ML101740432
| ML101740432 | |
| Person / Time | |
|---|---|
| Site: | University of Wisconsin |
| Issue date: | 06/21/2010 |
| From: | Agasie R Univ of Wisconsin - Madison |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| EA-09-141, RSC 1051, TAC MD9592 | |
| Download: ML101740432 (64) | |
Text
Nuclear Reactor Laboratory UWNR University of Wisconsin-Madison 1513 University Avenue, Room 1215 ME, Madison, WI 53706-1687, Tel: (608) 262-3392, FAX: (608) 262-8590 email: reactor@engr.wisc.edu, http://reactor.engr.wisc.edu June 21, 2010 RSC 1051 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555
Subject:
Docket 50-156, License R-74 Reactor Start-Up Report following Order to Convert from High to Low Enriched Uranium Fuel at the University Of Wisconsin Nuclear Reactor TAC No.
MD9592 (EA-09-141)
Dear Sirs:
By Order, dated June 11, 2009, the Commission modified facility operating license R-74 for the purpose of conversion of the reactor from high to low enriched uranium fuel.
In accordance with that order a reactor start-up report is required to be submitted to the NRC within 6 months following the return to normal reactor operations.
Normal reactor operations resumed on January 22, 2010 following extensive physics testing of the reactor.
Enclosed you will find a report detailing the results of this testing.
This report follows the suggested format for reactor startup reports found in attachment 3 to the above reference order, EA-09-141.
Sinc y
Robert J Agasie Reactor Director Enclosure
THE UNIVERSITY OF WISCONSIN NUCLEAR REACTOR LABORATORY REACTOR START-UP REPORT Prepared to meet reporting requirements of:
U. S. Nuclear Regulatory Commission Order to Convert from High to Low Enriched Uranium Fuel at the University Of Wisconsin Nuclear Reactor License R-74 June 21, 2010 0
THE UNIVERSITY WISCONSIN MADISON
UWNR LEU Conversion Reactor Start-Up Report Table of Contents A.
HISTORICAL BACKGROUND 3
B.
DISPOSITION OF OLD TRIGA FUEL 3
C.
APPROACH TO CRITICAL EXPERIMENT...............................................................
4 D.
LOADIN G TO J21-RO CORE.......................................................................................
12 E.
LOADING TO J21-R14 OPERATIONAL CORE........................................................
19 F.
OPERATIONAL CORE REACTIVITY MEASUREMENTS.....................................
29 F. 1 REFLECTOR REACTIVITIES (REFLECTED)........................................................... 29 F.2 COOLANT VOID COEFFICIENT..............................................................................
30 F.3 EXPERIMENTAL FACILITY REACTIVITY MEASUREMENTS.......................
32 G.
POWER CALIBRATIONS 34 H.'
POWER DEFECT MEASUREMENTS.................................................. 35 I.
FUEL TEMPERATURE MAPPING...........................................................................
40 J..
FLUX MEASUREMENTS 44 K.
PU L SE T E ST IN G.............................................................................................................
46 L.
SQUARE-W AVE TESTING...........................................
52 M.
SUMMARY
OF HEU MEASUREMENTS VS.
LEU PREDICTIONS AND M EA SU RE M EN TS........................................................................................................
54 N.
COMPUTING MCNP MODEL BIAS..........................................................................
58 Page 2 of 63
UWNR LEU Conversion Reactor Start-Up Report A.
HISTORICAL BACKGROUND The University of Wisconsin Nuclear Reactor achieved initial criticality on March 26, 1961 as an MTR-type flat-plate core. It was operated at a licensed maximum power level of 10 kW, with an upgrade to 250 kW on October 22, 1964. The MTR core was replaced with a Standard TRIGA core achieving first criticality on November 14, 1967 with a licensed nominal power of 1.0 MW.
MTR fuel was shipped off-site in 1986. The Standard TRIGA core was shutdown on March 1, 1974 after a total exposure of 145.0 MWd.
The Standard TRIGA core was converted in multiple phases to an all TRIGA FLIP (Fuel Life Improvement Plan) HEU core. The first mixed core with 9 FLIP bundles achieved criticality on March 12, 1974, the second mixed core with 15 FLIP bundles on January 3, 1978, and the final all FLIP core on June 15, 1979. The all FLIP core was shutdown on August 19, 2009 after a total exposure of 843.5 MWd and 1074.7 critical days. This includes the exposure of the Standard TRIGA core and mixed cores. The exposure of individual FLIP bundles ranged from 629.3 to 698.5 MWd.
B.
DISPOSITION OF OLD TRIGA FUEL During conversion to TRIGA FLIP fuel, the Standard TRIGA fuel was unloaded from the core into the fuel storage pit. This Standard TRIGA fuel remained in the storage pit throughout almost the entire FLIP core history, leaving just enough room to unload the FLIP operational core if needed. In order to convert to TRIGA LEU 30/20 fuel, the Standard TRIGA fuel had to be shipped off-site to make room in the fuel storage pit. In May 2009, 120 Standard TRIGA fuel elements were shipped to Idaho National Laboratory, leaving 8 Standard TRIGA fuel elements remaining. After the FLIP core was shutdown, it was unloaded into the fuel storage pit, aswere the lightly-irradiated FLIP fuel elements which had been stored in the fresh fuel vault. All TRIGA Standard and FLIP fuel elements (8 Standard and 101 FLIP) are currently being stored in the fuel storage pit awaiting shipment to Idaho National Laboratory sometime in 2010.
Page 3 of 63
UWNR LEU Conversion Reactor Start-Up Report C.
APPROACH TO CRITICAL EXPERIMENT TRIGA LEU 30/20 fuel loading began on September 8, 2,009. Fuel bundles were loaded based on 1/M plot predictions to the locations indicated in Figure 1 following procedure UWNR 143, which complies with standard ANS-1. The complete fuel loading sequence is included in Table
- 1. Subcritical multiplication data is shown in Table 2 and inverse count rate ration (1 /M) data in Table 3. Figure 2 through Figure 4 represent the 1/M plots for the varying core geometries. In the figures, FC is the installed Fission Counter, and AFC is the Auxiliary Fission Counter. Initial criticality was achieved on September 16, 2009 at 14:32 with core loading J 18-RO (18 fuel bundles, 0 reflectors). The critical bank height was 17 inches (full out on all control elements except the transient rod), and the excess reactivity was estimated as 0.074 %Ak/k by withdrawing the transient rod full out to 19.71 inches and measuring the doubling time. The shutdown margin with the most reactive control blade and non-scrammable control blade withdrawn, hereafter called the technical specification shutdown margin, was measured to be 5.26 %Ak/k, which satisfied the acceptance criteria of 0.2 %Ak/k.
Page 4 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 1, Fuel Loading Sequence Date Time Item To 9/8/2009 13:47 Bundle 61 D5 (SE) 9/9/2009 10:45 Install t-rod and drive 9/10/2009 14:20 Bundle 62 D4 (SW) 9/10/2009 14:25 IFE 11749 D4 (SW) 9/10/2009 14:37 Bundle 63 D6 9/10/2009 14:40 Bundle 64 E5 9/10/2009 - 14:43 Bundle 65 C5 9/11/2009,12:39 Bundle 66 E6 9/11/2009 12:41 Bundle 67 C4 9/14/2009 11:06 Bundle 68 E4 9/14/2009 11:08 Bundle. 69 C6 9/14/2009 14:13 Bundle 70 C3 9/15/2009 08:41 Bundle 71 E7 9/15/2009 11:12 Bundle 72 E3 (NE) 9/15/2009 11:16 IFE 11748 E3 (NE) 9/15/2009 11:31 Bundle 73 C7 9/15/2009 11:34 Bundle 74 F6 9/16/2009 07:53, Bundle 75 B4 9/16/2009 09:33 Bundle 76 F4 9/16/2009 11:06 Bundle 77 B6 9/16/2009 13:08 Bundle 78 B3 9/21/2009 09:31 Bundle 79 F7 9/21/2009 13:24 Bundle 80 F3 9/22/2009 09:55 Bundle 81 B7 Page 5 of 63
UWNR LEU Conversion Reactor Start-Up Report BEAM PORT
- 3 A
__ B L
BEAM PORT
- 4 D
E F
G w
H-I 0
LiJ 7-LIE 1 LOG N CIE LIE 2 LEGEND
- = Fuel Bundle R = Reflector TRANSIENT ROD Figure 1, J18-RO Initial Critical Core Map Page 6 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 2, Measured 1/M Count Data (no reflectors installed)
Bundle Total Bank full in Bank half out Bank full out Added FC AFC Min FC AFC Min FC AFC Min D5 1
26734*
5296 240 14867 2676 160 16773 3138 180 D4 2
D6 3
N/A N/A N/A N/A N/A N/A N/A N/A N/A E5 4
C5 5
24241 6644 240 11983 3651 120 12010 3882 120 E6 6
N/A N/A N/A N/A N/A N/A N/A N/A N/A C4 7
7766 2528 75 7887 2961 75 7669 3530 75 E4 8
N/A N/A N/A N/A N/A N/A N/A N/A N/A C6 9
6237 3095 60 6607 4248 60 3914 3039 30 C3 10 5045 3011 45 3482 2764 30 4189 4149 30 E7 11 5320 3192 45 3707 3130 30 4674 4586 30 E3 12 N/A N/A N/A N/A N/A N/A N/A N/A N/A C7 13 F6 14 6027 4599 30 3827 4195 15 7375 9202 15 B4 15 6377 5599 30 4048 5492 15 9331 15189 15 F4 16 4290 3061 15 4213 4556 10 15971 19670 10 B6 17 2856 3643 10 4599 10804 10 17610 53372 4
B3 18 2974 4027 10 2992 7589 6
Critical
- 5 bundle count data was used as Ro for FC Bank full in; suspect I bundle RO count is corrupted.
r.
Page 7 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 3, Calculated 1/M Values (no reflectors installed)
Total Bank full in Bank half out Bank full out Bundles 1/M FC 1/M AFC 1/M FC 1/M AFC 1/M FC 1/M AFC 1
1.000 1.000 1.000 1.000 1.000 1.000 2
3 N/A N/A N/A N/A N/A N/A 4
5 1.000*
0.797 0.931 0.550 0.931 0.539 6
N/A N/A N/A N/A N/A N/A 7
0.975 0.655 0.884 0.424 0.911 0.370 8
N/A N/A N/A N/A N/A N/A 9
0.972 0.428 0.844 0.236 0.714 0.172 10 0.901 0.330 0.801 0.182 0.667 0.126 11 0.854 0.311 0.752 0.160 0.598 0.114 12 N/A N/A N/A N/A N/A N/A 13 14 0.503 0.144 0.364 0.060 0.190 0.028 15 0.475 0.118 0.344 0.046 0.150 0.017 16 0.353 0.108 0.221 0.037 0.058 0.009 17 0.354 0.061 0.202 0.015 0.021 0.001 18 0.340 0.055 0.186 0.013 Critical
- 5 bundle count data was used as Ro for FC Bank full in; suspect I bundle Ro count is corrupted.
Page 8 of 63
UWNR LEU Conversion Reactor Start-Up Report IIM - Bank Full In (T-Rod Unlatched)
- FC AAFC 1 Oo 0.900 0,800 0.700 0
1'0.600 70.500 0
90.400 00 0.300 0.200 0.100 0.000
-- - *I
£ I
L
- i
'.. i A
.TU.
" " _ _L...........
2
_.* __Z_-7.L A-T--
T --- -- --
-T 7..
_2
,2 2 7 2 _
2.
.U 222 22 '
L *2 222 £/2
- T2T2T]
TTT TTTI TITTT.'T____-T--A----
U *T 0
1 2
3 4
5 6
7 8
9 10 11 12 13 14 Number of Bundles 15 16 17 18 19 20 21 22 23 24 25 Figure 2, Bank Full In 1/M Plot Page 9 of 63
UWNR LEU Conversion Reactor Start-Up Report I/M - Bank Half Out (9.00")
- FC AAFC 1.000 0.900 0.800 0.700 0
W0,600 E0.500 90.400 0.300 0.200 0.100 0.O00 7-----
AA 7.....
I.
,... A
.A 0
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 Number of Bundles 17 18 19 20 21 22 23 24 25 Figure 3, Bank Half Out I/M Plot Page 10 of 63
UWNR LEU Conversion Reactor Start-Up Report 11M - Bank Full Out UFC AAFC 1.000 0.900 0.800 0.700 10.600 S
- 0.500 0
0.400 0.300 0.200 0.100 0.000 AA 0
1 2
3 4
5 6
7 8
9 10 11 12 13 14 Number of Bundles 15 16 17 18 19 20 21 22 23 24 25 Figure 4, Bank Full Out I/M Plot The measured critical configuration is compared to the LEU conversion analysis prediction in Table 4.
Table 4, Predicted vs. Measured Critical Configuration LEU Predicted LEU Measured Critical bundles 19 18 Page 11 of 63
UWNR LEU Conversion Reactor Start-Up Report D.
LOADING TO J21-RO CORE Fuel loading continued utilizing inverse multiplication plots to assure each subsequent fuel bundle could be loaded safely. After each fuel addition absolute and technical specification shutdown margin measurements were made. The final unreflected 21 bundle core is shown in Figure 5. The subcritical count rate data and 1/M plot predictions are shown in Table 5 and Table 6.
BEAM PORT
- 3 A
_ B BEAM PORT G
E F
I 010 1 BEAM 3
Z<ý=_
4 5
6 Z
LL Lb H-FE.
LOG BEAM PORT 01 8
LEGEND
- = Fuel Bundle R = Reflector 9
T RANSIENT ROE Figure 5, J21-RO Core Map Table 5, Measured 1I/M Count Data Continued (no reflectors installed)
Bundle Total Bank full in Bank half out Added FC AFC Min FC AFC Min F7 19 3543 4347 10 3302 7295 5
F3 20 4328 5125 10 4504 9828 5
Page 12 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 6, Calculated 1/M Values Continued (no reflectors installed)
Total Bank full in Bank half out Bundles 1/M FC 1/M AFC 1/M FC 1/M AFC 19 0.285 0.051 0.141 0.011 20 0.233 0.043 0.103 0.009 Measurements of absolute and technical specification shutdown margin were made at each critical configuration, as indicated in Table 7. Reactivity was determined using the rod drop method.
Table 7, Rod Drop Measurements of Shutdown Margin (no reflectors installed)
Number Measured Bank Height Rod Drop True SDM Rod Drop T.S. SDM Bundles
[inches]
[%Ak/k]
[%Ak/k]
18 17.04
> 8.81" 5.26 19 14.83 8.41 4.40 20 13.66 8.43 3.36 21 12.83 8.08 3.19
- Insufficient excess reactivity to m'ake reactor critical without using a non scrammable control element After loading to the J21 -RO core, control element differential worths were measured using the rising period rod bump method. The measured data and worth curves are shown in Table 8 and Figure 6 through Figure 10, with the integral worths, shutdown margin and excess reactivity summarized in Table 9.
Page 13 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 8, J21-RO Differential Rod Worth Data Starting Ending Average Doubling Reactivity Differential Height Height Height Time Worth Worth
[inches]
[inches]
[inches]
[sec]
[%Ak/k]
[%Ak/k/in]
Blade 1 5.16 5.80 5.480 58.82 0.0785 0.1226 7.00 7.68 7.340 32.80 0.1190 0.1750 9.00 9.86 9.430 18.08 0.1722 0.2003 10.50 11.34 10.920 22.30 0.1522 0.1811 12.50 13.34 12.920 35.26 0.1134 0:1350 14.50 15.35 14.925 76.60 0.0639 0.0752 5.06 5.92 5.490 35.68 0.1125 0.1308 7.00 7.80 7.400 25.83 0.1390 0.1737 9.00 9.79 9.395 21.58 0.1552 0.1965 10.50 11.33 10.915 23.39 0.1478 0.1781 12.50 13.31 12.905 42.07 0.1003 0.1238 14.50 15.42 14.960 79.76 0.0619 0.0673 Blade 3 4.01 5.12 4.565 38.10 0.1075 0.0968 6.00 6.85 6.425 30.14 0.1259 0.1481 9.00 9.80 9.400 23.23 0.1485 0.1856 11.50 12.33 11.915 32.40 0.1200 0.1446 13.00 14.22 13.610 33.14 0.1182 0.0969 14.25 16.16 15.205 39.90 0.1041 0.0545 Reguilating Blade 1.50 3.43 2.465 343.72 0.0172 0.0089 5.00 6.86 5.930 105.85 0.0491 0.0264 8.00 8.87 8.435 204.11 0.0277 0.0319 11.00 12.08 11.54 210.13 0.0270 0.0250 13.50 15.16 14.33 288.24 0.0202 0.0122 Transient Rod Unlatched 10.64 N/A 9.88 0.2372 N/A 10.75 11.60 11.175 8.86 0.2499 0.2940 11.00 11.87 11.435 9.26 0.2448 0.2813 12.50 13.41 12.955 14.37 0.1957 0.2151 13.50-14.41 13.955 22.62 0.1509 0.1658 15.50 16.57 16.035 55.60 0.0819 0.0765 Page 14 of 63
UWNR LEU Conversion Reactor Start-Up Report Blade 1 Worth Curve - 09/23/2009 - J21-RO Equation of Fit: Differential Worth = cos^2[(pi/18.907874)*(x-9.407557)]*0.196548 R-Squared of Fit: 0.9965 Differential Fit N
Rising Period Data Points Integral 0.225 0.200 0,175 0.150 6 0.125
, '0.100 0.075 0.050 0,025 0.000 i-1.8 1.6 1.4 1.2 0.8 0.6 0.4 0.2 0
0 0
1 2
3 4
5 6
7 8
9 10 11 12 Length (Inches)
Figure 6, J21-RO Blade 1 Worth Curves Blade 2 Worth Curve - 09/23/2009 - J21-RO Equation of Fit: Differential Worth = cosA2[(pi/19.034443)* (x-9.173372)]*0.192884 R-Squared of Fit: 0.9950 0.225 0ý200 0.175 0.150 0.125 0,100 0.000 13 14 15 16 17 Differential Fit U Rising Period Data Points Integral 2
1.6 1.4 1.2 0.8 0.6 0.4 0.2 ii SR 0
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 Length (inches)
Figure 7, J21-RO Blade 2 Worth Curves Page 15 of 63
UWNR LEU Conversion Reactor Start-Up Report Blade 3 Worth Curve - 09/23/2009 - J21-RO Equation of Fit: Differential Worth = cos^2[(pi/18.908367)*(x-9.119543)]*0.183327 R-Squared of Fit: 0.9979 Differential Fit U
Rising Period Data Points Integral 0.200 0.175 0.150 0.125 0.100 0.075 0.050 0.025 0.000 1.8 1.6 1.4 1.2 0.8 0.6 0.4 0.2 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 Length (inches)
Figure 8, J21-RO Blade 3 Worth Curves Reg Blade Worth Curve - 09/24/2009 - J21-RO Equation of Fit: Differential Worth = cos^2[(pi/19.413792)* (x-8.665886)] *0.0319M5 R-Squared of Fit: 0.9985 Differential Fit N
Rising Period Data Points Integral 1 1 0.35 0.0350 0.0325 0.0300 0.0275 0.0250 0.0225 0.0200 0.0175 0.0150 0.0125 0.0100 0.0075 0.0050 0.0025 0.0000 0.3 0.25 0.2 0.15 10 11 12 13 14 15 16 17 0.05 0
1 2
3 4
5 6
7 a
9 Length (Inches)
Figure 9, J21-RO Regulating Blade Worth Curves Page 16 of 63
UWNR LEU Conversion Reactor Start-Up Report Transient Rod Worth Curve - 09/22/2009 -J21-RO Equation of Fit: Differential Worth = cos^2[(pi/25.435824)* (x-7.055959)] *0.384427 R-Squared of Fit: 0.9998 6C 0.350 0.325 0.300 0.275 0.250 0.225 01200 0.175 0.150 0.125 0.100 0.075 0.050 0.025 0.000 Differential Fit U Rising Period Data Points Integral 1.6 1,4 1.2 1
0.8 0.4 0.2 0
I I---
10 11 12 13 14 15 16 17 18 19 20 Length (inches)
Figure 10, J21-RO Transient Rod Worth Curves Table 9, J21-RO Control Element Calibration J21-RO Core Parameter Measured [%Ak/k]
Blade 1 worth (rod bump) 1.847 Blade 2 worth (rod bump) 1.827 Blade 3 worth (rod bump) 1.727 Regulating blade worth (rod bump) 0.308 Transient rod worth (rod bump) 1.390 Core excess reactivity 1.473 True shutdown margin (rod bump) 5.598 True shutdown margin (rod drop) 8.08 Technical specification shutdown margin (rod bump) 3.585 Technical specification shutdown margin (rod drop) 3.19 Based on the rod calibration data, the worth of each reflector position was measured by loading one reflector and noting the change in critical bank height. Reactivity worths are shown in Figure 11 for the unreflected configuration. During these measurements, some difficulty was encountered fully seating some reflector positions, and those remaining reflectors which had not been irradiated were shipped out to have the bases machined to a narrower width. In the meantime, reflector position B5 was chosen as a reference location to test the worth of several Page 17 of 63
UWNR LEU Conversion Reactor Start-Up Report reflectors due to its high reactivity worth. The testing demonstrated that not only were multiple new reflectors identical in worth, but they were also identical in worth to the old reflectors.
Therefore, reflector worth measurements resumed using reflector R 11 for the final five measurements.
A B
C D
E F
G 3
0.012 0.223 0.019 4
0.025 CD ca
.064 5
0.038 0.369
~0.419 0.064 6
0.032 V)M0.057 7
0.006
-~
0.237 0.025 Figure 11, Un-reflected Reflector Reactivity Worthis I%Ak/kI Page 18 of 63
UWNR LEU Conversion Reactor Start-Up Report E.
LOADING TO J21-R14 OPERATIONAL CORE After summing each individual reflector worth from Figure 11 to verify shutdown margin would meet requirements, reflectors were added in 4 steps, with a rod drop measurement of technical specification shutdown margin after each step. Reflector positions at each step are shown in through Figure 15. The measurements of technical specification shutdown margin at each step are summarized in Table 10. Old reflectors were used as place-holders for those new reflectors still being machined.
BEAM PORT
- 3 A
B C
BEAM PORT
- 4 D
E F
G Cý IQ BEAM PORT o2 Lj BEAM PORT 2
3 OIL 1 OON 5
6 Z
CD z
-9)
Ld FE LOG N CIE OlE 2 a
LEGEND
- = Fuel Bundle R = Reflector 9
TRANSIENT ROD Figure 12, J21-R4 Intermediate Core Map, T.S. SDM = 2.20 %Ak/k (rod drop)
Page 19 of 63
UWNR LEU Conversion Reactor Start-Up Report BEAM PORT
- 3 BEAM PORT
&4 A
G w
F-I 2
3 5
6 7
6 9
LEGEND
Page 20 of 63
UWNR LEU Conversion Reactor Start-Up Report BEAM PORT
- 3 A
B C
BEAM PORT t4 E
F G
Li 2
3 4
5 7
z 0L)
Fj OtE 1 LOG N CIE L z
E RIO.]
LEGEND
Page 21 of 63
UWNR LEU Conversion Reactor Start-Up Report BEAM B
EEAJM PORT PORT
- 3
- 4 A
I B
C D
E F
G
-]
Lw I
2 3
z 0Y LOG N C0C1 5
6 7
8 9
LEGEND
Table 10, SDM Measurements during Reflector Loading Core Loading Rod Drop Technical Specification Shutdown Margin
[%Ak/kl J21 -RO 3.19 J21 -R4 2.20 J21-R6 1.99 J21-RI0 1.62 J21-R14 1.39 Page 22 of 63
UWNR'LEU Conversion Reactor Start-Up Report The first J21-R14 core was loaded on September 30, 2009, with a technical specification shutdown margin of 1.390 %Ak/k measured with the rod drop method. The critical bank height was 10.19 inches at 100W. Control element differential worths were again determined using the rising period rod bump method on the J21-R14 core. The measured data and differential worth curves are shown in Table 11 and Figure 16 through Figure 20, with the integral worths, shutdown margin and excess reactivity summarized in Table 12.
Table 11, J21-R14 Differential Rod Worth Data Starting Ending Average Doubling Reactivity Differential Height Height Height Time Worth Worth
[inches]
[inches]
[inches]
[sec]
[%Ak/k]
[%Ak/k/in]
1.50 2.33 1.915 132.92 0.0405 0.0488 2.50 3.44 2.970 62.06 0.0753 0.0801 3.50 4.32 3.910 54.40 0.0832 0.1015 4.50 5.33 4.915 34.60 0.1148 0.1383 5.50 6.31 5.905 24.24 0.1446 0.1785 6.50 7.28 6.890 21.06 0.1575 0.2020 7.50 8.43 7.965 13.48 0.2026 0.2178 8.50 9.33 8.915 15.95 0.1849 0.2228 9.50 10.34 9.920 18.69 0.1690 0.2012 10.50 11.43 10.965 15.42 0.1884 0.2025 11.50 12.31 11.905 24.08 0.1452 0.1792 12.50 13.32 12.910 32.24 0.1204 0.1468 13.50 14.38 13.940 40.54 0.1030 0.1170 14.50 15.45 14.975 54.60 0.0830 0.0874 Blade2.
1.50 2.46 1.980 223.90 0.0255 0.0266 2.50 3.51 3.005 62.76 0.0747 0.0740 3.50 4.27 3.885 63.06 0.0744 0.0967 4.50 5.33 4.915 34.09 0.1160 0.1397 5.50 6.32 5.910 23.60 0.1470 0.1793 6.50 7.07 6.785 38.60 0.1065 0.1869 7.50 8.48 7.990 14.15 0.1974 0.2014 8.50 9.48 8.990 14.41 0.1954 0.1994 9.50 10.50 10.000 14.61 0.1940 0.1940 10.50 11.43 10.965 20.36 0.1607 0.1728 11.50 12.60 12.050 20.95 0.1580 0.1437 12.50 13.43 12.965 34.14 0.1159 0.1246 13.50 14.47 13.985 45.70 0.0945 0.0975 14.50 15.61 15.055 57.90 0.0794 0.0715 Page 23 of 63
UWNR LEU Conversion Reactor Start-Up Report Starting Ending Average Doubling Reactivity Differential Height Height Height Time Worth Worth
[inches]
[inches]
[inches]
[sec]
[%Ak/k]
[%Ak/k/in]
Blade 3 1.50 2.49 1.995 84.44 0.0591 0.0597 2.50 3.51 3.005 50.98 0.0873 0.0865 3.50 4.41 3.955 37.89 0.1079 0.1186 4.50 5.57 5.035 19.54 0.1646 0.1539 5.50 6.27 5.885 24.28 0.1444 0.1876 6.50 7.32 6.910 17.86 0.1734 0.2115 7.50 8.17 7.835 21.62 0.1551 0.2314 8.50 9.17 8.835 22.35 0.1520 0.2269 9.50 10.19 9.845 21.02 0.1577 0.2285 10.50 11.18 10.840 25.14 0.1413 0.2079 11.50 12.23 11.865 26.66 0.1362 0.1866 12.50 13.08 12.790 48.76 0.0902 0.1555 13.50 14.29 13.895 43.82 0.0974 0.1233 14.50 15.45 14.975 53.01 0.0848 0.0893 Regulating Blade 1.50 3.57 2.535 213.29 0.0267 0.0129 2.50 4.49 3.495 144.21 0.0377 0.0190 3.50 5.38 4.440 111.39 0.0471 0.0250 4.50 6.43 5.465 87.54 0.0574 0.0298 5.50 6.54 6.020 162.64 0.0340 0.0327 6.50 7.43 6.965 158.53 0.0347 0.0374 7.50 8.45 7.975 150.17 0.0364 0.0383 8.50 9.58 9.040 132.00 0.0407 0.0377 9.50 10.58 10.040 156.13 0.0352 0.0326 10.50 12.07 11.285 113.08 0.0465 0.0296 11.50 13.23 12.365 130.48 0.0411 0.0238 12.50 14.42 13.460 162.79 0.0339 0.0177 13.50 15.46 14.480 218.27 0.0261 0.0133 14.50 16.52 15.510 327.40 0.0180 0.0089 Transient Rod Unlatched 10.65 N/A 9.76 0.2387 N/A 10.75 11.50 11.125 13.01 0.2064 0.2752 11.50 12.19 11.845 17.74 0.1741 0.2524 12.50 13.15 12.825 26.85 0.1356 0.2086 13.50 14.22 13.860 32.81 0.1190 0.1653 14.50 15.42 14.960 37.55 0.1086 0.1180 15.50 17.34 16.420 31.47 0.1223 0.0665 17.50 19.14 18.320 160.79 0.0343 0.0209 Page 24 of 63
UWNR LEU Conversion Reactor Start-Up Report Blade 1 Worth Curve - 10/02/2009 - J21-R14 Equation of Fit: Differential Worth = cos^2[(pi/20.506812)*(x-9.026459)]*0.219009 R-Squared of Fit: 0.9897 Differential Fit N
Rising Period Data Points Integral 6e 0.250 0.225 0.200 0.175 0.150 0.125 0.100 0.075 0.050 0.025 0.000
~~~1~~~~
1.5 0
St C
2.5 2
t..
0.5 0
1 2
3 4
5 6
7 8
9 10 11 12 Length (inches)
Figure 16, J21-R14 Blade I Worth Curves Blade 2 Worth Curve - 10/02/2009 - J21-R14 Equation of Fit: Differential Worth = cos^2[(pi/19.61376S)*(x-8.793064)* 0.203733 R-Squared of Fit: 0.9766 0.225 0.2 0 0 0.175 0.15o0
" i 0.125 0.1000 0.075 0.025
+-
0.0 0 0 13 14 15 16 17 Differential Fit U Rising Period Data Points Integral St 2.5 2
1.5
'11 0.5 0
0 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 Length (inches)
Figure 17, J21-R14 Blade 2 Worth Curves Page 25 of 63
UWNR LEU Conversion Reactor Start-Up Report Blade 3 Worth Curve - 10/05/2009 - J21-R14 Equation of Fit: Differential Worth = cos^2[(pi/20.645064)* (x-8.931339)]*0.230594 R-Squared of Fit: 0.9956
-Differential Fit 8
Rising Period Data Points Integral 2.5 0.250 0.225 0.200 0.175 1.2 0.150
.6 8 0.125 0.100 0.075 0.050 0,025 0.000 0I 1
0.5 0
0 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 Length (Inches)
Figure 18, J21-R14 Blade 3 Worth Curves Reg Blade Worth Curve - 10/06/2009 -J21-R14 Equation of Fit: Differential Worth = cosA2[(pi/20.010091)*(x-8.378403)] *0.037422 R-Squared of Fit: 0.9870 6
OR 0.0425 0.0400 0.0375 0.0350 0,0325 0.0300 0.0275 0.0250 0,0225 0.0200 0.0175 0.0150 0.0125 0.0100 0.0075 0.0050 0.0025 0.0000 Differential Fit 0
Rising Period Data Points Integral 0.4 0.35 0.25 0.2 0.15 0.1 0,05 0
0 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 Length (Inches)
Figure 19, J21-R14 Regulating Blade Worth Curves Page 26 of 63 17
UWNR LEU Conversion Reactor Start-Up Report Transient Rod Worth Curve-10/01/2009 -J21-R14 Equation of Fit: Differential Worth = cos^2[(pi/28.432378)* (x-6.117435)]*0.383247 R-Squared of Fit: 0.9995 Differential Fit N
Rising Period Data Points Integral 0.325 0.300 0.275 0.250 0.225 0.200 0.175 0.150 0.125 0.100 0.075 0.050 1.6 1.A 1.2 1
0.8 6 1
0.A
-0 OA 0.2 0
0.025 0.000 10 11 12 13 14 15 16 17 18 Length (Indies) 19 20 Figure 20, J21-R14 Transient Rod Worth Curves Table 12, J21-R14 Control Element Calibration J21-R14 Core Parameter Measured [%Ak/k]
Blade 1 Worth (rod bump) 2.221 Blade 2 Worth (rod bump) 1.989 Blade 3 Worth (rod bump) 2.355 Regulating Blade Worth (rod bump) 0.372 Transient Rod Worth (rod bump) 1.363 Core Excess Reactivity (rod bump) 3.695 True Shutdown Margin (rod bump) 4.605 Technical Specification Shutdown Margin (rod bump) 1.779 Technical Specification Shutdown Margin (rod drop) 1.390 The hydraulic (whale) tube irradiation facilities were installed in positions C2, C8, and E8 and the critical bank height measured as 10.27 inches, a 0.08 inch increase corresponding to a reactivity worth of -0.052 %Ak/k. After receiving the re-machined reflectors, they were swapped with the old reflectors which had been loaded as place-holders. The critical bank height was measured as 10.26 inches, a 0.01 inch decrease corresponding to a reactivity worth of
+0.006 %Ak/k. It was unclear whether some of the new reflectors were fully seated, so the underwater camera was used to view the installed reflectors from the side. The reflector in Page 27 of 63
UWNR LEU Conversion Reactor Start-Up Report position A6 was sticking up about 1/8 of an inch and was pushed down until it was fully seated.
It was removed and reseated several times to verify that it did not get stuck. The critical bank height was then measured as 10.24 inches on October 9, 2009, a 0.02 inch. decrease corresponding to a reactivity worth of +0.013 %Ak/k. Therefore the net change from the initial J21 -R14 core at a bank height of 10.19 inches (before adding whale tubes) and the final J21 -R14 core at a bank height of 10.24 inches was a 0.05 inch increase corresponding to a reactivity worth of only -0.033 %Ak/k as a result of the displacement of water with aluminum structure.
Because of the small reactivity change observed, the control element calibrations for the J2 1-RI 4 core were not repeated after installation of the whale tubes.
Table 13, Predicted vs. Measured Cold Bank Height LEU Predicted LEU Measured Cold critical bank height [inches]
10.13 10.24 Page 28 of 63
UWNR LEU Conversion Reactor Start-Up Report F.
OPERATIONAL CORE REACTIVITY MEASUREMENTS F.1 REFLECTOR REACTIVITIES (REFLECTED)
Reflectors were removed one at a time from the reflected core (making a J21 core with only 13 reflectors) to measure the reflected reactivity worths. Measured reactivities are shown in Figure 21, with the unreflected worths duplicated from Figure 11.
A B
C D
E F
G 3R=0.025 R=019 R0.72 "U=0.012 U=0223U=.019 R=O.0o 44 4)
'a 1
1 7 R=0.052 R=0347R=049 R=Q. 150 6 R=.046 UQo3 R=O. 125 Figure 21, Reflector Reactivity Worths [%Ak/k, R=Reflected, U=Unreflectedi Page 29 of 63
UWNR LEU Conversion Reactor Start-Up Report F.2 COOLANT VOID COEFFICIENT Measurements were made of the coolant void coefficient of reactivity following the procedure in UWNR 002. A total of 19 locations were used to insert aluminum wands which could be filled with water or air in order to make the measurement, but only 6 wands were ever inserted at a time to map each quadrant of the core. Refer to Figure 22. Because 3 of the locations were duplicated as each core quadrant was measured, this led to a total of 22 wands used.
Figure 22, Air and Water Wand Locations Page 30 of 63
UWNR LEU Conversion Reactor Start-Up Report Static reactivity measurements were made by measuring the critical control element heights with the insertion of water filled and air filled wands. The table below shows the recorded blade heights and calculated reactivity insertion (from all blades in) based on control element calibrations.
Table 14, Water and Air Wand Reactivities Configuration Blade 1 Blade 2 Blade 3 Reg. Blade T-rod Reactivity Height Height Height Height Height Inserted
[inches]
[inches]
[inches]
[inches]
[inches]
[%Ak/k]
S WaeWad 10.34 10.34 10.34 10.37 106 4.705 SW Air Wand 10.61 10.61 10.61 10.77 10.66 4S88 NW Water Wand 10.34 10.34 10.34 10.24 10.66 4.700 NW Air Wand 10.61 10.61 10.61 10.08 10.66 4.861 NE Water Wand 10.34 10.34 10.34 9.43 10.65 4.672 NE Air Wand 10.61 10.61 10.61 7.41 10.66 4.763 SE Water Wand 10.34 10.34 10.34 9.51 10.65 4.674 SE Air Wand 10.61 10.61 10.61+
7.46 10.66 4.764 The measured reactivity change in each quadrant was calculated by subtracting the reactivity insertion with air wands from the reactivity insertion with water wands. The % void fraction was calculated according to the UWNR 002 based on the number of wands used in each quadrant, the cross-sectional area of void in each wand, and the core-wide coolant cross-sectional area. The void wands are sufficiently long in the vertical direction to cover the entire region of neutron importance, so the calculated % void fraction in terms of cross-sectional areas is equivalent to the % void fraction in terms of void volume. Results for each core quadrant are shown in Table 15.
Table 15, Void Reactivity Measurements Reactivity Reactivity Local Void Configuration Inserted Change
% Void Coefficient JAk/kj
[%Ak/k
[%Ak/k/%void]
SW Wate Wad 4.0-.79 0.9667
-0.185 SW Ater Wand 4.884 NW Water Wand 4.700
-0.161 0.9667
-0.167 NW Air Wand 4.861 SE Water Wand 4.674 SE War Wand 4.674
-0.090 0.8056
-0.112 SE Air Wand 4.764 Page 31 of 63
UWNR LEU Conversion Reactor Start-Up Report The average coolant void coefficient of reactivity is therefore -0.14 %Ak/k/%void. Note that the measurements for the SW and NW quadrants included a wand located between D5 and D6 adjacent to the center bundle, whereas the measurements for the NE and SE quadrants could not locate a wand adjacent to the center bundle (between D4 and D5) due to interference from the transient rod guide tube. The measured reactivity change for the west quadrants was noticeably larger than the east. Although the % void fraction does account for some of the difference, the worth of the missing wand location adjacent to the center bundle is higher than the average wand location, which explains why there is still a difference in the calculated local void coefficients.
Table 16, Predicted vs. Measured Void Coefficient LEU Predicted LEU Measured Void coefficient [%Ak/k/%void]
-0.149
-0.14 F.3 EXPERIMENTAL FACILITY REACTIVITY MEASUREMENTS Sample reactivity measurements were performed for the whale tube and pneumatic tube.
experimental facilities. See Table 17. The pneumatic tube was found to have very little reactivity worth with all changes well within the capability of the reactivity control system. The whale tube had higher worth as expected, but measured values were still small enough for reactivity control. Measurements were performed in whale tube E8, the highest worth location as shown in section J of this report. A number of cadmium samples weretested with varying thicknesses of cadmium wrapped along the interior of the whale bottle. The first sample had a thickness of 1 mm and a total mass of 86g cadmium, and the heaviest had a thickness of 5mm and a total mass of 372g cadmium. There was no noticeable change in bank height between samples, but some variability is introduced by having the diffuser and whale pumps running. In order to observe fine differences in bank height, the lighter cadmium samples which did not sink were loaded with small amounts of lead in order to make them sink so bank heights could be measured with the diffuser and whale pumps off. Because of the low reactivity worth of the large lead sample measured, the small amounts of lead added to the cadmium whales were neglected in calculating the reactivity worth. Even with the'diffuser and whale pumps off, there was no difference in bank heights between cadmium samples, demonstrating that the first 1mm layer of cadmium is essentially black to thermal neutrons.
Page 32 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 17, Predicted vs. Measured Experimental Facility Reactivities LEU LEU Predicted Measured
[%Ak/k]
[%Ak/k]
Pneumatlc Tube 4-30ml water (flooding)
+0.007
+0.004 75g B4C N/A
-0.007 148g Cd
-0.009
-0.009 i W hale Tub~e E8:
+ +++..,,
Air N/A
-0.030 237g Graphite N/A
+0.011 1464g Lead N/A
+0.003 86-372g Cd
-0.04
-0.087 Page 33 of 63
UWNR LEU Conversion Reactor Start-Up Report G.
POWER CALIBRATIONS The first power calibration by calorimetric heat balance was performed at 500kW as indicated by power level instrumentation, followed by a power calibration at 900kW, and finally a full power calibration at 1.00MW. Because of the results of the earlier heat balances, power level instruments werenot adjusted to agree with the measurements until after the 1.00MW power calibration. Measured power levels were reasonably close and lower than the indicated power levels. See Table 18.
Table 18, Power Calibration Data Date Indicated Power Actual Power 10/20/2009 500 kW 486 kW 10/26/2009 900 kW 883 kW 11/2/2009 1000 kW 979 kW During all three power calibrations, water and air samples were taken to look for any leaking fission products using a high purity germanium detector. All air particulate samples contained Ra-224, Bi-214, and Pb-214, all naturally occurring radon decay products which are routinely detected. No other isotopes were found. All water samples contained Mg-27, Na-24, and Ar-41.
Na-24 and Ar-41 are routinely detected in water samples taken at or recently after full power operations. Na-24 is produced from an n,u fast activation reaction from aluminum in the core grid box and end fittings. Ar-41 is produced from argon in the air dissolved in the pool water as it passes through the core. Mg-27 is not routinely detected, but is also produced from an n,p fast activation reaction from aluminum. It is not common to collect and count water samples during full power operations, which is why the short-lived Mg-27 is not routinely detected (half-life of 9.5 minutes). No fission products were detected in air or water samples, and no unusually high stack gas or continuous air monitor levels were observed.
After power calibrations and nuclear instrumentation adjustments were complete, the full power bank height was measured to be 12.58 inches.
Table 19, Predicted vs. Measured Hot Bank Height LEU Predicted LEU Measured Bank Height Bank Height
[inches]
[inches]
Hot critical bank height 11.73 12.58 Page 34 of 63
UWNR LEU Conversion Reactor Start-Up Report H.
POWER DEFECT MEASUREMENTS Power defect measurements were taken to record fuel temperature and excess reactivity vs.
power level. These measurements were done in three separate days, each day beginning with a cold clean core (free of xenon)..The three measurements were 0-50% in 50kW increments, 50-100% in 50kW increments, and 0-100% in 100kW increments. The raw data is shown in Table
- 20. The fuel temperature vs. power is shown in Figure 23 for bundle 62 and Figure 24 for bundle 72. By using the recorded bank heights along with the J21-R14 control element calibration curves, the excess reactivity vs. power was calculated and is shown in Figure 25. The maximum reading thermocouple, bundle 62 center, was also used to plot fuel temperature vs.
excess reactivity in Figure 26. The data from all three days shows good agreement indicating that no significant amount of xenon had built up during each day's measurements.
Page 35 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 20, Temperature vs. Power Measurements Power Level_[kW Bank Bundle 62 D4 SW [°C]
Bundle 72 E3 NE ['C]
kW pA p
LogN Height 1
k #1
- 2 L
in Bottom Center Top Bottom Center Top 0-'50% in 50kW increments '.,.
<1 0.30 0.32 0.40 10.36 50 47 50 50 10.64 71.2 70.2 67.7 57.1 56.7 54.1 100 98 100 90 10.82 107.3 106.2 101.0 88.3 88.3 83.9 150 150 150 160 10.98 140.9 139.8 132.6 115.8 115.6 108.6 200 200 200 210 11.13 167.3 166.2 156.8 137.9 137.4 129.8 250 240 250
- 280, 11.26 181.8 178.7 167.4 158.6 157.8 148.9 300 290 300 300 11.36 195.3 193.1 181.6 178.4 177.8 167.8 350 340 350 360 11.44 201.8 201.0 189.4 193.9 191.5 180.9 400 380 400 400 11.52 209.6 211.7 199.6 208.3 205.8 195.2 450 430 450 450 11.60 221.2 224.4 213.3 220.9 218.0 207.3 500 470 500 500 11.67 231.4 235.6 224.2 230.7 227.3 216.4 50-100% in 50kW increments
<1 0.28 0.32 0.40 10.39 25.9 25.8 25.8 25.3 25.2 25.2 500 470 500 500 11.66 226.0 229.9 220.2 222.2 224.8 215.8 550 520 550 550 11.74 235.9 241.9 232.5 232.6 233.8 225.0 600 570 600 600 11.82 246.5 253.8 245.2 239.3 240.5 233.7 650 620 650 650 11.89 255.9 265.8 257.6 247.4 248.7 242.1 700 670 700 700 11.96 265.3 277.8 270.7 252.5 257.1 252.9 750 730 750 800 12.04 274.4 288.4 281.4 262.4 268.6 263.9 800 780 800 850 12.12 282.6 298.7 291.8 272.0 278.2 273.7 850 840 850 900 12.21 293.4 311.1 303.3 285.2 291.9 286.0 900 890 900 950 12.30 303.0 322.3 314.4 297.6 303.0 296.0 950 940 950 950 12.38 312.4 333.2 325.0 309.7 314.3 306.1 1000 1000 1000 1000 12.47 320.6 342.6 334.1 318.6 323.4 314.2 0-100%inOOkWincrements
<1 0.36 0.39 0.50 10.37 25.8 25.7 25.6 25.2 24.9 25.0 100 100 100 100 10.82 112.7 111.2 105.4 88.2 89.6 87.7 200 190 200 200 11.11 165.8 162.4 152.4 131.0 134.0 131.6 300 290 300 300 11.35 194.4 190.1 178.4 170.6 174.6 169.3 400 380 400 400 11.51 207.3 206.3 196.3 200.0 204.3 196.2 500 470 500 500 11.66 225.2 231.2 221.8 223.6 225.5 216.0 600 570 600 600 11.81 247.2 255.1 246.9 240.7 242.8 234.2 700 670 700 700 11.96 265.7 277.7 270.1 252.7 256.5 251.7 800 780 800 850 12.11 283.0 298.9 291.4 272.1 277.6 273.1 900 890 900 950 12.29 303.4 322.6 314.1 295.4 301.4 294.7 1000 1000 1000 1000 12.47 320.2 342.8 334.4 318.2 322.3 312.7 Page 36 of 63
UWNR LEU Conversion Reactor Start-Up Report 350 U
300 U
250 C.
E i* 200 -
L-a 150 0
(4)00-50%
0100 050-100%
CO 0-100%
50 0
200 400 600 800 1000 Power (kW)
Figure 23, Bundle 62 Fuel Temperature vs. Power 350 300 U
U Z-250 E
200 LL*
0-50%
- 6 -150 M50-100%
A 0-100%
o100 04 50 0
0 200 400 600 800 1000 Power (kW)
Figure 24, Bundle 72 Fuel Temperature vs. Power Page 37 of 63
UWNR LEU Conversion Reactor Start-Up Report 4
3.5 3
,0-50%
- *050-o 100%
0 2.5 0
2 1.5 I
0 200 400 Power (kW) 600 800 1000 Figure 25, Excess Reactivity vs. Power 4
3.5 3
2.5 1U 1.5 0
50 100 150 200 250 300 Bundle 62 Center Fuel Temp (C)
Figure 26, Excess Reactivity vs. Bundle 62 Fuel Temperature Page 38 of 63 350
UWNR LEU Conversion Reactor Start-Up Report The linear trend of excess reactivity vs. power does not begin until power is greater than 350kW.
Using the data greater than 350kW, the power defect is calculated to be -1.22x 10-3 %Ak/k/kW with an R2 of 1.000, and the fuel temperature coefficient is calculated to be -5.55x 10-3 %Ak/k/°C with an R2 of 0.998 (using the highest reading thermocouple in bundle 62 center).
Table 21, Predicted vs. Measured Temperature Coefficients LEU Predicted LEU Measured Power defect N/A
-1.22x 10-3
[%Ak/k/kW]
Fuel temperature
-4.08x103 (300-400K)
-5 5 5 10-3 coefficient [%Ak/k/°C]
-6.58x 103 (400-600K)
.55x Page 39 of 63
UWNR LEU Conversion Reactor Start-Up Report I.
FUEL TEMPERATURE MAPPING The IFE in bundle 62 was moved into every available core location in order to map the temperature across the core. Only the D5 location was not mapped because of the transient rod bundle. The IFE in bundle 72 was also moved into 5 different locations to compare to the bundle 62 data. All measurements were performed with the thermocouple facing east for consistency.
Figure 27 through Figure 29 show the temperature distributions across the core for top, center, and bottom thermocouples in bundle 62. Table 22 compares measured temperatures between bundle 72 and 62.
Table 22, Fuel Temperature Mapping Data, Bundle 72 vs. Bundle 62 Position Bundle 72 [ C]
Bundle 62 10C]
Difference 72-62 [1C]
Top Center Bottom Top Center Bottom Top Center Bottom D4 SW 388.3 405.3 402.2 334.6 342.3 318.9 53.7 63.0 83.3 D4 SE 376.0 391.1 388.8 320.9 324.4 305.4 55.1 66.7 83.4 D4 NE 377.2 391.1 388.4 317.9 320.9 302.0 59.3 70.2 86.4 D4 NW 390.9 410.9 410.7 325.1 330.6 309.1 65.8 80.3 101.6 E3 NE 303.3 318.6 317.4 265.0 269.1 257.2 38.3 49.5 60.2 I
a I
I t,
I 3
Key IFE Temp Rod PPF Power Peaking ad*nr 4
5 1
6 306i.1 046316
.1J320.6 315.47 1.01107 1.2231.233 1.155 284.7l 7
Figure 27, Bundle 62 Top Measured IFE Temperature Distribution at 1.0MW Page 40 of 63
UWNR LEU Conversion Reactor Start-Up Report I
a I I 3
Rod PPF Power Peaking Factors 4
1307.3 295.71320.91 324.4 I1.211 1.217 5
1 6
3I22..7 324.4 1.230 1.246 313.7 309.2 319.3 322.8 317.9 1.101 1.107 1.223 1.233 1.155 286.4 1.006 7
Figure 28, Bundle 62 Center Measured IFE Temperature Distribution at 1.0MW Key IFE Temp Rod PPF 3
Power a
Peaking I=ne frtaI 4
5 6
7 Figure 29, Bundle 62 Bottom Measured IFE Temperature Distribution at 1.0 MW Page 41 of 63
UWNR LEU Conversion Reactor Start-Up Report The graph of center thermocouple temperature vs. pin power peaking factor is shown below, with the expected linear fit.
380 360 340
,- 320 E 300 0
I-,
280 t 260 C 240 N
220 200 180 U_
U~
Tem 16.03PPF+ 17.1 R1
~
=.95 E
IF Uil MU 0.4 0.6 0.8 1
1.2 1.4 1.6 MCNP calculated pin power peaking factor Figure 30, Bundle 62 Center Fuel Temperature vs. Pin PPF During measurements, a typical variability of 1 'C was observed during a single full power run, while the variability in the same location during multiple full power runs was as much as 5 'C.
Furthermore, in two locations the IFE was rotated to see how the observed temperature varied with thermocouple direction. One location was near the center of the core where the flux gradient is relatively flat, and rotating the IFE varied the temperature by less than 5 'C. The second location was on the periphery of the core where the flux gradient is large, and rotating the IFE varied the temperature by as much as 15 'C. In addition to the variability in IFE thermocouple readings, the pin power peaking factors were calculated using MCNP for the entire pin, rather than the local power being produced at the thermocouple location which also accounts for some of the scatter observed in Figure 30.
For the five locations in which both IFEs were measured, the IFE in bundle 72 measured higher than the IFE in bundle 62 by 68 'C on average, with a scatter of+/- 17 'C. See Table 22. It is likely that this discrepancy is due to differences in fuel swelling and contact with the thermocouples within each IFE. It is also possible that the difference in measured temperatures was due to unequal burnup and samarium buildup between IFE bundles. The critical bank height during temperature mapping ranged from 12.30 to 12.69 inches as fuel was shuffled, even when Page 42 of 63
UWNR LEU Conversion Reactor Start-Up Report there was no xenon in the core. Although the temperature mapping was an attempt at measuring BOL conditions, there was a total of 68.87 MWh of core exposure when temperature mapping began due to previous power calibrations and other required operations, including steady-state operation to verify no trending in IFE temperatures.
Table 23, Predicted vs. Measured IFE Temperatures LEU Predicted LEU Measured Bundle 62 Bundle 72 D4 SW Temp (C) 391-480 342.3 405.3 E3 NE Temp (°C) 264-317 269.1 318.6 Page 43 of 63
UWNR LEU Conversion Reactor Start-Up Report J.
FLUX MEASUREMENTS The neutron flux was measured in most of the irradiation facilities and in the core using neutron activation of samples. For the experimental facilities, multiple irradiations of gold, gold wrapped in cadmium, and nickel were used to calculate thermal, epithermal, and fast neutron flux. The results are given below.
Table 24, Experimental Facility Flux Measurements LEU Measured 2/secl In/cm.
Ic Whale C8 Thermal 1.20x10' Epithermal 6.10xlO" FnAt
') ARx 1 n"12 Thermal Pneumatic Tube Epithermal Fast Thermal Columin Thermal
-F
~4.14D A I V 1.34xlU i*!iiiii!!i For the in-core axial flux profile, it was desired to use a miniature fission chamber probe which could be raised and lowered inside a fuel coolant channel while at full power, but the probe failed and could not be readily repaired. Instead of using the miniature fission chamber, a gold wire attached to an aluminum rod was activated in the coolant channel following the UWNR 002 procedure. The procedure limits maximum reactor power to 1 kW. The rod was positioned between D5 and D6, the hottest available position. After irradiation, the gold wire was cut into 1/2 inch long segments and counted individually in order to construct the axial flux profile, shown in Figure 31. During measurement, the control elements were at a bank height of 10.42 inches compared to the full power bank height of 12.58 inches.
Page 44 of 63
UWNR LEU Conversion Reactor Start-Up Report 7E+09 6E+09 5E+09 E
U.
4E+09 3E+09 2E+09 1E+09 0 0 5
10 15 Axial Position above Bottom of Lower Reflector (inches)
Figure 31, LEU Axial Flux Profile 20 Page 45 of 63
UWNR LEU Conversion Reactor Start-Up Report K.;
PULSE TESTING Pulse testing began after 79.73 MWh of core exposure. Pulses ranged from the minimum operationally allowed 1.000 %Ak/k to the maximum achieved of 1.363 %Ak/k insertion. LEU measured pulse data is shown in Table 25. Figure 32 shows peak fuel temperature vs. reactivity insertion, Figure 33 shows integrated pulse power (energy) vs. reactivity insertion, and Figure 34 shows peak power vs. the square of reactivity insertion. All graphs show the expected linear trend, except for some of the larger pulses which may deviate from linearity due to slow transient rod ejection times with respect to pulse width. The graph of fuel temperature shows that the first 1.2 %Ak/k pulse data point is well above the linear fit, therefore this point was re-taken with a much better result. The initial outlier point is assumed to be attributed to electrical noise which was not reproducible.
During pulse testing, no increased stack or continuous air monitor readings were detected, and routine air and pool water samples continue to indicate no fission product release.
Table 25, LEU Pulse Data Reactivity Peak Peak Fuel Integrated FWHM Period Insertion Power Temperature Pulse Power FM Peio
(%Ak/k) AMW)
° (ms)
(mns) 1.000 80.26 186 2.77 39.30 9.81 1.026 89.72 185 3.35 36.17 8.16 1.050 144.75 193 4.21 29.03 7.81 1.075 159.99 197 4.52 26.12 6.71 1.100 206.89 206 5.32, 22.83 5.66 1.125 229.06 206 5.55 21.56 5.60 1.150 280.87 217 6.28 19.87 5.19 1.150 281.93 216.
6.19 19.77 5.22 1.175 306.98 217 6.36 18.91 4.32 1.200 361.59 277 7.24 17.55 3.52 1.200 345.87 221 6.96 17.42 4.63 1.225 385.26 227 7.39 17.05 3.96 1.250 435.92 236 7.97 16.20 4.23 1.275 470.85 235 8.35 15.49 4.04 1.300 496.00 239
'8.63 15.28 4.01 1.325 551.42 243 9.03 14.54 4.51 1.350 589.16 251 9.58 14.30 3.81 1.363 594.41 251 9.67 14.24 3.36 1.363 595.16 250 9.59 14.25 3.83 Page 46 of 63
UWNR LEU Conversion Reactor Start-Up Report The LEU peak powers are noticeably smaller than for the previous HEU core, which can be explained by the reduced total core heat capacity due to the reduction in fuel pins.
Peak Fuel Temperature vs. Pulse Reactivity C
0 43
'U I-aE 43 U-
'U 43 290 270 250 230 210 190 170 1
1.05 1.1 1.15 1.2 1.25 1.3 1.35 Reactivity Insertion (%Ak/k) 1.4 Figure 32, Peak Fuel Temperature vs. Pulse Reactivity The linear fit equation shown in Figure 32 was made after excluding the outlier point at 1.2
%Ak/k.
Page 47 of 63
UWNR LEU Conversion Reactor Start-Up Report Integrated Power vs. Pulse Reactivity I..
0 a-
'U 1~
U, C
12.00 10.00 8.00 6.00 4.00 2.00 0.00 y = 18.309x - 15.088 R' = 0.9892 A
1 1.05 1.1 1.15 1.2 1.25 1.3 1.35 Reactivity Insertion (%Ak/k)
Figure 33, Integrated Power vs. Pulse Reactivity 1.4 Page 48 of 63
UWNR LEU Conversion Reactor Start-Up Report Peak Power vs. Pulse Reactivity Squared 700 y =6E+06x - 546.65 R2 = 0.9971 600 500I I-0I 0
a.
(U a.
400 300 200 100 0
1.OE-04 1.1E-04 1.2E-04 1.3E-04 1.4E-04 1.5E-04 1.6E-04 1.7E-04 1.8E-04 1.9E-04 2.OE-04 Reactivity Insertion Squared ([Ak/k]2 )
Figure 34, Peak Power vs. Pulse Reactivity Squared The value of 13, *p, and a can be estimated from the measured data using several methods. First, from reactor kinetics the following equation relates P3 and tp to the measured period.
_ ep p-fl or
_1 _p-f/?
P.
T ep Where:T = period [ms]
p = pulse reactivity insertion [Ak/k]
03 = delayed neutron fraction p= prompt neutron lifetime [ms]
o= inverse period [ms-1]
Therefore, by plotting the inverse period vs. reactivity insertion the slope of the line is the inverse of the prompt neutron lifetime and the y-intercept is related to the delayed neutron fraction. This is shown in Figure 35 where the R2 value is 0.82, and the calculated values of Ep = 22.2ps and P3
= 0.0072.
Page 49 of 63
UWNR LEU Conversion Reactor Start-Up Report 0.35 0.30 0.25 r-0.20
- a. 0.15 0
0
.=
0.10 tp=22.2us y = 44.98x - 0.3263 03=0.0072 R2 = 0.8258 0.05 0.00
- 1. 00%
1.05%
1.10%
1.15%
1.20%
1.25%
Reactivity [%Ak/k]
1.30%
1.35%
1.40%
Figure 35, Inverse Period vs. Reactivity Alternatively, by applying the Fuchs Nordheim model for reactivity feedback the pulse full width at half maximum (FWHM) is related to the reactivity insertion according to the following equation. This is shown in Figure 35 where the W2 value is 0.97, and the calculated values of fp = 24.4gts and P = 0.0074.
1 p
ft fl 4 cosh-1 F2-ep 4 cosh-1 -vfep Where: p = pulse reactivity insertion [Ak/k]
03 = delayed neutron fraction fp = prompt neutron lifetime [ms]
= FWHM [ms]
Page 50 of 63
UWNR LEU Conversion Reactor Start-Up Report LOr4)
C 0.080 tp=24.4us Y 11.619x - 0.0858 0.070 P=0.0074 R= 0.968 0.060 0.050 0.040 0.030 0.020 0.010 0.000 1.00%
1.05%
1.10%
1.15%
1.20%
1.25%
1.30%
1.35%
1.40%
Reactivity [%Ak/k]
Figure 40, Inverse Full Width at Half Maximum vs Reactivity Insertion The values of 03 and Cp, as determined both from reactor kinetics and the Fuchs Nordheim model, were averaged and summarized below in Table 26.
Table 26, Predicted vs. Measured P* and tp LEU Predicted LEU Measured Delayed neutron fraction P3 0.0078 0.0073 Prompt neutron lifetime t.
27.1 tsec 23.3 sec Page 51 of 63
UWNR LEU Conversion Reactor Start-Up Report L.
SQUARE-WAVE TESTING Performing a square-wave operation involves firing the transient rod while at low power to insert a large amount of reactivity, but less than beta, to allow the power to quickly rise to its peak value, then withdrawing control elements in order to maintain the power level. Before performing a square-wave, the relation. of peak power vs. transient rod height must first be measured to allow square-waving to the desired power. This is done by firing the transient rod at a variety of heights and observing the peak power achieved, without withdrawing any control elements. Therefore power is allowed to decrease after reaching the peak and then the reactor is shutdown. The data is in the table below, followed by the square-wave curve to be used for future square-wave operations.
Table 27, Square-wave Data T-rod Height (inches)
Peak Power (kW) 10.65 66 10.90 110 11.25 222 11.75 556 11.95 840 12.03 962 12.05 971 Page 52 of 63
UWNR LEU Conversion Reactor Start-Up Report 100 pA of 1MW =
2 90 80 (5 70 S60 CO 50 40 v o-
&. 20 10 10 10.50 11.00 11.50 12.00 12.50 Transient Rod Position (inches)
Figure 36, Square-wave Curve Page 53 of 63
UWNR LEU Conversion Reactor Start-Up Report M.
SUMMARY
OF HEU MEASUREMENTS VS. LEU PREDICTIONS AND MEASUREMENTS Table 28 is a summary of all predicted vs. measured values reported earlier in this report.
Comparisons are also made to measured values on the HEU core, where available. Fluxes are compared in Table 29 and Table 30.
Table 28, Predicted vs. Measured Values HEU Measured LEU Predicted LEU Measured Critical bundles 18 19 18 Cold critical bank height 9.50 10.13 10.24
[inches]
Void coefficient
[%Ak/koperivoid]-0.2
-0.149
-0.14
[%Ak/k per %void]
Flooding pn tube [%Ak/k]
+0.0002
ý.+0.007
+0.004 B 4C in pn tube [%Ak/k]
N/A N/A
-0.007 Cd in pn tube [%Ak/k]
-0.0005
-0.009
-0.009 Air whale in E8 [%Ak/k]
N/A N/A
-0.030 Graphite whale in E8 [%Ak/k]
N/A N/A
+0.011 Lead whale in E8 [%Ak/k]
N/A N/A
+0.003 Cd whale in E8 [%Ak/k]
N/A
-0.04
-0.087 Hot critical bank height 11.50 11.73 12.58
[inches]
Power coefficient
[%k//k]N/A N/A
-1.22E-3
[%Ak/k/kW]
Fuel temperature coefficient 126E2
-4.08E-3 (300-400K) 5.5E-3
[%Ak/k/°C]
-6.58E-3 (400-600K)
D4 SW Temp [°C]
405 (Bundle 41) 391-480 342.3 (Bundle 62) 492 (Bundle 42) 405.3 (Bundle 72)
E3 NE Temp [°C]
277 (Bundle 41) 264-317 269.1 (Bundle 62) 375 (Bundle 42) 318.6 (Bundle 72)
Delayed neutron fraction 03 0.0070 0.0078 0.0073 Prompt neutron lifetime Cp I
22gsec 27.1 gsec 23.3gsec Page 54 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 29, Flux Comparison HEU vs. LEU TIITi'T T 1,A" I l 1.20x 10"'
6.10xlO" Whale C8 Epithermal 3.11xx10" I
t Fast 2.34xxlO"L 2.68xlOl" Thermal 8.06 x10'-
1.72 x 10 Fs 2.68x4O'2 2.89xll)'2 Thermal 3.32x 1012 4.45 x 1012 Pneumatic Tube Epithermal 1.23x10"l 1.34x10I" Fast 5.64xl0" 7.28x0I" The LEU Conversion SAR estimated the change in flux by calculating the ratio of LEU flux to HEU flux values (where the flux used was total neutron flux in the entire energy range). To compute this ratio, the sum of thermal, epithermal, and fast fluxes were compared.
Table 30, Predicted vs. Measured Experimental Facility Fluxes LEU Predicted LEU Measured Flux ratio LEU/HEU Whale C2 1.27 1.09 Flux ratio LEU/HEU Whale C8 1.26 1.50 Flux ratio LEU/HEU Whale E8 1.25 1.85 Flux ratio LEU/HEU Thermal Column 1.05 1.03 The LEU measured axial flux profile is shown again in Figure 37 compared to the HEU measured flux profile. Note that the HEU measurement was in relative units of probe current and not flux, but the HEU axial profile was arbitrarily adjusted in the vertical axis in order to overlay the LEU measurement to make a comparison of the shape.
Page 55 of 63
UWNR LEU Conversion Reactor Start-Up Report 7E+09 E
U.
6E+09 SE+09 4E+09 3E+09 2E+09 jL${ KTI
-- *LEU TT T
H E U_
T 711 IE+09 0
+/-
0 5
10 15 Axial Position above Bottom of Lower Reflector (inches) 20 Figure 37, Axial Flux Comparison HEU vs. LEU Table 31 shows a comparison of pulsing values for the HEU and LEU cores. The number of HEU pulses was more limited so not all LEU pulses from Table 25 were compared.
Page 56 of 63
UWNR LEU Conversion Reactor Start-Up Report Table 31, Pulse Comparison HEU vs. LEU Reactivity Peak Peak Fuel Integrated FWHM Period Insertion Power Temperature Pulse Power (ms)
(ms)
(%Ak/k)
(M
)
(°C)
(MJ)
HEU 1.000 111 198 2.64 39.42 10.58 LEU 1.000 80.26 186 2.77 39.30 9.81 HEU 1.100 284 216 5.02 22.26 6.22 LEU 1.100 206.89 206 5.32 22.83 5.66 HEU 1.200 506 233 6.53 16.59 4.79 LEU 1.200 345.87 221 6.96 17.42 4.63 HEU 1.300 746 252 8.38 13.60 4.10 T
Page 57 of 63
UWNR LEU Conversion Reactor Start-Up Report N.
COMPUTING MCNP MODEL BIAS In addition to the cold and hot bank height measurements noted above, a number of cold critical control element configurations were measured and added to the already existing rod calibration and reflector reactivity database in order to establish the criticality bias of the MCNP model of the LEU core (calculated below). In addition, the original MCNP model used for the LEU Conversion SAR used nominal fuel material definitions since the fuel had not yet been fabricated, but these fuel material definitions were updated to reflect as-built fuel data received from the manufacturer. The resulting bias in both the cold and hot models was surprisingly large. This led to an investigation of the MCNP model and possible sources of inaccuracies leading to the large bias. Several changes were made to the model as a result, summarized below:
" Transient rod was incorrectly defined as 1.25in diameter for the poison section, instead of the correct value of 1.25in diameter for the outside clad, and the fairly large bottom fitting had been neglected in the model; the fitting was added to correctly account for the displacement of water volume. Also the density of the poison section was slightly modified based on as-built documentation from the manufacturer.
" Shim safety blades were incorrectly defined as 3/8in thick boral with 1/8in clad, instead of the correct value of 1/8in thick boral with 1/8in clad for 3/8in total thickness. Also the rounded comers of the blades were added into the model.
" Reflector dimensions were updated using as-built drawings of the newly fabricated reflectors. Also, a small boron impurity was added to the graphite based on a material analysis report from the reflector manufacturer indicating level of equivalent boron content.
- A small boron impurity of 1 Oppm was added to the aluminum material definition. This is a typical value used in research reactor modeling when no aluminum material analysis is available.
" Graphite density in the axial reflectors within fuel pins was reduced or "smeared" to account for the smaller diameter of graphite slugs compared to fuel slugs.
" Cross-section libraries for uranium isotopes were switched to UTXS libraries to be consistent with erbium isotopes.
" A small hafnium impurity was added to the zirconium pin material consistent with the Hf/Zr ratio present in the fuel.
- Water density in the hot model core was reduced to the equivalent of a water temperature of 50'C (instead of 300C); 50'C is the approximate core averaged water temperature at 1.0MW.
Prior to changing the transient rod and shim safety blades in the model, the measured bias had a large degree of scatter and showed some trends when plotted vs. control element heights; after implementing the changes the bias was very consistent with little scatter, indicating that the control elements were being accurately modeled. See Figure 38 through Figure 42.
Page 58 of 63
UWNR LEU Conversion Reactor Start-Up Report U
0.=
z U
1.014 1.012 1.010 1.008 1.006 1.004 1.002 1.000 40 Oft L
0 2
4 6
8 10 Blade 1 Height (inches) 12 14 16 18 Figure 38, Calculated k vs. Blade 1 Height 1.014 1.012 4ftýýý w
IllplIr rw.
V V
rw
.x L.U.LU 0
1.004
- o.
1.006 z
U 1.004 1.002 1.000 0
2 4
6 8
10 Blade 2 Height (inches) 12 14 16 18 Figure 39, Calculated k vs. Blade 2 Height Page 59 of 63
UWNR LEU Conversion Reactor Start-Up Report U
2 1.014 1.012 1.010 1.008 1.006 1.004 1.002 1.000 0
2 4
6 8
10 Slade 3 Height (inches) 12 14 16 18 Figure 40, Calculated k vs. Blade 3 Height 5
~1a.
2~.1 1.014 1.012 1.010 1.008 1.006 1.004 1.002
.L.VUU 0
2 4
6 8
10 12 Regulating Blade Height (inches) 1 14 16 18 Figure 41, Calculated k vs. Regulating Blade Height Page 60 of 63
UWNR LEU Conversion Reactor Start-Up Report
_W V
GO 0.
1.014 1.0 12r g-1.010 1.008 1.006 1.004 1.002 1.000 10 11 12 13 14 15 16 17 18 19 20 Transient Rod Height (inches)
Figure 42, Calculated k vs. Transient Rod Height By using all available cold critical configurations for the operational core J21-R14, the average k bias is 1.01222 + 0.00040. The hot bias can only be calculated from the single full power data point as 1.00971 +/- 0.00040, since operational practice is to maintain control elements banked while at power. Therefore the cold bias is 1.01222 and the hot bias is 1.00971.
The following figure shows the change in bias as fuel and reflectors are added to the core.
Page 61 of 63
UWNR LEU Conversion Reactor Start-Up Report 1.014 J21R14 1.012
- gIIW
" 1.010 -
4J21R4 1.008 u
1.006 z
J18RO J19RO J20RO J21RO 1.004 1.002 1.000 0
0.5 1
1.5 2
2.5 3
3.5 4
Excess Reactivity (%Ak/k)
Figure 43, Calculated k vs. Core Loading As the figure above shows, there is little change in bias as fuel is loaded indicating an accurate modeling of the fuel pins. However, there is a large jump in bias as reflectors are loaded, increasing linearly with reactivity insertion. Clearly the reflectors are being modeled as more reactive than they physically are, and this alone accounts for about half of the cold model bias.
This conclusion was confirmed by review from staff at Argonne National Laboratory. However, because no reasonable solution could be found to reduce the bias in the cold or hot MCNP models, and because of the excellent consistency of cold model results, the investigation into the model bias was closed.
Using the updated MCNP model dated 2/9/2010, and implementing the known model bias, the burnup curve for the J2 1-RI 4 core was re-calculated. As seen in Figure 44, the revised curve predicts a much shorter core life. According to the LEU Conversion SAR, an operational core should have at least 0.5 %Ak/k hot excess reactivity to permit continued operation with high-worth experiments installed. The original bumup curve in the LEU Conversion SAR predicted continued operation up to 1800 MWd before falling below 0.5 %Ak/k hot excess reactivity, whereas the revised burnup curve predicts operation only until 50 MWd. Based on the revised MCNP model and well established bias, and revised burnup calculations, it is now apparent that a core reshuffle or fuel addition will be required at some point in the near future to extend core life. The analysis for this core shuffle is currently underway and will be performed as a 10 CFR 50.59 analysis or, if required, a license amendment to be submitted at a later date.
Page 62 of 63
UWNR LEU Conversion Reactor Start-Up Report Preliminary analysis favors shuffling to a compact rounded 21 bundle core with 6 reflectors.
With this rounded core, preliminary designation N2 1-R6, the calculated burnup curve predicts adequate operation up to 1700 MWd. Figure 44 compares the original predicted bumup curves for the HEU core and LEU core, the revised bumup curve for the LEU core, and the preliminary bumup curve for the N2 1 -R6 LEU shuffled core.
4 4-
'U 8I VI VI0
'I 4-0 3
2.5 2
1.5 1
0.5 0
0 500 1000 1500 2000 2500 3000 Burnup (MW-days)
Figure 44, Burnup Curve Comparison Page 63 of 63