ML20128C782
| ML20128C782 | |
| Person / Time | |
|---|---|
| Site: | University of Missouri-Rolla |
| Issue date: | 01/31/1993 |
| From: | MISSOURI, UNIV. OF, ROLLA, MO |
| To: | |
| Shared Package | |
| ML20128C779 | List: |
| References | |
| NUDOCS 9302040136 | |
| Download: ML20128C782 (35) | |
Text
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e. l 1 i INITIAL LOADING AND TESTING OP. i LOW ENRICHED URANIUM FUEL-AT THE UNIVERSITY OF MISSOURI-ROLLA , REAC'f0R FACILITY - -! (January,.1993) f s I 4
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TABLE OF CONTENTS 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . 1 2.0 Receipt and Inspection of the LEU Fuel . . . . . . . . . 2 3.0 LEU Core Loading . . . . . . . . . . . . . . . . . . . . 4 3.1 Initial Loading to Criticality . . . . . . . . . . 4 3.2 Core 100W and 100T . . . . . . . . . . . . . . . . 9 3.3 Core 101W and 101T (Without Rabbits) . . . . . . . 11 3.4 Core 101W and 101T (With Rabbits Inserted) . . . . 13 3.5 Adjusted Parameters for Core 101W and Core 101T . . 15 4.0 Excess Reactivity, _ Control Rod Worths and Shutdown-Margin . . . . . . . . . . . . . . . . . . . . . . . . . 16 5.0 Partial Fuel Element Worth . . . . . . . . . . . . . . . 21 6.0 Critical Hans . . . . . . . . . . . . . . . . . . . . . 21 7.0 Power Calibration . . . . . . . . . . . . . . . . . . . 22 8.0 Void Coefficient of Reactivity . . . . . . . . . . . . . 23 9.0 Temperature Coefficient of Reactivity . . . . . . . . . 25 10.0 Thermal Neutron Flux Distribution . . . . . . . . . . . 25 11.0 Delayed Neutron Fraction . . . . . . . . . . . . . . . . 25 12.0 Conclusions . . . . . . . . . . . . . . . . . . . . . . 26 ATTACHMENT A: SOP 816 " POWER CALIBRATION" . . . . . . . . . 27 l
.. 1
1 ', LIST OF TABLES l *able 1. Comparison of LEU and llEU Fuel Parameters . . . . . 1 J l Table 2. LEU Fuel Receipt and Inspection Results . . . . . . 3 i Table 5. Critical Rod Positions - Initial Criticality . . . 4 Table 3. Initial LEU Core Loading: Auxiliary Fission Chamber System 1/H Data . . . . . . . . . . . . . . . . . . 5 4 Table 4. Initial LEU Core Loading: Reactor Start-Up channel 1/M Data . . . . . . . . . . . . . . . . . . . . . 6 Table 6. Core 100W Parameters . . . . . . . . . . . . . . . 9 Table 7. Core 100T Parameters . . . . . . . . . . . . . . . 9 f i- Table 8. Coro 101W Parameters, Without Rabbits . . . . . . . 11 Table 9. Core 101T Parameters, Without Rabb!.ts . . . . . . . 11 Table 10. Core 101W und 101T Parameters, With Rabbits Inserted . . . . . . . . . . . . . . . . . . . . . 13 Table 11. Adjusted Core Parameters for Core 101 . . . . . . . 15
- Table 12. IIEU vs LEU Core Parameters . . . . . . . . . . . . 16 Table 13. Measured Void Coefficients as a Function of Selected Position for LEU Core 101W and ilEU Core 67W. . . . . . . . . . . . . . . . . . . . . . . . 24 Table 14. Measured and Calculated Values for Void Coefficient . . . . . . . . . . . . . . . 24 I
i t w , . - . - . ..,_.<~-y. .
. , , .- - . . - _ . -- o. . , - - - -n
- - . . - - -- -. . . ~ . . ._ - -- -_- - . - . . - . - -
+
l j LIST OF FIGURES 0 : Figure 1. 1/M Plot - Approach to criticality Based on Spare Fission Chamber System. . . . . . . . . . . . . . 7 Figure 2. 1/M Plot - Approach to Criticality Based on Reactor 1 Start-Up Channel. . . . . . . . . . . . . . . . . 7 l Figure 3. Core Loading Diagram for Initial Criticality of LEU Core . . . . . . . . . . . . . 8 j Figure 4. Core Loading Diagram for Core 100W . . . . . . . . 10 i Figure 5. Core Loading Diagram for Coro 101W (Without Rabbits) .. . . . . . . . . . -. . . . . . 12 ^ Figure 6. Core Loading Diagram for Core 101W 14 ! (With Rabbits) . . . . . . . . . . . . . . . . . . Figure 7. Core Loading Diagram for Core 67W . . . . . . . . 17 Figure 8. Regulating Rod Integral Rod Worth Curve, Core 101W . .- . . . . . . . . - . . . . . . . . . . 19 i Figure 9. Rod 1 Integral Rod Worth Curve, Core 101W . . . . . . . . . . . . . . . . . . . . 19 Figure 10. Rod 2 Integral Rod Worth Curve, Coro 101W . . . . . . . . . . . . . . . . . . . . 20 Figure 11. Rod 3 Integral Rod Worth curve Core 101W . . .. . . . . . . . . . . . . . . . - . 20 5 J 4 i
,-..w, r e ,,w - , , - r - uw- -r ,. ~- w-- , . , - - - . ~
1 1.0 Introdug. tion On July 18, 1992 the University of MissouJi-Rolla (UMR) licactor received low enriched uranium (LEU) fuel. The fuel was inspected and subsequently loaded into the reactor coro. At 15:13, July 23, 1992, the first LEU core achieved criticality. Calculations had shown that the expected core geometry, rod worths and kinetics responso would be very similar to the high enriched uranium (IIEU) fueled cores. Operationally, little change was expected other than increased U-235 loading to compensato for the higher resonance absorption and greater undermoderation associated with the LEU fueled coro. This was; however, expected to have a deleterious offect on the thermal flux in our experimental facilities. , The UMR Reactor is a pool-type reactor licensed at 200 kW. The fuel is standard MTR plate-type fuel. The UMR Reactor had been operating with its initial batch of IIEU fuel since 1965.. Table 1 below summarizes the major differences between the old llEU fuel and the new LEU fuel. Tablo 1. Comparison of LEU and IIEU Fuel Paramotors PARAMETER LEU llEU
- 1. Element Dimensions 3"x3"x36" 3"x3"nJ6"
- 2. Plates / Element 18 10
- 3. Enrichment 19.75% 90%
- 4. U-235/ Element 225 gram 170 gram
- 5. Fuel U.S i,- Al U,0,- Al LEU core parameters have been characterized and compared to predicted and measured HEU core parameters. This information is provided in the sections that follow.
2 2.0 Receipt and Innocction_Qf__the LEU Puol Twenty LEU fuel elements were received at UMRR on Saturday, July 18, 1992. Eight more LEU elements were received on Thursday, August 27, 1992. The combined shipments totaled 28 elements containing 5,248 grams of U-235. The 28 elements consists of 18 standard elements, 5 control oloments, 4 half elements, and 1 irradiation fuel element. Elements were unloaded from Type GM shipping containers, inspected, and placed in dry storage racks. No problems or difficulties were encountered. Standard eighteen plate elements are designated as "MTR-F ". Control elements (ten fueled plates per element) are designated "MTR-C ". . Half elements (nine fueled plates per element) are designated as "MTR-HF " or "MTR-HR ". The irradiation fuel element (nine fueled platea) is designated "MTR-IF ". Table 2 presents the element identifications and inspection results. (It should be noted that all incoming elements were judged to be satisfactory. Comments in Table 2 are observations, not statements of unacceptability.)
3 Table 2. LED Puol Receipt and Inspection Results ELEMEW* INSMCTION RESULTS MTR-C-001 Ssalllengthwisescratchnotedonfirstfuelplatenexttorearguideplate. MTR-C-002 Satisfactory. MTR-C-(0) Satisfactory. MTR-C-004 Satisfactory. MTR-C-005 Satisfactory. _ MTR-F 001 Satisfactory. MTR T-002 Satisfactory. MTR-T-003 Satistactory. MTR F-004 Satisfactory. MTR-T-005 Satisfactory. MTR-r-006 Smalldentnotedonthefrontfuelplateabout6inchesfromthetopoftheplateand1/2Inchfrom thesideplatewiththeID. MTR-T-007 Smalldentnotedonthefrontfuelplateabout6inchesfromthetopoftheplateand1/2inchfrom thesideplatewiththeID. MTR T-008 Satisfactory. MTR-F-009 Satisfactory. MTR T-010 Satisfactory. MTR-F-011 Smalldivot(hole)notedinfrontfuelplate(Plate 184-047-17)11/8 inchfromthelefthandsideand 7/8 incher down frce the ball. The area appears to have been " buffed'. MTR-T-012 Satistactory. MTR-F-013 Satisfactory. ! MTR-T-014 Satisfactory. HTR-T-015 Satisfactory. MTR-F-016 Satisfactory. _ MTR-F-017 Satisfactory. MTR T-018 Satisfactory. MTR-IIR-001 Smalldentnotedonthefrontfuelplateabou:6inchesfromthetopoftheplateand1/2inchfrom thesideplatewiththeID. MTR-BR-002 Satistactory. MTR-HF-001 Smalldepressionnotedontherearfuelplatsabout151/8inchfromthetop,7/8inchfromtheright side. MTR-HT-002 Satisfactory. MTR-lf-001 Satisfactory.
~~
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4 3.0 12U_CQIO_LQM11Dg 3.1 Initial LoadiDg to criticality The initial LEU core was loaded in accordance with SOP 207,
" Fuel Handling" and SOP 106, " Critical Experiment Procedures".
Subcritical multiplication data was collected after the addition of each fuel element. A spare fission chamber was placed near the core region. This chamber provided the increased sensitivity needed to soo multiplication at very low core loadings. Count data was collected on both the reactor Start-Up Channel and the spare fission chamber system. Control fuel elements were assumed to constitute one-half of an " effective" fuel element for the purpose of 1/M plots. The 1/M data and predicted critical loadings from the auxiliary firsion chamber system and f rom the Reactor Start-Up Channel are presented in Table 3 and Table 4, respectively. Data collected from both systems correctly predicted criticality with the addition of the fifteenth "offective" element. Figure 1 and Figure 2 present plots showing the expected critical loading for both systems. Initial criticality was achieved at 15:13 on July 23, 1992 with the addition of LEU element F-14. The critical rod positions are shown below in Table 5. Tablo 5. Critical Rod Positions - Initial Criticality ID POSITION (INCH) Hod 1 24.00 Rod 2 24.00 Rod 3 24.00 Reg Rod 18.78 The U-235 loading was 3420.63 ' grams. Figure 3 shows the loading for initial criticality of the LEU core. Because of the low excess reactivity (estimated at 0.045% Ak/k), no loading number was designated. l 1
i-t i i I '. 4
- I i
5 1, i Table 3. Initial LEU Core Loading: Auxiliary Fission Chamber ! System 1/M Data < j ' 3 1 Number Shim Rods at 12.5 inch Shim Rods at 24.0 inch j of I Elements Counts 1/M Predicted Counts 1/M - Predicted n - Critical critical ! Loading Loading j __ 2"' 5,774 1.00 -- 6,182 1.00 -- i 1 3 6,332 0.912 13 7,888 0.784 7-l- 4 9,401 0.614 6 : 10,675 0.579 7 ! 5 12,177 0.474 9 15,324 0.377 7 1 ! 6 1E,782 0.307 9 - 23,789 0.260 8
- 7 20,415"' 1.00 -- 26,57148 1.00 --
! 8 21,879 0.933 -- 29,157 0.911 -- 9 -29,352 0.696 12 42,415 0.626 12 ; i 10 38,961 0.524 13 62,547 0.425 13 > l 11 95,270 0.214 12 175,862 0.151 12 ! 12 107,838 0.189 17 '233,7541 0.114 15 -l l 13 206,037 0.099 15' 580,243 0.046 14
- 14 270,745 0.075 17 1,968,843 0.013 15 f 15 490,989 0.042 16 critical
") Four Control Rod Einments C-1, C-2, C-3,-C-4 and Source Inserted.- -;
- "i New C. Values.
I 1 I i l i i 4 s N
--w v.,,e . ,-+ - .~. -a, .m ry , -we.<pe g ..-r,w--w -e,-,ye=,--w3-e---c. , ,o. . . , - -,.,w,- r-, ., e- e, -e -r-r-,,ew.-e=,m,-- e e ,- w< r
j_ _ _ _ _ _ _ _ . . . _ . . .._...___-.___._-__.___.._.___..__m.- a 1 1 Tcble 4. Initial LEU Core Loading Reactor Start-Up Channel j 1/M Data I Number Shim Rods at 12.5 inch Shim Rods at 24.0 inch a of 3 Elements Counts 1/M Predicted Counts 1/M l Predicted
- Critical Critical-
; Loading Loading-I 2 48' 488 -- -- 316 -- --
1 i 3 277 ~~ -- 463 -- -- i i 4 474 -- -- 365 -- -- 5 412 -- -- 423 -- -- 6 2,710- -- -- 366 ~-- --
! 7 460 -- --
382 -- -- I-8 - 408 -- -- 374 -- -- l i
- 9 409 -- ---
409 1.00- -- 1 10 388 -- -- 4214" O.972 24 11 380 1.00 --- 472 0.866 19 ._. 12- 50 7 - 0.750 15 1,088 0.376 , 13 -- 13 606- O.627 18- 1,827 - 0.224 15 { s 14 1,067 0.356 16 8,479 0.048 15- l 6 l 15 1,163 0.327 -- criticair (*) Four. Control Rods C-1,.C-2, C-3, C-4 and Source Inserted. (*' Recorder, for first time, is not bouncing on'2 cps relay, f i i s I
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7 o,9,l.1% m% . _ _ _ . . . . _ _ - _ . . . . _ _ . _ _ _ . . ___ O.8 - - - - - - 0.7 - - - 0.6 e\N -- 3 0.5
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0.4 0.3
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02 7 8 9 10 12 D 13 14 15 16 17 18 19 20 No. of Elements Leaded
-e- Shim Rods at 12.5in Shim Rods at 24 Oin Figure 1. 1/M Plot - Approach to Criticality Based on Sparo Fission Chamber System.
=
1 w 0.9 - N\ - 0.8 --- - g m 0.7 0.6 + d 0.5- -\ \- -- ---
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\ 4
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0.2 0.1 -- 0 , . > . . 9 10 11 12 13 14 15 1'6 1'7 1'8 1'9 20 No. of Elements loaded
-=- Shirr Rods at 12 Sin Shim Rods at 24.0in Figure 2. 1/M Plot - Approach to Criticality Based on Reactor Start-Up Channel.
_ _ _ _ . . _ . _ . . . _ . _ _ . . _ ~ . _ . . _ . _ _ . _ _ . ._. _ _ _ _ . _ . _ . . - . . . .. _ __. _ . j i l 8 ! 3 ! i A MItoPitri1ES I T - Standard Elements : } B s CrControlElements ! ! Bf Balf front Elestat l' C 18 F4 0-4 BR Half Rear Element ; S Source Bolder 4 D F 13 01 T-3 F-2 f 12 3 1 E T-10 C-2 .[1 C-3 -F 9 T 14 < ! F F-5 T-6 T7 4
-2 3 4 5 6 .7 -8 9 ;
UNRR 00RE DIACRAM. ; i i I t i ! i Eles. Pos 0-235 Eles. Pos U-235: Eles. Pos' U-235-Ness Mass Nass I C-1 D4 124.86 F-5 F4 224.59. T-14 E8 224.76 C-2 E4- 124.88- T-6 T5- 224.63 C-3 E6 - 124.88~ -F 7 T6 224.66- ; ! C-4 C6 - 124.87 T 8~ C4- '224.66-i F1- E5 224.79 'F-9 E7 : 224.67-i
- j. T-2 D6- 224.83- F-10 E3 224.68-
- F D5- 224.79 T-12 D7 - 224.69 F4 C5 224.65 f D3 224.74 6
Total U-235 Mass _ (Grams) -3420.63 Figure-3'. ! Core-Loading Diagram for Initiall Criticality'of. LEU' Core , n i i - , F l q , s a
- m. _ _ , . -_ ; ;_. . .-
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). 9 3 j 3.2 Core 10.QM_And_.10.0T t llalf olomont HR-1 was added to Grid Position D-8 to increaso coro excess reactivity. The core was designated 100W. Excosa l ll reactivity, rod worths, and SDM were measured. The results :
; are shown below in Table 6. The U-235 loading was 3533.04- '
! grams. Figuro 4 shows the core loading. 3 j t* Table 6. Core 100W Parameters-
- Parameter Worth 1
! Excess Reactivity 0.450% Ak/k l: -! SDM,i,, 4. 59%- Ak/k-l- 3 Rod 1 2.52% Ak/k i
- Rod 2 2.52% Ak/k i
- Rod 3 3.25% -Ak/k l Reg Rod 0.338% Ak/k 1
! Similarly, measurement of the excess reactivity and reg rod worth wera made in the T mode. Table 7 lists the results for core 100T. l I Table 7. Core 100T Parameters-i
- Parameter Worth Excess Reactivity 0.845t Ak/k--
- Reg' Rod 0.330% Ak/k-T Column- O.395%- Ak/k >
c t i I i i l' i i 5 e n- ,-e -% - o-,,,. -, , ,,n., y- ens,e, - . - , . , , .' ,.gy .c.,,-- g ,.<,_,,.,h,eyg..c..g+,- ag w r, .,h3% , wy,4,, y vw-e-e --er.
4 10 a A KEY TO PREflIES T - Standard Elements B $ C - Control flements Er - Half front Element C F8 f4 0-4 ER - Half Rear Element S - Source Bolder D F-13 0-1 T-L F-2 1-12 HP1 E f-10 0-2 T-1 0-3 T9 T-14 f F-5 T-6 T-7 1 2 3 4 5 6 7 8 9 CNRR Coke DIAGRAM 4 Eles. Pos 0-235 Eles. Pos U 23$ Eles. Pos 0-235 Mass Bass Kass C-1 D4 124.86 T-5 f4 224.59 F-14 E8 224.76 C-2 T4 124.88 T-6 F5 224.63 ER-1 D8 112.41 T-7 C-3 E6 124.88 T6 224.66 C-4 C6 124.87 T-8 C4 224.66 T-1 E5 224.79 T-9 E7 224.67 T-2 D6 224.83 F 10 D 224.68 F-3 DS 224.79 T-12 D7 224.69 T4 05 224.65 T-13 D3 224.74 Tota 1 U-235 Mass (Grams) 3633.04 Figure 4. Coro Loading Diagram for Coro 100W
Il 3.3 Cnrc_10lW_Dnd 101T (Without Rabbits) on July 20, 1992 einment ilR-1 was replaced with F-15 to increase core excess reactivity. The U-235 loading was 3645.4 grams. Figure 5 shows the core loading. Both cores 101W and 101T were characterized; results are summarized below in Tables 8 and Tablo 9. Tablo 8. Coro 101W Paramotors, Without Rabbits Paramotor Worth Excess Reactivity -C.874% Ak/k SDM., 4.34% Ak/k Rod 1 2.63% Ak/k Rod 2 2.58% Ak/k Rod 3 3.30% Ak/k Rog Rod 0.350% Ak/k l
=
Table 9. Coro 101T Paramotors, Without Rabbita Paramotor Worth Excess Reactivity 1.28% Ak/k Reg Rod 0.356% Ak/k T Column 0.406% Ak/k l l I
,e ,--- - -
12 4 A __ KEY TO l'IllllES I Standard Elerents i B S C - Control Elements 4 HF - Half front Element C F8 T-4 C-4 HR-HalftearElement S - Source Holder D T-13 C-1. F-3 T-2 T-12 T-15 BR-fureRabbit CR-CadalusRabbit E f-10 C-2 T1 C-3 T-9 l-14 F F-5 T-6 T-7 t 1 2 3 4 5 6 7 8 9 l'KIR COPE DIACMX i Eles. Pos 0-235 Eles. Pos U-235 Eles. Pos 0-235 Kass Mass Mass C-1 D4 124.86 T-5 T4 224.59 F-14 E8 224.76 , 0-2 E4 124.88 I-6 f5 224.63 T-15 D8 224.77 j C-3 E6 124.88 T-7 f6 224.66
- C-4 C6 124.87 T-8 C4 224.66 i
F1 ES 224.79 T-9 E7 224.67 F-2 D6 224.83 f-10 E3 224.68 f-3 E6 224.79 T-12 D7 224.69 T-4 C5 224.65 T D3 224.74 ', 1 Tot al U-235 Mass (Grams) 3645.40 Figure S. ' Core Loading Diagram for Core 101W-(Without Rabbits) 4
'I
r-- - 13 3.4 CDI.0_101H_und_101T (With Rabbits Inserted) The Bare and Cadmium Rabbit Facilities were installed in the gridplate on July 28, 1992 as shown in Figure 6. Table 10 presents the measured parameters. Table 10. Core 101W and 101T Parameters, With Rabbits Inserted PARAMifrER CORE 101W CORE 101T Excess 0.429% Ak/k 0.812% Ak/k Reactivity SDM,i, 4.95% Ak/k 4.11% Ak/k Rod 1 2.71% Ak/k 2.46% Ak/k Rod 2 2.67% Ak/k 2.46% Ak/k Rod 3 3.20% Ak/k 3.23% Ak/k Reg Rod 0.355% Ak/k 0.355% Ak/k-The worth of the thermal column was 0.383% Ak/k. Core 101W was determined to be tentatively acceptable. I J e j
14 i - A EfYTOPIEfilES 4 f - Standard Elements B S . C-ControlElements HF - Half front Ilement i C T-8 T-4 C4 HR - Half Rear Element j S - Source Bolder l D T-13 C-1 1-3 T-2 T-12 T-15 l l E T-10 C2 T-1 C-3 T9 T-14 f ___ Q T-5 T-Q_ T-7 ft 1 2 3 4 5 6 7 8 9 4 CMRR 00RE DIACRAM 4 Eles. Pos U-235 Eles. Pos U-235 Eles. Pos U-235 Mans Mass Hus C-1 D4 124.86 T-5 F4 224.59 T-14 [8 224.76 C-2 E4 124.88 T-6 f5 224.63 T-15 D8 224.77 C3 E6 124.88 T-7 f6 224.66 C-4 C6 124.87 T-8 C4 224.66 I-1 ES 224.79 T-9 E7 224.67 T-2 D6 224.83 f 10 E3 224.68 T3 DS -224.79 T 12 D7 224.69 I-4 C5 224.65 T-13 D3 224.74 Total U-235 Hass (Grams) 3645.40 Figuro 6. Coro Loading Diagram for Coro 101W (With Rabbits)
1b l 3.5 Atijunted ParapatcrQ_for_Coro 101W and_Coro 101T It should be noted that all reactivity measuromonta mado using the positivo period method in Sections 3.1 through 3.4 woro
- based on J,,f=0.00755. Noutronics studies have yloided a calculated J,ff=0.0079 for the LEU fuel. This value is presented in Tablo VIII of the SAR. To be consistent, measured reactivity values using the positivo period method i need to be adjusted accordingly.
Additionally, the spare fission chamber used for coro loading was repositioned from the core sido to a new position above-and behind the core. The repositioning was measured to havo a small positivo reactivity offect of 0.045% Ak/k. j Table 11 presents the adjusted core parameters for Coros 101W I and 101T which account for the new value of J,,, and the movement of the detector. . Tablo 11. Adjusted Coro Paramotorn for coro 101 ( J,rt=0.0079 ) PARAMETER CORE 101W CORE 101T Excess 0.496% Ak/k 0.697% Ak/k - Reactivity _ SDM,i,, 4.92% Ak/k -4.10% Ak/k Rod 1 2.73% Ak/k 2.50% Ak/k Rod 2 2.69% Ak/k 2.50% Ak/k Rod 3 3.22% Ak/k 3.27% Ak/k Reg Rod 0.371% Ak/k 0.371% Ak/k The thermal column worth was 0.401% Ak/k.
48 e 16 4.0 Excena_ Reactivity,_ Control Rod Worths and Shutdown _MaE91D HEU Core 67W presented in Figure 7 was similar in geometry to LEU Coro 101W. The geometry differences are as follows:
- 1. The positions of the cadmium and bare rabbit facilities are reversed; and
- 2. Core 67W had an extra half-element in position C-3.
Although the geometries are not identical, the cores are very similar and merit comparison. Table 12 below compares excess-reactivities,-rod worths, and shutdown margites (SDM) for the two cores.
-Table 12. HEU vs LEU Core Parameters PARAMETER LEU (101W) HEU (67W)
- 1. -U-235-Loading- 3645 grams 2870 grams
- 2. # of " Effective" 16.0 16.5-Elements *
- 3. Excess 0.496% Ak/k 0.43% Ak/k Reactivity
- 4. Rod Worths Rod 1 2.73% Ak/k 2.64% Ak/k
-Rod 2 2.69%-Ak/k: 2.65% Ak/k-Rod 3 3.22% Ak/k 3.36% Ak/k' Reg Rod 0.37% Ak/k 0.35% Ak/k
- 5. SDM 4.92% Ak/k 4.86%-Ak/k-
-As can be seen in the comparison in T a b l'e - 1 1 , - the only parameter significantly affected by the conversion is the U-235 gram loading. The LEU core contains.about 27% more-U-235:
than the HEU core. This additional loadingEin required to overcome. the deleterious; effects- of increased. resonance absorption and, stronger under. moderation. The' LEU core has an excess reactivity of:0.496% Ak/k achieved-with 16. "ef f ective*" elements while the HEU core had an excess : reactivity 1of 0.43% Ak/k using 16.5 " effective" elements..
- control elements.are treated as one-half <of'an " effective"?
. element.
0 17 A EILIO PIEFIIU f - Standard Ele >ents B S C Control Elements HT Half front Element C _ , . D-1 T 14 T1 C-4 D - Half Rear Element S - Source Holder D F-8 C-1 T 16 T-9 T-4 T 10 E T-6 C2 T 19 C-3 T-12 T-11 i BR F+17 T lh CR 1 2 3 4 5 6 7 8 9 UKRR 0)RE DIACWl Eles. Pos U-235 Eles. Pos U-235 Eles. Pos U-235 Kass Mars Mass ER-1 C3 84.912 f 16 DS 170.270 T-12 E7 168.774 F8 D3 170.229 T 19 E5 170.264 T-10 D8 170.193 T-6 E3 169.160 f-15 F5 168.889 T-11 E8 168.969 T 14 C4 170.210 C-4 C6 102.112 C-1 D4 102.112 T-9 D6 170.178 C-2 E4 102.125 C-3 E6 101.978 T-17 F4 169.111 T-7 f6 170.154 T1 C5 170.223 T4 D7 170.206 Total U-235 Mass (Grams) 2870.069 Figuro 7. Coro Loading Diagram for Coro 67W 1 I ____-------_---____-__-_A
I 18 l This comparison demonstrates that the reactivity worth pot ! "offective" olomont in the LEU core is only slightly higher j than with the HEU core. I Rod worths in the LEU core are (but not appreciably) dif ferent i i from the HEU core. Tho slight shift in rod worths shown in ] Tablo 12 may be as much a result of the slight changes in core configuration geometry as with the chango in fuel type. Finally, becauso excess reactivities and rod worths woro very similar for the two cores, then necessarily the shutdown j margins are similar. Overall, the measured coro paramotors presented in Tablo 12 l did not change appreciably with the LEU conversion, with the i exception of the U-235 gram loading. Neutronics calculations woro perforped for both the HEU and
- LEU coros for various core configurations. This work is presented in Covington". Coro 101W is very similar to the 2 targot core with the following exceptions:
4
- 1. The target core had an extra half-olomont in Grid Position C-3.
, 2. The positions of the cadmium and bare rabbit j facilities were reversod. Covington predicted an excess reactivity of 0.87% Ak/k. Coro 3 101W has an excess reactivity.0.496% Ak/k. As presented in > Section 5, a half element is estimated to be worth ! approximately 0.424t Ak/k. If we assumed a half element was added to Coro 101W making it consistent with the target core configuration, the resulting excess reactivity _is estimated to be 0.469% Ak/k + 0.424% Ak/k = 0.92% Ak/k. This.value is in close agroomont with the calculated value of:0.87% Ak/k. Integral rod worth curvos for Coro 101W are presented in Figures 8 through 11. As was soon in Table 12, rod worth values changed very little with the fuel' conversion. This was
- as expected.
8Covington, Lorne J., "Neutronics. Study.of the Conversion of 3 the' University of Missouri-Rolla Reactor to Low Enriched Uranium Fuel". M.S. Thesis (1989) University of Missouri-Rolla. y y e . - -
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21 5.0 Partial Punl Element Worth The worth of a partial (or half) fuel olomont at the core , periphery was determined by the change in core excess j reactivity between core 100W and 101W. This core change ' , involved replacing the half element (HR-1) located in Grid Position U-8 with a full fuel element (F-15). The swing in excess reactivity was 0.424% Ak/k (W mode) and 0.435% Ak/k (T mode). 6.0 Critical Mass The critical mass of U-235 for the LEU fuel was estimated to bo 3409 grams based on actual core loading data. The critical mass of U-235 for the HEU core was estimated at 2,797 grams based on previous core loading information. Thus, the critical U-235 mass of the LEU core is about 22% larger than the HEU cores. i The estimated critical mass for the LEU core was determined by l comparing the excess 'r eactivities and U-235 loadings of cores 100W and 101W. Core 100W had a loading of 3533.0 grams U-235 and excess reactivity of 0.450% Ak/k. Core 101W had a loading of 3645.4 grams U-235 and excess reactivity of 0.874% Ak/k. The reactivity worth por gram of U-235 can be estimated as: 1 (0.874 % Ak/k - 0.450% Ak/k) = 0. 003 8 %Ak/k (3645.4 - 3533.0) grams gram The initially critical LEU core contained 3421 grams of U-235 and had an excess reactivity of 0.045% Ak/k. This implies an excess of about (0.045% Ak/k) + (0.0038% Ak/k-g) = 12 grams. The estimated LEU mass is therefore 3421 grams - 12 grams = 4 3409 grams U-235. The critical mess of the HEU core was estimated based-on Core 67W. Core 67W contained 2870 grams of U-235 and had a measured excess reactivity of 0.43% Ak/k. Experiments at the UMRR with dif ferent core configurations
- have shown that the
< worth of a fuel element (170 grams U-235) is between 0.5% and 1.5% Ak/k depen ding - on its position at the core periphery. Taking an average of 1.0% Ak/k and using the same methodology as above, the reactivity worth per gram of U-235 can be estimated as:
' Safety Analysis Report for the University of Missouri-Rolla Reactor. Docket Number 50-123, Page 9-13. (1984)
3-22-1.0%bk/k , 0 00588 g$g/g 170-grams gram The HEU core 67W thus had a surplus.of about-73 grams of- . U-235 above the. critical-mass. 1Therefore, the minimum ~ critical. mass of.U-235 for the HEU fuel is.approximately 2,797. grams. 7.0 Power Call'bration Af ter low power core characterization,-the reactor operated ~ at: s an indicated power of -1 kW on the Linear Channel for -one hour. - b The' wre was checked andithe' operation appeared normal. 1 A power -calibration was perf ormed on August- 4, .1992.- -The- - procedure involves operating the< reactor-atian intermediate
~
indicated power'of 40'kW:for-one' hour. The.resulting " heat balance" Mis 'obtained by : measuring Kthermal expansion of the? - pool. ~ The power calibration procedure'(SOP 816) is provided as Attachment A. . Detectors had . to be positioned- slightly ' 4 closer-to-the core as a: result of the calibration.- It was expected that: nuclear instruments would road:.a bit high due,to the higher core - ~ leakage 1 ' associated with the -harder e" spectrur.. A second power calibration,. performed on Aug.i't 1992 verified that the. detectors were' correctly-positir '4 _ _ _ _ 1
1 . 1 23 8.0 Void Coef ficient of Reactivity An aluminum void tube fabricated to fit into the grid plate was used to measure the void coefficient of reactivity. The tube is hollow and sealed. Core reactivity is measured with the tube both filled with air and with water. These measurements are used to calculate the void coefficient.- The void coef ficient (a,) is calculated by taking the ratio of the change in reactivity to the effective volume of the void:
, bPair Ellied ~b0t.ater tilled
Vol The volume of the tube is 1300 cm'.
3 The average void coef ficient measured for the new LEU Core 101W was ~7 .1 X 10-5 ( % Ak/k) . The previously measured value cm for HEU Core 67W was -6 . 9 X 10-5 ( %AM) . The' void Cm' coefficient for the LEU core appears to be slightly rore negative than the HEU core. This trend is as expected due t; the harder flux spectrum. In fact, measurement showed that the void coefficient of the LEU core is approximately 20% more negative than the HEU core at-similar peripheral positions. This is illustrated in -Table 13, which presents void coefficients measured in various locations for both the LEU and HEU cores. It should be noted th t the cores. geometries are not identical, as mentioned in Section 4. Neutronic calculations
- for the void coefficient of reactivity were also performed for both proposed HEU and '. LEU cores.
Table 14 presents neasured and calculated values. i
'Covington, Lorne J., "Neutronics Study of the Conversion of the University of Missouri-Rolla Reactor to Low Enriched . Uranium Fuel". H.S. Thesis (1989), University of Missouri-Rolla.
. e c , ,
I - i. ' 24 4 Table 13. Measured-Void Coefficients as:a Function of Selected Position for LEU! Core'101W and HEU Core'67W.- i n- -
- POSITION LEU CORE' HEU CORE (t Ak/k-ca') ( % ' Ak/k-cm*) .
E -8.62 x'10-* -7 . 2 5 - X 10-* ~ (- ; I C-8 -5. 7 5 x 10-* -4.90 x'10-* i D-9 -5.44~x 10-* ~4.80 x 10-* E -5.69 x 10-* -4.45 A 10-5 , a d !. Table 14. Measured and Calculated-: Values !' for Void Coefficient - ( k/R) cm 3 i ' LEU CORE DESCRIPTION ~ HEU' CORE' i Measured 1 -6. 8 5 x 10-* - -7.08 x 10-* l Calculated -7.0 x 107 5 -9.0 x.10-5 4 i The calculated value for the LEUIoid. coef ficient is- about 28%
~ '
more' negative -than f or ~ the HEU - f uel . . Comparing : measured values for selected locations-presented in' Table 13-showed , that the LEU void coef ficient. ranged from 'about- 13% - to 28% more negative than the.HEU values, i. l a E
- l P
i I L.
)
25 9.O Temperature Coefficient of Reactivity The temperature coefficient of reactivity will be measured'in Senior the Winter, 1993 semester _as a class Nuclear' Engineering project. It was planned to perform the experiment . during the Christmas break of '92/'93; however, required reactor maintenance lasted longer than planned, thus the experiment has been postponed. 10.0 Thermal Neutron Flux Distribution These measurements are somewhat time consuming and have'not yet_been obtained for the new core. We plan to characterize the flux'in the beamport ~ facility, bare' and cadmium rabbit facility, and at some various_ core positions during this-next-semester. As we have a very small staff and because these measurements make excellent student projects we. plan to obtain these measurements via student labs and projects.- -We anticipate an M.S.' Thesis project for the completion of the core flux profile measurements., 11.0 Delayed-Neutron-Fraction The new value of B.,,-is 0.0079 as presented in the SAR.- This 3 is about 4.6% larger than - the previously hssumed , value of - O.0075 used with the HEU core.
~
! No kinetics measurements have been made with'the LEU core and-to our knowledge, there - . is - no . documented measurement : of kinetics. parameters with any of: the HEU cores. Operationally, , the LEU core behaves very similarly.to the_HEU1 cores.- At present, we have assigned two-Senior Nuclear Engineering. i students to explore ways of determining kineticEJparameters !. with computer noise analysis. This will involve interfacing ' 7 l- a computer data collection J to the isolated = outputs of ' our-i" newly.(and as-yet'not installed) nuclear instrument drawers.- ' I We feel this will be a suitable _ topic forJan M.S. Thesis. . . . H l
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s k i I 26 [ 12.0 Conclusiong [ The receipt and inspection of the new elements went smoothly l and all elements were determined tc, be satisfactory. Initial i core loading to_ criticality went smoothly with 1/M data correctly predicting. criticality.. The target core configuration based on neutronics calculations differed from the actual final core configuration by only one-- i half element. Excess . reactivity, control rod _ worths and ' l shutdown margin for the new core are very.similar to those in the HEU cores. As expected, _ the critical mass in the LEU core i was significantly higher than with the HEU fuel although the overall core geometry was negligibly affected.- Detector positions changed very little but- did have to be-positioned 'slightly closer to the core as -a result' of - the_ core change.-_ This - was as. expected -due to- the greater leakage
~
- anticipated with the harder flux spectrum. The void coefficient was found to be slightly_more negative than with the :HEU fuel. This was ' ' consisteni. with calculations. Still to 'be further characterized are Lthe temperature coefficient, flux distribution, and kinetics parameters which are to be characterized in the near future. No abnormalities or surprises have been. uncovered with regard- - to the new fuel. We plan to shipiout our remaining balance of
~
HEU fuel-as soon as possible. We. hope this: fuel will-last'us well-~into the~21st century, a s i
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2'I ATTACHMENT A SOP 816-
" POWER CALIBRATION"
4 2a
*aa UMR REACTOR STANDARD OPERATING PROCEDURES ***
SOP 816 Titlei UNRR POWER CALIBRATION Revised: August 30, 1988 Page 1 of 4 A. Purcose: To ensure that the power indicated c.n the linear and log channels is the power generated in the reactor. B. Precautions. P;3reauisites, or Limitations
- 1. In accordance with Technical Specification 4.2.2(3) all console instrument s and saf ety system shall be REV.
- alibrated twice each year, not to exceed / months. I 1 '/1 F)
- 2. The power generation in the UMR Reactor is limited by Technical Specifications to 200 kW. It is, therefore.g,74 ^@3 $S important that the reactor power is less or, in an ideal case, equal to the power indicated in the reactor control room. The calibration of the power instru-ments is performed by the calibration procedure described below. (For more details see the report UMRR/85-i.) Stable atmospheric condition are helpful
, f or a successful calibration. C. Procedure
- 1. Turn on both nitrogen dif fusers and the pool lights.
- 2. Set up pool level measuring equipment. It is recom-mended that two gauges be used in order to have redun-dant measurements. (Minimum recommended scale division is 0.001 inches.)
- 3. After the diffusers have been on for at least 30 minutes st ar t to take level readings every 15 minutes.
Continue for at least one hour prior to the reactor startup te determine the average pool level drop. Be sure to note accurately the time of each reading. Record also the temperature cf the pool water using i all three reactor thermocouples.
- 4. Take the reactor to some intermediate power level, e.g.
20, 30 or 40 kW. Note the time the reactor re ches that power level. After running the reactor at this power for a time t, such that the reactor thermal out-put is between 30 and 50 kW hr . shut down the reactor and note the shutdown time. For example, it is recom-Written By: Milaq Straka Approved By: Albert lon A
{.. 29 , 4
* *
- UMR REACTOR STANDARD OPERATING PROCEDURES * *
- SOP: 816
Title:
UMRR POWER CALIBRATION Revised: August 30, 1988 Page 2 of 4 i
- mended that the reactor power be chosen'40 kW and the
. operational time t, i hr. j 5. Once all control rods and magnets are fully inserted, note time and pool level every 15 minutes until level ', decreases equal the rate of decrease bef ore the power run. During this time also continue to take tempera-l ture readings using all reactor thermocouples.
- 6. Plot the data measured with both relative h'eight gauges such as to construct the time-dependent plot of_h.-i.e.
the relative change in height of the pool water sur-face before, during, and after the power run. (Use 4 units of cm f or the plot of h.)
- 7. Determine ah as_shown in the sketch below h [cm]
. A L l 7 sh i
~
i % ! U 's i L = Time clime _t [hr] l at at Power Shutdown. I- p 4
- 8. Calculate the average pool water. temperature T, using the data taken immediately bef ore the beginning of the
- - power run and after the reactor shutdown. (Use only the inlet temperature readings. )
Written By: Mila Straka Approved By: Albert Bolon
/Ott ALh
30
*** UMR REACTOR STANDARD OPERATING PROCEDURES ***
SOP: 816
Title:
UMRR POWER CALIBRATION Revised: August 30, 1988 Page 3 of 4
- 9. Using Figure 1 and data determined in step 7 and 8 determine the amount of heat Q generated in the reac-tor during the calibration run. (The fact that the coefficient of the thermal volumetric expansion is to be taken at the temperature which is 1 K higher than the average pool temperature has already been taken into account while constructing the plot in Figure 1.)
- 10. Calculate the re actor power using the . relationship P(kW) =0 (kuhrl t, thr)
- 11. If the power indicated on the linear and/or Log N recorder is equal to or greater than the calculated power P by not more than 5% no further action is needed. In any other case the position of the per-tinent neutron detector needs to be adjusted so as to satisfy the above condition. .
pp
& zluin
- 12. After both power channels (linear and og N) have been IJSA properly adjusted take the reactor to 100 h" and adjust, if necessary, both safety channels so as to in-dicate the reactor power of y100 hM.-
N 2.ccH W/ W 3 Zlac/9Z.((C Written By: Milan Str aka Approved By: Albert B lon L WO
__ _ . .. . _ _ _. _ . _ ___ _ .. ~ ~ _ - - - _ _ . __ - . __ k
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*** UMR REACTOR STANDARD OPERATING PROCEDURES ***
! SOP: 816 Tit 1e: UMRR POWER CALIBRATION > 2 Revised: August 30, 1988 Page 4 of 4 t a . , ,r. - , _. s
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