ML20128A148

Analyses for Conversion of Ga Tech Research Reactor from HEU to LEU Fuel

ML20128A148
Person / Time
Site:	Neely Research Reactor
Issue date:	09/30/1992
From:	Matos J, Mo S, Woodruff W ARGONNE NATIONAL LABORATORY
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ML20127P210	List: ML20127P207 ML20127P263 ML20128A148
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NUDOCS 9302020139
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ATTACIIMENT 1 Analyses for Conversion.of the Georgia Toch Research Reactor from IIEU to LEU Fuol 9302020139 930121 PDR ADOCK 05000160 PDR

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ANALYSES FOlt CONVEllSION OF Tile GEOltGIA TECII ILESEAltCil ItEACTOlt FitOM IIEU TO LEU FUEL -

i J. E. Matos, S.C. Mo. and W.L. hdrufT REltTR Program Argonne National Laboratory  :

Argonne,IL 60439 September 1992 h

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SUMMARY

l This report contains the results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory (ANL) for conversion of the Georgia Tech 3 Research Reactor (GTRR) from the use of HEU fuel to the use of LEU fuel. The objectives of this study were to:(1) maintain or improve upon the present reactor performance and margins of safety, (2) maintain as closely as possible the technical specifications and operating procedures of the present HEU core, and (3) utilize a proven fuel assembly design that is economical to manufacture. Extensive collaboration with Dr. R. Karam, Director of the Neely Nucicar Research Center at Georgia Tech, took place on all aspects of this work.

The LEU fuel assembly has the same overall design as the present HEU fuel assembly,  :

except that it contains 18 fueled plates with LEU U3 SirAl fuelinstead of16 fueled plates with IIEU U Al alloy fuel. This LEU silicide fuel has been approved by the Nuclear-Regulatory Commission for use in non power reactors,  ;

Documents that were reviewed by ANL as bases for the design and safety evaluations- 9 were the GTRR Safety Analysis Reports, the GTRR Technical Specifications, and responses by the reactor organization to AEC questions in licensing the reactor for 5 MW operation.

The methods and codes that were utilized have been qualified using comparisons of calculations and measurements of LEU demonstration cores in the Ford Nuclear Reactor at' ,

the University of Michigan and in the Oak Ridge Research Reactor at the_ Oak Ridge National Laboratory. Additional qualification has been obtained via international benchmark comparisons sponsored by the IAEA for heavy water research reactors, Only those reactor parameters and safety analyses which could change as a result of replacing the HEU fuelin the core with LEU fuel are addressed. The attached summary.

table provides a comparison of the key design features of the HEU and LEU fuel assemblics and'a comparison of the key reactor rc.3d safety parameters that were calculated for each core. The results show that all of the objectives of this study were fully realized and that the GTRR reactor facility can t,c operated as safely with the new LEU fuel assemblics as .

P with the present HEU fuel assemblies.

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SUMMARY

TABLE HEU and LEU Design Data. Core Physics, and Safety Parameters for Conversion of the Georgia Tech Research Reactor DESIGN DATA HEU Colg LEU Core Minimum Numberof Fuel Assemblies 14 14 Maximum Number of Fuel Assemblies 19 19 Fuel Type U Al Alloy U3 Si2-At Enrichment, % 93 19.75 Uranium Density, g'em 3 0.65 3.5 Number of Fueled Plates per Assembly 16 18 Number of Non Fueled Plates per Assembly 2 2 235U per Fuel Plate, g 11.75 12.5 235U per Fuel Assembly, g 188 225 Fuel Meat Thickness, mm 0.51 0.51 Cladding Thickness, mm 0.38 0.38 Cladding Material 1100 Al 6061 Al Number of REACTOR PAR AMETERS HEU Core LEU Core Assembiics Cold Clean Excess Reactivity, % ak/k 11.7 ! 0.4 9.4 0.4 17 Coolant Temperature Coeflicient, % Ak/k/*C 0.0076 0.0067 14 Doppler Coefficient, % ak/k/*C -0.0 0.0017 14 Whole Reactor Isothermal Temp, Coeff., % ak/k/ C 0.0224 0.0232 14 Coolant Void Coefficient, % ak/k/% Void 0.0383 0.0333 14 Limiting Power Peaking Factor 1.54 1.58 14 Prompt Neutron Ufetime, s 780- 745 14 Effective Delayed Neutron Fraction 0.00755 0.0075 0.0076 14 Shutdown Margin, % ak/k 7.1 0.2 8.8t0.2 17 (Max. Wonh Shim Blade and Reg. Rod Stuck Out)

Top D20 Re!!ector Worth, % ak/k 2.1 0.3 - 2.4 !0.3 17 (For D20 2" Above Fuel Meat)

Reactor Power umits 1625 gpm Flow Rate Based on Departure from Nucleate Boiling, MW 11.5 10.8 14 Based on Flow Instability Criterion, MW 10.6 10.6 14 Umiting Reactor Inlet Temperature, 'F 172 170 14-Umiting Reactor Outlet Temperature,'F 188 187- 14 -

Umiting Safety System Settings Forced Convection Reactor Power, MW 5.5 5.6 14 Coolant Flow Rate, gpm 1625 < 1625 14 Reactor Outlet Temperature. *F 139 145- 14 Margin to D20 Saturation Temperature, 'F 8 11 14 Max. Fuel Plate Temp for LOCA atter 8 Hours Coohng, C 425 400 14 Maximum Positive Reactivity insertion, % ak/k > 2.2 > 2.2 14

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TABLE OF CONTENTS ,

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1. I N T R O D U C TI O N . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .I, t

2. R E ACTO R D E S C R I P TIO N . . . . .. .. .. . . .. . . .. . . . . .... .... . .. . ... .. . . . . .. . . .. . .. . . . .. . .. . . .. . . .. .. ... . .. . .

3. F U EL AS S EM B LY D E S C RIPTION S.. ........... ... . ....... ...... ............. .... ......... ..... ....... ....... . ... 5 g

4. C A L C U L ATI O N A L M O D E LS . .. .. . . . . . . .. . . .. . . . .. ... . .. . . .. . . .... .. . . .. . . .. . ... . . . . . . . . . .. . ... ....

m 4.1 Nuclear Cross Sections for Diifusion Theory Models.......................................... 6 4.2 Re a c to r M o d el s. . . . . . . . . . . . . . . . . . . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . . . . . . . . . -

5. N E U T R O N I C P A R A M ET E R S .. .. . .. . . .. .. . . . . .. . .. . . .. . . . . . . .. . ... . . .. . . .. . . .. . .. . . .. . . . . . . . . . .!

5.1 - Critical Ex peri m en t for H E U Core .............................. ..................................... .. 11 5.2 Col d Cle an E xcess Heactivi ti es ........................................................................... 11 5.3 B u rn u p Cal cula tio n s .. .... . . .. . . ... . ... .. .. .. . .... ... . . .. . . .. . ... .. . . . .. .... .. .. . . . . .. .. . . .. . . . . . . .. . .. . . ... . .. . 12 L 5.4 Power Distributio'ns and Power Peaking Factors ............................................. 13 5.5 Reactivity Coeflicienis and KineticsParametors................................................ 16

6. - S H U TD O WN M A R G I N S .. .... . . .. . . .. . . .. . . . . . .. . ... .. .... . . . . .. ...... ... .. . . .. . .. .. . .. . . .. . ... . . . . . . ... .... .. . . .. . .

. 7. TH ERM A L-HYD R AU LIC S A F ETY P A R AM ETER S ..................................................... 19 ^

7.1 Safety Limits in the Forced Convection Mode ................................................... 19 7.2 Safety Limits in the Natural Convection Mode ................................................. 22 l 7.3 Limiting Safety System Settings in the Forced Convection Mode.................... 22 -

7.4 - Limiting Safety System Settings in the Natural ('onvection Mode.................. 25f

8. - C OO LI N G TI M E R E Q U I R EM E NTS ...... . ... ... ..... .. ..... ......... ... .... .... ................ ....... . ......... . 25 9 t

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9. LIMIT ATI O N S O F EX P E R I M ENTS.. ... .. . . ... ... . ... ... . ....... . ... .. .. ... . ... . .... . .. . . . . . . ...... ....... . . .. .... 26 9.1 Comparison of Calculations with SPERT.II Experiments................................ 27 9.2 Inadvertent Reactivity Insertions Due to Experiment Failure......................... 27 -

10. A C C I D E N T A N A LY S E S .. . . .. . . .. . . .. . . .. . . . . . . . . . . .. . . . . . .. . . . .. . .. . . . . . .. . . .. . . . .. . ... . .. . . .. . . . .. . .. . . . . .

10.1 S tartu p A cci d e n t .. . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . .. . . ... . .. . . . . . .. . . .. . . .. . . . . . . .. . . .. . . . . . . . . . .. . .. . . .. . . .. . . . 29 10.2 Reaclivity Effects of Fuel Plate Melting............................................................. 30

10. 3 Fu el loa di n g A cci d e n t .. . . . . ... . .. . . .. . . .. . . . . . .. . . .. . . .. . . . . . . . .. . .. . . .. ... . .. . . .. . . .. . . . . . . ... . .. . .. . . . . . . .. 3 0 -

10.4 Ma xim um Positive Reactivity Insertion. ........................................................... 31 -

10. 5 D e s i gn B asi s A cci d e n t. . . . .. . . . . . . .. . . . . . .. . . .. . . . . . . .. . . .. . . . . . . .. . ... . .. . . .. . . .. . . . . . . . . . . . .. . .. . . .. .. . . . .. . . . 3 2 l

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11. F U E L H AN D LIN G A N D STO R A G E .... . ... ..... .. . . .. . . ... . . . . ... . .. .. .. . ... ..... . . . . .. .. .. . ... .. .. . . .. .... . . .. 34 REFERENCES.......................................................................................................................35 ATTACHMENT 1: Isothermal Reactivity Change Components For An HEU Core with 17 Fresh Fuel Assemblies....................................... 38 ATTACHMENT 2: E n gin cering Uncert ai n ty Factors................. .... ... ........ .................... .. 4 0

p 4 ANALYSES FOR CONVERSION OF THE GEORGIA TECH RESEARCH REACTOR FROM HEU TO LEU FUEL J.E. Matos, S.C. Mo, and W.L. Woodruff RERTR Program Argonne National Laboratory Argonne,IL 60439 September 1992

1. INTRODUCTION This report contains the results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory (ANL) for conversion of the Georgia Tech Research Reactor (GTRR) from the use of HEU fuel to the use of LEU fuel.

The objectives of this study were to: (1) maintain or improve upon the present reactor performance and margins of safety, (2) maintain as closely as possible the technical '

specifications and operating procedures of the present HEU core, and (3) utilize a proven fuel assembly design that is economical to manufacture.

The design and safety anrilyses in this report provide comparisons of reactor parameters and safety margins for the GTRR HEU and LE.U cores. Only those parameters which could change as a result of replacing the HEU fuel.in the core with LEU fuel are addressed. Documents that were reviewed by ANL as bases for tho' design and safety evaluations were the GTRR Safety Analysis Reports,1 the GTRR Technical Specifications 8, and responses A by the reactor organization to AEC questions in licensing the reactor for 5 MW operation.

The LEU fuel assembly has the same overall design as the present HEU fuel assembly, except that it contains 18 fueled plates with LEU U.SirAl fuel and two non-fueled plates instead of 16 fueled plates with HEU U Al alloy fuel and 2 non fueled plates.

l A detailed safety evaluation of LEU U3 Sir Al fuel can be found in Reference 5.

The methods and codes that were utilized by ANL have been . qualified using comparisons of calculations and measurements of LEU ~ demonstration cores 8 80 in the Ford Nuclear Reactor at the University of Michigan and in the~ Oak Ridge Research Reactor (ORE) at the Oak Ridge National Laboratory. - Additional qualification has been

-obtained via international benchmark comparisonsit.18 sponsored by the IAEA.

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2. REACTOR DESCRIPTION I i

The GTIllt is a heterogeneous, heavy water moderated and cooled, tank type reactor fueled with 93% enriched MTit type U Al alloy fuel. IIorizontal and vertical sections f

through the reactor are shown in Figs.1 and 2, respectively. Provision is made for up to -

19 fuel assemblies spaced 6 inches apart in a triangular array. The current core consists ~

of 17 fuel assemblies. Each assembly consists of 10 fueled and two non fueled plates with a fissile loading of about 188 g 235U. The total fissilo loading of a fresh 17 assembly coro would be about 3.2 kg 235U.

The fuel is centrally located in a six foot diameter aluminum reactor vessel which l provides a two foot thick D2 0 reflector completely surrounding the core. 'The reactor l l

vessel is mounted on a steel support structure and is suspended within a thick walled graphite cup. The graphite provides an additional two feet of reflector both radially and beneath the vessel. The coro and reflector system is completely enclosed by the lead and concrete biological shield.

The reactor is controlled by means of four cadmium shim safety blades and one

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cadmium regulating rod. The four shim safety blades are mounted at the top of.the reactor vessel and swing downward through the coro between adjacent rows of fuel assemblies. The regulating rod is supported on the reactor top shield and extends downward into the radial D 20 reflector region. This rod moves vertically between the horizontal midplane and the top of the core, s

( The heat removal system is composed of a primary heavy water system and a secondary light water system. The heavy water system includes the reactor vessel, the primary D 0 coolant pumps, the D 0 2 makeup pump, the heat exchangers, and the associated valves and piping. The light water secondary system is composed of the circulating water pumps, the cooling tower, and associated valves ar.d piping.

The LEU reference core used in this analysis consists of 17 fuel assemblies with the same arrangement as the present IIEU core. Each fuel assembly contains 18 fueled l plates with 225 g 235U when fresh. The LEU core will use the same control system, heat L removal system, and auxiliary systems as the current IIEU core.

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, 3. FUEL ASSEMBLY DESCRIPTIONS The geometries, materials and fissile loadings nf the current HEU fuel assemblies and the replacement LEU fuel assemblics are described in Table 1. A schematic diagram "

of the HEU fuel assembly is shown in Fig. 3. The LEU fuel plate is the standard DOE plate containing U 3SirAl fuel with ~3.5 g U/cm8 and 12.5 gn8U. The external dimensions '

and structural materials of both assemblies are identical, except that the LEU assemblies util!ze 6061 Alinstead of 1100 A1.

Table 1. Descriptions of the HEU and LEU Fuel Assemblies lieu LER Number of Fueled Plates / Assembly 16 18 Number of Non-Fueled Plates / Assembly 2 2 '

Fissile Loading / Plate, g 23sU 11.75 12,,5 Fissile Loading / Assembly, g ?35U 188 225 Fuel Meat Composition U Al Alloy U3 SirAl Cladding Material 1100 Al' 6061 Al2 Fuel Meat Dimensions Thickness, mm 0.51 0.51 -- .

Width, mm 63.5 58.9 62.8 Length, mm 584 610 572 610 Cladding Thickness, mm 0.38 0.38 1 to ppm natural boron was added to the composition of the cladding and all fuel assembly structural materials to represent the alloying materials, boron impurity, and other impurities in the 1100 At of the HEU assemblies, r 20 ppm natural boron was added to the composition of the cladding and strurural materials of the LEU assemblies to represent the alloying materials, boron impurity, and other impurities in 6061 A1. Aluminum with no boron or other impur. ties was used in the fuel meat of both the HEU and LEU assemblies.

Fig. 3. HFU Fuel Assembly Schematic d51=g ,

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4. CALCULATIOt1AL MODELS  ;

4.1 t1uclear Cross Sections for Diffusion Theory Models l Microscopic cross sections in seven energy groups (Table 2) were prepared at 23'C using the EPRI CELL code" for the llEU and LEU fuel assembly geometries and fissile loadings. The integral transport calculations in EPHI CELL were performed for 69 fast groups and 35 thermal groups (<L855 eV), which were then collapsed to seven broad energy groups for use in diffusion theory calculations.

Table 2. Seven Group Energy Group Boundaries Group Upper Lower Group Upper Lower M. Enerov Energy M. Eneroy Eriergy  ;

1 10.0 MeV 0.821 MeV 5 0.625 eV 0.251 eV 2 0.821 MeV 5.531 kev 6 0.251 eV 0.057 eV 3 5.531 kev 1.855 eV 7 0.057 eV 2.53 x 10 eV 4 1.855 eV 0,625 eV Figure 4 shows the dimensions of the HEU and LEU fuel assemblies and the fuel assembly models that were used in the diffusion theory calculations for the reactor. The fueled and non fueled regions were modeled separately. A non fueled region consists of a sideplate and the fuel plate aluminum (plus asseciated water) between the fuel meat and the sideplate.

Figure 5 shows the unit cell geometry and dimensions that were used in EPRI CELL-to generate microscopic cross sections for the fueled and non fueled regions of the HEU and LEU assemblies. The non fueled region inside the assembly is represented by the

'" extra region 1" containing calculated volume fractions of aluminum and heavy water associated with each fuel plate. " Extra region 2" was modeled to represent the heavy water outside the assembly that is associated with each fuel plate. Its thickness was chosen to preserve the water volume fraction in the physical unit cell of each fuel assembly. All cell calculations were done using a fixed buckling of 0.00373 cm 2, which corresponds with the anticipated axial extrapolation length of about 21 cm in each fuel assembly in the reactor diffusion theory calculations.

Each EPRI-CELL case was run three times using the local fine group spectra over I

the fueled region and the two extra regions to collapse the fine group cross sections into 7 broad groups. This procedure was performed because the fueled region, the non fueled region inside the fuel assembly and the water outside:each fuel assembly were modeled . ~

as separate regions in the diffusion theory model of the reactor. Cross sections for the heavy water and graphite reficciors and for the fuel assembly end fittings were calculated using a unit cell model consisting of a pure mU fission spectnim on a 10 cm thick slab'of water.

_- - - - _ ~ - = . . _ - _ . -_ .- ,_ _ - - _ - ,.. - . .

,' .' 7

/e Fig.4 Modols for HEU and LEU Fuel Elements (Dimensions in mm)

--+  : 4.8 --+

• 1.05

- ~~

0.508 I i 0.381 I g 2.692 ,,

i  ; e t~

~ '

63.5  : 63.5

65.6

= 75.2 +-- 5.85 75.2 HEU Fuel Element 16 Fueled Plates DIF3D Model for 2 Non Fueled Plates HEU Fuel Elomont

--* +-- 4,8 +-- 2.375

- 0.508 =

I- ,

1 0.381 l 2.250 I v  ;-

I g s 60.85- 00.95 65,6 ,

i

---*  : 7.175 75.2 75.2 LEU Fuel Element 18 Fueled Plates DIF3D Model for 2 Non Fueled Plates LEU Fuel Element

_____r- - _ _ _ _ _ _ -

____--________.---__am .- ____

.' 8 i

Fig. 5. EPRI CELL Model for Generating Fuel Element Cross Sections (Dimensions in mm) l Fuel Half Half Hall Extra Regioni Plates Fuel Clad Moderator Extra Extra Al/D20  !

C m g (Elen1, Redon (F) Begbn.[CJ Recion (M) Recioni (Ei) Realon2 (E2) Volume Fractions HEU 16 0.254 0.381 1.346 0.6244 7.2934 0.6499/0.3501 LEU 18 0.254 0.381 1.125 0.6567 6.7654 0.6611/0.3389 Number of Mesh Points in Each Region 4 1 7 4 10 i

i 'M VA i o Extra Extra a S Fuel l. Cladh Moderator '

Region 1 Region 2 S

.y  ;

(Al (D20) $ D20+Al (D20) ,y e

/

/ y

., m

~F~ ~C~  : M  : +-- E t -

• - E2-+ ,

Un:t-Cell Specifications for Fueled and I.'on-Fueled Portions of Fuel Element (Fuel Region Cross Sections: Collapse using Fluxos over F. O, and M; Non-Fuel Cross Sections: Collapse using Fluxes over E1 Only) r3

'hi- '

Region 1: Homogenized Fuel

.' r1 = 35.8 mm 3 2 fj#yg ~

Region 2: Mixture of Al + D20

.]"r11 r2 = 41.1 mm i! Region 3: D2O r3 - 80.0 mm Unit Cell Specificati0ns for D20 Between Fuel Elements (Collapse Cross Sections usin0 Fluxes over Region 3 Only.)

( 9 4.2 neactor Models Reactor calculations were performed in three dimensions using the VIM continuous energy Monte Carlo codem5 and the DIP 3D diffusion theory codel5 A detailed Monte Carlo model of the reactor was constructed including all fuel assemblies, the shim safety rods, the regulating rod, beam tubes and experiment penetrations, the bio-medical facility, and the top 'and bottom reflector regions in order to obtain absolute excess reactivities and shutdown margins for comparison with limits specified in the Technical Specifications. Nuclear cross sections were based on ENDF/B V data. The experiment facilities that were modeled are shown in Table 3.

In diffusion theory, the reactor was modeled in rectangular geometry with a heteror;eneous representation of the fueled and non fueled portions of the fuel assemblies and the water between fuel assemblies (see Fig. 6). The four shim safety rods (control arms) that swing between the fuel assemblies, the regulating rod, and the various reactor rme' ations (reactivity worth ~4.5% Ak/k) were not included in diffusion theory model.

. bottom axial reficcior and the radial ref'ector were also simplified. The LEU modelis identical with the HEU model except for the fuel assembly materials.

A simplified Monte Carlo model corresponding with the diffusion model was also constructed in order to verify that the difTusion theory model was correct.

Table 3. Experimental Facilities included in the Detailed Monte Carlo Model.

8 vertical experiment tubes filled with air in the D20 reflector 2 vertical experiment tubes filled with air in the graphite reflector 12 vertical experiment tubes filled with graphite in the graphite reflector 14 horin mal beam tubes filled with air and penetrating the D 20 and graphite reflectors 8 horizontal beam tubes filled with graphite and penetrating both reflectors 2 horizontal beam tubes filled with 12" graphite, remainder air and penetrating both reflectors Biomedical Facility: A portion of the graphite reflector between the vessel and the biomedical heility consists of a bismuth shield and air (see Fig.1).

Thermal Column

.' 10 Fig. 6. Radial and Axial Modols for Diffusion Theory Calculations Radial Model O

O O O O O O O O O O OOO O O

D20 AlTank Graphite 0.0 S4.9 105.0 106.1 217.0 271.8 Al + D2O 183.8 e .3 i D20  :

D20 5.

core j g

124.1 A' + 020 62.5 Graphite 0.0 Dimensions in cm

11

5. NEUTRONIC PARAMETERS 5.1 Cutical Experiment for HEU Core In 1974, a critical experiment was built using 9 fresh HEU fuel assemblies. The core was made critical at different shim-saf f blade positionst? with the regulating rod nearly fully withdrawn and nearly fully inserted. The kerfs calculated for these critical configurations using the detailed Monte Carlo model were 0.99110.002 and 0.988 0.002.

The corresponding reactivity values were 0.9110,20% Ak/k and 1.22 0.22% Ak/k, respectively. The reactivity bias of about 1.0 0.3% Ak/k in the calculations is attributed to uncertainties in the nuclear cross sections and uncertainties in the reactor materials.

5.2 Cold Clean Excess Reactivities Calculated excess reactivities (including reactivity bias) for the reference HEU and LEU cores with 17 fresh fuel assemblies are shown in Table 4. The Technical Specifications limit the excess reactivity to a maximum of 11.9% Ak/k. The LEU core is -

expected to satisfy this requirement.

Table 4. Excess Reactivities of HEU and LEU Cores with 17 Fuel Assemblies Calculated Excess React. , % Ak/k i 1cr Fresh HEU Core Fresh t EU Core Detailed Monte Carlo Model 11.7 0.4 9.4 0.4 Simplified Monte Carlo Model2 16.8- 0.4 14.310.4 Diffusion Theory Model2 16.6 14.6 i The reactivity bias of 1.0 0.3% auk was added to calculated values.

2 Without experiment penetrations, shim safety blades, and regulating rod.

Differences between the detailed and simplified Monte Carlo models were described in Section 4.2. The reactivity effect of(1) replacing all vertical and horizontal experiment facilities inside the heavy water vessel with D2 0, (2) replacing all air-filled experiment .

facilities in the graphite reflector with graphite, and (3) replacmg the bismuth shield and air in front of the biomedical facility with graphite was calculatedl8 to be 4.5 0.3% Ak/k.

The worth of replacing the control absorbers in their fully withdrawn position with D2 0 was calculated to be 0.1 i 0.3% Ak/k, a value consistent with zero worth. Thus, the simrP 1 Monte Carlo mcdel and the diffusion theory model . are reasonable repre .iations of the reactor if the reactivity worth of the experiment fac. ties is taken into acct, cat.

l

)

12 5.3 Burnup Calculations Burnup calculations were run using the REBUS ceden for HEU and LEU cores with 17 fuel assemblies to estimate fuel lifetimes. Reactivity profiles (including the 1% Ak/k reactivity bias) are shown in Fig. 7 over a limited burnup range. Excess reactivity values for fresh cores computed using the diffusion theory model are shown in Table 4. The dashed lines show the end-of cycle exceos reactivity range that accounts for reactivity losses due to experiment facilities (4,510.3% Ak/h), cold to-hot swing (~0.3% AM), and control provision (~0.5% Ak/k) that are not included in the difTusion theory burnup model.

Reactivity losses due to equilibrium Xe and Sm are included in the curves. We conclude that the lifetime of the LEU core will be about the same as that of the HEU core when absolute errors in the calculations are taken into account.

Fig. 7. Burnup Reactivity Profiles 11 . .- . . . . .

, 10 HEU 16P/17 Ass. -

g 3--- LEU 18P/17 Ass.

g . .

gt .

A 8 - -

= .

2 U 7 - -

E E

6 - -

g __________________ _____________.

8s ----f----------------------------

y EOC Excess 4 _ React. Rango, .

Bumup Model ,

3 40 60 80 100 120 140 160 180 200 Full Power Days at 5 MW 1

13 5.4 Power Distributions and Power Peaking Factors Power distributions and nuclear power peaking factors were calculated using the difrusion theory model for HEU and LEU cores with 14 and 17 fuel assemblies. As stated previously, the shim safety rods, regulating rod, and experiment penetrations were not represented. The results are shown in Fig. 8 for the 14 element cores and in Fig. 9 for the 17 element cores. The reason for calculating cores with 14 fuel assemblies is that this is the minimum GTRR core size and cores with 14 assemblies will be used .to compute the thenaal hydraulic safety margins.

From the point of view of thermal hydraulic safety margins, the most important neutronic parameter is the total 3D power peaking factor (the absolute peak power density in a fuel assembly divided by the average power density in the core). The total power peaking factor is defined here as the product of two components:(1) a radial factor defined as the average power density in each assembly divided by the average power density in the core and (2) an assembly factor def'med as the peak power density in each assembly divided by the average power density in that assembly. The assembly factoris a pointwise factor computed at the mesh interval edge and includes both planar and axial power pecking.

The data in Figs. 8 and 9 show that the power distributions and power peaking factors are nearly the same in fresh HEU and LEU cores with 14 fuel assemblies and in fresh HEU and LEU cores with 17 fuel assemblies. The percentages of reactor power shown in Figs. F and 9 do not add to 100% because ~2.5% of the energy is deposited outside ,

the fuel assembles.

14 Fig. 8. Power Distributions and Power Peaking Factors -

HEU and LEU Cores with 14 Fuel Assemblies HEU Power Distributions LEU 14 Fuel Assemblies 0.32 0.32 6.40 6.42 0.34 o,34 0.34 0.34 6.84 6.08 _ 6.88 6.84 0.35 0.35 6.98 6.90 0.37 0.37 0.36 0.36 7.32 742 728 728 MW--- 0.35 035 046 046 7.08 7.18- 7.18

% of--- 7.08 Total 0.3d 0.37 037 0.36 7.32 7.28 7.28 7.32 0.35 0.35 6.98 6.90 0.34 0.34 0.34 0.34-6.84 6.84 6.88 -6.88 0.32 0.32 -

6.40 6.42 Power Peaking Factors HEU 14 Fuel Assemblies LEU-l

][" I 1.43 0.92 1.51' 1' 9 1.31 199 0.99 ;

0.98 1,54 -

1.47 1.54 9 47 I- 0.99 1$44 1.00 1.55 1.49 I'N 1'48 1.04 - 1.04 -

1.05 1.05 1.51' 1,51 1.47 1.47 1.03 1.58 1.58 1.03 Radial 1.02 1.54 1.54- 1,c2 1.49 1.49 Element 1.43 1.43 1,54 -

1.4s 1.54 Total 1.4s 1.04 1.04 1.05- 1.05 1.51 1.51 1.47 1.47 ~

1.54 1M 0.99 1.00 1 33 1.49 1.53 1.49 0.99 0.99 -

0.98 0.98 1.54 1,54 1,47 1,47 -

1.44 1.4 0.92 W

0.92 1.51 1.43 1.39 1.31

15 Fig. 9, Power Distributions and Power Peaking Factors HEU and LEU Cores with 17 Fuel Assemblies Power Distributions LEU HEU 17 Fuel Assemblics 0.27 0.27 5.44 5.48 0.28 0.29 0.28 0.29 5 60 5.76 5.60 5.82 0.28 09 0.28 0.30 5.60 $J8 5.64 5.90 0.31 0.30 0.31 0.30 5.98 6.16 5.92.

6.16 0.29 028 MW --- 0.29 0.28 5.78

% of --- 5.72 5.64 5.64 Totti 0.29 0.30 0.30 0.30 6.02 5.86 5.96 5.90 0.27 0.30 0.30 0.28 0.28 5.98

48 5.92 5.00-5.44 5.64 0.28 020'

._ 028 018 5.68 5.68 5.60 5.60 0.28 0.28 5.60 5.64 Power Peaking Factors LEU HEU 17 Fuel Assemblies 0.95 0.95 1.41 1.50 0.97 0.97 1.00 1.01 1.46 1.54 1.45 1.52 1,45 1.54 1.02 0.98

, 1.03 0.98 1.44 1.54 1.51 l 1.48 1.40 1.57 1,03 1.49 1.07 1.52 1,04 1.07 1.52 1.50 1.57 1.46 i

1.57 1.58 .1,61 W 0.98-R:dlit 1.00 0.98 1.01 1.47 1.55 1,48 Elimat 1.42 Tctal 1.47 3.4n ,,,,Lig,_, j 1.48 1.03 1.05 1.02 1.04 1.52 1.51 1.57 1.56 1,56 "0 Om 1 60 1.03 0.95 1.04 0.98 O.99 1.44 1.51 1.53 1.52 1.48 1.44 1.38 1.54 1 44 1,58 1.50 1 42 0.97 0.99 0.97 0.99 1.46 1.53 1,54 1.46 l- 1,42 1 44 1'49 0.96 M2 0.97 1.43 1.50 1.39 1.47 l .

i

16 5.5 Reactivity Coefficients and Klriotics Parameters Reactivity coefficients were computed for HEU and LEU cores with 14 and 17 fresh fuel assemblies as functions of temperature and void fraction using the 3D diffusion theory model. Also computed were the whole-core void coefficient, the reactor isothermal temperature coefficient, and the prompt neutron lifetime. Fresh cores were calculated because they are limiting cores. As fuel burnup increases, the neutron spectrum becomes softer and the reactivity coefficients become more negative.

Reactivity changes were calculated separately for changes in coolant temperature, coolant density, and fuel temperature while holding all heavy water outside the fuel assemblies at 23 C. Slopes of the reactivity feedback components at 45 C are shown in Table 5 along with the void coefficient for a uniform 1% change in the coolant density in all fuel assemblies. These reactivity feedback coeflicients will be used in the transient analyses in Sections 9 and 10 because the transients considered involve heating of the fuel and coolant. Heating of the heavy water outside the fuel assemblies would have only a small efTect because of the time constants involved in the transients.

Table 5. Reactivity Coefficients (% Ak/irC at 45*C) and Kinetics Parameters hf.il LEll 14 Ass. 17 Ast. 14 Ass. 17 Ass.

-0.0062 -0.0055 -0.0055 0.0053 Coolant Temperature

-0.0014 -0.0014 -0.0012 -0.0013 Coolant Density

-QA -0 0017 -0.0020 Fuel Doppler -ILQ

-0.0076 -0.0069 -0.0084 -0.0086 Sum

-0.0224 0.0201 -0.0232 -0.0215 Whole Reactor Isothermal 1 Void Coefficient 2, -0.0383 -0.0392 -0.0333 -0.0350 3 780 704 745 695 1p , ps 0.00755' O.00755' O.0075 - 0.00765 ott 1 includes fuel, coolant, inter assembly water, and reflector.

2 % Ak/k/% Void. Uniform voiding of coolant in all fuel assemblies.

3 Calculated prompt neutron lifetime.

4 Measured effective delayed neutron traction.

5 Estimated value.

,l*

_.- .?-

17 l

.i

- The sum of the coolant and fuel Doppler reactivity coefficients in Table 5 are slightly more negative in the LEU cores than in the HEU cores. - The Doppler coefficient'actually- '{

has a larger weight than shown in Table 5 because the fuel temperature normally increases more rapidly than the coolant temperature. The coolant void coemeient for all fuel assemblies in the core is slightly more negative in the HEU cores than in the LEU cores.

)

The reactor isothermal temperature coeflicient for the 5 MW clean core with 16 HEU fuel assemblies was calculated in the GTRR Safety Analysis Report (Ref.1, p. 98) to be 0.0232% Aldid*C at 45'C. The reactor isothermal temperature coefficients shown in Table 5 for clean cores with 14 and 17 HEU assemblies are in' good agreement with this ,

value. The corresponding reactor isothermal temperature coefficients for LEU cores with 14 and 17 assemblies are slightly more negative than those for the HEU cores. A breakdown of calculated isothermal reactivity feedback components for the coolant, inter-assembly waf.er, and reflector of an HEU core with 17 fresh fuel assemblies is shown in Attachment 1.

In April 1992, the whole-reactor isothermal temperature coefficient was measured to be - 0.0338 ok/k/ C in a 17 assembly HEU core with about 10,000 MW-hr burnup over the-period 19741992 (R. Karam, GTRR; private communication). Although these measured and calculated data cannot be compared directly (temperature coefficients normally become more negative with increasing -burnup), it does indicate that measured temperature coefficients in the GTRR may be more negative than calculated values.

The calculated prompt neutron lifetimes shown in Table 5 for the LEU cores with 14 -

fuel assemblies and with 17 fuel assemblies are slightly smaller than those in the corresponding HEU cores because the LEU cores have a slightly harder neutron spectrum.

The fission component of thc delayed neutron fraction in both the_ HEU and LEU cores was calculated to be 0.0071. The difference between this value and the peg of 0.00755 measured in the HEU core is attributed to delayed neutrons resulting from dissociation of -

heavy water by neutrons and gamma rays. The latter component of p g has not been computed here. Since the fission components of p.g were computed to be the same in the

+ HEU and LEU cores, we expect that the heavy water components of p.g and thus the total effective delayed neutron fractions will be very similar as well.

_l' .

18

6. SHUTDOWN MARGINS The Technical Specifications require that the reactor have a shutdown margin of at-least 1% Ak/k with the most reactive shim safety blade and the regulating rod. fully withdrawn. Measured reactivity worths 2a of the shim-safety blades in the present HEU core are shown in Table 6. The blade with the highest reactivity worth is blade #3.

Table 6. Measured Reactivity Worths of Shim-Safety Blades in HEU Core (9/26/90).

Shim Safety Blades Reactivity Worth. */e Ak/k Blade #1 5.55 Blade #2 4.66 Blade #3 6.21 Blade #4 4.41 Table 7 compares shutdown margins calculated using the detailed Monte Carlo model for HEU and LEU cores with 17 fresh fuel assemblies. The regulating rod and shim-safety blade #3 were fully withdrawn and the other three shim-safety blades were fully-inserted. The results show that both cores satisfy the 1% Ak/k shutdown margin requirement of the Technical Specifications.

Table 7. Calculated Shutdown Margins for HEU and LEU Coros with 17 Fresh Fuel Assemblies.

fdgg Shutdown Marcin *4 Ak/k HEU -7.14 0.25 LEU -8.8410.21 In addition to the automatic protective systems, manual scram and reflector drain provide backup methods to shut the reactor down by operator action. The top of the core is covered by 29.75 inches of D2 0, measured from the top of the fuel meat. The top 28 inches of D2O can be drained through a 4 inch pipe which connects the reactor vessel to the storage tank of the primary D 20 system. The reactivity worth of the top 28' inches of reflector was measured" to be 2.75% Ak/k in an HEU core composed of15 fuel assemblies with 142 g 23sU per assembly.

Monte Carlo calculations using the detailed Monte Carlo model were done to compare reactivity worths of the top D2 0 reflector in HEU and LEU cores with 17 fresh fuel assemblies (188 g 2asU HEU,225 g SU LEU). Several calculations were first done for each core to determine shim-safety blade positions that would bring the reactor near critical. Results in Table 8 for cases with 1" and 2" of D 20 reflector above the top of the fuel meat show that the top reflector worths of the HEU and LEU cores are very similar.

Thus, the shutdown capability of reflector drain in the LEU core will be very similar to -

that in the present HEU core.

f ..

' Table B. Calculated Top Reflector Worths (%Ik/k) of HEU and LEU Cores with 17 Fuel Assemblies and Control Blades near Critical Positions Too D;O Reflector HEU Core igLQgg D201* Above Fuel Meat - 2.58 0.29 (10) - 2.7310.31 (1a)

D2O 2" Above Fuel Meat -2.0510.28 2.42 0.30

7. THERMAL HYDRAULIC SAFETY PARAMETERS -

Thermal-hydraulic safety limits and safety margins calculated using the PLTEMP code 22 for the LEU core with 14 fuel assemblies (see Fig. 8) were cornpared with' the thermal hydraulic safety parameters used as bases for the current _ Technical Specifications. The analyses by ANL for the LEU core used a combined multiplicative and statistical treatment of a revised set of engineering uncertainty factors. Attachment 23 2 lists the engineering uncertainty factors used by Georgia Tech for analyses of the HEU l core and discusses the factors used by ANL, the rationale for their choice, and the inethod -

used to combine them. Results for the HEU core obtained using ANLis statistical treatment of the engineering uncertainty factors agree well with the analyses performed by Georgia Tech.

7.1 Safety Limits in the Forced Convection Mode The current Technical Specifications utili::e departure from nucleate boiling (DNB) as a basis for establishing safety limits on reactor power, coolant flow, and coolant inlet (or outlet) temperature. This report evaluates these limits based on flow instability as well as DNB criteria. The modified Wheatherhead correlation 2m was used for DNB and the Forgan Whittle correlatien 2s .2s was used for flow instability.

Calculated reactor power limits ~ based on DNB and flow instability.nre shown in Table 9 for 14 assembly HEU and LEU cores with the minimum coolant flow of 1625 gpm and with the coolant lowflow limit of 760 gpm. A maximum inlet temperature of 123*F-was used in all cases. Power limits based on the flow instability criterion are smaller than those based on DNB, but are still adequate to ensure the safety of the facility. The .

main reason for the difference in reactor power limits in the HEU and LEU cores is that the manufacturing specifications for LEU silicide dispersion fuel plates contain a factor of 1.2 for homogeneity of the fuel distribution while the HEU alloy. fuel has a corresponding factor of 1.03.

r - - - _ . -- _ _ _ _ _ _ _ _ _ _ _ ..

A . '

20-Table 9.- Reactor Power Umits in 14-Assembly Cores for a Maximum inlet Temperature of 0 123*F Based on Departure from Nucleate Boiling and Flow Instability. .!

Reactor Coolant . .

Elow? aom GTRR-HEUS -  : ANL. LEU 2  ;

Reactor Power Level (MW) for DNB2324 .

760 5.5 5.3 1625 11.5 10.8 ~

2 Reactor Power Level (MW) for Flow instabilit/ s2s l

760 5.3 5.0 -

1625 10.6 10.6 1

Calculated by ANL using GTRR engineering uncertainty factors in Ref. 23.

2 Calculated by ANL using revised engineering uncertainty factors (see Attach. 2).

Figure 10 shows the calculated reactor power limits as functions of reactor coolant-flow based on DNB for the HEU core and on flow instability for the LEU core. In the LEU core, we recommend a power limit of 10.6 MW based on the flow instability criterion for the minimum coolant flow of 1625 gpm and the maximum inlet temperature of 123*F..

Fig. 10. - GTRR Safety Limit for Forced Convection 16 .

BASES: Moderator Within 12 inches of Overflow -

Tin - 123'F Max When the Flow is Minimized -

-- 14 & Power is Maximized: Applicable for Mode 2 Only -

-- 12 GTRR HEU h /

- Departure from

" Nucleate Boiling 2 ,....................,

10

$ - Line:GTBR HEU ' [j,' h ow Fl lodability .

$ ,e" Line: ANL LEU f [',*#i

,r SAFE OPERATING REGION .

(e# . Mode 2 C- Nominal Operating E

y Conditions. Tin .114*F 2' .Mmel NominalOperating

, 3 _ Condition,s. Tin .114*F, ,

0 500- 1000 1500- 2000 2500-Reactor Coolant Flow l(GPM)

..o 4:

~*--

4> ,

More detailed data for the minimum coolant flow rate of1625 gpm and the maximum inlet temperature of 123'F are shown in Table 10. The LEU fuel assembly has reduced

-- power per plate, a smaller flow area, a higher coolant velocity, and a larger pressure drop due to friction. The peak cladding surface temperature is larger by about 5*F and the >

margins to DNB and flow instability are adequate.

Table 10. Thermal-Hydraulic Data for 14 Ass mbly Cores with the Minimum Coolant Flow of 1625 GPM and the Maximum Inlet Temperature of 123*F.

GTRR-MEll' ANL-LEU 2 Coolant Velocity, m/s 2.44 2.61-Friction Pressure Drop 3, kPa 10.9 15.0 4 Power / Plate 4, kW 21.2 18.8-Outlet Temperature of Hottest Channel, 'F 157 156 Peak Clad Surface Temperature, *F 219 224 Minimum DNBR5 2.29 2,17 Limiting Power Based on Min. DNBR, MW 11.5 10.8 Flow instability Ratio (FIR)e 2.12 2.11 Limiting Power Based on FIR, MW 10.6 10.6 1 Calculated by ANL using engineering uncertainty factors used in Ref. 23.

2 Calculated by ANL using revised engineering uncertainty factors (see Attachment 2).

3 Pressure drop across active fuel only.

4 Assuming 95% of power deposited in fuel.

6 Using modified Weatherhead Correlation 23.24 for DNB.

6 Using Forgan Whittle Correlation 25.2e with 11 = 25.

Safety limits for the reactor inlet temperature were calculated at the maximum reactor power of 5.5 MW and the minimum coolant flow of 1625 gpm. The results are -

shown in Table 11. Data for the GTRR HEU core are based on DNB. ANL results for the LEU core are based on both DNB and flow instability criteria. A safety limit for the reactor outlet temperature was then established by adding the average temperature rise across the core to the limiting inlet temperature. These results show that the HEU and:

LEU cores have nearly identical safety limits on the reactor inlet and oulet temperatures.

Table 11. Sately Limits on Reactor intet and Outlet Temperatures, GTRR-HEU1 ANL-LEU 2 Parameter M M Flow Instability Limiting Reactor Inlet Temp., 'F 172 171 170 Ave; Coolant Temp. Rise across Core, 'F 16 17 17 Limiting Reactor Outlet Temp., 'F 188 188 187 5 Data from Ref. 23 based on DNB cntenon.

2 Calculated using ANL engineering uncertainty factors in Attachment 2.

,= ..- .gg 7.2 Safety Limits in the Natural Convection Mode The current Technical Specifications state that the reactor thermal power shall not exceed two (2) kW in the natural convection mode. This specification is based on GTRR experience showing that no damage to the core and no boiling occurs without forced convection coolant flow at power levels up to 2 kW. We expect that this specification will also hold in the LEU core because the average power per fuel plate will be lower in the LEU core. Each LEU fuel assembly will contain 18 fuel plates while each HEU assembly contains 16 fuel plates.

7.3 Limiting Safety System Settings in the Forced Convection Mode The safety system trip setting in the current GTRR Technical specifications for power levels >1 MW and for power levels s 1 MW are shown in Table 12.

Table 12. Safety System Trip Settings Reador Power Reador Power Parameter Level >1 MW Level 51 MW Thermal Power 5.5 MW 1.25 MW Reactor Coolant Flow 1625 GPM 1000 GPM Reactor Outlet Temperature 139'F 125'F These safety system trip settings are based on a criterion 3 that there shall be no incipient boiling during normal operation. The criterion is applied by ensuring that the surface temperature at any point on a fuel assembly does not exceed the coolant saturation temperature at that point. This criterion is conservative because there is an additional margin of ~26*F between the D2 0 saturation temperature and the temperature at which onset of nucleate boiling occurs.

Figure 11 shows the combinations of reactor power, coolant flow rate, and reactor inlet temperature that were calculated to have zero subcooling (fuel surface temperature

= coolant saturation temperature) for HEU and LEU cores with 14 fuel assemblies. Data for the HEU core were reproduced from Fig.1 of Ref. 3. Table 13 provides the parameter combinations which correspond with the safety system trip settings shown in Table 12.

The trip setting of 139 F on reactor outlet temperature was obtained by adding the 16*F temperature rise across the core to the maximum inlet temperature _of 123*F. Similar considerations based on operation during the period 1964 to 1973 were applied to determine the safety system trip settings for power levels equal to or less than 1 MW.

23 Parameter combinations that have ::cro subcooling in the LEU core are shovm in-Table 13 and in Fig.11. Since the values for the LEU core are more conservative than those for the HEU core, the current safety system trip settings for the HEU core can also be used for the LEU core.

Table 13. Parameter Combinations for Zero Subcooling with 14 Assembly HEU and LEU Cores GTRR HEU ANL LEU Reactor Power, MW 53 5.0 5.0 5S 5.0 5.0 Coolant Flow Rate, gpm 1800 1625 1800 1800 <1f25 1800 Reactor inlet Temp., *F 114 114 123 114 114 128 Temp. Rise Across Core,'F 16 16 16 17 17 .17 Reactor outlet Temp., *F 130 130 132 131 131 145-Fig. 11. Thermal Hydraulic Limits- Based on 2ero Subcooling For Operation at Power Levels s 5 MW.

7.0 . . .. . .. .. . , , . ,, , .

6.5 --

/ -

1625 - GPM 6.0 j '%

3:

s

's '*s*,

. '*% GTRR HEU GTRR HEU  ! **,

N'*,j , N *%, <,

o 5.0

.g ANL LEO l l

, i,-

1 4.5  ;

i I

I a

N**' %'

! l .

I I I I

.0 ,

Nominalinlet Maximum inlet .

Temp.:114'F i li Temp.: 123*F I

3.5 90 95 100 105- 110 115 120 125 130 135 140 Reactor inlet Temperature, 'F

24 The results in Table 14 show that the degree of subcooling (ATsub) at the hottest spot of the limiting fuel assembly under normal operating conditions is expected to be 11'F in the LEU core and 8'F in the HEU core. Another criterion that is often used in research -

reactors is that the margin to onset of nucleate boiling (ONB) should be equal to or greater than 1.2. ONB occurs at a temperature of about 246*F, which is ~26'F- above the D2 0 saturation temperature of 220*F. The margin to ONB in the LEU core was computed by increasing the reactor power until ONB occured and dividing by the nominal reactor power of 5 hRV. These margins are adequate to ensure that the LEU core can be operated safely at a power level of 5 hBV.

Table 14. Margins to D 2O Saturation Temperature and ONB for 14 Assembly Cores Parameter GTRR-HEU1 ANL. LEU 2 Thermal Power, MW 5,0 5.0 Reactor Coolant Flow, gpm 1800 1800 ,

Reactor inlet Temp., 'F 114 114 -l ATsub, 'F 8 11 Margin to ONBS - 1,44 Limiting Power Based on ONB, MW -

7.2 5 Data from RefsI3 and 23, 2

8 Calculated using ANL. engineering uncertainty > factors in Attachment 2.

Using the Oergies and Rohsenow correlation .

Calculations were also done to examine the adequacy of the current safety system trip settings shown in Table 12 for operation at power levels equal to or less than 1 hnV.

Since data from analyses of the HEU core by Georgia Tech' were not available, calculations were done using the GTRR HEU and the ANL-LEU engineering uncertainty factors shown in Attachment 2, a thermal power of 1.25 MW, a reactor coolant flow of 1000 gpm, and an inlet temperature of 123'F. The results shown in Table 15 for the degree of local subcooling (ATsub) and the flow instability ratio indicate that the current trip settings on reactor power and coolant flow are conservative and are adequate to ensure the safety of the facilib- for operation at power levels that are s 1 MW.

Table 15. Selected h iydraulic Safety Margins with 14-Assembly Cores and Power s 1 MW.

Pararr* GTRR-HEU1 ANL-LEU 2 Thermal ser, MW 1.25 1.25 Reactor Coolant Flow, gpm 1000 1000 Reactor Inlet Temp., 'F 123 123 Peak Surface Clad Temp., 'F 162 164 ATsub, 'F 58 56 Flow instability Ratio 5.4 5.3 1

Calculated using GTRR HEU engineenng uncertainty factors in Attachment 2.

2 Calculated using ANL LEU engineering uncertainty factors in Attachment 2.

j'.*. j 25 7.4. Limiting Safety System Settings in- the Natural Convection Mode H The Technical Specifications state that the reactor thermal power safety- system setting shall not exceed 1.1 kW when operating in the natural convection mode. This specification is based on GTRR experience showing that the reactor can be operated at l one kW indefinitely without exceeding a bulk reactor temperature of 123'F. We expect 4 that this safety system trip setting will also be adequate for the LEU core.

8. COOLING TIME REQUIREMENTS The Technical Specifications for the HEU core state that containment integrity shall-be maintained when the reactor has been shutdown from a power level greater than 1 MW for less than eight hours. In addition, a minimum cooldown time of twelve hours is required before fuel assemblies are transferred out of the reactor.

Fuel melting and subsequent release of fission products could result from n' loss-of-coolant accident following reactor shutdown if sufficient decay heat is present.

Containment integrity is therefore required until the decay heat generation rate is less-than that required to melt the fuel plates. A limit of 450*C was set in the Technical Specifications as the upper value for a fuel plate temperature to preclude melting of the plates. The decay time needed to ensure that this temperature would not be reached was calculated in Ref. 23.

The analysis method and input parameters described in Ref. 23 were used to

! repr duce the results for the .HEU core. The same methodology was then used for the-L LEU core, with modification' of the input parameters appropriate for. the LEU fuel assembly design. A standard 3 week operating history consisting of 4.33 days at full pouar of 5 MW and 2.67 days shutdown was used for 14 assembly cores with HEU and-LEU fuel. The analysis in both cases was applied to a fuel assembly.which has been

subjected to a power peaking factor of 1.5 (see Fig. 8).- As in Ref. 23, the peak power was increased by 17% to account for the incremental heat contribution due to additional gamma heating from surrounding fuel assemblics in the core and was decreased by 15%

to take credit for an improved convection condition in the reactor vessel.

Three input parameters that were used for the HEU fuel assembly in Ref. 23_were modified for the LEU fuel assembly design: (1) the parameter hAcTan was reduced from

- 3.03 x 104 kWrC for an HEU plate to 2.88 x 104 kWrC for an LEU plate based on the heat transfer areas of the HEU and LEU fuel meat shown in Table 1, (2) the mass of-

___-___I___________._.m_____.i.____...-_____.__.__.-____

.g ,*:

26 aluminum associated with one fuel plate was reduced from 0.418 lbm for an HEU plate to 0.377 lbm for an LEU plate, mainly because U 3Si, fuel particles occupy approximately 31% -

of the fuel meat volume in an LEU plate; no credit was taken for the specific heat of the U3Si2 particles, and (3) most importantly, the maximum power per fuel plate in the LEU' assembly was reduced by a factor of 16/18 since an HEU assembly contains 16 fueled plates and an LEU assembly contains 18 fueled plates.  ;

The results for loss of-coolant from the reactor vessel after eight hours of cooling s showed a maximum plate temperature of 425*C in the HEU coro and 400*C in the LEU core. The maximum temperature occurred 45 minutes after loss of-coolant in the HEU core and 50 minutes after loss-of coolant in the LEU core. For the more confined-heat transfer situation, without gamma rays from other fuel assemblics, but with a restricted-i heat transfer volume, the maximum fuel plate temperatura after a twelve hour cooldown was calculated to be 361*C for an HEU plate and 340 C for. m LEU plate. The maximum temperature occurred 60 minutes after removal from the Hl!U core and 50 minutes after removal from the LEU core.

i We conclude that the current Technical Specification requirements on cooling times

[ are more conservative for the LEU core than for the HEU core. The most important factor L is the reduced power per plate in the LEU core. However, any. reduction of technical -

specification cooling time requirements for the LEU core should be based on-measurements in the GTRR.

9. LIMITATIONS OF EXPERIMENTS The Technical Specifications contain three limitations of experiments that could be affected by changing the fuelin the core from HEU to LEU:

l L a) The magnitude of the potential reactivity worth of each unsecured-i experiment is limited to 0.004 ok/k.

L b) The potential reactivity worth of each secured removable experiment-L is limited to 0.015 Ak/k.

c) The sum of the magnitudes of the static reactivity. worths of all-

, unsecured' experiments which coexist is limited to 0.015 ok/k.

l The objective of these specifications is to prevent' damage to the reactor and to limit I

radiation dose to personnel C N public in event of experiment failure. Qualification of -

L the PARET code that was used c:r the transient analysis is discussed first, followed by the calculated results.

27 j l

Comparison of Calculations with SPERT il Experiments 9.1 The PARET code 28 was originally developed at the Idaho National Engineering Laboratory for analysis of the SPERT III experiments, which included both pin type and plate type cores and pressures and temperatures in the range typical of power reactors.

The code was modified by the RERTR Program at ANL to include a selection of flow instability, departure from nucleate boiling, single- and two phase heat transfer j correlations, and properties libraries for light water and heavy water that are applicable to the low pressures, temperatures, and flow rates encountered in research reactors.

l To validate the PARET code for use with heavy water reactors, calculated and- I measured data were compared 29 for the SPERT-II BD-22/24 HEU corea (24 MTR-type fuel elements with 22 plates per element). This core is similar to the GTRR in design. The -

tests performed in the BD 22/24 core included only nondestructive transients. Calculated transient parameters shown in Ref. 29 are in very good agreement with the measured data and validate the PARET code for use in calculating transients in heavy water research reactors.

9.2 Inadvertent Reactivity insertions Due to Experiment Failure The consequences ofinadvertent step reactivity insertion of 0.4% Ak/k and 1.5% Ak/k in HEU and LEU cores with 14 fuel assemblies were evaluated. The model and methods that were used for analysis of the SPERT-II BD 22/24 HEU cores were also used to analyze the HEU and LEU cores of the GTRR.

Inputs to the code for analysis of the GTRR included the prompt neutron lifetime, ,

effective delayed neutron fraction, temperature coefficients of reactivity, and power distributions discussed in Sections 5.4 and 5.5. Temperature coefficients included contributions from only the coolant and the fuel. Axial power distributions for the average channel of the HEU and LEU cores were represented by chopped cosine shapes having peak to-average power densities of L19. In the hot channel, these axin.1 shapes were scaled to produce peak power densities in the limiting fuel assemblies of the HEU and the LEU cores that are consitent with the power distributions shown in Fig. 8.

Calculations were performed for step reactivity insertions of 0.4% and L5% &Jk with the reactor at nominal operating conditions of 5 MW thermal power, a coolant flow rate of 1800 gpm, and a reactor inlet temperature of 114 F. A scram signal was initiated when the reactor power reached the safety system overpower trip setting of 5.5 MW. A time

1

]* ..*I . .

28 delay of 100 ms was assumed between introduction of the scram signal and release of the.

shim safety blades. The results of these calculations are shown in Table 16.~

Table 16. Results of Assumed Step Reactivity insertions Due to_ Experiment Failure Parameter HEU Coro LEU Core Step Reactivity insertion, % Ak/k 0.4 1.5 0,4 1.5 Asymptotic Period, s 0.18 0.05 0.10 0.05- -;

Peak Power, MW 7.4 27.5 7.4 27.2 Peak Surface Cladding Temp., 'F 184 277 179 267 Peak Coolant Outlet Temp., 'F 135 - 135 -

A positive step reactivity change less than 0.4% Ak/k caused by the ejection or 1 insertion of experiments would result in transient behavior that' would not exceed the safety limits for the HEU or LEU cores that were discussed in Section 7.1. The peak power -

of 7.4 MW in both cores is well below the safety limits of 11.5 MW in the HEU core and 10.6 MW in the LEU core. Similarly, the peak coolant outlet temperatures are well below the limiting reactor outlet temperature of 188'F.

Step reactivt> insertions of 1.5% Ak/k would result in peak surface cladding temperatures that are far below the solidus temperature of1220*F (660*C)in the 1100 Al~

cladding of the HEU core and far below the solidus temperature of 1080 F (582*C)in the 6061 Al cladding of the LEU core. Thus, no damage to the fuel and no release of Sssion products is expected.

_-. _ _s - -

29

10. ACCIDENT ANALYSES A spectrum of accident scenarios .was evaluated by Georgia Tech in its safety documentationL3A for 5 MW operation. These scenarios included (1) failure of electrical power, (2) failure of various reactor components, (3) a startup accident in which one shim blade and the regulating rod were withdrawn simultaneously, (4) reactivity ciTects resulting from the melting of fuel plates, (5) assumed maximum positive reactivity l

insertion, and (6) the Design Basis Accident. A review of these scenarios concluded that only scenarios (3)-(6) could be affected by changing the fuel assemblies from HEU to LEU, and only these scenarios are addressed here. l 10.1 Startup Accident The worst case for a possible startup accident in the current HEU core was determined 3 to result from the simultaneous withdrawal of one shim blade and the regulating rod. An experiment was done in the GTRR to simulate reactor behavior when  ;

reactivity was added at rate of approximately 0.005 Ak/k per second starting from a power level of 5 kW. Within 3 seconds, the reactor was automatically scrammed by a positive period trip. The power level at the scram point was 6.5 kW. On this basis, it was concluded) that if the reactor were operating at 5 MW, the reactor would be scrammed by the overpower trip at 5.5 MW or the log N period systems would scram the reactor at a power level of no more than 7 MW. Since this is well below the 11.5 MW burnout power level of the GTRR, no fuel plate melting would be expected.

Calculations were done here using the PARET code for the HEU and LEU cores with 14 fuel assemblies in which reactivity was added at a rate of 0.005 ok/k per second starting from a power level of 5 MW Except for the reactivity addition rate, inputs to the code were the satne as those described in paragraphs 2 and 3 of Section 9.2. Both the HEU and LEU cores were scrammed by the overpower trip at 5.5 MW. A time delay of 100 ms was assumed between introduction of the scram signal and release of the shim-safety blades. Both cores reached a peak power of 5.9 MW at a time of 0.335 s after the transient was initiated. Peak surface cladding temperatures of 177 F and 172*F were reached in the limiting fuel assembly of the HEU and LEU cores, respectively. The peak power is well below the safety limits of11.5 MW in the HEU core and 10.6 MW in the LEU core. The peak surface cladding temperatures are far below the solidus temperature of 1220*F in the 1100 Al cladding of the HEU core and far below the solidus temperature of 1080 F in the 6061 Al cladding of the LEU core. Thus, no damage to the fuel and no release of fission products is expected.

E

30-10.2 Reactivity Effects of Fuel Plate Melting The reactivity ofrect of melting individual fuel plates within an assembly due to the blockage of individual flow channels was analyzed 4 for the current GTRR HEU core by estimating the reactivity change caused by removing the two central fuel plates in a fuel assembly at the core center. It was concluded that the loss of one or more fuel plates would result in a negative reactivity effect, Calculations were done for HEU and LEU cores with 14 and 17 fresh fuel assemblies using the reactor diffusion theory model described in Section 4.2 and in Pigs. 8 and 9. The results in Table 17 show that the reactivity effect of removing one or two fuel plates from a fuel assembly near the center of the HEU and LEU cores and replacing the fuel plate volume with D 20 is expected to be negative.

Table 17. Calculated Reactivty Effect of Removing Fuel Plates from a Fuel Assembly Near  !

the Center of the HEU and LEU Cores. '

Reactivity Change, % Ak/k 14 Assembly Cores 17 Assembly Cores BEU LEU HEU LEU 1 Fuel Plate Removed - 0.060 - 0.037 - 0.043 0.028 2 Fuel Plates Removed 0.127 - 0.078 - 0.090 0.060 10.3 Fuel Loading Accident During refueling operations, all control blades are required to be fully inserted and the top D2 0 reflector drained to storage. Calculations in Section 6 indicated that the shutdown margin with the blade of maximum worth stuck out of the core is expected to be - 7.1 i 0.3% Ak/k in the HEU core and - 8.8 0.2% Ak/k in the LEU core. The shutdown margins will be more negative with all shim safety blades inserted. In addition, the reactivity worth of the top reflector is at least 2% Ak/k.

The current GTRR surety analysis reportl analyzed a hypothetical fuel loading accident scenario assuming, in violation of established startup procedures, that the shim-safety blades are withdrawn so that the reactor is just sub-critical and that the D2 0 is at the normal operating level. A fresh fuel assembly was then assumed to be dropped into the center core position, resulting in a sudden reactivity insertion of 2.5% Ak/k. We consider this postulated scenario to be incredible and no analysis of this scenario is presented in this report. The maximum positive reactivity insertion is addressed in-Section 10.4.

i

]

c{ . . . ..

31  ;

10.4 Maximum Positive Reactivity insettlon The Technical Specifications limit the potential reactivity worth of each secured removable experiment to 1.5% Ak/k and the sum of the magnitudes of the static reactivity worths of all unsecured experiments which coexist to 1.5% Ak/k. The purpose of this analysis is to show that there is a sufficient margin between the maximum allowable reactivity worth of a single experiment and the maximum step reactivity insertion that can be tolerated without fuel damage, assuming failure of reactor scram systems.

Analysist for the current IIEU core used SPERT Il experimental data # as a basis for estimating the step reactivity insertion that would result in the onset of steam blanketing in the GTRR. In the present analysis, the PARET code was used to compute the step reactivity insertion required to initiate steam blanketing (film boiling)in both the SPERT.

11 B22/24 core and 14 assembly GTRR cores with HEU and LEU fuel. Some of the kinetics parameters and key PARET results are provided in Table 18. Power peaking factors are similar in the SPERT II and GTRR cores. The inverse period corresponding to the onset of steam blanketing as determined from the SPERT experimental datat a is about 13 s 1, The PARET code predicts the onset of film boiling for a step insertion of $2.0 (1.5% Ak/k) with an inverse period of 12 s4, in good agreement with experiment.

The same methodology was used to compuie GTRR cores with 14 fuel assemblies.

These cores have smaller coolant void coefficients than the SPERT II B22/24 core, but the step insertions needed to initiate film boiling (~$2.0) and the peak surface cladding temperatures (250-260*C) at the onset of steam blanketing are nearly the same. At the time of peak power, the energy deposited per plate is about the same in the SPERT and GTRR cores. The peak surface cladding temperature at the time of peak power is about 220*C in the GTRR cores and about 204*C in the SPERT core.

The SPERT-II B22/24 tests" indicate that even more extensive film boiling (or steam blanketing) does not result in temperatures that exceed the solidus temperature of the cladding. The most extreme case in the test series with a reactivity insertion of $2.95 (2.2% Ak/k) resulted in a peak surface cladding temperature of 337'C, a temperature far below the solidus temperature of 582 C for 6061 Al cladding. The GTRR SAR1 also notes that the maximum temperature for large insertions is primarily limited by the energy deposited in the plate with very little effect from the boiling heat transfer.

Since the behavior of the SPERT-II B22/24 and GTRR 14 assembly cores is very similar, a step reactivity insertion greater than 2.2% Akik would be required to initiate melting of the GTRR LEU core. The margin of at least 0.7% Ale /k above the maximum .

3 ?- .*

32 allowed reactivity wcrth of 1.5% Ak/k for a single experiment is sufficient to ensure that l

the facility is safe in the unlikely event that the maximum allowed reactivity were inserted in a step and the reactor scram system failed to function. J Table 18. Comparison of Kinetics Parameters and Onset of Steam Blanketing Results l

SPERTIl 14 Assembly GTRR l B-22/24 g '

1, Elf Prompt Neutron Generation Time, s 660 780 -

745 1 Beta Effective 0.0075 0.00755 0.00755 i Coolant Temperture Coeff., $/'C -0.00867 0.00874 -0.00689' ]

Void Coefficient, $/% Vold 0.0729 0.0509 -0.0442  !

1 Doppler Coefficient, $/*C -0.0 -0.0 0.00096 Operating Pressure, kPa 122 127 127 Step Reactivity insertion, S (% ok/k) 2.00 1.99 1.95 Inverse Period, s 1 12 19 19-  !

Energy / Plate at tm. kWs 31.8 31.2 32.0  :

Peak CladdingTemperature at tm. *C 204 218 225- 1 Peak Cladding Temperaturc at 252 257 257  !

Onset of Steam Blanketing. 'C where im is the time of peak power.

10.5 Design Basis Accident The Design Basis Accident for the HEU core was determined 4 to be the melting and release of the fission products from one fuel assembly into the containment atmosphere.

This accident was assumed to occur during a fuel transfer operation in which an irradiated fuel assembly was being moved from the core to the fuel storage area using a shielded transfer cask. Fuel assemblies are not normally discharged from the reactor-until'at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor shutdown. This ensures that sufficient fission product decay heat has been removed from the assembly - and that the. surface temperature of the fuel plates will not reach 450 C when the assembly is moved into the cask.

In spite of administrative controls, it is conceivable that a. fuel assembly could be withdrawn from the reactor prior to a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> cooldown period. Some or all of the fuel

~

plates within the assembly could then melt and release'some of their fission products into the containment atmosphere.

The source term for evaluating the radiological consequences of this accident was obtained4 by assuming that an HEU fuel assembly with equilibrium burnup was removed

p 3. =

33 4

from the core before the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> cooldown period. All of the plates in the fuel assembly melt and the isotopes ofiodine, krypton, and xenon were released to' the containment. The-methodology for the dose calculations and the results are shown in Ref. 4. The limiting.

dose is the thyroid dose from the iodine isotopes.

Since the HEU and LEU cores operate at 5 MW, neutron flux levels and equilibrium concentratious of iodine, xenon, and krypton will be about the same in the two cores.

Burnup calculation results shown in Section 5.3 concluded that the lifetime of the LEU core will be comparable to but probably less than that of the HEU. core. As a result, concentrations of the other fission products in LEU fuel assemblies will be the same'or less than those in HEU fuel assemblies. The exception is that the LEU assembly will:

contain' larger concentrations of plutonium isotopes. Reference 31 contains a detailed analysis comparing the radiological consequences of a hypothetical accident in a generic -

10 MW reactor using HEU and LEU fuels. This analysis concluded that the buildup of -

plutonium in discharge fuel assemblics with 235U burnup of over _50% does not

'significantly increase the radiological consequences over those of HEU fuel, Because-fission product concentrations in the GTRR HEU and LEU cores are expected to be comparable, the thyroid dose shown in Ref. 4 will be the limiting dose for both cores.-

. . _ . _ . ~. _ , . _ _

34

11. FUEL HANDLING AND STORAGE Three Technical Specifications apply to the handling and storage of fuel assemblies.

The objective of these specifications is to prevent in' advertent criticality outside of the reactor vessel and to prevent overheating ofirradiated fuel assemblies.

Irradiated fuel assemblics are stored in aluminum racks fastened to the side walls of a light water pool. There is one rack along each of the two walls and each rack can accomodate up to 20 assemblics in a linear array. The center to center spacing of'the assemblies is six inches and the separation between assemblies is about three inches.

A systematic nuclear criticality assessment 32 been done for infinite by infinite arrays of fresh LEU fuel assemblics with 235U contents between 225 and 621 grnms using the ORR fuel storage rack spacing specifications 33 oi 0.7 inch assemlity separation and 6.8 inch row separation. An assembly similar to the GTRR LEU assembly with a 235U content of 225 grams gave a kefT of 0.72, well below the maximum keft of 0.85 needed to ensure an adequate margin below criticality for storage of irradiated fuel assemblies.

The GTRR storage configuration discussed above will have keftless than 0.72.

Calculations 1 with HEU fuel assemblies have shown that four unirradiated fuel asserrblies cannot achieve criticality. Calculations of HEU and LEU cores shown in Section 5.2 indicate that a grouping of four LEU assemblies will be less reactive than the same configuration of HEU assemblies. Thus, the current specification that no more than four unirradiated fuel asse nblies shall be together in any one room outside the-reactor, shipping container, or fuel storage racks will also hold for the LEU assemblies.

Acknowledgment The authors wish to acknowledge significant contributions to this work by K. E. Freese, Argonne National Laboratory, who passed away in June 1991.

s

.35 REFERENCES

1. Safety Analysis Rt port for the 5 MW Georgia Tech Research Reactor December 1967. Final Safeguards Report for the Georgia 'rech Research Reactor, February 19G3.

2. Appendix A to Facility License No. R 97: Technical Specifications for the Georgia Tech Research Reactor, Docket No. 50-160, June 6,1974.

3. Letter, R. S. Kirkland, GTRR Reactor Supervisor, to USAEC, October 22,1971.

4. Letter, R. S. Kirkland, GTRR Reactor Supervisor, to USAEC, June 23,1972.

5. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Evaluation of Low Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Non Power Reactors", NUREG 1313, July 1988.

6. M. M. Bretscher and J. L. Snelgrove, "Comaarison of Calculated Quantities with Measured Quantities for the LEU Fue:.ed Ford Nuclear Reactor," Proc.

International Meeting on Research and Test Reactor Core Conversions from HEU to LEU Fuel, Argonne National Laboratory, Argonne, IL, November 8 10, 1982, ANL/RERTR/rM-4, CONF-821155, pp. 397 425 (1983).

7. M. M. Bretscher, " Analytical Support for the Whole-Core Demonstration at-the ORR," Proc.1986 International Meeting on Reduced Enrichment for Research and Test Reactors, Gatlinburg, TN, November 3-6,' 1986, ANI/RERTR/TM 9 , CONF-861185, pp. 287 301(1988).

8. R. -J. Cornella and M. M. Bretscher, " Comparison of Calculated and Experimental Wire Activations," Proc.1986 International Meeting on Reduced Enrichment for Research and Test Reactors, Gatlinburg, TN, November 3-6,1986, ANI/RERTR/rM-9, CONF-861185, pp. 302-309 (1988). -

9. M. M. Bretscher, J. L. Snelgrove, and R. W. Hobbs, "The ORR Whole-Core LEU Fuel Demonstration", Trans. Am. Nucl. Soc. 5fi,579 581 (1988).

10. M. M. Bretscher and J. L. Snelgrove, "The Whole-Core U3Si2 Fuel Demonstration in the 30-MW Oak Ridge Research Reactor', ANL/RERTRfrM-14, July 1991.

11. IAEA Guidebook on Research Reactor Core Conversion from the .Use of Highly Enriched Uranium to the Use of Low Enriched. Uranium Fuels, Addendum' on-Heavy Water Moderated Reactors, IAEA-TECDOC-324, Appendix F, pp. 182 250 (1985).

~

12 .J. E. Matos, E. M. Pennington, K. E. Freese, and W. L. Woodruff, " Safety Related Benchmark Calculations for MTR-Type Reactors with HEU, MEU, and LEU Fuels,"

paper included in IAEA Safety and Licensing Guidebook on Research: Reactor Core Conversions from HEU to LEU Fuel, Volume 2, Analytical Verification, Draft #7, June 1985,

13. B.A. Zolotar et al., "EPRI-CELL Code Description," Advanced Recycle Methodology Program System Documentation, Part II, Chapter 5 (Oct.1975).

b f I F

36 i

14. E. M. Gelbard and R. E. Prael, " Monte Carlo Work at Argonne National Laboratory,"

in Proc. NEACRP Mtg. Monte Carlo Study Group, July 1-3, 1974, Argonne, Illinois, ANlr75 2 (NEA-CRP L 118), Argonne National Laboratory , p. 201 (1975).

15 R. Blomquist, " VIM - A Continuous Energy Neutronics and Photon Transport Code", Proc. Topl. Mtg. Advances in Reactor Computations, Salt Lake City, Utah, March 28-31,1983, p. 222, American Nuclear Society (1983).

la K.L. Derstine, "DIF3D: A Code to Solve One , Two , and Three Dimensional Finite-Difference DiTusion Theory Problems," ANIr82-64, April 1984.

17. GTRR Operation Log Book, June 10 and June 11,1974.

IR ANL Internal Memorandum, K.E. Freese to J.E. Matos, " Monte Carlo Calculations for the GTRR Reference Core", March 12,1991.

l 19. R. P. Hosteny, " REBUS-2, Fuel Cycle Analysis Capability," ANL 7721, October 1978, m " Measured Reactivity Worths of Shim Safety Blades and Regulating Rod", GTRR Work Order #90317, Record No.188,26 September 1990.

21. GTRR Operation log Book, " Top Reflector Worth", May 17,1966.

22 K. Mishima and T. Shibata, " Thermal hydraulic Calculations for KUHFR with Reduced Enrichment Uranium Fuel," KURRI TR-223 (1982), and K. Mishima, K.

! Kanda, and T. Shibata, " Thermal hydraulic Analysis for Core Conversion to Use of Lo v enriched Uranium Fuels in the KUR," KURRI TR-258 (1984).

o 21 Appendix A to Facility License No. R 97: Draft Technical Specifications for the Georgia Tech Research Reactor, Docket No. 50-160, May 26,1972.

24. R. J. Weatherhead, " Nucleate Boiling Characteristics and the Critical Heat Flux Occurrence in Subcooled Axial Flow Water Systems," ANL-6674,1963.

l 21 R. H. Whittle and R. Forgan, "A Cor. elation for the Minima in the Pressure Drop Versus Flow Rate Curves for Subcooled Water Flowing in Narrow Heated Channels," Nuclear Engineering and Design, Vol. 6, (1967) pp. 89 99, m " Guidebook on Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels", IAEA TECDOC-233 (1980) pp.99-106.

L 27. A.E. Bergles and W.M. Rohsenow, "The Determination of Forced-Convection l- Surface-Boiling Heat Transfers," Transactions of the ASME Bfi(Series C - Journal of

Heat Transfer), pp. 365-371 ( August 1964).

2a C.F. Obenchain, "PARET - A Program for the Analysis of Reactor Transients." IDO-17282, Idaho National Engineering Laboratory (1969).

29. W. L. Woodruff, " Additional Capabilities and Benchmarking with the SPERT l Transients fo Heavy Water Applications of the PARET Code," Proc. XIIth l International Meeting on Reduced Enrichment for Research and Test Reactors, Berlin,10.-14. September 1989., Konferenzen des Forschungs-zentrums Julich(1991).

l . o 1 37 l

l

n J. E. Grund,"Self Limiting Excursion Tests of a flighly Enriched Plate-Type D20-Moderated Reactor, Part 1. Initial Test Series", USAEC lleport IDO 16891, Phillips -

Petmleum Co., July 12,1963.

31. W.L. Woodruff, D.K. Warinner, and J.E. Matos, "A Radiological Co isequence Analysis with IIEU and LEU Fuels," Proc.1984 International Meeting on Reduced Enrichment for Itesearch and Test Reactors, Argonne National Laboratory, Argonne, IL, October 15 18, 1984, ANI/RERTilffM 6, CONF 8410173, pp. 472 490 (July 1985).

32 It.B. Pond and J.E. Matos," Nuclear Criticality Assessment of LEU and IIEU Fuel Element Storage," Proc.1983 International Meeting on Reduced Enrichment for llescarch and Test Reactors, Japan Atomic Energy Reeearch Institute, Tokai, Japan, October 24 27,1983, JAERI M 84 073, pp. 416-425 (May 1984).

In J.T. Thomas, " Nuclear Criticality Assessment of Oak Ridge Research Reactor Fuel Element Storage," ORNI/CSDfrM 58, Oak Ridge National Laboratory (1978).

at W.L. Woodruff, "Evaluntion and Selection of Ilot Channel (Peaking) Factors for Research Reactor Applications," Proc. X International Meeting on Reduced Enrichment for Research and Test Reactors, Buenos Aires, Argentina, September 28

- October 1,1987, pp. 443-452.

__m.___-_____------ --

.o o* 38 ATTACHfAElli 1 ISOTHERfAAL REACTIVITY CHANGE COMPONENTS FOR AN HEU CORE WITH 17 FRESH FUEL ASSEMBLIES The purpose of this attachment is to analyze the components of the reactor isothermal temperature coemeient for the heavy water in various regions of the reactor tank. The calculations were done for an HEU core with 17 fresh fuel assemblics.

Reactivity component values for heavy water outside of the fuel assemblies are expected to be very similar in LEU cores. Reactivity coefficients for the fuel and coolant shown in Table 5 of Section 5.5 are also very similar in IIEU and LEU cores.

The reactor was divided into three regions: (1) the heavy water inside the fuel assemblies, (2) the heavy water between fuel assemblies, and (3) the heavy water reflector.

On the outer edges of the core, a heavy water thickness equal to one half the water thickness between fuel assemblics was included as part of the inter assembly water. The remaining heavy water in the tank is referred to as the reflector. Calculations were performed by separately changing the water temperature and density in each region while holding the water in the other two regions at 23 C. Least squares fits were then done to obtain reactivity values at intermediate temperattles.

Reactivity changes relative to 20 C for water temperature and density changes in each region are shown in the attached figure. Increasing the heavy water temperature and decreasing its density in the fuel assemblies and between fuel assemblies results in negative reactivity changes for both the temperature and density components. In the reflector, the water density component is negative, but the water temperature component is positive. Combined temperature and density efTects for each heavy water region show that reactivity changes with increasing water temperature are negative for the fuel assembly and inter assembly water. In the reflector, net reactivity changes are slightly positive for heavy water temperatures up to about 60"C and then become negative with further increnses in temperature.

The sum of the temperature and density components over the three heavy water regions is negative for the entire temperature range between 20'C and 100*C, A direct calculation of the isothermal temperature coeflicient in which all changes were made simultaneously gave results which are in good agreement with those obtained by summing the various cotaponents.

- - ___.____-._.m.m .,___.________.,______ __

, .. 39 ATTACHME!JT 1 Figure 1 1. Calculated Reactivity Changes (in % ok/k) with Temperature for a Fresn GTRR HEU Core with 17 Fresh Fuel Assemblies FA Fut xtsently Wator: lA = Inter Assembly Wator; Reft . Reflector Water HEU 17: Reactivity Chango Components for Fuel Assembly, Inter Assembly, and Reflector Water 0.6 , , , .

. HEU 17: Reactivity Changes with Temperature for Fuel Assernbly, inter Assembly, and Reflector Water 0.5 , -

0.4 , Reft Temp only , , .

i<

0.0 **************'-

0.2 -

% s ,.s****...,,

N * * *. Refl Only' nP N 0.5 -

N *

~'%

t c  %

O g x-0.2 -

g kg FA Density only g1.0 j m

- Ns ,

3 4 g \%

M

" g( FA Temp C'nly x

3 \

FA Only -

C -0.4 w B 1.5 -

\\ g

\ m \ \

\

0.6 -

\g  %

lA Temp only 4 2.0 -

'IA Only *

\%IA Density Only

' Curves Show Sum of Temperature and '

hofi Dons)ity Only Density Components by Water Region 0.8 ' '

20 40 60 80 100 120 Water Temperature, 'C Water Temperature. 'C

[_

,,,,,,_-7.g v.-mypr--++r- - ' + - - " ' ' " ^ '~

. *%* 40 ATTACHMEllT 2 El1GINEERll1G Ut1 CERTAINTY FACTORS This attachment addresses the engineering uncertainty factors (or hot channel factors) that were used to compute the thermal hydraulic safety limits, safety margins, and safety system trip settings in IIEU and LEU cores with 14 fuel assemblies. The rationale for choosing these factors and the method used to combine them are outlined along with a summary of results for the lieu and LEU cores. -

The PLTEMP code 22 used in the ANL analyses allows for introduction of three separate engineering hot channel factors as they apply to the uncertainty in the various parameters (as opposed to a single lumped factor). The three hot channel factors are:

Fq for uncertainties that influence the heat fhtx q 14 for uncertainties in the temperature rise or enthalpy change in the coolant Ph for uncertainties in the heat transfer coefficient h.

The co6e also allows introduction of nuclear peaking factors for the radial, Fr, and axial, Fz, distributions of the heat flux.

While there is no generally accepted method for the selection of hot channel factors, these factors are normally a composite of sub factors, and the sub factors can be combined either multiplicatively, statistically [ rd -1 + k (1 - r et ) 2 ), or as a mixture of the two. A detailed description of methods for calculating hot channel factors is contained in Ref. 34. The multiplicative method of combining the sub factors is very conservative and somewhat unrea l istic. The statistical method recognizes that all of these conditions do not occur at the same time and location.

The engineering uncertainty factors that were combined multiplicatively and used by Georgia Tech in analyses 3m of the 11EU core are shown in Table 21. The factors that were combined statistically and used by ANL for calculations of the IIEU and LEU cores are shown in Table 2 2.

Key thermal hydraulic safety limits and safety margins for the 11EU and LEU cores computed using the Georgia Tech factors and the ANL factors are compared in Table 2 3.

Results for the IIEU core obtained using ANL's statistical treatment of the engineering uncertainty factors agree well with the analyses performed by Georgia Tech. Except for the reactor power limit, data for the LEU core are comparable to or more conservative than those for the llEU core. An LEU core power limit of 10.6 MW based on the flow instability criterion is considered to be adequate.

I

,

• D

Table 21. GTRR HEU Engineering Uncertainty Factorsrs Uncertainty Fq Fb Fh Equivalent Diameter -

1.09 -

Fuel Distribution 1.03 1.03 -

Axial Flux Peaking 1.19 - -

Power Level Measurement 1.03 1.03 -

Fbw Distribution - Plenum -

1.07 -

Flow Distribution Channel -

1.10 -

Multiplicative Combination 1.26 1.36 1.0 Table 2 2. ANL-HEU and ANL-LEU Engineering Uncertainty Factors ANL-HEU Factors ANL LEU Factors Uncertainty Fq Fb Fh Fq Fb Fh Fuel Meat Thickness a 1,04 - -

1.04 - -

23sU Loading 1.03 b 1.03 b -

1.03 e 1.03 e .

23sU Homogeneity 1.03 d 1.03 d -

1.20 e 1,10 e .

Coolant Channel Spacing -

1.17 I 1.03 I - 1.22 9 1.04 9 Power Level Measurement d 1.03 1.03 -

1.03 1.03 -

Calculated Power Density h 1.10 1.10 -

1.10 1.10 -

Coolant Flow Rata h 1.10 1.00 -

1.10 1.08 Heat Transfer Coefficient h 1.20 - -

1.20 l

l Statistical Combination 1.12 1.23 1.30 1.23 ' 28 1.31 Muttlplicative Combination 1.26 1.55 1.33 1.41 1.72 1.35 l a Derived from fuel plate thickness specification of 5012 mils.

I b Assumed to be the same as for the LEU plate.

c From LEU fuel plate loading specification of 12.5 0.35 g M5U.

l d GTRR HEU value from Table 21.

l e From LEU plate fuel homogeneity specification.

i f Computed based on coolant channel spacing of 106 10 mils and fuel plate thickness specification of 50 2 mils in HEU assembly (see Ref. 34 for calculation method).

g Computed based on coolant channel spacing of 89 10 mils and fuel plate thickness specification of 50 2 mils in LEU assembly (see Ref. 34 for calculation method).

h Assumed values.

{

l The ANL factors for Fq andb F were combined statistically using the relation r - 1+ YL1 - rp 2.

I The corresponding factor for Fh was obtained by statistically combining the factors for the coolant l

channel spacing and the coolant flow rate and multiplying the result by the factor for the heat l transfer coefficient.

1

toe i 42 Table 2 3. Comparison of Key Thermal Hydraulic Safety Parameters for HEU and LEU Cores wth 14 Fuel Assemblies.

Reactor Power Limits for a Maxiruum inlet Temperature of 123'F Reactor Coolant flow nom GTER HEU NJL HEU NLLQ Reactor Power level (MW) for DNOM24 760 5.5 5.7 5.3 1625 11.5 11.9 10.8 Reactor Power Level (MW) for Flow instabilityrs2s 760 5.3 5.1 5.0 1625 10.6 11.0 10.6 Thermal Hydraulle Data with Min. Coolant Flow of 1625 GPM and Max, Inlet Temp. of 623'F.

GTRR HEU ANL-HQ NJL.1 FU Coolant Velocity, m/s 2.44 2.44 2.61 Friction Pressure Drop1 , kPa 10.9 11.0 15.0 Power / Platea , kW 21.2 21.2 18.8 Outlet Temperature of Hottest Channel,'F 157 154 156 Peak C!ad Suriace Temperature,'F 219 229 224 Minimum DNDR3 2.29 2.37 2.17 Umaing Power Based on Min. DNBR, MW 11.5 11 9 10.8 Flow instabliny Ratio (FIR)4 2.12 2.19 2.11 Umning Power Based on FIR, MW 10.6 11,0 10.6 1 Pressure drop across active fuel only. 3 Using modified Weatherhead CorrelationM24 for DNB.

2 Assuming 95% of power deposited in fuel. 4 Using V(nittle Forgan Conelation 25 2s wnh n - 25.

Safety Limits on Reactor Inlet and Outlet Temperatures, GTRR44EU ANL HEU ANL-tFU Parameter DM DM Flow intt DE Flow Inst-Limning Reactor inlet Temp.,'F 172 175 172 171 170 Aw. Coolant Tomp. Rise across Core. 'F 16 17 17 17 17 Limning Reactor Outlet Temp.,'F 188 192 189 188 187 Margins to D20 Saturation Temperature and OND Parnmeter GTAR-HEU ANL-HEU ANL-LEU Thermal Power, MW 5.0 5.0 5.0 Reactor Coolant flow, gpm 1800 1000 1800 Reactor Intet Temp., 'F 114 114 114 A Tsub,*F 8 5 11 Margin to ONB1 - 1.34 1.44 Umrung Power Based on ONB, MW -

6.7 7.2 1

Using the Bergies and Rohsenow correlation??.

Power Levels and Inlet Temperatures for Zero Subcooling at a Coolant Flow of 1800 GPM Parametci GTER HEU MitHEU ANL LEU Thermal Power, MW 5.45 5.35 5.6 Reactor intet Temp., 'F 114 114 114 Thermal Power, MW 5.0 5.0 5.0 Reactor intet Temp., 'F 123 122 128

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