ML20127P278

Initial Loading & Testing of Low Enriched U Fuel at Univ of Missouri-Rolla Reactor Facility

ML20127P278
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Site:	University of Missouri-Columbia
Issue date:	01/31/1993
From:	MISSOURI, UNIV. OF, COLUMBIA, MO
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ML20127P274	List: ML20127P271 ML20127P278
References
NUDOCS 9302010232
Download: ML20127P278 (35)

Text

. . .

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INITIAL LOADING AND TESTING OF LOW ENRICilED URANIUM FUEL AT Tile UNIVERSITY OF MISSOURI-ROLLA REACTOR FACILITY (January, 1993) s 9

'V302010232 930120 PDR ADOCK 05000286'

.P. ppg

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TABLE OF CONTENTS ,

6 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 Coro 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 Mass . . . . . . . . . . . . . . . . . . . . . 21 7.0 Power Calibration . . . . . . . . . . . . . . . - . . . 22 e

8.0 Void Coefficient of Reactivity . . . . . . . . . . . - . . 23 9.0 Temperaturo Coefficient of Reactivity . . . . .. . . . . 25 10.0 Thermal Neutron Flux Distribution . . . . . . . . . . . 25 11.0 Dolayed Neutron Fraction . . . . . . . . . . - . . . . . . 25 12.0 Conclusions . . . . . . . . . . . . . . . . . - . . . . . 26 ATTACHMENT'A: SOP 816 " POWER CALIBRATION" . . . . . . . . - . 27 r

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I-LIST OF TABLES  !

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Tablo 1. Comparison of LEU and HEU Fuel Parameters . . . . . 1  !

Table 2. LEU-Fuel Recolpt and 2nspection Results . . . . . . 3-f i

Table 5. Critical Rod Positions - Initial Criticality .- . . 4 Tablo 3. Initial LEU Coro Loading Auxiliary Fission Chambor System 1/M Data . . . . . . . . . . . . . . . . . . 5 a

Table 4. Initial LEU Core Loading: Reactor Start-Up Channel 1/H Data . . . . . . . . . . . . . . . . . . .. . . 6 ,

t Table 6. Coro 100W Paramotors . . . . . . . . . . . . . . . 9 Table 7. Core 100T Parameters . . . . . . . . . .: . .- . . . 9

. Table 8. Coro 101W Paramotors, Without Rabbits . . . . . . . 11 Table 9. Coro 101T Paramotors, Without Rabbits . . . . .- . . 11 Table 10. Core 101W and 101T Parameters, With Rabbits i Inserted . . . . . . . . . . . . . . . . . . . . . .13 Table 11. Adjusted Coro Parameters for Core 101 . . .. . .' . .. 15 Tablo 12. HEU vs LEU Core Parameters . . . . . . . . . . . .- 16 Table 13. Measured Void Coefficients as a Function of Selected Position for LEU Coro 101W and HEU Coro 67W. . . . . . . . . . . .. . . .. . . . . . . . . 24 ,

Table 14. Measured and Calculated Values for Void Coefficient. . . .. . . . . . . . . . .- . . 24-i k

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LIST OF FIGURES  ;

i Figuro 1. 1/H Plot _- Approach to criticality Based on Spare -i Fission Chamber System. . . . . . . . . . . . . .- 7 t

5 Figure 2. 1/M Plot - Approach to criticality Based on Reactor ,

Start-Up Channel. . . . . . . . . . . . . . . . . 7 l

f Figure 3. Coro Loading Diagram for Initial Criticality of LEU Coro . . . . . . . . . . . . . 8 Ffgure 4. Core. Loading Diagram for Coro 100W . . . . . . . .. 10 Figure 5. Core Loading Diagram for Core 101W 12 (Without Rabbits) . . . . . .. . - . . . . . . . . -

7 r

Coro Loading Diagram for Coro 101W M Figuro 6.

(With Rabbits) . . . . . . . . . . . . . . . . -.-. 14  ;

Figure 7. Coro Loading Diagram for Coro 67W . . . . . . - . _. Regulating Rod Integral Rod Worth Curvo, l Figuro 8.

Core 101W . . . . . . . . . . . . . . . . . . - . . . 19 Figuro 9. "od 1 Integral Rod Worth Curvo, Coro 101W . . . . . . . . . . . . . . . . . . . . . 19.

T Figure 10. Rod 2 Integral Rod Worth Curvo, Coro 101W . . . . . , . . . . . . . . . _. . . . . 20 j Figure 11. Rod 3 Integral Rod Worth Curve Coro 101W . . . . . . . . . . . . . . . . . . . . . 20 I

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1 1.0 Introduction On July 18, 1992 the University of Missouri-Rolla (UMR)

Reactor received low onriched uranium (LEU) fuel. The fuel was inspected and subsequently loaded into the reactor curo.

At 15:13, July 23, 1992, the first LEU core achieved criticality.

Calculations had shown that the expected core goometry, rod worths-and kinetics responso would bo very similar to the high onriched u: anium (!!EU) fuolod cores. Operationally, little change was expected other than increased U-235 loading to compensate for.the highor resonanco absorption and greater  !

undermoderation associated with the LEU fueled coro. Thin was; however, expected to have a doloterious of fect on - the thermal flux in our experimental facilities.

The UHR Reactor is a pool-type reactor licensed at 200 kW.

The fuel is standard MTR plato-typo fuel. The UHR Reactor had boon operating with its initial batch of HEU fuel since 1961.

Tablo 1 below summarizes the major dif ferences between the old

!!EU fuel and the now LEU f uel.

Tablo 1. Comparison of LEU and !!EU Fuol Paramotors PARAMh"ITJ1 LEU IIEU

1. Element Dimensions 3"x3"x36" 3">3"x36"

2. Plates / Element 18 10'

3. Enrichment 19.75%- 90%

4. U-235/ Element 225 gram 170 gram

5. Fuel U.Si -Al U,0, - Al LEU core paramotors have boon characterized and compared'to predicted and measured HEU core parameters. This information-is provided in the sections that follow.

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2.0 Ennalp.t and innocction of_the LEU Puel l Twenty LEU fuel elements were received at UMRR on Saturday,.

July 18, 1992. Eight woro LEU olomonts 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 elements, 4 half alomonts, and 1 irradiation fuel element.

Elements were unloaded from Type 6M shipping containers, inspected, and placed in dry storage racks. No problems-or i difficulties were encountered. Standard eighteen plato' elements are designated as "MTR-F ". Control elements-(ton fueled plates per element) are designated "MTR-C ". Half .

elements (nino fueled platos per eloment)-are designated as_

"MTR-HF " or "MTR-HR ". The irradiation fuel element -(nino fueled plates) is designated "MTR-IF ". Table 2 presents the element identifications and inspection results. (It should be noted that all incoming elements were judgod to be- ,

satisfactory. Comments in Table 2 are observations, not statements of unacceptability.)

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_,' 3 fable 2.' LL7 fuel Receipt and Inspection Results ELEMDT INSPECTION RECULi$  ;

MTk C-001 Small lengthwise scratch noted on first fuel plate next to rear guide plate.  ;

MIR-C-002 Satisfactory.

MTP-C 'X)) Satisfactory. l MTR-C-004 Satisfactory. .

MTR-C-005 Satisfactory. ,

Mik F-001 Satisfactory.

MIR T 002 Satisfactory.

MTRF003 Satisfactory.

MTR t-004 Satisfactory.

MTR-F-005 Satistactory. ,

HTRf-006 Smalldentnotedonthefrohtfuelplateabout6inchesfromthetopoftheplateand1/2inchfrom the side plate with the ID.

MTR T-007 Smalldentnotedonthefrontfuelplateabout6inchesfromthetopoftheplateand1/2inchfros thesideplatewiththeID. ,

HTR F 008 Satisfactory. ,

MTR T 009 Satisfactory.

MIR T-010 Satisfactory.

MTR T-011 Sulldivot(hole)notedinfrontfuelplate(Plate 184-047-17)11/8 inchfromthelefthandsideand 7/8inchesdownfromtheball. The area appears to have been ' buffed".

MIR F-012 Satisfactory.

.MIR-T-013 Satistactory.

MTR T-014 Satisfactory. ,

MTR T-015 Satisfactory.

MTR T-016 Satistactory.

MIR-T-017 Satisfactory.

MTR T-018 Satistactory.

MTR-ItR-001 Sull dent noted on the front fuel plate about 6 inches from the top of the plate and 1/2 inch from thesideplatewiththeID.

MTR-HR-002 Satisfactory.

MTR HF-001 Small depression noted on the rear fuel plate about 151/8 inch from the top, 7/8 inch from the right side.

MTR-HF-002 Satisfactory.

HTR IF-001 Satisfactory.

_+,.u., . . . . . - .=, - , .- - - ..--:.... . . . . - . _ . - .

4 3.O LEU _ fore._JaDiling 3.1 initial Lomling.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 oloment. A spare fission chamber was placed near the core region. This chamber provided the increased sensitivity needed to see multiplication at very low core loadings. Count data waL 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 fission chamber system and from the Reactor Scart-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 " effective" olomont. 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.

Table 5. Critical Rod Positions - Initial Criticality ID POSITION (INCH)

Rod 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.

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Table 3. Initini LEU Coro Loading: Auxiliary Fission Chambor '

System 1/M Data Number Shim Itoda at 12.5 inch Shim Itoda at 24.0- inch .

of Elements Counts 1/H Prodicted Counts 1/M Predicted Critical Critical Loading _ Loading 2'" 5,774 1.00 -- 6,182 1.00 --

3 6,332 0.912 13 7,888 0.784 7 4 9,401 0.614 6 10,675 0.579 7 5 12,177 0.474 9 15,324 0.377 7 6 18,782 0.307 9 23,789 0.260 8 7 20,415 1.00 -- 26,571'8' 1.00 --

0 21,879 0.933 -- 29,157 0.911 --

9 29,352 0.696 12 42,415 0.626 12-10 38,961 0.524 13 62,547 0.425- 13 11 95,270 0.214 12 175,862 0.151 12 12 107,838 0.189 17 233,754 0.114 15 1 13 206,037 0.099 15 580,243 0.046 14 14 270,745 0.075 17 1,968,843 0.013 15 15 490,989 0.042 16 critical a

'" Four Control Rod Elements C-1, C-2, C-1, C-4 and Source Inserted.

'" New C, Values.

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6 i Table 4. Initial LEU Coro Loading: Reactor Start-Up Channel 1/M Data i Number Shim Roda at 12.5 inch Shim Rods at 24.0 inch .

of Elements Counts 1/H Predicted Countu 1/H Prodlcted Critical Critical Loading Loading _

2"' 488 -- -- 316 -- --

3 277 -- -- 463 -- --

4 474 -- -- 365 -- --

5 412 -- --

423 -- ~~

6 2,710 -- -- 366 -- --

7 460 -- --

382 -- --

8 408 -- -- 374 -- --

9 409 -- --

409 1.00 --

10 388 -- --

4 21 O.972 24 11 380 1.00 --

472 0.866 19 12 507'" O.750 15 1,088 0.376 13 13 606 0.627 18 1,827 0.224 15

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14 1,067 0.356 16 8,479 0.048 15 15 _. 1,163 0.327 -- critical

"' 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.

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0.1 -l N-O 7 8 9 1'O 1'1 12 1'3 1'a 15 1'6 17 1'8 19 20 No. of Elements Leaded I -t- Snim Rods at 12 Sin Snim Rods at 24 Cin Figure 1. 1/M Plot - Approach to Criticality Based on Spara Fission Chamber System.

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9 10 11 12 13 14 15 16 17 18 19 20 No. of Elements leaded

-+- Shim Reds at 12 Sin Shim Rods at 24,0in Figure 2. 1/M Plot - Approach to Criticality Based on Reactor Start-Up-Channel.

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8 A - KEYTOPUIIIIS i f - Standard Elerents 14 __

S C - Control Elements ET - Half front Element C __

T-l F-4 C-4 ER - Half tear Element S Source Bolder D F-13 .C-1 T-3 f*2 T 12 t f-10 0-2. f-1 C-3 F9 f 14 Ti F-5 T-6 T-7 1 2 3 4 5 6 7 8 9 CMIR 0)RE DIACF).H Eles. Pos U-235 Eles. Pos F 235 Elen. Pos D-235 Mas.s Mass Mass C1 D4 124.86 T-5 F4 224.59 F-14 E8 224.76 C-2 E4 124.88 T-6 15 224.63 C-3 E6 124.88 F-7 f6 224.66 C-4 C6 124.87 T-8 C4 224.66 T-1 ES 224.79 F9 E7 224.67 T-2 D6 224.83 T-10 E3 224.68 T-3 DS 224.79 T-12 D7 224.69 T-4 C5 224.65 T-13 D3 224.74 Total U-235 Mass (Grams) 3420.63 Figuro 3. Coro Loading Diagram for Initial criticality of LEU Core

.= .-- - . .

9 3.2 Coro 100.W and 100T l

Half element HR-1 was added to Grid Position D-8 to increase i coro excess reactivity. The core was designated 100W. Excess reactivity, rod worths, and SDM were measured. The results are shown below in Table 6. The U-235 loading was 3533.04 grams. Figure 4 shows the core loading.

Table 6. Coro 100W Paramotors Paramotor Worth Excess Reactivity 0.450% Ak/k SDM,a 4.59% Ak/k Rod 1 2.52% Ak/k Rod 2 2.52% Ak/k Rod 3 3.25% Ak/k Reg Rod 0.338% Ak/k Similarly, measurement of the excess reactivity and reg rod worth woro mado in the T modo. Table 7 lists the results for core 100T.

Table 7 Coro 100T Paramotors Parameter Worth Excess Reactivity 0.845%. Ak/k Reg Rod 0.330% Ak/k T Column 0.395% Ak/k l-l 5

.- ~ ._ .

10 A TR TO PIEflXLi f - Standard Elements B $ C-ControlElements ET - Half front Element C _

F8 T-4 C4 ER - Balf Evar Elemet.t S - Source Holder D f-13 C-1 T-3 T-2 f-12 M-1 E f-10 C-2 _f-l C-3 T-9 T-14 f i f-5 f-6 T-7 1 2 3 4 5 6 7 8 9 1:Mk? C0il DIACTAM 9l Elen. Pos 0-235 Eles. Pos 0-235 Eles. Pos 0-235 Macs Mass Mass C-1 D4 124.86 T5 F4 224.59 T-14 E8 224.76 C-2 E4 124.88 T-6 F5 224.63 H-1 D8 112.41 0-3 E6 124.88 T-7 f6 224.66 C-4 C6 124.87 T-8 C4 224.66-F-1 E5 224.79 ' F -9 E7 224.67 F-2 D6 224.83 T-10 E3 224.68 F-3 D5 224.79 T-12 D7 224.69 f-4 c5 -224.65 T-13 03 224.74 Total U-235 Mass (Grams) 3533.04 Figure 4. Core Loading Diagram-for Coro 100W-l l;

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11 3.3 Carg_l01)f_ond_101T (Without Rabbits)

On July 28, 1992 element HR-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 Table 9.

Tablo S. Coro 101W Parameters, Without Rabbits Parameter Worth Excess Reactivity 0.874% Ak/k SDM,a 4.34% Ak/k Rod 1 2.63% Ak/k Rod 2 2.58% Ak/k Rod 3 3.30% Ak/k Reg Rod 0.350% Ak/k Table 9. Core 3 1 Parameters, Without Rabbits Parameter Worth Excess Reactivity :1.28% 'Ak/k Reg Rod 0.356% Ak/k T Column 0.406% Ak/k .

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12 A _y FIY TO PLETIIES T - Standard Elecents B 1 0-ControlElements ET-HalffrontElement C fB T _Ll-4 HR - Half fear Element S - Se Tce Bolder D -_.

f-13 C_L F-3 F-2 F-12 f-15 BR - hate Pabbit CR-CadalusRabbit E .E10 C-2 T-1 C3 T-9 T-14 f ,_ _ _

F-5. F-6 F7 1 2 3 4 5 6 7 8 9 (MRR COPE DIACP).M Eles. Pos U 235 Eles. Pos 0-235 Eles. Pos 0-235 Kass Kass Mass C1 D4 124.86 T-5 f4 224.59 T 14 E8 224.76 C-2 E4 124.88 F-6 f5 224.63 F-15 D8 224.77 C-3 E6 124.88 F-7 f6 224.66 C-4 C6 124.87 7-8 C4 224.66 T-1 E5 224.79 F-9 E7 224.67 F-2 D6 224.83 F-10 E3 224.68 T-3 DS 224.79 T-12 D7 224.69 F-4 C5 224.65 F-13 D3 224.74 Tota 1 U-235 Mass (Grams) 3645.40 Figure 5. Core Loading Diagram for Coro 101W (Without Rabbits)

4 13 3.4 Cnrn_10.1H_und 101T (With Rabbits Inserted)

The Baro and Cadmium Rabbit Facilities were installed in the gridplate on July 28, 1992 as shown in Figuro 6. Tablo 10 i presents the measured parameters.

Tablo 10. Coro 101W and 101T Paramotors, With Rabbits Insorted PARAMI7PER 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 /b .

The worth of the thermal column was 0.383% Ak/k. ,

Coro 101W was determined to be tentatively acceptable.

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. l 14 A KTY TO FPITIIES T - Standard Elements B S_._ C-ControlElements HT - Balf front Elenett C T-3 T-4 C-4 ER - b lf Pear Element 5 - Source Bolder D T-13 C-1 T3 f-2 T-11._[15 E f-10 C-2 f-1 03 T-9 f-14 ,_

f CR f-5 -F-6 T7 BP 1 2 3 4 5 6 7 8 9 EKit CORE DIACT).X Eles. Pos 0-235 Eles. Pos 0-235 Eles. Pos 0-235 Kass Mass Rats C-1 D4 124.86 T-5 F4 224.59 T-14 E8 224.76 C-2 E4 124.88 T-6 f5 224.63 F-15 D$ 224.77 C-3 E6 124.88 f-7 f6 224.66 C4 C6 124.87 F-8 + C4 224,66 T-1 E5 224.79 T-9 E7 224.67 F-2 D6 224.83 T-10 E3 224.68 F-3 DS 224.79 F-12 D7 224.69 F-4 c5 224.65 T 13 D3 224.74 Total U-235 Mass (Grams) 3645.40 Figure 6. Coro Loading Diagram for Coro 101W (With Rabbits) i I m- w ---m , . , , , , . ,

15 3.5 AdjuntetLParnautorLI.or_Cnrs_191W.Jind_Coro 10LT i

It should be noted that all reactivity measuromonts mado using the positive parlod method in Sections 3.1 through 3.4 woro ,

based on I,,f=0.00755. Heutronics studios have yielded a  ;

calculated p.,f f = 0 . 0 07 9 for the LEU fuel. This value is presented in Tablo VIII of the SAR. To be consistent, neasured reactivity values using the positive period method need to be adjusted accordingly.

Additionally, the spare fission chamber used for core loading was repositioned from the core side to a now position above and behind the core. The repositioning was measured to have a small positivo reactivity offect of 0.045% Ak/k.

Table 11 procents the adjusted core paramotors for Coros 101W and 101T which account for the now value of J,ff and the movement of the detector.

Tablo 11. Adjunted Coro Paramotors for Coro 101 ( J,ff=0.0079 )

PARAMETER CORE 101W CORE 101T Excess 0.496% Ak/k 0.897% Ak/k Reactivity SDM.m 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 ,

Roy Rod 0.371% Ak/k 0.371% Ak/k fh"1"."Enermal column worth-~iTas 0. 4 01% Ak/k.

. - , . . . , . . , - - , . , - . , , _ , , , , , , - - . . _ , - , . . . . , . , , . . -- .,_n.. - - . - , , , .

16 4.O ExcssE_Reantlyltya_Contral_RosLiior_thn_and Shutdown _liargin llEU Core 67W presented in Figure 7 was similar in geometry to LEU Core 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 margins (SDM) for the two cores.

Table 12. IIEU vn LEU Core Paramotors PARAMETER LEU (101W) IIEU (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 Wortha 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 Table 11, 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 llEU core. This additional loading is 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 octive'" elements while the HEU core had an excess reactivity of 0.43% Ak/k using 16.5 " effective" elements.

2 control elements are treated as one-half of an "offective" element.

_ . -. ._ .. . _ _ . _ . . _ _ _ . _ _ _ ___.. m-_ . - _ . _ . . _ __ _ ..

t 17 A FEYtoPREflI[$

f - Standt.rd Elements B $ C - Control Elements EF - Half front Element C  !!R-1 f-14 T-1 C-4 HR - Half Rear Element S - Source Bolder D f-8 C-1 T-16 F-9 F4 f-10 E T-6 C-2 I-19 0-3 T 12 T-11 i BR F-17 F15 CR 1 2 3 4 5 6 7 8 9 UKRR CORE DIAGP1H fles. Pos 0-235 Eles. Pos (1-235 Eles. Pos 0-235 Xass Xass Mass ER-1 C3 84.912 F-16 DS 170.270 T-12 E7 168.774 T8 D3 170.229 T-19 ES 170.264 T-10 Db 170.193 f-6 E3 169.160 T-15 F5 168.889 f-11 E8 168.969' F 14 C4 170.210 C-4 C6 102.112 ,

C-1 D4 102.112 F-9 D6 170.178 C-2 E4 102.125 C-3 E6 101.978 T-17 F4 169.111 F7 f6 170.154 f-1 C5 170.223 F-4 D7 170.206 Tota 10-235 Mass (Grams) 2870.Q69 Figure 7. Coro Loading Diagram for Core 67W

4 18 This comparison demonstrates that the reactivity worth per

" effective" element in the LEU core is only slightly higher than with the HEU core.

Rod worths in the LEU core are (but not appreciably) dif f erent from the HEU core. The slight shift in rod worths shown in Table 12 may be as much a result of the slight changes in core configuration geometry as with the change in fuel type.

Finally, because excess reactivities and rod worths were very similar for the two cores, then necessarily the shutdown margins are similar.

Overall, the measured core parameters presented in Table 12 did not change appreciably with the LEU conversion, with the exception of the U-235 gram loading.

Neutronics calculations were performed for both the HEU and LEU cores for various core coafigurations. This work is presented in Covington'. Core 101W is very similar to the target core with the following exceptions:

1. The target core had an extra half-element in Grid Position C-3.

2. The positions of the cadmium and bare rabbit facilities were reversed.

Covington predicted an excess reactivity of 0.87% Ak/k. Core 101W has an excess reactivity 0.496% Ak/k. As presented in Section 5, a half element is estimated to be worth approximately 0.424% 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 agreement with the calculated value of 0.87% Ak/k.

Integral rod worth curves fur Core 101W are presented in Figures 8 through 11. As was seen in Table 12, rod worth values changed very little with the fuel conversion. This was as expected.

'Covington, Lorne J., "Neutronics Study of the Conversion of the University of Missouri-Rolla Reactor to Low Enriched Uranium Fuel". M.S. Thesis (1989) University of Missouri-Rolla.

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Core 101W (Total Worth: 3.22% Ak/k)

4 21 5.0 Partial Fuel Element Wort;h ,

The worth of a- partial -(or half) fuel element - at the core periphery _was determined by the change -in core excess reactivity between core 100W and- 101W. This core change involved replacing-the half' element (HR-1) located in Grid-Position D-8 with a full fuel elettent (F-15).

The swing in excess reactivity was 0.424%_Ak/k_(W mode) and '

0.435% Ak/k (T mode).

6.0 Crlt;ical Mass The critical mass of U-235 for the LEU fuel was estimated to-be 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.

The estimated critical mass for the LEU core was determined by comparing the excess reactivities 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. Cort 101W had a loading.

of 3645.4 grams U-235 and excess reactivity of 0.874% Ak/k. -

The reactivity worth per gram of U-235 can be estimated as:

(0.87 4% Ak/A - 0.450% Ak/h) = 0.003 8 % A k/ k,

( 3 6 4 5. 4 - 3 5 3 3 . 0) grams gram The initially critical LEU core contained 3421 grams- of ~U-735 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.=

3409 grams U-235.

The critical mass 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 f erent core configurations

• have shown that -the worth of a fuel element (170 grams U-235) is between 0.5% and.-

1.5% Ak/k -- depending on . 'its position at the core periphery. .

Taking an _ AVE. rage of 1.0% Ak/k and using -the same methodology as above, the reactivity worth per gram of U-235 "can be estimated as:

'Saf ety Analysis Report for the University of Missauri-Rolla Reactor. Docket Number 50-123, Page 9-13. (1984)

y - . .- - .- .

.,- e ',

-22 1.0% Ak/k = 0.00588 %Ak/k gram 170 grams The HEU core 67W thus had a surplus of about 73 grams.of iU-- ,

235 above the critical mass. Therefore, the tinimum critical-mass of U-235 for the HEU fuel is approximately 2,797-grams.

7.0 Egwc G A11bration Af ter low power core characterization, the reactor operated at an indicated power of 1 kW on the Linear Channel for one hour.

The core was checked and the operation appeared normal.

A power calibration was perf ormed on August 4, 1992. The procedure involves operating the reactor at an intermediate indicated power of 40 kW for one hour. The resulting " heat balance" is obtained by measuring thermal expansion of_-the poc_. The power calibration procedure (Sol 816) is_provided as Attachment A. Detectors had to be positioned slightly clocer to the core.as a result of th.e calibration._..It was-expected that nuclear instruments would read a bit high due to-the higher core leakage associated. with : the harder flux spectrum. A second power calibration, performed on August'5, 1992 verified that the detectors were correctly positioned.

m .

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23 8.0 ypid coeffIclent;of Reactivity T

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 scaled. 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 y ) is calculated by taking the ratio of--the-change in reactivity to the effective volume of the void:

g*, b Paittilled ~6Vol 9 water fL D ed The volume of the tube is 1300 cm'.

The average void coefficient measured for the new LEU Core ,

101W was -7 .1 X 10-5 ( tak/k) . The previously measured value Cm for HEU Core 67W was -6 . 9 X 10 ~5 ( % Ak/k) . The . void cm) coefficient for the LEU core .appearc to. be slightly more negative than the HEU core. This trend is as expected-due to the harder flux spectrum. In fact, measurement showed that the void coef ficient 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 that the_ cores geometries are not identical, as mentioned in Section 4.

Neutronic calculations

• for the _ void coef ficient of reactivity were also performed - for both proposed HEU and LEU cores. .

Table 14 presents measured and calculated values.

'Covington, Lorte J,, "Neutronics Study of the. Conversion of-the - University of Mi* ao iri-Rolla Reactor to Low Enriched Ur'anium -

Fuel". M.S.-Thesis ( aJ), University of Missouri-Rolla.

n a w <.

t 2 24

Table 13. Measured Void Coefficients as n' Function of Selected Position for LEU Core 101W and HEU Core 67W.

POSITION- LEU CORE IIEU CORE -

-( % Ak/k-cm*) (% Ak/k-cm')

B-6 -8. 62 x 10-* -7.25 x 10-*

C-8 -5.75 x 10'* -4 . 9 0 x 10**

D-9 -5.44 x 10-* ~ 4 . 8 0 x 10-*

E-9

-5. 69 x 10** -4. 4 5 x 10-5 Table 14. Heasured and Calculated Values -

for Void Coefficient (*ON/N) cm' DESCRIPTION llEU CORE LEU CORE Heasured -6 . 8 5 x 10-* -7.08 x'10-*

Calculated -7.0 x '0~' -9.0 x 10

The calculated value for the LEU void coefficient is about 28%

more negative than for the HEU fuel.- C o m p a r i n g -- m e a s u r e d values for selected locations presented in Table 13 shcwed that the LEU void coefficient ranged from about 13% to.28%'

more negative than the HEU values.

c+ ..

i.

25 >

9.O TgancratMrc_Coe.ffigient of Reactivity ,

The temperature coefficient of reactivity will be measured in the Winter, 1993 semester as a Senior -class Nuclear.

Engineering project. It was planned to perform the experiment during the Christmas break of '92/'93; however,~ required reactor naintenance 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 completi.on of the core flux profile measurements.

11.0 Delaved Nggtron Fraction The new value of B.,, is 0.0079 as presented in -the SAR. This is about 4.6% larger than the previously assumed . value of 0.0075 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 tho' HEU cores. Operationally, the LEU core behaves very similarly to'the HEU cores.

At present, we have assigned two Senior-Nuclear Engineering-students to explore-ways of determining kinetics parameters with computer noise analysis. This will involve interfacing a computer data collection to the isolated outputs of our newly (and as yet not installed) nuclear instrument drawers.

We feel this will be a suitable topic for an M.S. Thesis.

.-4 s

9 26 12.0 CQDelusions The receipt and inspection of the new elements went smoothly-and all elements we.re determined to be satisf actory. Initial core leading to criticality went smoothly with 1/M data 4 correctly predicting criticality.

The target core configuration based on neutronics calculations dif fered from the actual final core configuration by only one-half element. Excess reactivity, control rod worths and shutdown margin for the new core are very similar to those in the HEU cores . As expected, the critical mass in the LEU core 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 consistent with calculations.

Sti'l to be further characterized are the temperature coefficient, flux distribution, and kingtics parameters which .

are to be characterized in the near fuEtfre.

W No abnormalities or surprises have been uncovered with regard to the new fuel. We plan to ship out our remaining balance of HEU fuel as soon as possible. We hope this fuel will last us well into the 21st century.

f' p

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.4 .

27-ATTACilMENT A SOP 816

" POWER CALIBRATION" J

28 _!

• a* UMR REACTOR. STANDARD OPERATING PROCEDURES ***

SOP: 816 Title UMRR POWER CALIBRATION Revivedi August 30, 1988 Page 1 of 4 A. Purcose To ensure that the power indicated on the linear and log ch anne ls in the power generated in the reactor.

B. Precautione. Prerecuisites. or Limitationst

1. In accordance with Technical Specification 4.2.2(3) all console instruments and safety system shall be RD*

calibrated twice each year, not to exceed / months.

T/L F1

2. The power generation in the UMR Reactor is limited by Technical Specificatior.s to 200 kW. It is, therefore,g,73s%S ofb-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 in performed by the calibration procedure decer ibed below . (For more details see the report UMRR/85-i.) Stable atmospheric condition are helpful for 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 divinion-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 to determine the average pool level dr op . Be sure to note accurately the time of each reading.

Record also the temperature of the pool water using 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 reaches .

that power level. After running the reactor at this power for a time to such that the reactor thermal out-put is between 30 and 50 kW hr . shut coun the reactor and note the shutdoun time. For-example, it is recom-Written By Milaa Straka Approved'By: Albert- lon 0

e

w- ,

29 a** aa*

UMR REACTOR STANDARD OPERATING PROCEDURES SOP: 816

Title:

UMRR POWER CALIBRATION -i Revised: August 30, 1988 Page 2 of 4 mended that the reactor power be chosen 40 kW and the 4 operational time t. I hr.

5. Once all control rods and magnets are fully inserted, note time and pool level every 15 minutes until level decreases equel the rate of decrease bef ore the power run. During this time also continue to take tempera-ture readings using all reactor thermocouples.

6. Plot the data measured with both relative height 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 units of cm for the plot of h.)

7, Determine ah as shown in the sketch below h [cm] .

A N

ah I's jf  %.

= Time = Time t [hr]

at at Power Shutdown l- t p -l

8. Calculate the average pool water tempere'iire T. using the data taken immediately bef ore the bewaning of the power run and after the reactor shutdown. (Use only the ir ' et- temperature readings. )

Written By: Mila Straka Approved By: Albert Bolon

/bt3 (WAk

, s. ,

30 a== *aa UMR REACTOR STANDARD OPERATING PROCEDURES SOP: 816 Tit 1e: UMRR POWER CALIBRATION Revisedi 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-tot 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 hiaher than the average pool temperature has already been +Then into account while constructing the plot- in Firsure 1.)

10. Ca;culate the reactor power using the relationship P(kul =0 t guhr L

t. ihr)

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 satisiy the above condition. pp I

f'og N) have been zlujeq(,dA

12. Af ter both power channels (linear and properly adjusted take the reactor to 100 kU and adjust, if necessary, both safety channels so as to in-dicate the reactor power of -100 y h" N 2 ccH W/ W 3 Z/2e/92.((f.

Written By: Milan Straka. Approved By: Albert B lon L,

4

,, 4

s 31'

• • • UMR REACTOR ST AND ARD OPERATING PROCEDURES * *

• SCP: 816 Tit 1e UMRR POWER CALIBRATION Revised: August 30, 1983 Page 4 of 4

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