Text

Initial Loading & Testing of Low-Enrichment U Fuel in Ohio State Univ Research Reactor
	ML20248J091
Person / Time
Site:	Ohio State University
Issue date:	10/03/1989
From:	Redmond R; OHIO STATE UNIV., COLUMBUS, OH
To:	; Office of Nuclear Reactor Regulation
References
	NUDOCS 8910130036
	Download: ML20248J091 (31)
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T- H E Engineering Experiment Station 142 Hitchcock Hall 2070 Neil Avenue O,,]y,; Columbus, OH 43210-1275 4 #~m Phone 614-292-4903

-d FAX # 614-292-9021

.. -e UNIVERSTFY October 3, 1989 Director, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Sir:

please find enclosed the reactor start-up report for The Ohio State University Research Reactor, Docket No. 50-150. This report is being submitted as required by our Technical Specifications, Section 6.6.2(4) and as requested in Item #1 of Enclosure 4 of the " Issuance of Order Modifying l

License No. R-75 to convert from High to Low-Enr.iched Uranium - Amendment No. 12 - Ohio State University dated September 27, 1988. If you have questions on the content of this report please contact Mr. Richard Myser, Associate Director of the Nuclear Reactor Laboratory or Mr. Joseph Talnagi, Senior Research Associate (614-292-6755).

Sincerely,

-/sa2 Robert F. Redmond, Director Engineering Experiment Station Enclosure c: with enclosure Nuclear Regulatory Commission Region III Office of Inspection and Enforcement 799 Roosevelt Road Glen Ellyn, IL 60137 l Don W. Miller, Director l The Ohio State University l Nuclear Reactor Laboratory 8910130036 891003 kOY l i

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I Initial Loading and Testing of Low-Enrichment Uranium Fuel in The Ohio State University Research Reactor A Report Submitted To The U.S. Nuclear Regulatory Commission By The Ohio State University Nuclear Reactor Laboratory 1

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1.0 Introduction and Background The contents'of this report describe the activities related to the conversion of The Ohio State University Research Reactor (OSURR) from utilization of high-enrichment uranium (HEU) to low-enrichment uranium (LEU) fuel. Major activities included storage, in preparation for shipment, of the HEU material, receipt of the LEU fuel, loading of the initial LEU core, and subsequent startup testing and calibration. These activities will be described in detail in'the following sections and will include discussion of the results of the startup testing and comparisons with previous results from testing of the HEU core, and with calculated ~

LEU core parameters.

The OSURR is licensed to operate at a steady-state thermal power output from shutdown to 10 kilowatts (KW), as authorized by the Nuclear l Regulatory Commission (NRC) in facility license number R-75, docket number 50-150-. Initial operation began in 1961, with the facility being operated on an on-demand duty cyclc, with no fixed total hours of i operation per week or year. Average yearly energy output is about 2000 kilowatt-hours (KWH). Typical facility utilization involves, among other things, instructional and research activities such as demonstration of reactor operation, measurement of various core parameters and operating characteristics, irradiation of materials for activation analysis or damage studies, and demonstration of neutron radiography techniques.

The OSURR was initially fueled with HEU fuel elements, supplied by the reactor vendor (Lockheed Corporation, Nuclear Products Division),

with ownership of the fuel retained by the Department of Energy (then the Atomic Energy Commission). The HEU fuel elements were inspected periodically, with no failures observed or detected. Total HEU fuel burnup through 1988 was estimated to be slightly less than 2 grams.

Total on-site inventory of HEU at the end of 1988 was about 3575.81 grams.

In July 1985 funding was obtained from DOE to begin conversion of the OSURR to LEU fuel. Analyses were performed to specify an LEU core design using " standardized" LEU fuel, which specified a nominal uranium enrichment of 19.5%, and a fuel matrix based on uranium silicide-aluminum (US S19 -A1) dispersion. Nominal U-235 loading was specified as 12.5 grams 7 fuel plate, with a uranium density of 3.47 grams /cc. An 18-plate standard fuel element was specified, with the control rod fuel elements having 10 plates. The standard fuel element would have 16 fueled plates, with two " dummy" plates of aluminum as the outside plates. The LEU fuel elements were specified to be outwardly identical to the HEU elements.

No modifications to the existing control rod drive system, grid plate, experimental facilities, or associated core structures were planned.

Results of the core analyses were used to write a Safety Analysis Report (SAR) and Technical Specifications for the OSURR operated with LEU fuel. The SAR was received at the NRC offices in October, 1987.

Questions related to this SAR submittal were received in February, 1988, and those questions related to the fuel change were answered and returned to NRC in April, 1988. Subsequently, NRC issued an order, dated September 27, 1988, to convert the OSURR to LEU fuel. Information 1

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- relatedLto thisLorder was published in the Federal Register.in the beginning of October,' 1988; After.the required 30 days had. passed, the staff of the OSURR facility. began. conversion activities;in November, 1988. .r

-The LEU fuel elements 'were received at the OSURH facility site on.

11/8/88. .These elements were visually inspected and placed in temporary storage at;the OSURR.pending placement int,0 the' reactor pool storage pit; HEU-fuel unloading began on 11/9/88 and was completed on 11/28/88.

Removal and storage of the HEULfuel elements took longer than expected rJ because of minor mechanical modifications.(not safety-related) required on.the two HEU fuel element storage racks. Loading of the LEU core began on 12/7/88. The initial core geometry, denoted as core LEU-1, was' fully loaded on 12/15/88 ,with'criticalityLattained at 14:53 EST on that date.

Testing followed, with a. subsequent small reduction in' total uranium loading for the purpose of excess reactivity adjustment. This modified core loading was denoted as core LEU-2.

2.0. Procedures and Methods 2.1 HEU Fuel Unicading and Storage

. Handling of the. irradiated HEU fuel elements was done in accordance l' with ' approved fuel: handling procedures. A shielded transfer cask, .

designed and fabricated by the University, was used to contain individual HEU' fuel elements as they were moved from the reactor pool to the storage pool. Appropriate radiation safety precautions were observed, consistent with the As Low As Reasonably Achievable (ALARA) policy of the University. Individual dose monitoring devices were worn by personnel involved in the fuel handling and transfer operations. These dosimeters included whole-body monitors such as-film badges and self-reading pocket dosimeters, as well as extremity monitors such as TLD-type ring badges.

Subsequent readout of personnel dosimeters indicated that the ALARA l.

approach employed was effective as no significant exposures were recorded for any person involved in the fuel handling operations.

. Portable radiation survey instruments were also used to measure doses from irradiated HEU fuel elements when contained in the transfer cask.

Maximum cask-surface dose rates for gamma radiation were generally less than 50 millirems / hour, with the maximum measured dose rate being about 80 millirem / hour for the most heavily-loaded HEU fuel element (denoted LNP-Z-1) .which was also positioned at or near the flux peak of the HEU-fueled OSURR core for most of its operational history.

Upon completion of the HEU fuel transfer, the storage pool was L . secured. The pool is periodically monitored for pool water radioisotope content, fuel-inventory, and pool security. HEU fuel element shipment is pending, dependent upon shipping cask availability, NRC approval, and DOE receiving facility schedule.

I 2.2' LEU Fuel Loading LEU fuel elements were first placed in the fuel storage pit at the east end of the reactor pool. Single fuel elements were moved from the 2

storage pit to designated locations in the grid plate. Existing handling l tools and procedures.were used for this operation.

An approach-to-critical experiment was performed to determine critical core size, excess reactivity, and shutdown reactivity margin.

Several" instrument channels were used to calculate the inverse multiplication factor for a series of core loadings. The inverse multiplication factor was plotted as a function of core loading and used to predict successive fuel element additions. Shutdown margin and excess reactivity were estimated after the final fuel addition.

Initial startup source count rates were obtained first. The OSURR startup source, a 5-curie Pu-Be source, was positioned in its normal startup location. Two pulse-mode instrument channels, both utilizing fission counters, were used to obtain initial count rates. Because of the relatively low source count rates, overnight counts were taken over several days. Fuel loading commenced after these initial rates were established with reasonable certainty.

Figure 1 shows a plan view of the OSURR grid plate. Fuel loading procedures require that the control rod fuel elements be loaded first.

Shim safety rods were placed in positions 2B, 2D, and 4D, with the regulating rod placed in positien 4B. The CIF is located in the central position, 3C. Neutronic analyses done earlier indicated that a reasonable core size would be obtained if the grid plate corner positions lA, 1E, 5A, and 5E, as well as the entire row 6A-6B were left unfueled.

These unfueled positions were eventually filled with plugs similar in overall shape to fuel elements, but with no fuel or dummy plates in their interior. For the initial core loadings, however, these unfueled grid plate positions were simply left unfilled until some fuel was loaded.

After loading the control elements, the control rods were re-installed, and an initial series of tests were conducted to assure that the control system was operating within Technical Specification limits.

These tests include rod withdrawal and insertion rates, drop times from various heights, and rod release times. Control rod magnet tests were also performed to assure that various magnet parameters had not changed.

Count rates were then measured for the core configuration with only the control elements and the CIF in place. Readings were taken with the control rods fully inserted and fully withdrawn. Fuel was added in amounts predicted from the inverse multiplication curves. using the curve from the instrument channel predicting the smallest fuel addition (i.e.

most conservative). Initial fuel additions were made in amounts equal to one-half that needed to attain criticality with the control rods fully withdrawn. As criticality was apnroached, single fuel elements were loaded. Table 1 shows the loading order and U-235 mass at each loading.

In this table, loading number 0 denotes just the startup source and CIF installed, while loading number 5A indicates installation of the filler plugs noted above.

Inner grid plate positions were loaded after control rod positions.

Following these loadings, positions forming a symmetric core geometry were loaded, where possible, filling relatively high reactivity worth 3

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[ position, near shim safety rods first. The relatively.Iow-worth position, 58, was-designated as the.finalLposition to be loadedi As more fuel was loaded-into the core and the neutron' multiplication increased, l useful information was obtained on other' instrumentation channels f utilizing current-type ionization chambers, When available, the information from.these channels was included withlthe inverse

multiplication graphs drawn for the pulse' mode' fission counter channels.

As fuel elements werelloaded, estimated critical loading was determined to be 3479.63 grams of U-235. The closest approximation to this loading was to have all grid plate positions. filled with standard

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fuel elements.(except for the CIF and control rod locations) excluding

.This'gave a loading of- 3496.91 grams of U-235, and an.

f position 5B.~

estimated excess reactivity of 0.119% A k/k ($1~ 00=0.0075, as specified by the OSURR vendor), which meant that the reactor was barely critical-with the control rods mostly withdrawn. Critical banked control rod. ,

height for this loading was 58.7 centimeters. Since the OSURR attained criticality with this loading, the core geometry was designated LEU-0.

Since it is undesirable to operate'the OSURR with such a low reactivity excess, additional. fuel was added to the core. A standard fuel, element was' loaded in position 6B, and various tests performed.- The loading.for this core configuration, denoted as LEU-1, was 3696.73 grams of U-235. The excess reactivity fc< LEU-1 was measured to be = 1.5% A k/k, with a total- shutdown margin of 5.7% A k/k. with a minimum shutdown margin (most reactive' shim rod and regulating rod fully withdrawn) of 2.6% A'k/k, both of which are within Technical Specification limits.

-It-was decided to adjust the excess reactivity downward,.since that of; core LEU-1 was close to the Technical Specification limit. Core loading was reduced to 3621.81 grams of U-235 by placing a partially loaded fuel element (124.90 grams of U-235) into pos.ition 5B. This configuration, denoted as core LEU-2, has a measured excess reactivity of 1.2% A k/k,-with a total shutdown margin of 6.0% 6 k/k, and a minimum shutdown margin of 2.9% A k/k with the regulating rod and maximum worth control rod fully withdrawn. Again, both of these values are within acceptable limits. LEU-2 is the current standard core configuration.of the OSURR, pending further testing and modification. ,

Core configurations LEU-1 and LEU-2 were used to measure control rod worths. Rod worths were measured using standard methods including suberitical multiplication, positive period, and rod drop. Table 2-summarizes these measurements.

A measurement of the OSURR transfer function was made using a reactivity oscillator. The $/t cutoff frequency was estimated to be about 16.8 Hz, or 105.6 radiana/second. The transfer. function measurement uses the regulating rod calibration curve obtained during control rod calibration testing, which, for calibration by the method of positive period assumes a period-reactivity relationship shown in Table 1-18 of ANL-5800 [1]. Therefore, this'is not a unique measurement for j the $ of the LEU core. A delayed' neutron fraction of 0.0072 is obtained j from the transfer function, which, if used with the cutoff frequency i estimate, yields a prompt neutron lifetime of 68.2 microseconds.

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The worth of a graphite isotope irradiation element (GIIE) placed in position 6C was measured using the change in regulating rod critical height and the regulating rod calibration data. The worth was found to be +0.21% A k/k.

Worth of fuel in position 5B was estimated from epproach-to-critical data. The excess reactivity of core LEU-0 (3496.91 grams of U-235) was estimated to be 0.12% A k/k, while adding *199.82 grams of U-235 to obtain the LEU-2 configuration resulted in an excess reactivity of 1.5% A k/k.

This gives a worth of 0.0069% A k/k/ gram of U-235 at position SB.

The moderator temperature coefficient of reactivity was estimated by observing the change in reactor period for both slow and fast power transients, and measuring the temperature of the water rising above the top of the core. This enables one to separate moderator temperature from fuel heating effects. The moderator temperature coefficient of reactivity was estimated to be -6.19x10 Ak/k/jC.,whichsatisfiesthe Technical Specification limit of at least -2x10 A k/k/0C.

The void coefficient of reactivity was measured by inserting air-filled plastic inserts of known volume between fuel plates at various locations in the core. The void coefficient avegaged across the core per 1% voidable core volume was found to be -7.9x10 A 3

, m ets the Technical Specifications limit of at least -1.8x10 A k/k.

Thermal power output was calibrated using calorimetry techniques.

pool water temperature as a function of operating time was observed over the course of eight hours. Heat losses were reduced by placing insulating material (styrofoam blocks) over the surface of the pool.

Uniform pool temperature was assured by operating the water purification system pumps during the operation. Temperature sensors at various locations in the reactor-pool system were monitored. Observed power output on the nuclear instrument channels was then matched to that calculated from thermal data.

3.0 n Measurements a.n.d Results

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1 3.1 Critical Mass Based on the original Hazards Summary Report (HSR) for the OSURR, the minimum critical mass using HEU fuel is 2840 grams of U-235 [2]. For the LEU-fueled OSURR, the measured minimum critical mass is 3479.63 grams j of U-235. This was determined as outlined below. j The reactor was barely critical with the control rods at 58.7 cm !

withdrawal. The total control rod length is 61 cm. Excess reactivity ]

for this 3496.91g loading was determined to be = 0.119% A k/k. Excess reactivity for a 3696.73g loading was determined to be = 1.495% A k/k.

The difference is = 1.376 A k/k/199.82g. This equals .0069% A k/k/g.

Therefore the initial LEU critical core had approximately 0.119 %

0.0069%/g = 17.28g of U-235 5

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in excess of critical, i I

i' As far as.we have determined there were no calculations performed to explicitly predict the expected critical-mass for either the HEU fuel or LEU fueled cores. ;

3.2 Excess Reactivity HEU Standard Core #1, loaded to 3181.'07g of U-235 in-August, 1988, i had a measured excess reactivity of = 0.555% A k/k. LEU Standard-Core

1, loaded to 3696.73g of U-235 on 12/20/88, had an excess reactivity =

1.495% A k/k. LEU Standard Core #2, loaded to 3621.81g of U-235 on 1/3/89, has an excess reactivity = 1.203% A k/k. These were determined by subtracting the rod worth at the minimum critical height from the total rod worth (Table 3).

No calculations were made for the HEU core for excess reactivity. j In the HEU-LEU Fuel Conversion response to question 14 dated April 19, 1988, a core with a nominal loading of 3700g was predicted to have an excess reactivity of 1.5% A k/k. The actual loading of 3G96.73g gave a measured value of 1.495% A k/k.

Since Technical Specification 3.1.1(1) limits excess reactivity to 1.5% A k/k including installed experiments under any operating condition it was decided to reduce total excess reactivity. This would allow the flexibility of positive reactivity experiments not available with LEU #1 core. Knowing that the last loading of 199.82g into position B-5 was worth about 1.376% A k/k or 0.0069%/g, an estimate of a partial element (124.9g of U-235) reactivity in the same position was made. This was about 0.862% A k/k. Added to the already existing estimate of 0.119% A k/k for the 3496.91g core the LEU #2 core was estimated to have about 0.981% A k/k. After control rod calibrations were completed, the measured value was about 1.2% A k/k.

3.3 Control Rod Worth Table 2 lists the results of the control rod calibrations and compares them with the rod worths from HEU Standard Core I, and with predicted values for LEU Core 1. Control rod calibration curves are attached for the last HEU core calibration. These were determined by full-length rod drops for the three shim safety rods and by positive period for the regulating rod. The curves for LEU #1 and #2 were determined from rod drops at the minimum critical height for each shim safety rod and positive periods for the upper part of each shim safety rod and the regulating rod. Calibration curves for these two cores are (

also attached. Predicted control rod worths for core LEU-1 are obtained from Table B-1 of the responses sent to NRC on 4/19/88.

3.4 Reactor Power Calibration Thermal power output was determined using calorimetry techniques, as has been done in previous power calibrations. Pool water temperature as a function of operating time was observed over the course of eight hours.

Heat losses were reduced by placing insulating material (styrofoam ,

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blocks) over the surface of the pool. Uniform pool temperature was assured by operating the water purification system pumps.during the operation. Temperature sensors at various locations in the reactor-pool system were. monitored. Observed power output on the nuclear instrument channels was then matched to that calculated from thermal data. No changes in instrument setpoints were made. Detector positions changed relatively little as a result of the power calibration.

3.5 Shutdown Margin In the Technical Specifications for our HEU core, the requirements for shutdown. margin were at least 6% in' total reactivity and at least 1%

when the maximum worth safety rod and the regulating rod were fully withdrawn. The requirements for the LEU Core operated at 10KW are simply

.1% with the maximum worth shim safety and regulating rod full out (T.S.

3.1.1(2). . Table 3 shows a summary of the control rod worths measured for cores LEU-1 and LEU-2, as well as those for the most recent HEU core calibration, and the shutdown margin calculations. In this table, the column headed " Rod Worth At Minimum Critical Height" indicates the amount of reactivity available for insertion with the reactor critical, i.e.,

the total figure on the bottom line for each core geometry is the total shutdown margin for that core. The right-hand column in the table shows the calculation for minimum shutdown margin, starting with the total shutdown margin and deducting the worths of the most reactive shim safety rod and the regulating rod. In all three cases the Technical Specification requirements of a 1% shutdown margin with the most reactive !

rod.plus the regulating rod full out are met. The most reactive rod is Shim Safety #1 for all three cores. Calculations for LEU #1 had predicted a shutdown margin of 3.09% A k/k.

3.6 Partial Fuel Element Worth As of this date we have determined the reactivity worth of our highest loaded partial fuel element in position B-5 (Figure 2). This position was picked since it was the same as the last one loaded with a standard element when loading the original core with LEU. After calibration of LEU #1 control rods, it was determined that excess reactivity was 1.495% A k/k. Using the same. control rod calibration curves, it was determined that the initia))y critical core of 3496.91g (with position B-5 empty) had an excess reactivity of 0.119% A k/k.

LEU #2 had an excess reactivity of 1.203%. Therefore we estimate the partial element loaded with 124.9g (62.5%) was worth approximately 1.084%

A k/k in position B-5.

3.7 Thermal Neutron Flux Distributions These measurements are somewhat time-consuming and are currently in progress. Results will be provided as available under separate cover.

We have determined the peak flux position in the Central Irradiation Facility (CIF). Spectrum measurements have been completed using bare and cadmium covered foils in the following experimental facilities:

1. Beam Port #1,

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3. Thermal Column and

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Analysis of these is not completed. We anticipate an M.S. thesis'to complete the core flux profile measurements.

3.8a Void Coefficient o_f,Reactivitv The method used for measuring the moderator void coefficient of reactivity at the OSURR requires two plastic " void boxes" fabricated to L fit into the coolant flow channels in a reactor fuel element. One box is air filled and one box is water filled. The critical regulating rod position with the air filled box inserted into a particular fuel. element flow channel and the position with the water filled box inserted are compared. The difference in critical rod height-is due to the reactivity worth of the air void inside the box. This was done for each element- .

except'the control. rod elements. The' air filled void was calculated to be 12.1675 ml. The voidable water volume of the core including the water surrounding the CIF and the water below the control rods'is 48,993 ml.

Technical Specifications are based on 1% of this water volume.

The average negative reactivity for all elements with the void box inserted in the north side of the fuel element, four water gaps from the I outside edge was measured to be -0.0197%/12.1675 ml.

- 0.0107%/12.1675 ml = 0.0016%/ml 48,993 ml x 1% voidable volume = 489.93ml l l

- 0.0016%/ml x 489.93ml/1% void =

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- 0.7915%/1% void = - 0.007915/1% void

= - 7.915x10' /1% void Technical Specification limi found in 3.1.1(6) is negative with a minimumabsolutevalueof1.8x10'g/1%voidacrosstheactivecore.

Previous measured values for the HEU core are as follows.

1965 - .34%/1% void l 1972 - .72*/1% void 1975 - .82%/1% void 1981 - .33%/1% void Differences in these numbers are due to differences in the f measurement of actual core floodable volume. Comparisons between these !

values and those for the LEU core may not be meaningful since the actual l core floodable volume has changed with the new core. More importantly we !

have demonstrated that there is a negative moderator void coefficient of reactivity which meets our Technical Specifications.

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The predicted average value for all LEU core geometries analyzed was

-0.32% a k/k/1% void.

3.8b Moderator Temperature Coefficient,ol Reactivity

.To determine the moderator temperature coefficient of reactivity the following measurements were made:

1. Long (slow) period (100-300 sec.) starting at low power and

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continuing to = 10KW, making the following measurements 1

a.-Reactivity measurement at low power from doubling

-time measurement.

b. Reactivity measurement at high power from doubling

. time measurement.

c. Moderator temperature change above the core.

2. Short (fast) period (10-12 sec.) starting at low power and continuing to = 10KW, making the same measurements.

The reactivity change for the transient of the long, slow period was 0.046%. It was 0.074% at the start and 0.028% at the finish. The moderator temperature change was 4.2*C. It is assumed that the observed reactivity change of 0.046% was due to both mcderator and fuel heating.

The reactivity change for the transient of the short, fast period was 0.02%. It was 0.30% at the start and 0.28% at the finish. There was no moderator temperature change measured during this short period. The reactivity change was therefore only due to fuel heating.

- 0.046% Moderator + Fuel Feedback

- 0.020% Fuel Feedback only

- 0.026% Moderator Feedback only

- 0. 026*/4. 2

C = - 6.19x10~ a k/k/*C This meets Technical Specification 3.1.1(5) which requires a negative moderator temjyrature coefficient of reactivity with an absolute value of at least 2x10 a k/k/*C.

Previous measurements for the HEU Core are as follows:

1965 - 4.32x10~ /DC 1972 - 5.00x10~ /0C 1575 ~ /0C 1981 - 5.00x10~

- 5.68x10 /0C The predicted value for the LEU Core was calculated to be about 5

5.5x10 / *C.

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3.9 Comparison of Results Comparisons of the various results are contained within each section where appropriate. We have not determined any significant differences which impact on normal operations or potential accidents.

3.10 Initial Londing Measurements Measurements made during initial loading of the LEU fuel are summarized below. The loading sequence shown in Table 1 and Figure 1 may be referenced. SU refers to the normal start-up Channel of the reactor. j It was positioned near grid 2E. TC refers to another fission chamber positioned near grid 1C inside the thermal column. The PuBe start-up- i source is near grid 6B. BP refers to a CIC in Beam Port #1 near grid 3E l which we were able to use starting with loading number 8.

BKG determined by total cts / total seconds = No fuel and source down inside Cd shield I SU Channel = 0.0538 c/s TC Channel = 0.0026 c/s C determined by total cts / total seconds = No fuel and source up next to core grid SU Channel = 0.3368 c/s TC Channel = 0.2722 c/s Loading #1.

Control rod elements were loaded:

l OHC-001; OHC-002; OHC-003; OHC-004 These elements contained 499.53g. 1/m measurements indicated we could add three additional standard elements.

Loading #2.

Three standard elements were loaded:

GH-007; OH-010; OH-011 These elements contained 599.53g. 1/m measurements indicated we could add two additional standard elements.

Loading #3.

Two standard elements were loaded:

OH-003; OH-014 These elements contained 399.68g. 1/m measurements indicated we could add two additional standard elements.

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Loading #4.

Two. standard elements were loaded:

OH-009; OH-012 These elements contained 399.69g. 1/m measurements indicated we could add two additional standard elements.

Loading #5.

Two standard elements were loaded:

OH-002; OH-018 These elements contained 399.54g. 1/m measurements indicated we could add one additional element.

Loading Sa, Nine filler elements were loaded:

OHF-001; OHF-002; OHF-003; OHF-004; OHF-005; OHF-00G; OHF-007; OHF-008; OHF-009 These elements contained 0.0g. 1/m measurements indicated we could add one additional element.

Loading 86.