the DTRR for the 2011 core configuration with a reactor power level of 300 kW, a bulk pool temperature of 25 'C, and a pool level of 16 feet above the top of the core. Case #2 again utilizes the 2011 core configuration, but this time assumes the limiting thermal hydraulic conditions with a power level of 300 kW, a pool inlet temperature of 60 0C, and a pool level of 15 feet above the top of the core. Case #3 is based upon the limiting core configuration (LCC) which was analyzed in the neutronics analysis of the DTRR.

This core represents extreme power peaking in the DTRR due to possible reconfiguration of the DTRR core. The LCC core configuration was used in conjunction with a reactor power level of 300 kW, a pool inlet temperature of 60 'C, and a pool level of 15 feet

2. Bernath, L., "A Theory of Local Boiling Burnout and Its Application to Existing Data," Heat Transfer -

Chemical Engineering Progress Symposium Series No. 30, 56, pp95-116 (1960).

6

above the top of the core. The three cases are summarized in Table 3 below, where the hot rod power was as determined in the detailed neutronics analysis of the DTRR, the details of which are provided in a separate report.

Table 3. Summary of conditions for the three thermal hydraulic cases considered in the analysis of the thermal hydraulic performance of the DTRR.

Case #1 Case #2 Case #3 Hot Rod Power [kW]

5.91 5.91 6.08 Pool Inlet Temperature [°C]

25 60 60 Pool Level Above DTRR Core [ft]

16 15 15 For the 2011 core configuration, the normalized axial and radial power profiles for the hot rod are shown in Figures 5 and 6, respectively. While for the LCC, the normalized axial and radial power profiles are shown in Figures 7 and 8, respectively.

1.4 1.2 C

.12

" 0.8 0

E Zo 0.4 I.

0.2 0

-20

-15

-10

-5 0

5 Distance From Axial Centerline [cm]

Figure 5. Normalized axial power profile for the hot rod 2011 core configuration of the DTRR.

10 15 20 (element B-5) in the 7

1.4 1.2 021

_ 08 0

-o 0.6 Z 0.4 0.2 0

Figure 6.

1.4 1.2 C

.2 1

.2 3 0.8

°:

• 0 0.6 ZO0.4 0

0.2 0.4 0.6 0.8 1

1.2 1.4 1.6 1.8 2

Distance From Radial Centerline [cm]

Normalized radial power profile for the hot rod (element B-5) in 2011 core configuration of the DTRR the 0.2 0

-20

-15

-10

-5 0

5 Distance From Axial Centerline [cm]

10 15 20 Figure 7. Normalized axial power profile for the hot rod (element C-6) in the LCC core configuration of the DTRR.

8

1.4 1.2 C0 1 is 0.8 0 0.6 E

ZO 0.4 0.2 0

0 0.2 0.4 0.6 0.8 1

1.2 1.4 Distance From Radial Centerline [cm]

1.6 1.8 2

Figure 8. Normalized axial power profile for the hot rod (element C-6) in the LCC core configuration of the DTRR.

For Case #1, the axial profile for the fuel centerline temperature, the clad outer temperature, and the bulk coolant are shown in Figure 9. The results of the analysis are summarized in Table 4, where it can be seen that the DNBR predicted by the Bernath correlation is 9.99.

Table 4. Summary of thermal hydraulic analysis for Case #1.

Rod Power [kW]

5.91 Hot Channel Fuel Element Peaking Factor [PmajPavg]Element 1.576 Hot Channel Fuel Axial Peaking Factor [Pma./Pavg]Axial nodes 1.322 Hot Channel Fuel Radial Peaking Factor [Pmax/Pavg]Radial nodes 1.254 Maximum Fuel Temperature [°C]

240.30 Maximum Outer Cladding Temperature [°C]

121.06 Minimum Predicted DNBR (Bernath Correlation) 9.99 9

300 250

• Fuel Centerline n Outer Cladding A Bulk Coolant a

200 +

a Co 4) 4)

150 100 E

a 0

N M 8 a

U a

a a

50 +

A A

A A

A A

A a

A A

A A

A A

A A

A A

A A

01

-0.

ý... i I... 1

ý.

ý.

ý.

1 ý,

1 ý,

1.

2

-0.15

-0.1

-0.05 0

0.05 0.1 i

n n

0.15 0.2 Distance From Fuel Axial Centerline [m]

Figure 9. Fuel centerline, outer cladding, and bulk coolant temperature for the hot-channel in Case #1.

For Case #2, the axial profile for the fuel centerline temperature, the clad outer temperature, and the bulk coolant are shown in Figure 10. The results of the analysis are summarized in Table 5, where it can be seen that the minimum DNBR predicted by the Bernath correlation is 6.81.

Table 5. Summary of thermal hydraulic analysis for Case #2.

Rod Power [kW]

5.91 Hot Channel Fuel Element Peaking Factor [Pmax/PavglElement 1.576 Hot Channel Fuel Axial Peaking Factor [PMrax/Pavg]Axial nodes 1.322 Hot Channel Fuel Radial Peaking Factor [PmaP/PavglRadial nodes 1.254 Maximum Fuel Temperature [°C]

241.69 Maximum Outer Cladding Temperature [°C]

122.56 Minimum Predicted DNBR (Bernath Correlation) 6.81 10

300 250 200

• Fuel Centerline n Outer Cladding A Bulk Coolant 150 +

I-U a

a a

a a

a U

N U

a U

0 a

a 0

E 100-A A

A A

A A

A A

A A

A A

A A

A A

A A

A A

50 +

0 4-

-0.2

-0.15

-0.1

-0.05 0

0.05 0.1 0.15 0.2 Distance From Fuel Axial Centerline [m]

Figure 10. Fuel centerline, outer cladding, and bulk coolant temperature for the hot-channel in Case #2.

For Case #3, the axial profile for the fuel centerline temperature, the clad outer temperature, and the bulk coolant are shown in Figure 11. The results of the analysis are summarized in Table 6, where it can be seen that the minimum DNBR predicted by the Bernath correlation is 6.76.

Table 6. Summary of thermal hydraulic analysis for Case #3.

Rod Power [kW]

5.91 Hot Channel Fuel Element Peaking Factor [Pmax/Pavgl]Element 1.539 Hot Channel Fuel Axial Peaking Factor [PmaxlPavg Axial nodes 1.312 Hot Channel Fuel Radial Peaking Factor [Pmax/Pavg]Radial nodes 1.196 Maximum Fuel Temperature [°C]

246.70 Maximum Outer Cladding Temperature [°C]

122.84 Minimum Predicted DNBR (Bernath Correlation) 6.76 11

300 250 200

• Fuel Centerline

• Outer Cladding

• Bulk Coolant B Co 150 +

U 0

U a

a a

M U

U 0

U 0

100 +

U A

M U

a A

A A

A A

A A

A A

A A

A A

A A

A A

A 50 0

-0.2

-0.15

-0.1

-0.05 0

0.05 0.1 0.15 0.2 Distance From Fuel Axial Centerline [m]

Figure 11. Fuel centerline, outer cladding, and bulk coolant temperature for the hot-channel in Case #3.

Note that in all cases, the minimum DNBR predicted by the Bernath correlation is well above the limiting minimum DNBR of 2.0 which is typically applied to research reactors.

3. Reactivity Insertions To verify the validity of the DTRR Safety Limit which states that "The temperature in any fuel element in the DOW TRIGA Research Reactor shall not exceed 500 'C under any condition of operation.", calculations were performed to assess the response of the DTRR to postulated reactivity insertion events. Two such events were considered.

The first of these was a step insertion of reactivity corresponding to the maximum worth of a moveable experiment in the DTRR ($0.75, DTRR Technical Specification 3.7.3).

The second reactivity insertion considered was the uncontrolled withdrawal of a control rod at the maximum allowed reactivity insertion rate ($0.20/s, DTRR SAR Table 4). In each case a point-kinetics model was developed and solved numerically to determine the peak fuel temperature observed in response to the event. The point-kinetics model is as follows:

6 dt

[

A IJk.+Ltk+*tJ g=1 dC((t)

[

t)

- X1C(t) dt A

12

p(t) = PReactvity event(t) + PFeedback(t) k (t) 1 k(t) 1 - p(t)

PFeedback (t) aFuel (TFuel (0 - TFuel (0))

and dTFuel(t)

[P(t) - P(0)]

dt mFuelCpFuel (T) where p(t) is the reactivity at time t, f# is the effective delayed neutron fraction, 1 is the prompt-neutron lifetime, A is the mean generation time, Xi is the decay constant for the ith delayed neutron precursor group, Ci(t) is the density of delayed neutron precursors in the ith group, S(t) is an external source, k(t) is the time-dependent multiplication factor, TFueI(t) is the average core fuel temperature, mFuel is the mass of the fuel, and Cp(T) is the specific heat capacity of the fuel.

For the calculations performed herein, the following additional assumptions were made:

1. External sources, S(t), can be ignored

2. The mean generation time, A, was taken to be the same as the prompt-neutron lifetime, 1.

3. The fuel elements were treated adiabatically (i.e. no heat rejection to the coolant) resulting in conservative estimates of the fuel temperature in response to a reactivity event.

4. The product mFuelCp_Fuel was taken to be 825 + 1.6 1*(TFuer25) W.sec/(0 C.fuel element), where TFuel is in units of oC.3 The step insertion of the maximum worth of a moveable experiment ($0.75) was analyzed first.

The initial fuel temperature, TFueI(O), was taken to be 2000C, a value of 0.0070 for fl was used, the prompt-neutron lifetime was taken to be 60 psec, and a value of -$0.0181/°C was used for the fuel temperature reactivity coefficient, OtFuel. For an initial power of 300 kW, the maximum observed average fuel temperature was 255 'C, corresponding to a temperature increase of 55 °C. Based on the results presented in Section 2 of this report, the maximum peaking factor

3. M. T. Simnad, F. C. Foushee, and G. B. West, "Fuel Elements for Pulsed TRIGA Research Reactors," Nuclear Technology 28 (1976) 31-56.

13

for the highest power fuel element was 2.613 for the 2011 core configuration (as determined by taking the product of the three peaking factors stated in Tables 4 or 5).

Using this, the calculated temperature rise at the hottest point in the highest power fuel element would be 2.613

• 55 'C = 143.7 'C. Adding this to the maximum fuel temperature of 246.7 'C observed for any of the three thermal hydraulic conditions considered in Section 2, the maximum predicted fuel temperature for the step removal of an experiment with the maximum moveable reactivity worth of $0.75 is found to be 390.4 'C.

It should be noted that use of the highest fuel temperature (thermal hydraulic Case #3) in conjunction with the peaking factor for thermal hydraulic Cases #1 and #2 is conservative since the peaking factor for Case #3 is lower than for the other cases (2.415 versus 2.613).

It should also be noted that there is >100 °C margin between the predicted maximum fuel temperature in response to this reactivity insertion and the DTRR fuel temperature Safety Limit of 500 °C. Initiation of this event from lower power levels results in correspondingly lower peak fuel temperatures and is not considered further in this report.

The uncontrolled withdrawal of a control rod at the DTRR maximum reactivity insertion rate of

$0.20/s was considered next. (NOTE: In the recent MCNP5 neutronics calculations of the DTRR, the most reactive rod was found to be SHIM 1 with an integral rod worth of $3.85. During the most recent rod calibrations at the DTRR, the rod withdrawal times were measured, and the fastest rod withdrawal time was measured to be 41.53 seconds, corresponding to a rod speed of 21.67 inches/minute.

Using a conservatively faster rod speed of 22 inches/minute and the MCNP5 calculated differential rod worth curve for SHIM 1, the maximum reactivity insertion rate for the 2011 DTTR core configuration was found to be $0.15/s which is well below the Technical Specification maximum reactivity insertion rate of $0.20/s which is used in the present analysis.) It was further assumed that upon insertion of the remaining two control rods that the core would be subcritical by the minimum shutdown margin of $0.50. The limiting transient was determined to occur for low initial reactor power levels due to the increased time to reach the SCRAM setpoint. The model parameters where the same as those described above, with the following exceptions: 1) the reactivity insertion was modeled as a linear ramp at a reactivity addition rate of $0.20/s, 2) a reactor scram was initiated 6.825 s after the initiation of the uncontrolled rod withdrawal (this corresponds to the time at which power reaches 300 kW plus an additional second to account for control rod insertion time), and 3) the rod experiencing the uncontrolled withdrawal was assumed to not SCRAM and continue its withdrawal.

Strictly speaking, the latter of these exceptions assumes the simultaneous failure of the rod control circuit for the affected rod as well as the magnet which couples the rod to the control rod drive.

Hence, continuing the rod withdrawal following the SCRAM will produce conservative results for this analysis due to the continued reactivity insertion of the rod experiencing the uncontrolled withdrawal.

An additional measure of conservatism in this analysis is the use of a constant reactivity insertion rate of $0.20/s. In reality, the differential rod worth curve for the affected rod will have its largest value near the center of the core and will decrease as the rod approaches the end of its travel, as such the reactivity insertion rate will decrease as the transient proceeds. For the transient described above, the peak power was 10.3 MW, producing 14

a core average fuel temperature increase of 72.15 °C.

Using a peaking factor of 2.613, this results in a maximum temperature in the highest power rod of 213.5 'C which is well below the DTRR fuel temperature Safety Limit of 500 'C. Even with the several layers of conservatism built into the present analysis, there is ample margin to the Safety Limit.

4. Summary A detailed analysis of the thermal hydraulic analysis of the DTRR has been performed using RELAP5-3D in conjunction with information from the recent neutronics analysis of the DTRR using MCNP5. The analysis demonstrates that natural circulation cooling provides adequate cooling of the DTRR core under all operating conditions. The minimum DNBR (based upon the Bernath correlation) for the current DTRR core was 6.81 under the most limiting thermal hydraulic conditions which are permitted by the DTRR Technical Specifications. The minimum DNBR for a hypothetical arrangement of the DTRR core which results in exaggerated power peaking (denoted as the LCC in the neutronics report) was 6.74 (based upon the Bernath correlation).

Reactivity transients corresponding to 1) the step insertion of the maximum reactivity worth of a moveable experiment ($0.75) and 2) the uncontrolled withdrawal of a control rod at the maximum reactivity insertion rate ($0.20/s) were both simulated.

The step insertion of reactivity was determined to be the bounding transient, which when initiated from the maximum licensed power level for the DTRR of 300 kW, resulted in a peak fuel temperature of 390.4 °C in the highest power rod.

The uncontrolled rod withdrawal resulted in a peak temperature in the highest power rod of 213.5 °C. Both of these results are well below the DTRR fuel temperature Safety Limit of 500 °C, and as such they confirm the adequacy of the current DTRR licensing basis.

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