ML20082N914

Forwards Estimated Uncertainity for Univ of VA Reactor Reactivity Measurements

ML20082N914
Person / Time
Site:	University of Virginia
Issue date:	04/19/1995
From:	Mulder R VIRGINIA, UNIV. OF, CHARLOTTESVILLE, VA
To:	Alexander Adams NRC
References
NUDOCS 9504250447
Download: ML20082N914 (8)

Text

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April 19,1995 & APPUED SCIENCE DEPARTMENTOFMECil4NICAL, AEROSPACE (

Mr. Alexander Adams AsouvCtc4aEscisEEnzsc U.S. Nuclear Regulatory Commission Mail Stop 10-D-21 university of virginia Thornton Hall Washinbton D.C. Charlottesville, VA 22903-2442 20555 go4/924-7421 FAX: 804/982-2037 TDD: 804/982 HEAR Phone: (301) M4-1127 NRC Facsimile: (301) 415-2260 Phone then Fax: (301) 415-1032 Re: Estimated uncertainty for UVAR reactivity measurements

Dear Mr. Adams,

In reply to (INEL) Mr. James Miller's request for reactivity coefficient values and uncertainties associated with the LEU UVAR core, we have extracted pertinent information listed in the IIEU to LEU Conversion Report, UVN527367/ MANE 95/101, dated October 1994. The reactivity measurements furnished in attachment can be accepted at an estimated +- 10% uncertainty level.

I am still trying to obtain supporting information concerning the HEU-18 Expt data listed in Table 3 of the " Safety Analysis for the University of Virginia Reactor LEU Conversion" paper presented at the XII International RERTR Meeting, in September of 1989. It should be noted that an LEU fuel element has the same external dimensions, and same plate dimensions, as that of the old HEU element. However, LEU elements have 22 fuel plates, rather than 18 plates, each.

Sincerely, W'W Robert U. Mu der, Director U.VA. Reactor Facility enc: HEU and LEU reactivity parameters

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,,J SHUTDOWN MARGIN - 1.2 % delta k/k June 13. 1994

.- Date EXCESS REACTIVITY + 3.27 % delta k/k U-235 4807 GRAMS EXPERIMENT WORTH 1.63 % delta k/k I

F - Normal Fuel Element P - Grid Plate Plug

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v. PF - Partial Fuel Element HYD RAB - Hydraulic Rabbit

{ CR - Control Rod Fuel Element THER RAB - Thermal Pneumatic Rabbit

'a- G - Graphite Element EPI RAB - Epithermal Pneumatic Rabbit S - Graphite Source Element RB - Radiation Basket F'

REG - Control Rod Fuel Element with Regulating Rod Er Rod Vorths #1 - 2.80 % #2 - 2.63 % #3 - 1.84 % Ree - 0.401 %

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MINERAL IRRADIATION FACILITY

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, P VS-015 VS-009 VC-001 VS-013 VS-004 C P 11 12 13 14 15 16 17 A 18 N

F F-CR1 PF F F I P VS-006 VC-002 VP-001 VS-007 VS 008 S P 27 T 21 22 23 24 25 26 28 E

F F F. F-CR2 F R G VS-001 VS-010 VS-011 VC-003 VS-012 P 31 32 33 34 35 36 37 I 38 R

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• THER HYD C G RAB G G RAB G G 71 72 73 74 75 76 77 78 G G G G G G G G 81 82 83 84 85 86 87 88 Figure 9. LEU-2 Core Loading Diagram W -

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The excess reactivity of LEU-2 was measured experimentally to be + 3.27 %

Ak/k, well below the + 5.0 % Ak/k limit required by the Technical Specifications.

i-b) Experimental Facilities m-Before operating at any appreciable power, various experimental facilities were installed in the reactor and the reactivity worth of each facility was determined with the following results:

Exnerimental Facility Worth (%Ak/kT

'" Epi-Thermal Rabbit, grid position 64 - 0.01 Thermal Rabbit, grid position 73 - 0.01 Hot Thimble #1, with specimens, grid position 53 - 0.28

" Hot Thimble #3, without specimens, grid position 55 - 0.54 Canister Irradiation Facility, east side of core - 0.04 The sum of the absolute worth of the experimental facilities is 0.88 %Ak/k, well below the 2 %Ak/k limit for all experiments required by the Technical Specifications.

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c) Approach to Full Power The initial approach to power with the LEU-2 core was begun on the afternoon of May 12,1994. The power was leveled out at 200 kW,500 kW,1 MW, and I i.

finally 2 MW to check instrument response and to perform radiation surveys.

Full power of 2 MW was achieved at about 5:20 P.M. Radiation surveys at the reactor bridge and around experimental areas showed no increase in radiation levels as compared to the HEU core.  ;

., d) Power Coefficient Measurement The reactor was taken critical at low power in a xenon-free condition and the critical rod positions were noted. Rod #2 was withdrawn about 0.5 inch to put

" the reactor on a positive period. Doubling times were measured with a stop watch using the linear instrument. The power was allowed to rise until it leveled off at 870 kW due to negative temperature effects. The average temperature rise across

" the core was noted. The doubling times were converted to reactivity and matched ahnost exactly with the reactivity worth of rod #2 when it was withdrawn as ,

i determined from the rod worth curve. The power coefficient was experimentally verified to be -0.139 %Ak/k/MW or in terms of the change in average core temperature - 0.0275 %Ak/k/F. It is noted that this measurement includes several effects such as the fuel doppler, fuel expansion, and moderator temperature i g

1 coefficients. The moderator temperature coefficient was measured separately.

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y e) Moderator Temperature Coefficient Measurement The reactor was operated all day at full power on May 13,1994 and the pool temperature was 97.9 'F when the reactor was shut down. The cooling

~ systems were secured and the reactor remained shut down for three days to allow Xenon to decay. The reactor was taken critical at low power on May 16,1994.

The pool temperature was 86.9 'F and the AT was 0.0 'F. The secondary cooling system was energized and the pool was allowed to cool down. The power level was maintained by adjusting Rod #2. The pool was cooled for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The Core outlet temperature changed by 11.24 *F. Rod #2 moved from 14.39 inches at the beginning of the test to a final position of 14.16 inches. The reactivity

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associated with this change was obtained from the rod worth curves and determined to be 0.05 % Ak/k.

~~ Rod #2 was then moved back to it's original position of 14.39 and doubling times were taken to determine the reactivity associated with the rod movement as a

, check against the rod curves. The reactivity associated with the doubling times was 0.039 % Ak/k. Due to incomplete mixing from the cooldown the moderator l, ,j temperature increased 2.06 F during this measurement giving a temperature 3 differential 9.18 *F as compared with the original temperature. Ar average of

_ _l these measurements yielded a value of- 0.0044 %Ak/k/ F for the moderator temperature coefficient.

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_ f) Experimental Facilities Between May 23,1994 and June 10,1994, several other experiments were loaded in the reactor, giving the following experiments and experimental facilities now in the reactor.

Worth (% Ak/k) l

1. Epi-Thermal Rabbit, grid position 64 - 0.01 7 2. Thermal Rabbit, grid position 73

- 0.01

3. Hot Thimble #1, with specimens, grid position 53 - 0.28

_ 4. Hot Thimble #3, without specimens, grid position 55 - 0.54

5. 1.2 kg of Fe samples in H.T. #3 + 0.05

6. Cannister Irradiation Facility (East side of core) - 0.04

- 7. MIF Stand, front of core on pool floor 0.00

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8. MIF Lead Shield, front of core on MIF stand + 0.33

9. MIF Box, Model 3, #2 (Gas cooled), North side of core - 0.37 w

l The sum of the absolute worth of the experimental facilities is 1.63 %Ak/k, which  !

. is below the Technical Specification limit of 2.0 % for all experiments.. '

20

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1 VI. Comparison Between LEU and IIEU Cores The LEU-1 core was similar to an HEU core operated in 1975. They were both

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4x4 configurations surrounded by graphite. Due to the increased U-235 content of the LEU fuel, with changes from 18 to 22 plates per element, partial elements were loaded in grid positions 23 and 26 in the LEU core. Comparisions of the two cores are as follows:

, HEU Core LEU-1 Core (2.73 kg U-235) (3.57 ke U-235)

Rod Worth (% Ak/k)

Rod #1 3.56 3.46

~~

Rod #2 3.75 3.86 Rod #3 2.30 2.41 Reg Rod 0.428 0.375 3 Shutdown Margin (% Ak/k)

__ ] (Rod #2 and Reg withdrawn) -2.7 - 1.21 Excess Reactivity (% Ak/k) + 3.35 + 4.66

~~~

The control rod worths that were computed for both the HEU and LEU-1 cores were in good agreement with these measured values.

A comparison of the measurements made with the LEU fuel and predictions in the LEU SAR, and HEU comparisons, where available, are presented below.

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The " Power" temperature coefficient was measured during a core-heat-up experiment. It is expressed as the change in Ak/k per average change in core temperature. This is not

_ the same conceptually as the isothermal moderator temperature coefficient, which was measured during a pool-cool-down experiment at low power.

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LEU Core L EU-SAR HEU Core Power Coeff. (% Ak/k/MW) - 0.139 NA NA

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" Power" Temperature

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Coefficient (% Ak/k/'F)(LEU-2) .0275 NA .0288 Moderator Temp. Coefficient

(% Ak/k/'F) (LEU-2) .0044 .0082 NA

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~ Void Coefficient (uniform)

(% Ak/k/% void)(LEU-1) - 0.19 - 0.22 NA The most significant observation is that the " Power" temperature coefficients for the old HEU and new LEU-2 cores are almost identical. signifying that the safety characteristics

_ of the two cores are essentially the same.

The isothermal moderator temperature coefficient value is clearly different from the

" Power" temperature coefficient value, but it was not expected to be comparable. The

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j measured isothermal moderator temperature coefficient is about a factor of 2 smaller than the value calculated for an ideal unrodded LEU core model. Considering experimental uncertainties in the measurement, and the lack of partially inserted control

_ rods in the computational model, the agreement is judged to be acceptable.

L-- The void coefficient measurements lie in the range of the calculated void coefficients.

Local experimental differences are expected te be caused by the presence of partially inserted control rods in the core.

VII. Shipment of All IIEU Fuel Elements From Facility All HEU fuel (32 elements) was shipped to Savannah River in the BMI-1 \

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" shipping cask in three shipments made on May 4,1994, June 6,1994, and

"_ July 8,1994, respectively.

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