ML081480498
| ML081480498 | |
| Person / Time | |
|---|---|
| Site: | Kansas State University |
| Issue date: | 05/16/2008 |
| From: | Whaley P Kansas State University |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML081480498 (35) | |
Text
Kansas State University Mechanical & Nuclear Engineering 16 May 2008
.5r. -/F9g Paul M. Whaley Kansas State University Nuclear Reactor Facility Manager 112 Ward Hall Manhattan, Kansas 66506 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Gentlemen:
Attached is a report of activities accomplished for initial startup and testing for the renewed facility license No. R-88 for the Kansas State University TRIGA Research Reactor (TAC NO. MC9031). Please contact me if there are any questions at 785-532-6657 or whaley@ksu.edu.
Thank ou-fo. your attention, Paul M Whaley Nuclear Reactor Facility Manager cc Daniel E. Hughes, Project Manager Research and Test Reactors Branch A Division of Policy and Rulemaking Office of Nuclear Reactor Regulation
STARTUP REPORT FOR THE MARCH 19, 2008 KSU REACTOR LICENSE This report describes activities conducted to implement a new facility operating license, effective March 19 2008. Implementation was accomplished under a special test procedure approved by the Reactor Safeguards Committee (Attachment I).,
I.
PRELIMINARY ACTIVITIES In accordance with the special test procedure, the status of all Technical Specification surveillances was verified (Attachment I, Appendix I). Two surveillances (calculation of Argon 41 release, verification of reactor bay negative differential pressure) were not required under the previous license. The calculation was performed and documented in an email to the Reactor Safeguards 'Committee, and is being incorporated into the Semiannual Reactor Manager Report.
The negative differential pressure check was added to the daily preoperational checklist.
II.
FUEL MANIPULATION Core configuration was modified to increase excess reactivity in accordance with approved facility procedures on the effective date of the license. The in-pile rabbit assembly was removed and replaced with a partially-burned fuel element from pool storage. One (of three) instrumented fuel element was configured for the 3-control rod core, and interfered with fuel manipulation in the 4-rod core; the element was replaced with a new instrumented fuel element. One partially burned fuel element located between the safety, shim, and regulating rods was replaced with a new fuel element to enhance excess reactivity.
III.
REACTIVITY LIMIT VERIFICATION On March 20-21, 2008 the control rods were calibrated in accordance with facility procedures to verify reactivity limits are met in the new configuration. The reactor achieved criticality with the pulse rod removed and the safety-rod withdrawn to 781 units. Data collection within the special test procedure assumed a control rod different configuration, and minor changes were required to process reactivity data under the actual configuration. Control rod worth curves are attached.
- Total worth of all control rods:
$7.279
- Required reactivity addition for critical:
$4.317
- Minimum shutdown margin: Required--
$0.50 Actual-
$1.495
- Maximum excess reactivity: Required--
$4.00 Actual--
$2.962 IV.
POWER CALIBRATION AND ASCENSION On March 24, 2008, a power level measuring channel calibration was performed in accordance with facility procedures at 200 kW for 100% indication of 1,000 kW. The calibration point is low in the operating range; during subsequent power ascension, power calculated from heatup rate was observed to be reasonably consistent but slightly higher than indicated power. A formal power calibration at a higher power level could not be performed during initial power ascension and testing since the 200 kW operation abrogated initial conditions for calibration. Since a maximum power level of only 720 kW could be obtained with total core excess reactivity, it is not physically possible to challenge license power level limits. Testing and operations were permitted untila power level calibration could be established on March 31, 2008.
KSU License Implementation/Startup Report V.
POWER ASCENSION On March 25 2008 reactor power was increased in accordance with the special test procedure to 400, 500, 600, and 700 kW. In a final step, power was raised to the maximum available of 720 kW. At each step, cooling system response and fuel temperatures were monitored and evaluated.
Cooling System Secondary cooling system and controls were previously modified to support operations at higher power levels. In automatic cooling tower fan mode, fan speed control is based on tower return temperature. The fans energize at a preset speed when return header temperature reaches a low temperature setpoint. Fan speed increases to the maximum preset speed at the high temperature setpoint as return header temperature increases.
The secondary cooling system was previously adjusted for maximum power level operations at 250 kW, complicating assessment of the cooling system; appropriate control setpoints for higher power levels could not be established until actual heat load could be increased under the new license. Equilibrium temperatures were not observed during testing, but temperature increases were controlled and not excessive. The pool was allowed to heat up to an administrative limit of 40'C (Technical Specifications limit of 130'F, or 54.4°C). The control system was subsequently adjusted so that 500 kW for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> results in pool temperatures lessthan 40'C.
Fuel Temperatures Three instrumented fuel elements (IFE) are in the B-ring. The IFE for measuring channel FT-i is position B-3. The IFE for measuring channel FT-2 is position B-2. The IFE for measuring channel FT-3 is position B-5.
Because reference temperature values are for pool water temperature of 20'C, comparison of. observed temperatures to reference values required correcting the observed fuel temperatures for elevated pool temperatures (indicated in Table 1).
Table1: Power Ascension Temperature Data I
POWER FT-1 (00)
FT-2 (°C)
FT-3 (°C)
IFE Corrected IFE Corrected IFE Corrected I 400 284 281.1 231 228.1 293 290.1 500 313 308.3 251 246.3 321.
316.3 600 340 328.7 273 261.7 341
. 329.7 700 358.
342:6 289 273.6 354 338.6
.720.
363 345.7 291 273.7 359 341.7 Temperatures plotted on graphs referenced in the upgrade-implementation procedure (Figure 1) fall within the limits of expected values. Although the observed values are higher than the average reference values for two channels, the difference between observed and reference values decreases as power increases; it is likely that at some power level between 700 and 1000 kW the observed power will be less than the reference.
Page 2 of 9
KSU License Implementation/Startup Report Maximum Fuel Temperature Versus Power level 500 400 0
300 200 100 0
100 200 300 400 500 600 700 Power Level (kW) 0 I-Reference 0
FT-1 A
FT-2 X
FT-3 --
- FT-1 (corr) - +- FT-2 (corr) --4-- FT-3 (corr) I Figure 1: Power ascension temperature data of Table I plotted against reference expected value. Symbols indicate fuel temperature measuring channel indication with data corrected for elevated pool temperature represented by lines.
The fuel element with the highest temperature is not instrumented, and fuel thermocouples in an IFE are not located at fuel centerline. Because fuel temperature values were higher than the average value, additional analysis was conducted to evaluate the maximum core temperatures.
The relationship between the IFEs and the unmonitored element was calculated, and the monitored temperature versus the peak centerline temperature within the IFE. The potential error in the power calibration was also evaluated.
IFE Compared to Fuel Element with Peak Power. The instrumented fuel elements designated by FT1 and FT3 are located in fuel positions adjacent to the pulse rod, which is fully removed during normal operations. An MCNP calculation was performed to determine the distribution of power across the core for a slightly supercritical configuration and also for all rods out. The ratio of individual B-ring fuel element fission heat generation to the core average was calculated (Figure 2).
The fuel position between IFEs for FT1 and FT3 has a slightly elevated power production.
Page 3 of 9
KSU License Implementation/Startup Report B-Ring Power Ratio (Core Ave:Fuel Position)
B01 B 06 B 05 FT2 B 02 FT3 B 03 FT1 B 04 IX Critical ECAII Rods Out Figure 2: The ratio of B-Ring element heat generation rate to the core average for allfuel rods calculated by MCNP5. Squares are data with control rods in an approximate critical configuration, and the alternate symbol with control rods fully withdrawn.
Thermocouple Location. Fuel rods have a diameter large with respect to neutron migration length; consequently there is a distribution of power biased towards the outer sections of the fuel rod. Power distribution shapes the temperature profile, with the variation from fuel centerline temperature to the outer radius of the fuel (neglecting heat transfer except in the radial direction) calculated by:
1 d
.dT]+q":O r dr drj MCNP calculations (originally performed for and reported in the Safety Analysis Report) show power distribution across an individual fuel rod. General Atomics indicates the thermocouples are located about 0.3 in. (0.76 cm) from the inner surface of instrumented fuel elements. As a consequence, the measured temperature is comparable to the peak temperature (Figure 3).
Page 4 of 9-
KSU License Implementation/Startup Report Temperature Distribution Across Fuel a)
E 0
a)
C.E 0
4-
.M a) 0 0.2 0.4 0.6 0.8 1
1.2, 1.4 1.6 1.8 2
Distance from Fuel Center (cm)
Figure 3: Decrease in temperature from centerline calculated for full power operations.
These results are within the fuel rod; indicated temperature correlates to temperature differences across the gap, cladding, and water as well as the fuel. The difference between the maximum temperature decrease within the fuel is about 10% of total temperature difference across the fuel.
The maximum fuel element temperature in the core is within about 8% of the highest reading indicated by the IFE measuring channels.
Power Level Measuring Channel Calibration. As previously noted, calculation of power from heat up rate following initial calibration indicated actual power is higher than indicated power.
The power level calibration completed for higher power operation showed an indicated 600 kW power level was actually 639 kW. After the calibration at 600 kW was complete, the observed fuel element ter iperatures correlated more closely to the expected fuel temperatures.
In conclusion, two fuel element temperatures were slightly higher than the expected average with one lower, but all well within the range of expected values.
VI.
RADIATION SURVEYS Radiation surveys were conducted during power ascension under supervision of the KSU Radiation Safety Officer. Additional radiation surveys were conducted on March 27-28, 2008, to validate experiment shielding.
It was observed during operation for calibration at 600 kW that if primary cooling is secured during high power operations the pool surface monitor exceeds 500 mR h-1.
With primary cooling operating, the radiation levels fall to less than 100 mR h-I.
Primary cooling is normally operated at high power levels. Access to the pool (22-foot level) is visible from the control Page 5 of 9
KSU License Implementation/Startup Report room, and the visual surveillance system can monitor the 22-foot level. Radiation levels on the 22-foot level are indicated in the control room.
Radiation levels were generally found to be acceptable, with one local area at the north west beam port experiment installation higher than desired. The experiment shielding was modified and subsequently demonstrated to be more effective. A revision to the calibration procedure will be submitted for Reactor Safeguards Committee review that restricts access to the 22-foot level during data collection for calibration.
VII.
POWER LEVEL CALIBRATION On March 31, 2008, the final power level calibration was completed, with an indicated power level of 600 kW corresponding to an actual power level of 639 kW. Linearity of the heatup rate was not as consistent as previously experienced. Bubbles were observed in the convection flow from the core were observed in the pool, and is the likely cause for the non linearity.
The KSU reactor facility does not have experience with the bubble phenomena that occurs at power levels available under the new license, and this effect was unanticipated. Two higher-power reactors were contacted, and indicated this phenomena is a characteristic of high power operations. The bubbles are likely related to three interrelated characteristics: changes in solubility of air in water with temperature, nucleate boiling, and pool boiling characteristics of water-air systems. There may an additional or contributing factor in the buildup of gas in pool water associated with radiolytic decomposition of core water.
The pool is open to the reactor bay environment, and dissolved air in the pool establishes equilibrium with reactor bay air based on water temperature and solubility constants for atmospheric gases. Heat transfer from fuel elements to coolant changes solubility locally in the core. Water temperature vertically along a fuel element during operations was analyzed using TRISTAN (FORTRAN code, ORNL RSICC PSR-537). Hydraulic parameters were taken from the Safety Analysis Report, inner-ring cooling inlet-aperture specified in GA-3399, and observed pool temperatures. At 400 °C, a 17 °C temperature rise is calculated along the fuel element; at 720 'C, a 24 'C is calculated along the fuel element. Over the range of observed operating temperatures, solubility decreases as temperature increases by approximately 1.4 mg 1-1 for oxygen and 12 mg 11 for nitrogen. Therefore, as water is heated by fuel elements nitrogen and oxygen gas are likely to evolve from the core region.
TRISTAN predicts boiling regime for specified thermal hydraulic parameters. Nucleate boiling is predicted to occur for the K-State core beginning at about 100 kW (for a small section of fuel rod).- At 400 kW nucleate boiling is predicted for the full length of the fuel rod. At 720 kW the minimum DNBR for the fuel rod is calculated at 5.60 (Bernath correlation) and 2.96 (compared to a nominal 137 W cm-2). At 1250 the minimum DNBR ratios are 4.68 (Bernath correlation) and 2.95 (compared to the nominal 137 W cm-2).
Therefore nucleate boiling is expected during high power operations, and the margin to DNB is large.
Lu and Peng (Nucleate Boiling Modes of Subcooled Water on Fine Wires Submerged in a Pool, Experimental Heat Transfer, 19:95-111, 2006) identify a regime for nucleate boiling in a system of water with air where heat transfer nucleation sites provide conditions permitting evolution of gas from solution. Large, stable bubbles of mostly gas are formed. When the heat Page 6 of 9
KSU License Implementation/Startup Report transfer transitions to fully developed nucleate boiling, small bubbles that collapse after leaving the surface are observed (i.e., a traditional "nucleate boiling") as the dominant characteristic.
Therefore bubbles that reach the pool surface are a normal characteristic of TRIGA reactor operation at high power.
Calibration procedures from three 'TRIGA reactors were obtained to determine if the methodology could be improved based on other research reactor experience. One lower-power facility uses the same method as KSU, with a different constant (because of a larger pool) correlating the rate of temperature rise to thermal power. One higher-power reactor uses a mixer in the reactor pool to support the calibration process. The other reactor uses cooling system temperature differences and flow. rates to calculate heat transfer. This information is under consideration for potential changes to the KSU reactor power level calibration procedure (which under I OCFR50.59 may require NRC approval prior to implementation).
Visible evidence of bubbles was somewhat unexpected, but (1) experience at other facilities and (2) methodology examining development of nucleate boiling independent of prior Safety Analysis Report work indicates this to be a normal condition, with a large margin to DNBR. The established method for power level measuring channel calibration was adequately implemented but the method may have room for improvement.
IX.
PULSING OPERATIONS A series of pulsing operations was performed over April 1-4, 2008 using reactivity additions of
$1.00, $1.50, $2.00 and the maximumavailable of $2.85 (Table 2). Two pulses were performed for each pulsed reactivity value. The columns in the table below labeled "energy," "Max Pwr,"
and "FWHM" were obtained through a LabView application.
Columns in the table below designated FT represent maximum temperature data for three instrumented fuel elements in the B-ring. The NV and NVT.columns are data from the NPP-1000, power and pulsing channel.
The LabView application did not record data on four of the pulses.
Table 2: Pulse Data Max Date No React Energy Pwr FWHM FT1 FT2 FT3 NV NVT Pwr 3-Apr-08 457.20 1.00 6.30 1.05 117 112 124 0
0 1-Apr-01 457.21 1.00 6.60 1.10 7.60E-3 119 114 127 0
0 3-Apr-08 458.00 1.50 25.51 52.79 6.08E-2 222 222 241 0
0 3-Apr-08 459.00 1.50 25.78 56.85 6.26E-2 224 224 243 20 0.5 3-Apr-08 460.00 2.00 221.30 3.03E-2 288 289 313 150 1.9 3-Apr-08 461.00 2.00 226.00 2.92E-2 289 291 313 150 1.9 3-Apr-08 463.00 2.50 296.36 515.66 2.08E-2 *356 359 383 240 3.5 3-Apr-08 464.00 2.50 296.53 514.00 1.04E-2 355 357 382 240 3.25 3-Apr-08 465.00 2.85 302.96 778.60 1.30E-2 394 405 428 245 4.75 4-Apr-08 466.00 2.85 402 406 431 240 4.8 4-Apr-08 467.00. 2.85 401 404 429 160 4.8 The LabView application monitors signal from a picoammeter connected to a detector inserted in the central thimble (added pulse channel): The picoammeter signal is linear, but the gain of the picoammeter required correction of the LabView-indicated power level (Figure 4).
Page 7 of 9
KSU License Implementation/Startup Report Pulse Calibration 0
y 2.5130E-04x + 7.2760E-17 R2 = 1.0000E+00
- y
- 3 Indicuate Power/
R 2 = 9.9986E-01'*
0 100 200 300 400 500 600 kw Figure 4: Calibration data for-the added pulse channel.
Correction for calibration will be implemented directly in LabView from calibration of the added pulse channel in a future revision to the pulse monitoring program. Maximum power and maximum fuel element temperature were compared to reference values (Figures 5 and 6 respectively); other reference parameters could not be compared.
Peak Power (Corrected for Detector Calibration) 1800 1600 1400 1200 0 1000 3-
)T 800 600 3-400 200 "0
dM W HHHH 11 ý I ý I MM fflffl 1111h
_--HM MW HM MH 0
i i ! ! i i i i i i i ý_i i i i ! ! J--r-i i i i i i i i i i i i i i
0.5 1
1.5 2
Reactivity Added ($-1)2 2.5 3
3.5 Pulse 1 -O-- Pulse 2 REF DATAI Figure 5: Peak power during pulsing operations compared to reference expected values.
Peak pulse power deviates from the reference value; the reference value is taken from a TRIGA reactor using a different core lattice, and may reflect different neutronic parameters.
Other parameters for which reference values were available are more sensitive, and therefore less comparable, possibly complicated by calibration and time-response characteristics of pulse monitoring instrumentation. The slope of the observed peak temperature is also different from the slope of the reference curve, which may be attributed to the neutron lifetime for the K-State Page 8 of 9
KSU License Implementation/Startup Report core.
Both peak power and pulse fuel temperature show stable, predictable behavior from pulsing operations.
Pulse Fuel Temperature 6UU 500 400 300 E
- 200 100 0
II I I I
I I I
Ir I
I I
r I I
I I I 1.00 1.25 1.50 1.75 2.00 2.25 2.50 2.75 3.00 Reactivity ($)
-0 FT1 FT2 -&FT3 -
REF DATA Figure 6: Peak fuel temperature during pulsing operations compared to reference expected values X.
CONCLUSION As expected, maximum operating power level with current core loading is about 700 kW.
Steady state operations are limited by excess reactivity, which is near the maximum physically possible with the K-State core grid plate and the type of fuel authorized under Technical Specifications.
One unfueled space is currently occupied by a neutron source,.and fuel (reactivity worth about $0.40) could be used to increase excess reactivity to about $3.20, well within the Technical Specification limit of $4.00.
Increasing excess reactivity to permit extended operations at full power may require a license amendment to permit the use of 12%
TRIGA fuel, with specific controls on the location of the elements within the core to ensure power peaking factors remain within current analysis.
All reactivity limits are met by a large margin. Fuel temperature limits are met during normal and pulsing operations to available excess reactivity.
Methods for conducting power level measuring channel calibration are being evaluated to determine if improvements in data collection and analysis or a different technique would be more appropriate.
Bubbles are observed during operations and (according to experience at other TRIGA reactors and by a code designed to analyze TRIGA reactor behavior) to be a normal characteristic.
Page 9 of 9
K-State Reactor Startup Report A TTA CHMENTS
- 1.
Implementation and Test Plan for KSU Reactor Power Upgrade
- 2.
Control Rod Worth Curves
- 3.
Power Ascension Radiation Surveys
Implementation & Test Plan for KSU Reactor Power Page 1 of 7 Upgrade 03/14/08 1 Rev 0 SCOPE This procedure provides direction for testing and initial operations under the licensed maximum power level of 1,250 kW operations.
This procedure addresses:
Performance of pre critical checks that verify all Technical Specifications have been met Control rod calibrations following fuel loading to increase excess reactivity Verification that reactivity limits have been met Power level calibration for 100% indication at 1,000 kW Adjustment of pool surface monitor alarm setpoint Surveys to verify acceptable radiation levels and appropriate controls Comparison of fuel temperature and reactivity deficit to expected values Verification that the cooling system functions A series of pulsing operations, starting with low pulse worth and increments to the maximum available reactivity within Technical Specifications limits DISCUSSION The KSU TRIGA Mark II reactor has historically required approximately $4.20 of reactivity to'achieve criticality; with the current core configuration, Core Il1-1 requires $5.40.
The 250 kW license required that the pulse rod have a nominal worth of $2.00; the 1,250 kW license does not have a Technical Specification limit for the pulse rod, but the pulse rod in the C ring is expected to have a worth of approximately
$3.00.
The worth of the pulse rod in the C ring while under the 250 kW license was artificially depressed by placing a water-filled aluminum tube next to the pulse rod. This arrangement permitted configuring the control rods to support the 1,250 kW license prior to issuance of the license.
Implementation & Test Plan for KSUI Reactor Power Page 2 of7 Upgrade 03/14/08 1Rev The worth of fuel elements in the C ring have been measured with respect to a water void to be approximately $1.00.
If the pneumatic tube is removed form the core to accommodate the addition of another fuel element, operations with the rabbit in place may require additional reactivity checks to ensure the control rod worth curves are properly calibrated.
Calculations conducted for the Illinois Advanced TRIGA (reference 2) are provided in Appendix I to provide values for comparison of expected fuel temperature, reactivity loss due *to fuel temperature, and reactivity loss for operation at power.
LIMITS AND PRECAUTIONS KSU Technical Specifications for operation with a maximum power level of 1,250 kW has. reactivity limits for excess reactivity of $4.00 and a minimum shutdown margin with the most reactive rod fully withdrawn of $0.87.
PREREQUISITES Approved license permitting a maximum of 1,250 steady state thermal power.
For 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> prior to calibration, the following conditions are required to be met:
- Operations limited to < 1 kWh, No secondary, cooling operation
- Pool water 20+/-5 IC INSTRUCTIONS 1.
VERIFY Technical Specifications are met
- 2.
STARTUP to 10 watts using OP-IS5, Reactor Startup
- 3.
SHUTDOWN the reactor using OP-16, Reactor Shutdown
- 4.
REMOVE the experiment thimble from position C-8 AND SECURE the tube to the pool wall
- 5.
REMOVE the in-core rabbit tube from position F723 AND SECURE the tube to the pool wall
Implementation & Test Plan for KSU Reactor Power Upgrade Pa e3of7 03/14/08 1Rev 0 Applies to Step 6-7 NOTE Inspection includes verification of serial number, visual inspection, and verification of elongation and bend meet Technical Specification requirements Steps 6 and 7 may be performed sequentially of each element, or inspections of both preceding loading
- 6.
INSPECT two fuel elements
- 7.
LOAD fuel positions C-Xand F-23
- 8.
STARTUP to 10 watts using OP-1 5, Reactor Startup Applies to Step 9 NOTE Fuel loading affects control rod worth calibration, so the reactivity worth based on previous calibration may not be accurate Technical Specifications limit on excess reactivity is $4.00; excess reactivity preceding Step 7 should be approximately
$1.1; if the reactivity difference in Step 9 exceeds $2.90, reactivity measurements in step 10 should begin with measurements required to verify excess reactivity
- 9.
RECORD reactivity difference between Step 2 and Step 8, based on previous control rod worth calibrations Applied to Step 10 NOTE The pulse rod should be worth more than
$3.00, and a combination of rod drops and positive period measurements may be required to get reactivity worth for the control rod at the fully withdrawn position
Implementation & Test Plan for KSU Reactor Power Page 4 of 7 Upgrade 03/14/08 Rev 0
- 10.
PERFORM reactivity verifications for critical rod positions
- 11.
VERIFY reactivity limits are met Excess reactivity (1) Transient Rod worth fully withdrawn:
4 2Ji (2) Safety Rod worth fully withdrawn:
Shim Rod:
(3a) Worth fully withdrawn:
t V16 Critical position:
_76
(
(3b) Critical Worth:,AA."3(3c) Difference (3b)}:
Regulating Rod:
(4a) Worth fully with drawn:
___(_-_
Critical position:
(4b) Critical Worth: j (4c) Difference {(4a)- (4b)}:
,)
Pool bulk temperature:
- Associated Reactivity:
(5) -0 f
V Source worth ($0.025 inserted)
(6)
Total Control Rod Worth:
(1) + (2) + (3a) + (4a)
(7)
X,-V*(
Critical Reactivity Addition:
(1) + (2) + (3b) + (4b)- (5)- (6)
(8)
Implementation & Test Plan for KSU Reactor Power Upgrade Page 5 of 7 03/14/081 Rev 0 Excess Reactivity (7)-(8) 7276 - q-,1 Minimum Shutdown Marain (2) + (3b) + (4b)- (5)- (6)
IF "Excess Reactivity" > $4.00, OR IF Minimum Shutdown Margin < $0.50, THEN REMOVE one fuel element AND REPEAT steps 10 and 11.
- 12.
CALIBRATE the shim and regulating rods over full span of rod movement Applied to Step 13 NOTE For 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> prior to calibration, the following conditions are required to be met:
- Operations limited to < 1 kWh,
- No secondary, cooling operation
" Pool water 20 +/- 5 °C
- 13.
STARTUP using OP-1 5, Reactor Startup, or INCREASE power to 200 kW
- 14.
RECORD data (fuel temperature, control rod position) AND COMPARE the associated reactivity deficit to data in Attachment II
- 15.
CALIBRATE NMP-1000, Nuclear Multi-range Power channel to indicate 20% power at 200 kW
- 16.
ADJUST pool surface monitor alarm setpoint as required to prevent spurious alarms during the following power increase
- 17.
WHEN pool surface monitor exceeds 100 mrem/hr, POST the 22 foot level as a high radiation area AND prohibit access WHILE the area is a high radiation area
Implementation & Test Plan for KSU Reactor Power Page 6 of 7 Upgrade 03/14/08 Rev 0
- 18.
ADJUST pool surface monitor:
"ALERT" setpoint 100 mrem/h "ALARM" setpoint 500 mrem/h
- 19.
INCREASE power to 500 kW Applies to Step 19 NOTE A complete calibration following Procedure 2, "Annual Power Level Calibration," is not possible because of operating history of the previous step.
- 20.
VERIFY power level calibration
- 21.
VERIFY cooling system function
- 22.
INCREASE power in 100 kW steps to a maximum of 1,000 kW AND at each increment, 22.1.
VERIFY cooling system function 22.2. VERIFY power level calibration
- 23.
At maximum power level for which steady state operation is possible, 23.1.
PERFORM radiation surveys 23.2. ENSURE applicable radiation area postings are accurate 23.3. ADJUST pool surface monitor setpoint to 150% of the steady state radiation level
- 24.
SHUTDOWN using OP-16, Reactor Shutdown
Implementation & Test Plan for KSU Reactor Power Upgrade Page 7 of7 03/14/08 1Rev 0_
Applies to Step 25 NOTE The new calibration of NLW-1 000, the wide range power level indicator, may place the detector so that the source interlock will not clear with the source in F-10; if the source interlock will not clear, the source may be placed in an alternate core or RSR position followed by verification that reactivity limits are met verification that power'level calibration is not chanaed
- 25.
PULSE the reactor using Experiment 23 and reactivity values from $1.00 up to the lesser of maximum available pulse rod reactivity or $3.00.
- 26.
COMPARE pulse data to expected pulse data in Appendix II DEFINITIONS None.
REFERENCES
- 1.
Kansas State University Research Reactor Technical Specifications, 2008
- 2.
Safety Analysis Report for the Illinois Advanced TRIGA,Section XIII (Initial Tests and Operation), 1967
- 3.
Amendment No 4 to the Safety Analysis Report for the Oregon State TRIGA Reactor (OSTR), 1975 Approved: KSU Reactor Safeguards Committee M. H. Hosni, RSC Chairman Date
Appendix 1: Technical Specifications Requirements Page 1 of 2 03/14/08 Rev 0 Safety Limit, fuel temperature -- 1150 peak, 750 steady state Limiting Safety System Setting -- 1,250 kW max excess $4.00 Verify excess reactivity semiannuai after experiments with measurable tnositive reactivity Excess < $4.00 OP 15 Core reactivity lja 16 SDM $0.5 most1 rectv rd mout 01(-
Verify SDM semiannual
> $0.5 SDM Manager Audit 3/(7/O.
reactive rod out ctnnfrn1 rod wo~rth hie~nrnialv N/A Pulsed mode transient rod Pulse worth <
i
'7 0 positioned to less
< $3.00 prior to pulsing
$3.00 Exp 23 than $3.00
$3.001____
Safety channel safety' channel &
control rod operability startup channel count rate > minimum sensitivity t"
Verify channel >
min sensitivity daily CR > min sensitivity Daily checklist 1jit /0o -
Channel
/l[Oo Channel test oY-daily operating Daily checklist operating: 2 power Calibration annual Calibration OP 2 L//g 7 levelChannel HV test OIL daily operating Daily checklist 35 /17/1) operaperatin 1_oo operating: 1 pool Calibration annual Calibration
,QP, U&kneM)0, operating: 1 bay Calibration annual Calibration
,.28.(new)
/
differential pressure operating: 1 fuel Calibration annual Calibration Q.28n.)zt
- [('[Og operating: 22 foot Channel test 0k..
daily Channel Daily checklist
-7 arardmntroperating area rad monitor Calibration e' V annual Calibration OP 3 519Y
-7 Channel
._iye~li~,*/l*0*
operating: 0 or 12 Channel test daily operating
.Dijyr...
is 5 /
footarea monitor Calibration annual Calibration OP 3 operating: continuous Channei test 6 k_
daily Channel air monitor operating Daily checklist "j17)0 Calibration__L annual Calibration OP 8 5Y,/a7 operating: exhanust Channel test 6-daily Channel Daily checklist 3117/atT plenum__monitor_
operating plenum monitor Calibration annual Calibration g
"g i9#
control rod drop time Measure drop
<Isec time annual
<1 sec OP 4 2 power level scram Test scram 6tI.
Daily SCRAM Daily checklist 3// 7/0./
manual scram bar Test scram oL Daily SCRAM Daily checklist 3 l 1ý(O.*
interlock for pulse Rod motion mode/standard Test interlock "i-semiannual inhibited OP 5, OP 12
, /;t*,q47 movement interlock for pulse LRod motion rod coupling except Test interlock semiannual inhibited OP 5, OP12
?J.
/X1/d7 in pulse mode ROD check control rods ROD for corrosion &
Biennial Visually OK OP 1 OPERABILITY damage o_
iceO PULSING OPERABILITY functional test nulse rod prior to pulsing System OK OP 12 tll T/*,g
Appendix 1: Technical Specifications Requirements
-E Page 2 of 2 03/14/08 1Rev 0 pulzvAU U IYv cylinder and air sunolv checks semiannual Visually OK OP 6 j,217167 in leakage negative bay dp Daily Inleakage Daily checklist 3 //7/0 9-Gaseous Ar 41, 30 Ci per year Calculate release Annual
<30 Ci effluent channel test air Daily
- ]aily_-hecklist 7 /17/,
monitor single experiment <
Estimate evaluate prior to Expt worth <
.$2.00 worth reactivity worth insertion
$2.00 sum of experiments if >$0.40, that can cause if >$0.40 measure evaluate prior to verify expt <
ýOPR15-5 A)14 experiments reactivity change <
& record worth insertion
$200
$2.00 irradiation holders Inspect irradiation evaluate prior to Holder will prevent release into holder insertion prevent release pool elongation inspection &
500 pulses or
<l18in OP13 Fuel measurement exceeding LSSS
__/_nP10____'
integrity bend inspection &
500 pulses or
< 1/8 in.
OP 10 3/11/07 measurement.
exceeding LSSS FUEL rNTEGRITY 1/3 core visual annual Visual OK 01.10.
- 3) 3 707
<130°F with Check channel operable
<1 300F, w demineralizer flow temperature depin f Console Logs f3lo7/w Check Pool water conductivity < 5 uSv conductivity daily
<5 uSv Daily checklist 3/ ]7105 Check at least 30 days
<5 uSv Surveillance 71".
conductivity Check sheet water level > 13 feet Check level daily Daily checklist l3evel over core 13 feet D
evaluate maintenance Evaluate Following Potential effect maintenance for potential to affect maintenance maintenance on operability
-T
.k retests operability applicable Retest Following Specific to retrvticomple e prior to maintenance..surveillance......15/-<
surveillanoperations requirements maintenance surveillance t"CLk1~4
Appendix I1: Expected Parameters, Steady State Page 1 of 3 Operations 3/14/08 j Rev O IF0 Th~-P ffh7 fftff4ti II4TTI:Wfl4l 6.0
-b~~hi~~h~h~h~hth l hh~~IThh~
5.0.
4~0 3.0
,-77A 11 J-'.i4 Fz-S 4 0 rEQ II,.; i..;.;,.;..;. ! I:.-.:1; I...
.. I,-"* e¢"i*"
f
-F LIEL W
7 jo
- 7.
7-7tj C
50 to0 i50 2*00 y.
.AE*AGE FUEL TIEMPERATURE R':
r *LOVs.
S R*EA;C~erVITY LO-SS
Appendix II: Expected Parameters, Steady State Operations
. Page 2 of 3 3/14/08 1 Rev 0 144 1.
-1,
'T
- .-F F. 1 "1411 TI T'- "ýF -+TT 11- -
7"'r.1 4-0.1 TA JýJ::Tl I M': I ff M4-ýrF -1..-T::
1-17"FETT :1ý 1. :1. 1.
f Ll I, I-I L. If
-14' 31' =-' 7! r ý
-3ý00 7 rý7 T
- 1 1 T I if I T; 1 1: 1 1; 1 1, 1 li 17 M I A I tLIMA 44+4-~4.4,4-
.ad 0L5 t,0
[5 2.0 2,5 3.0 REA.CTOR ROWER Va.
MA X.M
!'Mu d AVERAGEt FrUtL. TEtMPERATUR8
Appendix Ih: Expected Parameters, Steady State Operations Pa e3of3 3/14/08 1RevO0 1ig-29 8.0 7,0 t
4i0-3.0 0'
0.45 1a.5*
2.0O 2o,51 3.0 REACTOR POWE REACI:IVIlTY LOSS
Appendix IIl: Expected Parameters, Pulsing Operations Pae 1 of5 3/14/08 Rev0
- z.
111437
.~ ~ ~
~ ~ ~...I...
I 1.1,-.
.-1 I jo.
100
- d. : 1-1 1 71 J
I i!;
- t I:.
2w 1W
~t~J i-J
[3 SO.
80.
60.
2-NEDATA xý CALCULATEb 30,
.01
'.jt Ii 6-2 6
'r 8
& I 2 3 4
1 1
? I D
-I O*;
VLLARS 00EK POWER/F-UE'L ELEMENT
,, ~
Appendix II: Expected Parameters, Pulsing Operations 3 /14/0 2 of 5 13/14/08 1 Rev 0 I1I~9 i i ja;, - -
-w-0.48 0A4 2
~ QAO 4JJ w.
~ 0~36 I
4-.
'0.3a 0
I&3~ o.a~
z bJ LU
-Jw 0.24 tUE IL
~ o.ao
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bA z
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(
- ~~~.............
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r, 1 4..
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i m 11.. 11 1.
t,000 20G
(~J
- ~DOLLARS 00to
- , M, PROMTNRGY0.
RYLEASE Vs.
TI~iTY INSERTiON
Appendix IIl: Expected Parameters, Pulsing Operations Page 3 of 5 3/14/08 1 Rev 0 0.20
- ,,.... :-i, I ".* :.....
- i*:
t
- v,
If"!
0.18B t
0w X,
042
/
/
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/
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I-C
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0.08 1ý 4-It I
1.00 2,0;-OG.ý 30-0i-
- 4.O0
?QOLLARS
++
A' r.. iJLiKA 75' 04 VT V`
RECIPROCALW1DH IA -
qL~
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Pae 4 of 5 Appendix III: Expected Parameters, Pulsing Operations 31/
Rev 0 111-40 0
1.00 2.00 3.00 4.00
("",IK
,DOLLARS INVERSE PERIOD Vs.
REACTIVITY INSERTION
I Appendix III: Expected Parameters, Pulsing Operations Page 5 of 5 3/14/08 1 Rev0 1
LA 1 1-4.
.L-
..4, MUM00 100
'11 1:
900 too
,V rI eo0o'I
!+. :-W l}.; !*i P
- *i*
59 a 449
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+/-I -tH- -HI 4
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1 1 1 1i 1410 tO o0:
4A 5.0
.6.0 T.EMPEATLWE Vs. REACTIITY INSERTION
Attachment II: Control Rod Calibration Transient/Pulse Rod Safety Rod Worth 3.0 1.4 A.
1.2 0
0 0.8 O3 0.6 m A 0.2 j
_Jy = 1.579032E-12x' - 6.375674E-09x3 + 6.891884E-06x2 - 3.474520E-04x R' = 9.995596E-01 0
100 200 300 400 500 600 700 800 900 1000 1100 Position (Units withdrawn)
Shim Rod Worth 0
100 200 300 400 500 600 700 800 900 1000 1100 Position (Units withdrawn)
Regulating Rod Worth IA 1.4 o
1.2 0
0.8 I-0.6 C
0.7 0.6 0.5 0.4 0
0.3 0)
C 0.2 0.1 0.2
ýy = 4.999535E-13x4 - 4.830914E-09x3 + 7.168128E-06x 2 - 9.055775E-04x y = 1.203127E-12x4 - 3.718597E-09x3 + 3.280751E-06x2 - 1.064721E-04x 100 200 300 400 500 600 700 800 900 1000 1100 Position (Units Withdrawn) 0 100 200 300 400 500 600 700 800 900 1000 1100 Position (Units withdrawn)
..i*-.*.*.......... _
)
CD 0
fr~
12 FOl LEVEL I
& m""" re,,.d.ys V;CWdre.i A&/to 4 s'o SIN /4-7' Iveml'*,,,,'r-**. s:
ekdu.*[,4
/Al 1,-14-f SINJ f 7/38
-o
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vt 13-Z ý,e y 0,6.
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ie
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112 FOOT LEVEL vi~~ EVSA S7q7 Nel4-~
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PUAA~~~~Uq'i
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Attachment III: Radiation Surveys Measurements for Initial Environmental Surveillance for Core 111-2 Power: 500 kW Date: 25 March 2008 Taken by Meyer Gamma readings: Victoreen M/N 450P S/N 1474 Neutron readings: Ludlum M/N 12-4 S/N 47138 Site Gamma (mR/h)
Neutron (mremnh)
Description of Site 1
0.168 0.4 Window 2
0.190 0.4 Window 3
0.216 0.4 Window 4
0.218 0.4 Window 5
0.290 0.4 Window 6
0.212
.0.4 Window 7
0.248 0.4 Window 8
0.278 0.4 Window 9
0.388 0.4 Window 10 0.420 0.8 Window 11 0.660 0.4 Window 12 0.260 0.6 Window 13 0.230
.0,6 Control Room Door 14 0.134 0.6 0' Level Door 15 0.420 0.8 NEBP Door (secured) 16 8.8 4.0 SEBP Shutter (closed) 17 0.780 1.0 SWBP Door (secured) 18 0.620 1.0 NWBP Fence corner 19 9.0 8.0 NWBP Side (M-V configuration 20 214 300 NWBP Above (M-V configuration) 21 26.0 0.6 East Railing 22 20.0
'0.6 South Railing 23 22.6 0.6 West Railing 24 23.4 1.0 North Railing Page 6 of 6