ML102650326
| ML102650326 | |
| Person / Time | |
|---|---|
| Site: | 05000128 |
| Issue date: | 08/19/2010 |
| From: | Reece W Texas A&M Univ |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| DiMeglio A, NRR/DPR/PRTA, 301-415-0894 | |
| References | |
| 2010-0049 | |
| Download: ML102650326 (60) | |
Text
TEXAS ENGINEERING EXPERIMENT STATION TEXAS A&M UNIVERSITY 3575 TAMU COLLEGE STATION, TEXAS 77843-3575 NUCLEAR SCIENCE CENTER 979/845-7551 FAX 979/862-2667 August 19, 2010 2010-0049 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 I.
Subject:
Texas A&M University System, Texas Engineering Experiment Station, Nuclear Science Center Reactor (NSCR, License No. R-83, Docket 50-128) - Startup Report after conversion to LEU To Whom It May Concern:
Please find included the Texas A&M University Nuclear Science Center Reactor Startup Report describing the results of converting from HEU to LEU fuel.
If you have any questions, please contact me at 979-845-7551.
I declare under penalty of perjury that the foregoing is true and correct. Executed on August 19, 2010.
W. D.Reece NSC, Director Xc:
21 I/Central File Christian Cowdrey, Relicensing Manager Duane Hardesty, Project Manager Francis DiMeglio, Relicensing Project Manager NRC Document Control Desk RESEARCH AND DEVELOPMENT FOR MANKIND http ://nsc.tamu.edu A -')i 74'ý.2 0
REDACTED VERSION Texas A&M University, Nuclear Science Center Reactor Startup Report TRIGA 1 MW LEU Conversion Reactor Prepared by:
Edited by:
John Hernandez, Reactor Supervisor, NSC Dr. W.D. Reece, Director, NSC April 30, 2007 REDACTED VERSION
Texas A&M University Nuclear Science Center Reactor Startup Report TRIGA-: 1 MW LEU Conversion Reactor Prepared by:
John Hernandez, Reactor Supervisor, NSC Edited by:
Dr. W.D. Reece, Director, NSC Startup data and results were collected by the following individuals:
Tom Fisher, Electronic Shop Foreman, NSC Pamela Gondeck, Reactor Operator, NSC Alfred Hanna, Electronic Shop Assistant, NSC Aaron Heinrich, Senior Reactor Operator, NSC John Hernandez, Reactor Supervisor, NSC Ken Mushinski, TRIGA Reactor Projects Engineer Manager, GA Jerry Newhouse, Senior Reactor Operator, NSC Ilya Pavlenko, Duty Health Physicist, NSC Jared Porter, Reactor Operator, NSC Dr. W.D. Reece, Director, NSC Jim Remlinger, Assistant Director, NSC Travis Trahan, Reactor Operator, NSC Latha Vasudevan, Radiological Safety Officer, NSC April 30, 2007
Abstract Reactor Startup Report: TRIGA 1 MW LEU Conversion Reactor Prepared by John Hernandez, Reactor Supervisor, NSC On September 1, 2006, the Texas A&M University Nuclear Science Center (NSC) received an order from the United States Nuclear Regulatory Commission (NRC) to convert the NSC TRIGA research reactor from high-enriched uranium to low-enriched uranium. Also ordered was the submission of a startup report which provides the results of the startup testing procedures. This report serves as a response to that order.
Having completed the core conversion startup testing, data was accumulated and compiled describing the results of testing performed. The data requested by the NRC is included in the report along with other information which the NSC deemed use'ful or appropriate for NRC review. At the conclusion of the startup testing, the conversion core was declared Steady-State and Pulse Operational per Technical Specification definition.
Reactor Startup Report 4/30/07 Texas A&M University Nuclear Science Center Table of Contents Introduction.........................................................................................................................
I M ethods and R esults...........................................................................................................
3 1.0 Critical Mass Determination and Initial LEU Core Loading..........................
3 1.1 Calculations for the Approach to Criticality...............................................
3 1.2 Performance of Criticality Approach and Critical Mass Determination........ 7 2.0 Control Rod Calibration.................................................................................
10 2.1 Calculated Control Rod Worth 10 2.2 Performance of Control Rod Calibration.................................................
10 3.0 Excess Reactivity and Shutdown Margin Determination.............................
12 3.1 Calculated Excess Reactivity and Shutdown Margin...............................
12 3.2 Performance of Excess Reactivity and Shutdown Margin Determination... 12 4.0 Reactor Power Calibration..............................................................................
14 4.1 Expected Power Indication vs. Actual Reactor Power..............................
14 4.2 Performance of Reactor Power Calibration...............................................
14 5.0 Pulse M easurem ents......................................................................................
17 5.1 Pulse C alculations.....................................................................................
17 5.2 Pulse M easurem ents.................................................................................
18 6.0 Thermal Neutron Flux Distribution Measurements..................................... 21 6.1 Calculated BOL Thermal Neutron Flux...................................................
21 6.2 Measured Thermal Neutron Flux............................................................
21 7.0 Reactor Physics Measurements....................................................................
23 7.1 C alculated Param eters............................................
....................................... 23 7.2 Measured Reactor Physics Parameters.....................................................
27 8.0 Primary Coolant Measurements....................................................................
29 9.0 Description of Instrumented Fuel Element Anomaly...................................
31 10.0 D iscussion of Results...................................................................................
32 R eferences.........................................................................................................................
33 A ttachm ent to O rder....................................................................................................
A -i Outline of Reactor Startup Report..............................................................................
A-1 iii
Reactor Startup Report 4/30/07 Texas A&M University Nuclear Science Center List of Tables Table 1: Nuclide Densities of Fuel Used in the PRNC and TAMU MCNP Models.......... 4 Table 2: Material Composition Used in MCNP Models................................................
5 Table 3: Inverse M ultiplication Data..............................................................................
8 Table 4: C ontrol R od W orth..............................................................................................
11 Table -5: Excess Reactivity and Shutdown Margin Away from Thermal Column.....
13 Table 6: Excess Reactivity and Shutdown Margin against Thermal Column................ 13 Table 7: Calorimetric Calibration Thermocouple Measurements at 400 kW.........
5 15 Table 8: Reactor Pulse Calculated Parameters (23°C ambient)..................
18 Table 9: BOL Pulse M easurem ents..............................................................................
19 Table 10: Max. and Min. Flux in 1.8 cm Square Centered at Sample Positions.......... 21 Table 11: Fission Spectra Used for Calculation of 3ef...
24 Table 12: Reactivity Change with Temperature..........................................................
26 Table 13: Prompt Neutron Lifetime from Pulses at Various Reactivities.................... 28 Table 14: Reactivity Loss at Reactor Power...............................................................
29 Table 15: Pool Water Isotope Activities for First 30 Days of Operation.....................
30 Table 16: Initial Core Loading by Elem ent....................................................................
A-4 Table 17: Core Uranium Loading at Criticality...........................................................
A-5 Table 18: Transient Rod Period and Reactivity Per Iteration.........................................
A-7 Table 19: Shim'Safety Control Rod #1 Period and Reactivity Per Iteration.................. A-8 Table 20: Shim Safety Control Rod #2 Period and Reactivity Per Iteration.................. A-8 Table 21: Shim Safety Control Rod #3 Period and Reactivity Per Iteration.................. A-9 Table 22: A4'Flux Comparisons (Shim Safety Control Rods Banked)................... A-19 Table 23: A4 Flux Comparisons (Shim Safeties Control Rods Skewed)..................... A-19 Table 24: A6 Flux Comparisons (Shim Safety Control Rods Banked).................... At20 Table 25: A6 Flux Comparisons (Shim Safety Control Rods Skewed)........................ A-20 iv
Reactor Startup Report 4/30/07 Texas A&M University Nuclear Science Center List of Figures Figure 1: TAMU Axial Model for MCNP Calculations.................................................
6 Figure 2: TAMU Radial Model for MCNP Calculations -. 62 Elements...................... 7 Figure 3: Inverse Multiplication Plot Channel I...........................................................
9 Figure 4: Inverse M ultiplication Plot Channel 2...............................................................
9 Figure 5: TAMU Radial Model MCNP-90 Elements.................................................
10 Figure 6: Instrum ent Elem ent Position..........................................................................
18 Figure 7: Calculated Reactivity Loss vs. Reactor Power.............................................
27 Figure 8: Reactivity vs. Reactor Power.................................
29 Figure 9: Core Loading at Criticality..............................................................................
A -3 Figure 10: Transient Rod Integrated W orth..................................................................
A-11 Figure 11: Regulating Rod Integrated Worth..........................................................
A-12 Figure 12: Shim Safety #1 Rod Integrated Worth...................................................
A-13 Figure 13: Shim Safety #2 Rod Integrated Worth...................................................
A-14 Figure 14: Shim Safety #3 Rod Integrated Worth...................................................
A-l15 Figure 1"5: Shim Safety #4 Rod Integrated Worth...................................................
A-16 Figure 16: Pulse Energy vs. Akp.....................................
A-17 Figure 17: (
T,,,,ob) vs. Ak A-17 Figure 18: Peak Pow er vs. V - Tamb)2.......................................................................... A-18 Figure 19: Measured Temperature Rise vs. (T - Ta,,,b )..............................................
A-18 v
Reactor Startup Report 4/30/07 Texas A&M University Nuclear Science Center Introduction In accordance with Nuclear Regulatory Commission (NRC) Order EA-06-21 1, the Texas A&M University Nuclear Science Center (NSC) completed a reactor core conversion of the NSC TRIGA research reactor from high-enriched uranium to low-enriched uranium.
In addition to completing the core conversion, the NSC was ordered to submit a Reactor Startup Report of the core conversion containing the data listed in the "Outline of Reactor Startup Report" attachment to the order (See Attachment #1 of Appendix A).
This report and included data has been prepared in response to the NRC order. The data includes results from startup testing performed prior to declaring the reactor Steady-State and Pulse Operational from September to December of 2006. Along with the requirements of the order, the report also contains data collected that the NSC deems appropriate and useful for NRC review.
In accordance with the NRC order, the NSC began the reactor conversion on September 27, 2006. The procedures for conversion directed the NSC through performing pre-operational checks, inspecting fuel and attaining steady-state and pulse operability.
To verify the operability of all facility-safety equipment and reactor controls prior to core conversion, all routine maintenance and surveillances affecting facility and core safety and instrumentation were performed prior to core conversion. Fuel inspections were performed on all new fuel elements, instrumented fuel elements (IFE) and fuel-followed control rods. Inspections included checks for transverse bend, initial elongation measurements and visual inspection of fuel and cladding integrity.
The approach to criticality was performed and criticality was achieved within the range of the predicted number of fuel elements. After attaining criticality, an initial determination of excess reactivity was performed and verified to be was within design criteria. Initial control rod worths were then measured and determined to be within expected values and consistent with the location of the rods within the partial core configuration.
Additional fuel was added to allow for a second control rod calibration and shutdown margin determination prior to proceeding to the complete core loading. The core was completely loaded placing the core in the desired final configuration with all experimental facilities installed. A final control rod calibration and shutdown margin determination was then performed. The value of 'shutdown margin was much more conservative than the required $0.25.
The reactor was subsequently brought to power to perform a calorimetric calibration of the linear power channel at 400 kW; the results showed that no linear channel gain adjustment was required.
Following the calorimetric calibration, the commission plan directed determination of reactivity loss versus reactor power. Power was to be raised incrementally to full power I
Reactor Startup Report 4/30/07 Texas A&M University Nuclear Science Center to perform this determination. During this procedure the indicated fuel temperature was higher than expected for the indicated power level.
An investigation into the anomalous temperature indication followed. The temperature anomaly arose from a larger than design-basis gap. The-gaps in the IFE and the entire core were monitored for several months to assure that there were no safety issues as the fuel remodeled to close the gaps and to ensure that the new fuel would perform acceptably throughout the life of the core.
The NSC TRIGA LEU Conversion core was approved for Steady-State Operation as defined by the TAMUNSC Technical Specifications. The temperature anomaly postponed pulse testing until November 30, 2006.
Pulse testing commenced in late November. A series of pulses was conducted to determine various parameters of the new core. The data collected from the various tests suggest that the maximum prompt insertion should be at or below $1.91 to keep the maximum temperature in the core below 850'C. Measured and predicted data were found to be in excellent agreement. The reactor was approved for Pulse Operation as defined by the TAMU NSC Technical Specifications.
2
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Methods and Results 1.0 Critical Mass Determination and Initial LEU Core Loading 1.1 Calculations for the Approach to Criticality For the approach to initial criticality, calculations were performed using both diffusion theory and Monte Carlo codes. In general, multi-group diffusion theory was used for design calculations since it gives adequate results.for systems of this kind and its multi-group fluxes and cross sections are easily utilized in nuclide bum-up calculations. Monte Carlo calculations were used to evaluate the facilities around the core and also to compute the worth of core components and different core configurations.
The diffusion theory code used is DIF3D, a multi-group code which solves the neutron diffusion equations with arbitrary group scattering. The Monte Carlo code used is MCNP5 that contains its own cross 'section library.
Evaluation of the Puerto Rico Nuclear Center (PRNC) TRIGA HEU (FLIP) core was conducted to provide benchmark data for the computational technique to be used for evaluating the TRIGA LEU (30/-20) fuel in the Texas A&M University (TAMU) TRIGA core. It also provided the information required for the performance comparison of the fresh HEU (FLIP) and fresh LEU (30/.20) fuel for this report.
The computations produced operational parameters to be compared with the actual measured values from the commissioning tests for the PRNC TRIGA core loaded with FLIP (HEU) fuel including the 1/M approach to critical tests conducted by GA.
Reactor calculations were performed in three dimensions for the initial criticality and the full core loading of the PRNC core and the TAMU core using the MCNP 5, Version 1.3, continuous energy Monte Carlo code. The nuclide cross sections were based on ENDF/B VI data included in the MCNP 5 data libraries.
The PRNC and TAMU fuel meat nuclide densities used in the two models are shown in Table 1. The other materials beside the fuel used in the PRNC and TAMU MCNP models are listed in Table 2.
3
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Table 1: Nuclide Densities of Fuel Used in the PRNC and TAMU MCNP Models H
1.0079 C
12.011 Zr 91.224 Er-166 165.93 Er-167 166.932 5.4'59307E-02 1.49606E-03 3.529874E-02 1.0624E-04 7.295E-05 32.05 10.47 1875.81 10.27 7.09 4.91'5763E-02 1.78701 E-03
.3.227,955E-02 7.717E-05
-5.299E-05 28.86 12:50 1715.37 7.46 5.15 Hf 178.49 2.11792E-06 0.22 1.93677E-06 0.20 F Total,00258'=9F 4
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Table 2: Material Composition Used in MCNP Models SS 304(clad)
Cr-50 0.000778 7.98 Cr-52 0.015003 Cr-53 0.001701 Fe-56 0.056730 Ni-58 0.007939 Mn-55 0.001697 Graphite(TC)
C 1.7 Graphite (reflector in fuel)
C 1.75 Zirconium (rod)
Zr 6.51 6061 Al (grid plate and AI-27 0.058693 control rod clad)
Fe-56 0.000502 90% B4C (control rod)
B-10 0.020950 B-11 0.084310 C
0.026320 Boral ( 3'5wt% B4C, 65 B-10 0.06031 2.64 wt% Al) (detector channel)
B-i 1 0.24489 C
0.08725 AI-27 0.63581 AI+Water Mix 1 (2" lower H
0.028748 cluster adapter)
O 0.014374 AI-27 0.0334"55 Fe-56 0.000286 AI+Water Mix 2 (5" -grid H
0.030788 plate) 0 0.01"5394 AI-27 0.031663 Fe-56 0.000271 Water
'1.0 Air 0.000123 5
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Geometrical Models Each fuel rod was explicitly modeled such that 15 cells and 6 surface cards were constructed to properly represent one fuel rod. A total of 930 cells and 372 surface cards were made for the PRNC and a similar number of cells and surface cards for the TAMU O
in the approach-to-critical core.
A detailed MCNP model of the TAMU reactor was made including (30/20) fuel rods,
-rods, 1 void followed transient rod, I water-followed regulating rod, 11,graphite blocks around the core and 4 detector assemblies. This number of fuel rods was chosen to be close to rods in the critical configuration at PRNC.
For the initial critical,case the TAMU core was modeled to be close to the coupler box; a 0.5 inch water gap existing between the core and the coupler box. This configuration has been chosen since this is the most reactive arrangement.
Figures 1 and,2 are the XZ and XY plots of the MCNP model of the TAMU cold critical case.
Based on the results of the MCNP modeling and calculation, as well as practical experience derived from the criticality approach with several other TRIGA reactors, criticality should be expected with a loading of LEU elements including fuel-followed control rods.
Transient Rod Fuel Region Adapter Grid Plate Figure 1: TAMU Axial Model for MCNP Calculations 6
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Figure 2: TAMU Radial Model for MCNP Calculations -I Elements 1.2 Performance of Criticality Approach and Critical Mass Determination Method The loading of fuel elements to obtain criticality was accomplished using the standard inverse multiplication curve (l/M) approach. This approach is based on the fact that subcritical multiplication is given as M= 1/ (1-k) from which one obtains 1/M= 1-k where k ranges from 0 (no fuel) to 1 (at criticality). The experimental values for 1 /M subcritical multiplication are given by the count rate with no fuel Co divided 7
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center by C, for loading step n. However, for the present 1/M application for approach to critical, the value Co can start at any convenient loading point.
Results After completing pre-operational checks and fuel inspections, core loading was begun, leading to initial criticality. Using two fission chambers, the approach to criticality was monitored using the Inverse Multiplication (1/M) Method. The core loading sequence for the approach to criticality is listed on Table 17 of Appendix A.
In the early stages of the core loading procedure, the installed neutron source was positioned in various core positions depending on the existent core configuration.
The final source position used in the approach was external to the core east of the F-:5 core position (See Figure 9 in Appendix A).
For practical purposes, the core configuration deviated from the modeled core configuration during the approach to criticality, but every effort was made to maintain a rectangular configuration similar to the model., The shim-safety control rods and the transient rod were inserted in the core at the earliest practicable time; however the regulating rod remained out of the core until criticality was achieved.
The value for Co was determined witl in the core. As further count rates were determined, a half-step approach to criticality was performed where practicable, however, the bundle-style TAMU core did not allow for a single-element approach to criticality. Criticality was achieved wit el elements installed in the core. This was within the predicted range of&
elements. The core configuration at time of initial criticality is shown in Figure 9. Core loading at time of initial criticality was
ý of Uranium-235. The loading per element can be found on Table 17 in Appendix A. The core excess reactivity was determined to be $0.147 based on the asymptotic period of the reactor with all rods fully withdrawn.
Table 3: Inverse Multiplication Data 8
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Figure 3: Inverse Multiplication Plot Channel 1 Figure 4: Inverse Multiplication Plot Channel 2 9
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center 2.0 Control Rod Calibration 2.1 Calculated Control Rod Worth The Monte Carlo code used to calculate control rod worth is described in Section 1.1. The full core loading in the TAMU reactor contains 90 fuel rods including 4 fuel followed control rods. With the reactor core positioned against the thermal couple, the MCNP calculation for the combined control rod worth is $16.34.
Figure 5: TAMU Radial Model MCNP /
Elements 2.2 Performance of Control Rod Calibration Method After criticality was achieved, the reactor was fully loaded to the desired final configuration. Rod worth determination was performed on control rods using the Positive Period-Differential Worth Method.
For this method, a computer program was written to determine the reactor period, amount of reactivity inserted, and the reactivity per percent of control rod 10
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center withdrawal. The program CRCAL uses an input signal from the linear drawer that is proportional to reactor power, measures the rate of change of this signal, computes the log of it and plots reactor power versus time on the PC monitor. A linear fit of this curve is taken to generate a reactor period, which is used to calculate the amount of reactivity inserted and the reactivity per percent of control rod withdrawal. This program can also determine and print out the control rod position and corresponding differential worth.
To establish a positive reactor period, the reactor is brought critical with the rod being calibrated at the desired initial desired rod height and all other rods positioned as necessary to achieve criticality. The rod being calibrated is then withdrawn to establish a steady positive period. The CRCAL program collects power data until reactor power approaches the point of adding heat. This process is repeated until the entire length of the rod is calibrated.
The accuracy of the procedure is dependent on the number of iterations performed per rod. Each iteration was performed in a manner to establish a reactor period of approximately 1'5 - 20 seconds where possible. 'Significantly larger reactor periods sometimes occurred due to the low worth of the associated rod.
Results Table 4 contains the results of the control rod calibrations including the number of iteration performed. The reactivity and reactor period for each iteration are found on Tables 19 - 24 of Appendix A. Figure 10 - 15 in Appendix A show the associated integrated rod worth curves. The determined value of $17.197 was
$0.857 greater (5.2%) than calculated values.
Table 4: Control Rod Worth
=,NunRod odNumb t
2.035 Shim Safety #3 14 2.901 Shim Safety #4 19 4.609 Regulating Rod 7
1.021 Transient Rod 15 3.457
~.
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- ~ 7* 1 ~ r~Sz~.
1~
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INO-w-ft ý"
11
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center 3.0 Excess Reactivity and Shutdown Margin Determination 3.1 Calculated Excess Reactivity and Shutdown Margin The Monte Carlo code used to calculate excess reactivity and shutdown margin is described in Section 1.1. Excess reactivity was calculated with the reactor against the thermal column and 250 mm away from the thermal column.
The MCNP calculation gives an unrodded kef value with one sigma uncertainty of:
ken`= 1.0572-2 +/- 0.00018 This corresponds to a core reactivity of $7.73.
With the core assembly moved a distance of about 250 mm from the thermal column, the knff value with one sigma uncertainty is:
ken`= 1.04553 +/- 0.00017 This corresponds to a core reactivity of $6.22, a loss of $1 51 with a water reflector substituted for the thermal couple.
The reactor shutdown margin was calculated with the reactor core positioned against the thermal column. The MCNP calculation with all control rods inserted gives a k-ff value with one sigma uncertainty:
kef= 0.94314 + 0.00017 This is equivalent to reactivity shutdown of -$8.61. No shutdown margin was calculated for the reactor away from the thermal column.
3.2 Performance of Excess Reactivity and Shutdown Margin Determination Method The results of Section 2.2 were used in determining excess reactivity and shutdown margin. The results were measured with the reactor in a cold xenon-fee condition (reactor shutdown > 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). This procedure was performed twice:
once with the reactor against the thermal column and once with the reactor > 250 mm away from the thermal column.
To determine excess reactivity, the reactor was made critical at 300 W following an extended shutdown. 300 W. This is high enough to ensure that source neutrons are not making a significant contribution to neutron population, but low enough to ensure that the reactor is below the point of adding heat preventing negative reactivity insertion.
12
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center
(
h With the reactor critical, rod heights were recorded. Using the rod worth data from Section 2.2, the rod worth at the recorded rod height was subtracted from total rod worth to determine excess reactivity. With the reactor against the thermal column, excess reactivity was determined to be $8.641. With the reactor away from the thermal column, excess reactivity was determined to be $7.483. These values were greater than the calculated values by $0.911 and $1.263, respectively.
This demonstrated a thermal column reactivity of $1.158.
Shutdown margin was determined by subtracting total rod worth from excess reactivity. The values for shutdown margin were $8.566 against the thermal column and $9.714 away from the thermal column. These values do not correspond to the shutdown margin determination of the Technical Specifications which consider stuck rod criteria and maximum sample -reactivity. The results of excess reactivity and shutdown margin determination are displayed in Tables 5 and 6. Due to the $0.857 difference in control rod worth from calculated worth, the values for excess reactivity, and shutdown margin were considered acceptable.
Table'5: Excess Reactivity and Shutdown Margin Away from Thermal Column
- ..ont-pLI Roqd f-:>'
joro§-
(~PIto4' D!
or XS' Shim Safety #1 49.0 1.700 Shim Safety #2 49.0 1.159 Shim Safety #3 49.0 1.631 Shim Safety #4 49.0
.2.453 Regulating Rod 45.6 0.540 Transient Rod 100.0 0.000 Table 6: Excess Reactivity and Shutdown Margin against Thermal Column Shim Safety #1 43.5 1.986 Shim Safety #2 43.5 1.330 Shim Safety #3 43.5 1.881 Shim Safety #4 43.5 2.856 Regulating Rod 42.7 0.588 Transient Rod 100.0 0.000 otal soe EXcess 1ý
ý~toMkar~giný
ý 13
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center 4.0 Reactor Power Calibration 4.1 Expected Power Indication vs. Actual Reactor Power The power instrumentation used for the conversion core uses detectors and instruments which were utilized in the previously operating HEU core. Due to the graphite reflector feeding the linear detector, no change in detector position was expected following the core conversion. Following the conversion, power indication was expected to be consistent with actual reactor power based on previous instrument calibration. As a result, the startup plan only called for performing the annual routine power calibrations. Adjustments would be made to those instruments which deviated f'rom indicated power as directed by the annual maintenance requirements.
4.2 Performance of Reactor Power Calibration Method Calibration of the linear power instrument was performed using a slope-method calorimetric calibration. The-slope method, involves measuring the rate of '
temperature rise for the reactor pool water [dT/dt (°C/h)] while the reactor is operating at power P and the pool water is stirred. For the NSC reactor pool with a water volume of about 106,000 gallons (4.0125 x 105 liters), the pool constant is calculated to be 2.1429 °C/(MWh) as follows:
1,000,000 watts/ 4.1868 Joules/calorie 1
4.0125 x 108 cm 3 x 1/3600sec/h 2.1429 °C/(MWh)
The calorimetric calibration was performed at 400 kW using 5 thermocouples at varying depths in the pool. For the copper-constantan thermocouples used during the calorimetric, a.pool constant of 2.1429 'C/ (MWh) corresponds to 36.216 ýtV (average)/400 kWhr. Thermocouple voltage versus time was recorded for a period of two hours. Actual power was calculated using the following equation.
_(TCpeak -TCo).x(OOOOlV/mV)x400kWxlHrxL,,,,
'Patual
-36.216pVx (2Hr) x 106,000 where:
TCpeak = Thermocouple average voltage at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> TC0 = Initial thermocouple average voltage Lactual = Actual pool level in gallons 14
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center If indicated reactor power deviates from actual power by > 5%, the Linear instrument full powergain potentiometer must be adjusted until indicated power matches calculated actual power.
Following the completion of the calorimetric calibration of the linear channel, the remaining power indications were adjusted to the same indication as the linear drawer. This was performed in accordance with the annual maintenance procedures for the two Safety Power Channels, the Pulse Channel, and the Log Power Channel.
Results As expected, power indication was consistent with actual power on all five power indicating channels. The calorimetric calibration of the linear drawer was
'completed in accordance with the Linear Drawer Annual Maintenance Procedures. At 400 kW, indicated power was found to deviate from actual power by 1.31% and no adjustment of the Full Power Gain Adjust Potentiometer was required. Table 7 contains the thermocouple data accumulated during the two hourcalorimetric calibration procedure.
Table 7: Calorimetric Calibration Thermocouple Measurements at 400 kW 14:20 4.462 0.8924 72.6 14:35 4.497 0.8994 72.8 14:50 4:546 0.9092 73.3 15:05 4.596 0.9129 73.7 15:20 4.643 0.9286 74.2 15:35 4.688 0.9376 74.6 15:50 4.735 0.9470 75.0 16:05 4.77 0.9554 75.3 16:20 4.827 0.9654 75.8 15
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Using the data from Table 7, actual reactor power was calculated as well as the percent error from indicated power. The results of these calculations are as follow.
Actual Power Calculation:
P.C, I
=u (0.96 0.8924)m V x (I 000,uV/m V).x 400kW x 1Hr x 103800gal 36.216pV x (2Hrs) x 106,000gal Pactual 394.77kW
% Error Calculation:
%Error =nicae-Pac~ualI x 100% = 1.3 1%
Pindicated Following the calibration of the linear drawer, all remaining power indications were adjusted to coincide with the calculated reactor power. The results of these calibrations are retained on site at the NSC.
16
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center 5.0 Pulse Measurements 5.1 Pulse Calculations To predict pulse performance for the conversion core, a calculational procedure (TRIGA-BLOOST) based on a space-independent kinetics model was used by General Atomics. The following simplified relationships are -given to show qualitatively how the pulsing performance is influenced by the important reactor parameters:
r =
= reactor period Akp 2Ak AT=-
P =E/C a
AT= 2(Akp _E/
- = af
=peak pulsed power
.2ae E -,2CAkp =total energy release in prompt burst where e = prompt neutron life a = prompt negative temperature coefficient C = total heat capacity of the core available to the prompt pulse energy release AT = change in average core temperature produced by the prompt pulse Akp = that portion of the step reactivity insertion which is above prompt critical Several expected core parameters were determined for the 90-element TAMU conversion core configuration using the BLOOST procedure prior to conducting pulse operations. Table 8 provides various calculated pulse parameters for pulses ranging from $1.45 to $3.-2 1. Measured temperature is for the instrumented element located at position 5E4 (See Figure 6).
Per BLOOST calculations, a pulse reactivity insertion of $2.10 will result in a peak core temperature of 830'C.
17
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center
[
E F
Instrumented Element 1
2 3
4 5
6 7
8 9
Figure 6: Instrument Element Position Table 8: Reactor Pulse Calculated Parameters (23 0C ambient)
Peak Power P (MW) 227 1008 1873 2468 3434 3775 Total Energy Release E (MW-sec) 11.3 19.8 25.5 28.2
,29.6 35.5 Peak Temperature fT(°C) 459 755 921 1006 1144.
1182 Average Temperature T (°C) 115'5 264 330 365 422 438 Measure Temperature To.3(C) at'5E4 200 333 413 456
-526
-54'5 Per the calculated data of Table 8, peak core temperature can be estimated by the following equation:
T= 2.9445 x 10-7 x (0. 3 -Tmbient)2 -9.2520x 10-4 x (f,3 -
Tambie,,) + 2.6 18 3 + Trnbien, 5.2 Pulse Measurements Methods After certifying the LEU core for steady-state operation, the NSC conducted a series of pulses (all below $1.56) to determine various parameters of the new core and compare them to the calculated values. The data from these pulses are given in Appendix A.
Pulses were conducted per the NSC Standard Operating Procedures with reactor power and temperature collected by the NSC Reactor Controls Console Computer. The NSC PULSE program installed on the computer collected and plotted the pulse data.
18
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Total pulse energy was calculated by integrating pulse power from the time the pulse was initiated until the reactor was scrammed (approximately 2 seconds following pulse initiation). Peak core temperature was calculated using the peak measure core temperature and the peak temperature equation of Section 5.1.
Finally, total pulse energy release and peak power were plotted versus Akp and Akp2 respectively to verify linearity.
Results The measurements obtained from a series of pulses from $1.15 to $1.:56 yielded the results of Table 9. Plotted results are given in Figures 16-19 of Appendix A.
Table 9: BOL Pulse Measurements 0.15 33.2 33.2 7.67 33.33 137.78 104.44 297.17
.263.83 0.22 63.6
'54.7 10.5 32.33 175.11 142.78 388.11 355.78 0.24 64.2 64.2 10.4 31.44 176.78 145.33 393.39 361.94 0.42 30.2 222.7 14.73 30:56 211.94 181.39, 476.83 446.28 0.52 24.1 343.8 16.45 30.78 257.83 227.06 5580.94 50.17 0.33 40.5 129.9 13.06 36.89 212.39 175:50 469.44 432.56 0.47 25.9 294.9 15.95 34.78 249.67 214.89
'557.67 522.89 0.56 21.6 423.8 17.56 33.89 275.39 241.50 616.39 582.50 A plot of energy released during a pulse versus prompt reactivity insertion was a straight line as expected. A plot of maximum power in the core or in the instrumented element versus prompt reactivity insertion squared also gave straight lines as expected.
These -graphs predicted that the maximum prompt insertion should be $1.91 so as to keep the maximum temperature in the core below 830'C. BLOOST predicted
$2.10.
19
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center BLOOST predicted the energy released from a $1.45 pulse is 11.3 MW-sec.
Interpolation from our measured data gave 15.3 MW-sec, but this value contained part of the residual tail in the pulse. BLOOST predicted a 177'C rise as measured by the instrumented element during a $1.45 pulse; interpolated from measured data yielded a rise of 188TC.
The measured and predicted data are in excellent agreement. For $1.45 pulse the measured data agree with predicted data within experimental error. Error can arise both from the measured values themselves (i.e., thermocouple readings, and ion chamber readings) and from error in the assigned value of the prompt insertion. The transient rod calibration error can easily be a few cents and is sufficient to account for all the difference between the measured and predicted data.
20
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center 6.0 Thermal Neutron Flux Distribution Measurements 6.1 Calculated BOL Thermal Neutron Flux The DIF3D code described in Section 1.1 was used to calculate expected cross-sectional thermal neutron flux at the maximum axial power peaking position. The cross-sectional fluxes were calculated at 0.428 cm intervals within the core boundary and 2.470 cm intervals external to the core. Axial thermal neutron flux for the centerline A4 and A6 core positions was also calculated at six core height corresponding to typical sample positions utilized by the NSC.
Table 10 provides the minimum cross-sectional thermal fluxes at four core positions utilized for NSC-samples. These values are the maximum and minimum calculated fluxes within a 1.8 cm square centered at the sample position.
Table 10: Max. and Min. Flux in 1.8 cm Square Centered at Sample Positions i
vinimumk iux t
5.I6bt1Z I
I
.ZUVIZ Maximum Flux 4.03E+12 1.1OE+13 1.12E+13 4.50E+12 6.2 Measured Thermal Neutron Flux Methods Thermal and epithermal neutron flux was measured in the A4 and A6 core positions at six core heights corresponding to typical NSC sample positions using bare gold and cadmium covered gold flux foils. The measurements were performed first with the four shim safety control rods banked. A second set of measurements was performed with the shim safety control rod skewed to tilt the core flux to the core A-row.
After irradiating the foils for four hours, the foil activity was measured using a HPGe detector. Using foil irradiation and decay times, and comparing the bare foils to the cadmium-covered foils, the NSC was able to determine the thermal neutron flux, epithermal flux and the epithermal ratio. The measured thermal flux was then compared to the calculated thermal flux calculated by General Atomics for each of the 4 core positions at six sample positions.
21
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Results With the shim safety control rods banked, measured thermal flux differed from calculated flux by 21 - 38 percent. With the shim safety control rods skewed, measured thermal flux differed from calculated flux by 25 - 48 percent.
A primary contributor to this difference can be attributed to flux foil-sample placement. As demonstrated by Table 10, measured thermal flux at a sample position can differ from 20 to 50 percent by positioning the flux foil as little as 0.9 cm from the centerline sample position.
A second source of measurement error arises from the model used to calculate flux. The calculated thermal flux assumes no air void surrounding the foil samples, where in reality, the foil samples are surrounded by a -2.74 inch diameter air void. This can cause a significant difference in actual thermal flux due to reduced neutron attenuation and thermalization.
In conclusion, it was determined that the measured flux values are well within calculational tolerance. Since BOL, flux has also continuously trended upward and is well within. the range of usability for the purposes of the NSC.
22
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center 7.0 Reactor Physics Measurements 7.1 Calculated Parameters Prior to core conversion, the following reactor physics parameters were calculated by General Atomics:
- 1. Effective Delayed Neutron Fraction ([3etr)
- 2. Prompt Neutron Lifetime (t)
- 3. Prompt Negative Temperature Coefficient (aT)
- 4. Void Coefficient Additionally, Reactivity Loss at Reactor Power (above the point of adding heat) was calculated and is included in this discussion.
Effective Delayed Neutron Fraction The effective delayed neutron fraction, [3ef, was derived from diffusion theory reactor calculations where the reactivity is first computed with the prompt fission spectrum alone and then recalculated with the fission spectrum of both prompt and delayed neutrons. Seventeen groups in the fast energy range and the standard four thermal groups were used for these 3-D calculations in order to represent-the two fission spectra in -greater detail than is possible with only three fast groups.
The results of previous General Atomics (GA) SAR calculations indicate that detail in the -group structure is of much greater importance than geometric detail.
This is not unexpected since the calculation of [3enf is directly related to neutron energy effects.
The prompt fission spectrum is obtained from the GGC-5 spectrum calculation.
The delayed fission spectrum is obtained by integrating over the broad energy groups. The prompt and total fission spectra for each of the broad energy groups in the calculation are given in Table 11.
The computed values of Kt and Kp are used in the following expression to obtain Peft:
Peo = K,(I + 0.o)/Kp ] -I where:
Kt = core reactivity using prompt and delayed fission spectrum, Kp= core reactivity using prompt fission spectrum, P3. = intrinsic delayed neutron fraction for U-235 (0.0065) 23
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center The 3-D model used 21 total groups and very tight convergence criteria (I.0xl0 8 on leff, 1.0 x 10-6 point flux). The cases were run cold (23QC) with fresh fuel. The result for TRIGA LEU (30/20) fuel was efr = 0.0070 Table 11: Fission Spectra Used for Calculation of Perr Prompt Neutron Lifetime The prompt neutron lifetime, i, was computed by the 1/v absorber method where a very small amount of boron is distributed homogeneously throughout the system and the resulting change in reactivity is related to the neutron lifetime. This, calculation was done using the 3-D diffusion theory model for the core to allow very tight convergence of the problems. The boron cross sections used in the core were generated over a homogenized core spectrum. Boron cross sections used in all other zones were generated over a water spectrum.
The neutron lifetime is defined as follows:
.i
= Akeff/co 24
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center where Akfcf is the change in reactivity due to the addition of boron and o0 is related to the boron atom density and, NB = co/6o vo = 6.0205 x 10-7 where NB = boron density (atoms/b-cm) co = integer = 100 (the calculation is insensitive to changes in 0O between 1 and 100),
Vo = -2200 m/sec, 8,o = 755 barns = 8aB at 2.200 m/sec As described in the t3eff section above, the 3-D model used very tight convergence criteria (1.0 x 108 of ken', 1.0 x 10-6 point flux). The cases were run cold (23QC) with fresh TRIGA LEU (30/120) fuel. The result for the prompt neutron life (ý) in the unrodded TAMU core is the following:
f = 26.3 ptsec Prompt Negative Temperature Coefficient The definition of a, the prompt negative temperature coefficient of reactivity, is given as adp dT where p = reactivity
= (k-1)/k T
reactor temperature (°C) 1 dk k2 dT To evaluate (A p) from reactivity as a function of reactor core temperature, the finite differences can be written as follows:
k2 -1 ki - 1 _k 2 -kI k2 ki kik 2 25
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center T hus, (1,2
-k2 T]X I
k1k2 A1,2 The data in Table 12 were produced by.DIF3D for the listed core temperatures.
Table 12: Reactivity Change with Temperature 6W k
23 1.04496 0.01015 0.0093387
-5.303 x 105 200 1.03481 0.00585 0.005494 6.87 x 10-5 280 1.02896 0.01039 0.009913 8.26 x 105 400 1.01857 0.0.3271 0.032574 10.86 x 10-:
700 0.98586 0.03673 0.039254 13.1 x 10*
1000 0.94913 1
1 Void Coefficient The "void" coefficient of reactivity is defined for a TRIGA reactor as the negative reactivity per 1% void in the reactor core water. The void coefficient for the NSC reactor is 0.130% Ak/k per 1% water void. If a dry experimental region (of 281 cc in the 38.1 cm of fuel height) is inserted near core center (replacing a fuel rod) and is accidentally flooded with water, the prompt -gain in reactivity is about
$0.26. This reactivity addition is far less than $1.00 required for prompt critical.
The conclusion is that the very small void coefficient is not a source of safety concern.
Reactivity Loss at Power The prompt negative temperature coefficient of reactivity is active in all reactor operations for which the fuel temperature is elevated above ambient.
Consequently, core reactivity is lost at any power above a few kilowatts (when fuel temperatures begin to rise). Calculations of core reactivity were made for operating power levels up to 1.3 MW. From these calculated values of kff, the 26
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center loss in reactivity has been computed and is shown in Figure 7. The reactivity loss at 1.0 MW is $1.69 (cold-hot reactivity swing).
,2.0-I VI I-P4 1.0 -
0.5 1.0 1.5 Reactor Power (MW)
Figure 7: Calculated Reactivity Loss vs. Reactor Power 7.2 Measured Reactor Physics Parameters The NSC did not possess the means to calculate the effective delayed neutron fraction, the negative temperature coefficient and void coefficients. Measurements were made of prompt neutron lifetime and reactivity loss vs. reactor power.
Prompt Neutron Lifetime - Methods Prompt neutron (C) lifetime can be calculated from reactor pulse data using the following equation:
FWHM x (Akp) X/eff
_ FWHM x (Akp) Xlff 4coshý' V2 3.524 A series of pulses were performed as described in Section 5.2 to obtain the FWHM values for a range of reactivity insertions. For each pulse, prompt neutron lifetime-was calculated using the above equation. These values were then averaged to derive the NSC prompt neutron lifetime.
27
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Prompt Neutron Lifetime - Results Table 13 contains the results of the reactor pulse measurements and the calculated prompt neutron lifetime.
Table 13: Prompt Neutron Lifetime from Pulses at Various Reactivities 0.22 63.6 0.027793 0.24 64.2 0.030606 0.42 30.2 0.025195 0.52 24.1 0.024893 0.33 40.5 0.026548 0.47 25.9 0.02418
.0.56 21.6 0.024027 Reactivity Loss at Power - Methods To determine reactivity loss at reactor power, the reactor was first made critical at 300 W (well below the point of adding heat). Control rod reactivity was calculated at their criticality positions to serve as a reference point.
The intent was then to raise reactor power incrementally to full power collecting rod height data and calculating added control rod reactivity. The change in reactivity from the initial reference point represented the reactivity loss at the associated reactor power level. The power levels to be used were 300 W, 100 kW, 250 kW, 500 kW and 1 MW.
Reactivity Loss at Power - Results Prior to the performance of this procedure, the highest reactor power achieved was 400 kW which occurred during the calorimetric calibration of Section 4.
During the performance of this measurement, reactor temperature was observed to be greater than expected at 750 kW. The measurement at 1 MW was not performed for this procedure due to the abnormally high temperature indication.
The full discussion of the temperature anomaly are described in Section 9.
28
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Using the data for reactor power up to 750 kW, the calculated reactivity losses are displayed in Table 14. The shape of the plot of measured reactivity loss vs. reactor power (See Figure 8) is consistent with the calculated plot; however, the value of measured reactivity loss differs from the calculated value by $0.80. This value is consistent with the error calculated from control rod worth.
Table 14: Reactivity Loss at Reactor Power 0.3
$9.704
$0.000 100
$10.449
$0.745 250
$11.048
$1.344 500
$11.652
$1.948 750
$12.062
$2.358
$2.500
$2.000
$1.500 i
$1.000
$0.500
$0.000 0
100 200 300 400 500 600 700 800 Reactor Power (kW)
Figure 8: Reactivity vs. Reactor Power 8.0 Primary Coolant Measurements 29
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center To monitor the fuel cladding integrity, reactor pool samples are collected and analyzed for the presence of fission products. Samples are collected daily 30 minutes after the reactor is brought to power. A radioassay is then performed on each 3000 mL sample to determine nuclides present and their associated conductivity. Table 15 contains the results for the pool water samples for the first 30 day of reactor operations. Included are the isotopes identified in the pool water, the number of times each appeared in the first 30 days, and their associated activity concentrations.
Table 15: Pool Water Isotope Activities for First 30 Days of Operation Mg-27 2.99E-O6 1.02E-05 4.49E-05 8.33E-05 8
AI-28 4
6.85E-06 7.98E-06 Mg-28 1
2.67E-15 2.67E-1'5 Cl-38 3
1.87E-06 3.74E-06 K-40 8
1.23E-06 1.97E-06 Ar-41 26 1.34E-04 1.13E-03 Mn-54 6
.1.13E-07 2.98E-07 Mn-56 23 3.98E-06 9.3'5E-06 Co-58 9
1.21E-07 1.67E-07 Co-60 4
9.83E-08 1.68E-07 All isotopes encountered are typical impurities found in the pool water as demonstrated by decades of operating the NSC. There is no evidence to'suggest that the fuel cladding is in any way compromised.
30
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center 9.0 Description of Instrumented Fuel Element Anomaly With the concurrence of GA, the NSC performed a series of tests to determine the cause of the unusually high temperature indication. IFE #11452 and 114:53 were alternated between positions 5E4 and 6D4 (see Figure 6 of Section 5) It was determined that IFE #114"52 indicated consistently high-regardless of core position as compared to IFE #11453. Fabrication records were reviewed for both elements. The elements were determined to meet all quality assurance standard and design requirements, however, the cladding-fuel gap for the IFE's is larger than for the rest of the fuel as a whole due to manufacturing problems.
A lengthy study was performed to measure the closure of the gap during reactor operation over the next several months. GA has issued a separate report describing their analysis of the -gap (7). The NSC was adequately assured by GA that the IFE possessed no'safety-related concerns and was safe for-continued operation.
This concludes the Commissioning Program for the NSC reactor; the core has operated without abnormality or anomaly to date. IFE #11452 is now positioned in the core periphery (7B4) and data continues to be collected. IFE # 11453 currently provides indication for Safety Limit and Limiting Safety System Setting considerations.
31
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center 10.0 Discussion of Results As a result of the methods described, the LEU conversion core was declared steady-state operational on November 1, 2006. The core was declared pulse operational on December.12,2006. Comparing all core data to calculated and predicted values, the conversioncore parameters were consistent with expectations with the exception of the IFE indications. All measured values were within '5.0% of calculated values with the exception of rod worth and excess reactivity values for the operational core. These values were still within the design tolerance as presented by the Safety and Accident Analyses Report submitted by the NSC.
The total measured of rod worth of $17.197 was 5.2%.greater than the predicted value of $16.340. Since the initial control rod worth determination, several worth determinations have been performed. The rod worths measured through April 2007 have all been within design tolerances and have stabilized over the intervening months. Future rod worth determinations will continue according to the required annual periodicity.
After detailed analysis by both the NSC and GA, IFE #11453 is being used to provide indication for the Limiting Safety'System Setting. GA and the NSC determined that IFE # 11452 is safe for continued operation but will remain on the periphery of the core to be monitored.
The NSC has continued to operate the LEU conversion core to date without incident or anomaly in both steady-state and pulse-mode. The results of the startup plan have been reviewed and approved by the NSC Reactor Safety Board.
These results are hereby submitted for review.by the Texas A&M University NSC for review by the NRC.
32
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center References
- 1. Los Alamos X-5 Monte Carlo Team, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 5," LA-UR-03-1987, April 24,2003.
- 2. Mathews, D.R., et al., "GGC-5, A Computer Program for Calculating Neutron Spectra and Group Constants," Gulf General Atomic Report GA-8871, 1971.
- 3. NSCR License No. 83 (1983) Revised through Amendment No. 15 (1999).
- 4. NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors," USNRC, August 1987.
- 5. Safety and Accident Analyses Report TAMU Conversion from HEU to LEU, Texas A&M University Nuclear Science Center, December 2005.
- 6. West, G.B., et al., "Kinetic Behavior of TRIGA Reactors," Gulf Atomic Report, GA-7882, 1967.
- 7. Texas A&M University HEU to LEU Conversion, Fabrication and Operational Performiance, April 2007,Prepared by TRIGA Reactor Division 33
Commissioning Report Attachment #1 4/30/07 Texas A&M University Nuclear Science Center Attachment to Order Outline of Reactor Startup Report In accordance with NRC order EA-06-2 11, following the conversion of the Texas A&M University Nuclear Science Center (License Number R-83) reactor from high-enriched uranium to low-enriched uranium, The license holder is ordered to provide a reactor startup report to the NRC within six months following the completion of LEU fuel loading. As a minimum, the report shall contain the following information:
- 1. Critical Mass
- a. Measurements with HEU
- b. Measurements with LEU
- 2. Excess' (operation) reactivity
- a. Measurement with HEU
- b. Measurement with LEU
- 3. Regulating and Safety control rod calibrations
- a. Measurements of HEU and LEU rod worths and comparisons with calculations for LEU and if available, LEU
- 4. Reactor power calibration
- a. Methods and measurements that ensure operation within the license limit and comparison between HEU and LEU nuclear instrumentation'set points, detector positions and detector output
- a. Measurement with HEU
- b. Measurement with LEU
- 6. Pulse measurements
- a. Measurements of any test pulses performed
- 7. Thermal neutron flux distribution
- a. Measurements of the core and measured experimental facilities (to the extent available) with HEU and LEU and comparisons with calculations for LEU and if available, HEU
- 8. Reactor physics measurements
- a. Results of determination of LEU effective delayed neutron fraction, temperature coefficient, and void coefficient to the extent that A-l
Commissioning Report Attachment #1 4/30/07 Texas A&M University Nuclear Science Center measurements are possible and comparison with calculations and available HEU core measurements
- 9. Initial core loading
- a. Measurements made during initial loading of the LEU fuel, presenting subcritical multiplication measurements, predictions of multiplication for next fuel additions, and prediction and verification of final criticality conditions
- 10. Primary coolant measurements
- a. Results of any primary coolant water sample measurements for fission product activity taken during the first 30.days of LEU operation
- 11. Discussion of results
- a. Discussion of the comparison of the various results including an explanation of any significant differences that could affect normal operation and accident analyses A-2
Commissioning Report 4/30/07 Texas A&M University Nuclear Science Center Neutron.Source Neutron Detector SC Startup Count Rate Channel Graphite Reflector D Vacant Grid Position Fuel Followed Control Rod O Transient Rod SWater-filled Element Instrumented Fuel Element O Pneumatic Receiver Attachment #2 Figure 9: Core Loading at Criticality A-3
Commissioning Report Attachment #2 4/30/07 Texas A&M University Nuclear Science Center Table 16: Initial Core Loading by Element
Commissioning Report 4/30/07 Attachment #3 Texas A&M University Nuclear Science Center Table 17: Core Uranium Loading at Criticality A-5
Commissioning Report 4/30/07 Attachment #3 Texas A&M University Nuclear Science Center Table 17 (cont.): Core Uranium Loading at Criticality A-6
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center Table 18: Transient Rod Period and Reactivity Per Iteration I
0.0 8.0 27.69 2
8.0 14.0 22.42 26.81 3
14.0 19.0 19.13 29.21 4
19.0 23.5 19.06 29.29
'5 23.5 29.0 13.15 34.75 6
29.0 33.0
-21.77 27.25 7
33.0 37:5 21.35 27.54 8
37.5 42.0 25.26 25.07 9
42.0 46.5 35.09 20.62 10 46.5
'52.0 30.31 22:54 11 52.0
'58.0 36.17
-20.24 12 58.0 66.0 35.44 20.49 13 66.0 77.0 43.96 17.86 14 77.0 83.0 111.95 9.04 15 83.0 100.0 125.50 8.25 Table 19: Regulating Rod Period and Reactivity Per Iteration 1
0.0 15.0 169.72 6.42 2
15.0 30.0 46.51 17.21 3
30.0 42.0 42.40 19.29 4
42.0 52.0 47.25 17.03 5
52.0 65.0 38.87 19.34 6
65.0 80.0 57.83 14.83 7
80.0 100.0 109.53 9.20 A-7
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center Table 19: Shim Safety Control Rod #1 Period and Reactivity Per Iteration I
U.U I:.U 31./U IO.UL 2
15.0 26.0
.22.63 26.67 3
26.0 33.5 18.42
-29.79 4
33.5 39.0 29.31 23.00 5
39.0 44.5
-21.69 27.30 6
44.5 49.5 21.70 27.29 7
49.5 55.0 18.45 29.77 8
55.0 60.0 22.84 26.83 9
60.0 65.5 23.34 26.22 10 65.5 7 1.5 22.90
.2650 11 71.5 78:5 30.65 22.39 12 78.5 88.0 29.50 22.91 13 88.0 100.0 67.52 13.28 Table 20: Shim Safety Control Rod #2 Period and Reactivity Per Iteration 1
0.0
!16.0 91"14 10.44 2
16.0 30.0 30.47 22.47 3
30.0 39.0 28.43
.23.42 4
39.0 48.0 20.48 28.16 5
48.0 54.0 39.76 19.37 6
54.0 61.0 32.19 21.74 7
61.0 70.0 22.09 27.03 8
70.0 83.0 17.83 30.30 9
83.0 100.0 35.64 20.42 A-8
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center Table 21: Shim Safety Control Rod #3 Period and Reactivity Per Iteration I
U.U I Z.U
/ I.//
1z. 1 /
2 15.0 25.0 35.28
-20.55 3
25.0 32.0 33.00 21.42 4
32.0 38.0 27.60 23.83 5
38.0 43.0 39.69 19.08 6
43.0 48.0 27.79 23.73 7
48.0
'52.5 31.81 21.90 8
52.5 57.0 28.26 23:50 9
'57.0 61.5 45.65 17.42 10 61.5 66.5 32.13 21.77 11 66.5 72.0 32.50 21.62 12 72.0 78.5 30.85 22.31 13 78.5 87.0 31.15 22.18 14 87.0 100.0 43.42 18.00 A-9
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center Table 23: Shim Safety Control Rod #4 Period and Reactivity Per Iteration 1
U.U 15.0 33.U1 21.41 2
15.0 23.0 20.32 28.28 3
23.0 28.5
.20.69 28.01 4
28.5 33.0 24.53 25.50
'5 33.0 37.0 24.83 25.32 6
37.0 40.5 21.58 22.0 7
40.5 44.0 26.65 24.32 8
44.0 47:5 22.09 27.03 9
47.5 50.5 28.05 23.60 10
'50.5 54.8 1'5.78 32.25 11 54.8 57.5 35.74 20.40 12 57.5 60.5 36.07 20.27 13 60.5 63.7 34.75 20.75
.14 63.7 66.9 33.80 21.10 15 66.9 70.5 33.75 21.1-2 16 70.5 74.5 31.18 22.17 17 74.5 79:5 23.12 24.60 18 79:5 85.5 29.95 22.71 19 85.5 100.0 19.33 29.05 A-10
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center 400 350 300 250 200 150 100 50 0
0 10 20 30 40 50 60 70 80 90 100 Position (%)
Figure 10: Transient Rod Integrated Worth A-1I
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center I
100 80 60 40 20 0
10 20 30 40 50 60 70 80 90 100 Position (%).
Figure 11: Regulating Rod Integrated Worth A-12
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center 350 300 250 200 o
150 100 50 0
0 10 20 30 40 50 60 70 80 90 100 Position (%)
Figure 12: Shim Safety #1 Rod Integrated Worth A-13
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center 250 200 150 100 50 0
0 10 20 30 40 50 60 70 80 90 100 Position (%)
Figure 13: Shim Safety #2 Rod Integrated Worth A-14
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center 300 250 0
200 150 100 50 0
0
-10 20 30 40 50 60 70 80 90 100 Position (%)
Figure 14: Shim Safety #3 Rod Integrated Worth A-15
Commissioning Report 4/30/07 Attachment #4 Texas A&M University Nuclear Science Center 500 450 400 350 300 250 200 0/
0 150 -
100 50 -
0-0 10 20 30 40 50 60 70 80 90 100 Position (%)
Figure 15: Shim Safety #4 Rod Integrated Worth A-16
Commissioning Report 4/30/07 Attachment #5 Texas A&M University Nuclear Science Center 20 18 16 14 12
.10 6
0 0
0.1 0.2 0.3 0.4 0.5 (Pulse Reactivity - 1) (S) 0.6 Figure 16: Pulse Energy vs. Aký 1.
L 4J C.
2 1200 1000 800 600 400 200 0
0 0.1 0.2 0.3 0.4 0.5 (Pulse Reactivity-i) ($)
0.6 Figure 17: (-Tamb
) vs. Akp A-17
Commissioning Report 4/30/07 Attachment #5 Texas A&M University Nuclear Science Center 450 400 350 300 250 150 100 50 0
0 0.05 0.1 0.15 0.2 0,25 0.3 (Reactivity-l1)^ 2 (S) 0.35 Figure 18: Peak Power vs. V - T.am 500 450 400 350 300 250 200 150 100
- 50 0
0 0.1 0.2 0.3 0.4 0.5 0.6 (Reagctivity - 1) (S)
Figure 19: Measured Temperature Rise vs. (T - Ta..b )
A-18
Commissioning Report 4/30/07 Attachment #6 Texas A&M University Nuclear Science Center Table 22: A4 Flux Comparisons (Shim Safety Control Rods Banked) 2.77E+07 2.054E+11 2.63 2.75E+07 9.93E+07 3.62 9.52E+12 5.93E+12 37.7%
2.87E+07 2.131E+11 2.69 3.51E+07 1.05E+08 3.00 9.70E+12 6.38E+12 34.2%
2.73E+07 2.024E+11 2.72 3.16E+07 1.OOE+08 3.18 9.49E+ 12 6.18E+12 34.9%
2.63E+07 1.952E+11 2.81 3.42E+07 9.89E+07 2.89 8.93E+12 6.26E+12 29.9%
2.29E+07 1.701E+I1 2.74 2.77E+07 8.63E+07 3.12 8.04E+12 5.25E+12 34.7%
1.93E+07 1.432E+11 3.01 2.05E+07 7.18E+07 3.51 7.04E+12 5.09E+12 27.7%
Table 23: A4 Flux Comparisons (Shim Safeties Control Rods SkeWed) 3.51E+07 7.72E+07 2.20 3.80E+07 1.14E+08 3.00 1.05E+13 5.53E+12 47.2%
3.68E+07 8.51E+07 2.31 3.89E+07 1.18E+08 3.04 1.09E+13 6.35E+12 41.7%
3.50E+07 9.21E+07 2.63 3.81E+07 1.17E+08 3.08 1.09E+13 7.51E+12 31.2%
3.51E+07 8.76E+07 2.49 3.80E+07 1.14E+08 3.01 1.05E+13 6.89E+12 34.5%
3.19E+07 8.14E+07 2.56 3.29E+07 1.04E+08 3.16 9.80E+12 6.51E+12 33.6%0 3.28E+07 7.70E+07 2.34 2.88E+07 9.19E+07 3.19 8.66E+12 5.80E+12 33.1%
A-19
Commissioning Report 4/30/07 Attachment #6 Texas A&M University Nuclear Science Center Table 24: A6 Flux Comparisons (Shim Safety Control Rods Banked) 3 24E+07 2.403E+ 11 2.77 3.70E+07 1.07E+08 2.88 9.63E+12 7.54E+12 21.7%
2.81E+07 2.084E+11 3.10 3.65E+07 1.08E+08 2.96 9.90E+12 7.76E+12 21.6%.
3.22E+07 2.384E+11 2.57 3.47E+07 1.05E+08 3.03 9.74E+12 6.64E+12 331.8%
2.57E+07 1.906E+11 2.73 3.43E+07 1.O1E+08 2.93 9.13E+12 5.84E+12 36.0%
2.40E+07 1.780E+11 2.87 3.16E+07 9.21E+07 2.91 8.32E+12 5.91E+12 28.9%
1.79E+07 1.330E+11 3.39 2.29E+07 7.61E+07 3.33 7.30E+12 5.63E+12 22.9%
Table 25: A6 Flux Comparisons (Shim Safety Control Rods Skewed)
J. /UJrItUV 1 1./5IW 1 1
/.D:)
1~tJt
/
l ri 8IvtU Z..YD I.Uz/rI-tlj D 7.0~
I-g 3/
4.14E+07 3.072E+11 2.37 3.85E+07 1.20E+08 3.10 1.12E+13 7.44E+12 33.7%
4.18E+07 3.099E+11 2.19 4.29E+07 1.24E+08 2.90 1.12E+13 6.55E+12 41.8%
3.87E+07 2.872E+11 2.20 3.24E+07 1.11E+08 3.44 1.09E+13 6.12E+12' 43.9%
2.84E+07 2.106E+11 2.80 3.49E+07 1.08E+08 3.10 1.O1E+13 6.72E+12 33.8%
2.31E+07 1.711E+I 1 2.93 2.76E+07 9.39E+07 3.40 9.11E+12 5.86E+12 35.7%
A-20