ML062350107

University of Florida Training Reactor, Responses Dated August 4, 2006 to NRC Request for Additional Information for HEU to Leu Fuel Conversion

ML062350107
Person / Time
Site:	05000083
Issue date:	08/04/2006
From:	Vernetson W Univ of Florida
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML062350107 (29)

Text

UNIVERSITY OF FLORIDA TRAINING REACTOR LICENSE NO. R-56 DOCKET NO. 50-83 SUBMITTAL OF RESPONSES DATED AUGUST 4, 2006, TO NRC REQUEST FOR ADDITIONAL INFORMATION FOR THE CONVERSION FROM HIGH-ENRICHED URANIUM TO LOW-ENRICHED URANIUM FUEL REACTED VERSION SECURITY RELATED INFORMATION REMOVED IN ACCORDANCE WITH 10 CFR 2.390(d)(1)

Redacted text and figures blacked out or denoted by brackets

UNIVERIYp College of Engineering 202 Nuclear Sciences Bldg.

Department of Nuclear & Radiological Engineering PO Box 118300 UFTR Nuclear Facilities Gainesville, FL 32611-8300 352-392-1401 352-392-3380 Fax www.nre.ufl.edu August 4, 2006 ATTN: Document Control Desk Amendment 26 U.S. Nuclear Regulatory Commission (previously Amendment 25)

Washington, DC 20555 UFTR Technical Specifications Addendum 2 University of Florida Training Reactor, Facility License: R-56, Docket No. 50-83 Request for Change in Technical Specifications Approving HEU to LEU Conversion With Responses to Requests for Additional Information A proposed amendment to the UFTR Technical Specifications (R-56 License) for conversion from high enriched uranium (HlIU) fuel to low enriched uranium (LEU) fuel affecting pages 4, 5, 6, 7, 8, 9, 13, 15, 16, 21, 23, 24, 26 and 38 of the approved Tech Specs was submitted by letter dated June 19, 2006.

A second submittal with a letter dated July 20, 2006 corrected minor typographical errors in the Technical Specification pages of the June 19, 2006 submittal that were not intended to be changed per discussions with NRC Senior Project Manager Al Adams on July 19, 2006 plus other changes. Tech Spec pages in the July 20 submittal to replace those in the June 19 package were pages 4, 6, 8, 9, 15, 16 and 23.

The proposed changes, updated with this submittal, will continue to constitute Amendment 26 to the UFTR R-56 License as noted on the text pages. The corrected pages as submitted are considered to have minor safety significance. These two changed Tech Spec pages simply replace the corresponding pages in the previous two submittals to account for Limiting Safety System Settings necessary based on the analyses of the allowed tolerances in the fuel bundle fuel coolant channel spacing in Attachment HI to this letter. Tech Spec pages affected which replace those in the earlier submittals are pages 5 and 8 in Attachment Ito this letter.

First, on page 5, in Section 2.2, Limiting Safety System Settings, specification (2) is now split into two alternative specifications where the LSSS is set at 36 gpm for fuel channel spacing tolerances up to 15 mils and 41 gpm for fuel channel spacing tolerances up to 20 mils. Similarly, in specification (3),

the average primary coolant inlet temperature specification in (3Xa) is now split into two alternative specifications at 1090 F for fuel channel spacing tolerances up to 10 mils and 990 F for fuel channel spacing tolerances up to 20 mils.

Second, on page 8, in Table 3.1, Specifications for Reactor Safety System Trips, under Automatic Trips, the specification for Primary Cooling System, Low Inlet Water Flow is split into two alternative specifications at < 36 gpm for fuel coolant channel spacing tolerances at < 15 mils and < 41 gpm for fuel coolant channel spacing tolerances : 20 mils. Similarly, the specification for High Primary Coolant The Foundationfor The Gator Nation An E"qual Opprtunity Inution

T T'Nmr f' 1D. Aim tf A ')n b b- , W Average Inlet Temperature is split into two alternative specifications at > 1090 F for fuel coolant channel spacing tolerances at : 10 mils and Z: 990 F for fuel coolant channel spacing tolerances at : 20 mils.

These two corrected pages as submitted are considered to have minor safety significance. These two changed pages simply replace the corresponding pages in the previous submittals. These proposed changes have been reviewed in progress by UFTR management and some members of the Reactor Safety Review Subcommittee (RSRS), as well as formally by the RSRS Executive Committee prior to submittal, with all concurring on this evaluation.

Based on a telephone conversation on August 2, 2006 with project manager A] Adams, included as Attachment HfI is additional discussion and analyses showing that existing fuel storage facilities continue to meet the criticality requirements of Tech Spec Section 3.7(6), Fuel and Fuel Handling, requiring kl_ to be less than 0.8, as well as the criticality safety requirements of Tech Spec Section 5.8, Fuel Storage.

This supporting analysis is supplied following the format of NUREG-1537.

Finally, the question about consistency of control blade reactivity worths is addressed in Attachment IV including replacement pages for the earlier submittal.

This entire submittal consists of one signed original letter of transmittal plus Attachment Icontaining the two replacement Tech Spec changed pages to the requested Amendment 26 plus Attachment H and Attachment HI containing the reference analyses as well as Attachment IV.

We appreciate your consideration of this submittal. Please advise if further information is needed.

Sincerely, William G. Vemetson Director of Nuclear Facilities Email: vcrnel@ufl.edu Phone: 352-392-1408 ext. 317 WGV/dms Attachments I, I, III & TV cc: Al Adams, NRC Project Manager Craig Bassett, NRC Inspector Reactor Safety Review Subcommittee Sworn and subscribed this 4'day of August 2006.

TerriN*L Sparks NotaryPublicCommfission of DD346496 VVExpires August 12,.20M

A TTACHMENT I REPLACEMENT TECH SPEC CHANGED PAGES FOR ADDENDUM 2 TO THE TECH SPEC AMENDMENT 26 JUNE 19, 2006 SUBMITTAL

Specifications: The limiting safety system settings shall be (1) Power level at any flow rate shall not exceed 119 kW.

(2) The primary coolant flow rate shall be (a) greater than 36 gpm at all power levels greater than 1 watt Ifthe fuel coolant channel spacing tolerance Is :5 15 mils (b) greater than 41 gpm at all power levels greater than I watt Ifthe fuel coolant channel spacing tolerance Is .* 20 mils.

(3) The average primary coolant (a) Inlet temperature shall not exceed 1090 F when the fuel coolant channel spacing tolerance Is :5 10 mils (b) inlet temperature shall not exceed 990 F when the fuel coolant channel spacing tolerance is i- 20 mils (c) outlet temperature shall not exceed 1550 F when measured at any fuel box outlet.

(4) The reactor period shall not be faster than 3 sec.

(5) The high voltage applied to Safety Channels I and 2 neutron chambers shall be 90% or more of the established normal value.

(6) The primary coolant pump shall be energized during reactor operations.

(7) The primary coolant flow rate shall be monitored at the return line.

(8) The primary coolant core level shall be at least 2 in. above the fuel.

(9) The secondary coolant flow shall satisfy the following conditions when the reactor is being operated at power levels equal to or larger than I kW:

(a) Power shall be provided to the well pump and the well water flow rate shall be larger than 60 gpm when using the well system for secondary cooling.

or (b) The water flow rate shall be larger than 8 gpm when using the city water system for secondary cooling.

(10) The reactor shall be shut down when the main alternating current (ac) power is not operating.

(11) The reactor vent system shall be operating during reactor operations.

(12) The water level In the shield tank shall not be reduced 6 In.below the established normal level.

Bases: The University of Florida Training Reactor (UFTR) limiting safety system settings (LSSS) are established from operating experience and safety considerations. The LSSS 2.2.3 (1) through (10) are established for the protection of the fuel, the fuel cladding, and the reactor core Integrity. The primary and secondary bulk coolant temperatures, as well as the outlet temperatures for the six fuel boxes, are monitored and recorded In the control room. LSSS 2.2.3 (11) are established for the protection of reactor personnel In relation to accumulation of argon-41 In the reactor cell and for the control of radioactive gaseous effluents from the cell.

LSSS 2.2.3 (12) are established to protect reactor personnel from potential external radiation hazards caused by loss of biological shielding.

Amendment 15 Amendment 25 Amendment 26 5

Table 3.1 Specifications for reactor safety system trips Type of safety Specification system trip Automatic Trips Period less than 3 sec Full Power at 119% of full power Full Loss of chamber high voltage ( :10%) Full Loss of electrical power to control console Full Primary cooling system Rod-drop Loss of pump power Low-water level in core (< 42.5")

No outlet flow Low inlet water flow

(< 36 gpm for fuel coolant channel spacing tolerance at r 15 mils;

< 41 gpm for fuel coolant channel spacing tolerance at

• 20 mils)

Secondary cooling system (at power levels above I kW) Rod-drop Loss of flow (well water < 60 gpm, city water < 8 gpm)

Loss of pump power High primary coolant average Inlet temperature Rod-drop

(: 1099 F for fuel coolant channel spacing tolerance at <210 mils;

> 99° F for fuel coolant channel spacing tolerance at 120 mils)

High primary coolant average outlet temperature ( >1 55° F) Rod-drop Shield tank Rod-drop Low water level (6" below established normal level)

Ventilation system Rod-drop Loss of power to dilution fan Loss of power to core vent system Manual Trips Manual scram bar Rod-drop Console key-switch OFF (two blades off bottom) Full Amendment 15 Amendment 26 8

ATTACHMENT H ANALYSES OF ALLOWED TOLERANCES IN FUEL BUNDLE FUEL COOLANT CHANNEL SPACING FOR TECH SPEC AMENDMENT 26 CONSIDERATION

UNIVERSITY OF FLORIDA TRAINING REACTOR SUPPLEMENTAL INFORMATION August 3, 2006

References:

1. UFTR Conversion Proposal, December 2,2005

2. UFTR Responses to NRC Request for Additional Information, June 9,2006 INTRODUCTION Initial inspection and measurement of LEU fuel assemblies manufactured by BWXT in July 2006 showed a variation in the water channel thickness that was larger than expected. In this design, the ends of the plates are separated by aluminum spacers and are bolted together. Aluminum spacers are welded onto the edges of the plates at about half their height. This fuel assembly design has been modified by BWXT to include the "comb" design shown in Figure 1 that will physically separate the fuel plates at the nominal quarter-points along the fuel plate length. In this modified design, the tolerance on the minimum water channel spacing is expected to be a maximum of+/- 20 mils. The nominal water channel spacing at the bolted ends of the fuel assembly on the manufacturing drawings is 110 - 112 MRils, giving a minimum water channel spacing of 90 mils.

The actual minimum water channel spacing of each water channel in each finished fuel assembly will be measured by BWXT and made available by August 10, 2006. In case any one of the channel spacings is less than 90 mils, another comb will be added to maintain a minimum channel spacing of 90 mils.

Thermal-hydraulic analyses are provided in this supplement assuming tolerances on the water channel spacing of 10, 15, and 20 mils, along with repositioning of the fuel assemblies in the fuel boxes due to the 65 mil protrusion of the combs beyond the ends of the fuel plates. The same hot-channel-factor methodology described in Ref. 2, Appendix Q8 and related appendices, was used to compute curves of true reactor power versus the true coolant flow rate at which onset ofnucleate boiling (ONB) would occur for several values of the average coolant inlet temperature.

The three key parameters to be examined are the true minimum coolant flow rate, the Limiting Safety System Setting (LSSS) for the coolant flow rate, and the normal operating coolant flow rate. The corresponding three parameters for reactor power level are not affected and will remain the same as in Reference 2.

CALCULATIONAL MODEL AND RESULTS New Hot Channel Factors and Revised Model The steady-state thermal hydraulic analysis described in Ref. 2 assumed hot channel factors based on a 1 rail tolerance for the water channel spacing, as shown on the manufacturing drawings at the bolted ends of the fuel assemblies. In the model, the ends of the fuel plates were assumed to be in contact with the fuel box wall. These new analyses account for two effects - the larger tolerance on the water channel spacing described above and the repositioning of the fuel assemblies in the fuel boxes due to the thickness of the combs extending beyond the edges I

4....190 4.005--

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(~9) 1.793 (2.340) 1.t 30 Se Detail A 1.J 7 2X(10) 4-4-4-4-'-

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.510 DEEP I.000 +/-.010 45*X.020e.00 2XR.060,0-010 I- Notes 1.

2.

Remove bums and sharp edges Material b AN6061.

.Onwn 9 Tolerances Unless Specified Reviewed By:

.063 Approed By:.

Dim. *.002 ReviJ,. 0 Figure 1 - Modified fuel assembly design by BWXT to include the "comb" design 2

of the fuel plates. The combs introduce into the computer model another bypass channel pathway that is located between the fuel plate ends and the box wall. In these new analyses, the ends of the fuel plates are moved away from the wall of the fuel box by 0.065 inches - the height of the comb above the "teeth", as shown in Figure 1.

However, the newly created slot is actually blocked by the width of the comb (0.190 inches), at the quarter-points along the length of the assembly. There will be a small bypass flow through this pathway that is conservatively accounted for by including it as a clear channel, ignoring the blockage created by the presence of the combs. The pressure drop introduced by the combs is not accounted for because it is very small. As a result, the computed coolant flow through the heated channels will be reduced, and the ONB margin and other parameters will be conservative. The presence of the comb also reduces the width of the central channel slot between fuel assemblies. This is a significant beneficial change because it reduces the bypass flow through the central slot.

The hot channel factors shown in Table 1 were computed for coolant channel spacing tolerances of 10, 15, and 20 mils using the methodology described in Ref. 2, Appendix Q8 and related appendices. All other hot channel factors shown in Ref. 2, Appendix Q8 are unchanged.

Table 1. Hot Channel Factors as a Function of Tolerance on Coolant Channel Width A!Hot Channei Factor, im 0 mil "*.15 mil '.?---20 roil, Fb 1.21 1.39 1.58 1.84 Frim 1.23 1.24 1.26 1.29 Calculated Results For convenience, Figure 2 shows the curves from Ref. 2, Question 8 and Appendix Q8 of true reactor power versus true coolant flow rate at which ONB occurs for average coolant inlet temperatures of 86, 100, and 110 OF. The key parameters are shown for reference purposes in Table 2.

Table 2. Key Parameters for Reactor Power and Coolant Flow Rate in Ref. 2.

!Parameter Valfie. Paramieter-. Value True Maximum Power 125 kW True Minimum Flow Rate 34 gpm LSSS Power 119 kW LSSS Flow Rate 36 gpm Maximum Operating Power 100 kW Operating Flow Rate 43 gpm The new power versus coolant flow rate curves for water channel spacing tolerances of 10 mils, 15 mils, and 20 mils, along with repositioning of the fuel assemblies in the fuel box to account for the thickness of the combs are shown in Figures 3, 4, and 5, respectively.

3

275 250- Figure 2. Figure In Final Responses to NRC RAI 225 I I I linm110 I 200 175

-f I50 True ~x. Fw o: 125MN 125 1501- TrueMx oe:15k 100-I.-

75 True ni-um Flow Rate: 34 qpm t;- OPERATING REGION 25 0I I 15 20 25 30 35 40 45 5o 55 True Coolant Flow Rate, gpm Figure 2. True Power versus True Coolant Flow Rate at Which ONB Occurs for Three Inlet Temperatures. (Figure from Ref. 2, Question 8 and Appendix Q8.)

4

275 -

-1=-I= !- 4=1 1 1 1 -I I I - I I I I ý Figure 3. 10 mil tolerance on Water Channel 250 - 48pacing and 0.06W"Rapoltloning of Each SAsembly Due to Combs Tin 15SF Z25 175 LSS w~111[W Tru"x1 Power" 10 MW I.$pSafnPower.

10 MW I125, 2 100

75. SOPERATI? GREd S0. -iTrue Mrinmum Flow Raw: 34 gpm

- LSSSFkvwRako:30gpm

- peratng low Rato:43 gpm 25.

o * -+

20 25 30 3C 40 45 so S5 True Coolant Rlow Rate, gpmn Figure 3. True Power versus True Flow Rate at Which ONB Occurs with 10 mil Tolerance on Water Channel Spacing and Repositioning of Assemblies in Fuel Box to Account for Thickness of Combs.

5

250. 3 lqiz 4 15wiItdoramn~nV a~WarOhaoI

-4Spadrig andl M0.05 Rpdtngc(Ekkdi

-41 byDet ab I

-- 4-4 Tn-1WF0 k

-tTho Mm Pom125 MN 150 S LSSSRPDer.119kW M.QwftP.g .100 1jWj 1125 Io I100.

Tin -100 F44

75. FaPr K-etTe'n'100 6

-L SInaw~a 30 gprn

50. 0pro~ng Imbb 43Wgi ~j~$F5~AnIa3 ~CNfaFAx. T,~ -110

,PvrL~kWetTynwIv 11 ThesMdfrn~ovvM~

LSSSRONOW C0W8Urg Rcwt, a.j- ===CM=M; 15 2D 25 30 25 40 45 50D 55 Thm CodaztFlwRawI, gy.

Figure 4. True Power versus True Flow Rate at Which ONB Occurs with 15 mil Tolerance on Water Channel Spacing and Repositioning of Assemblies in Fuel Box to Account for Thickness of Combs.

6

I Figure 5. True Power versus True Flow Rate at Which ONB Occurs with 20 rail Tolerance on Water Channel Spacing and Repositioning of Assemblies in Fuel Box to Account for Thickness of Combs.

7

Based on these results, two sets of key flow rate settings are proposed, depending on the maximum value of the average inlet temperature:

True maximum inlet temperature = 100 *F (LSSS on Inlet temperature of 99 *F)

Table 3 gives the proposed settings for different flow rates depending on the different water channel spacing tolerances.

Table 3 - Flow Rate Settings for Different Water-Channel-Spacing Tolerances Type*of flowr rate True Min. (gpm) 34. 34 3 39 LSSS (gpm) . 36 3 36 41 Operatlng (gpm) 3 43. 43 48 Note that only the case with 20 mil tolerance requires changes to the values in Ref. 2.

True maximum Inlet temperature =1 10 F (LSSS on inlet temperature of 109 *F)

Table 4 gives the proposed settings for different flow rates depending on the different water channel spacing tolerances.

Table 4- Flow Rate Settings for Different Water-Channel-Spacing Tolerances

-Typeof rate .flo True Min. (gpm) 34..337' LSSS (gpm) 1 3T6 -36, . 3 45 Operating (gpm) 437 * *-."52 -

Cases with 15 mil tolerance and with 20 rail tolerance require changes to the values in Ref. 2.

CONCLUSIONS Since a measured minimum water channel spacing of 90 mils (20 mil tolerance) is anticipated on a number of the LEU fuel assemblies being re-worked by BWXT, the following values for key coolant flow rate parameters are proposed in Table 5. The UFTR Technical Specifications will need to be changed.

Table 5- Flow Rate Settings for 20 mif Water-Channel-Spacing Tolerance (measured minimum water channel spacing of 90 mils)

SLSSS .(gpm~ .

Oprting (g -2 8

The accident analyses in Ref. 2 were performed using a true minimum coolant flow rate of 34 gpm, which is consistent with the proposed Technical Specifications for maximum water channel spacing tolerances of 10 mils and 15 mils. For a maximum tolerance of 20 mils, the accident analyses in Ref. 2 provide results that are more conservative than with the true minimum flow rate of 39 gpm proposed in the Technical Specifications. Therefore, none of the accident analyses in Ref. 2 need to be changed.

It is recognized that the thermal-hydraulic analyses with the hot-channel factor method used in the analyses in this supplement give conservative results. ANL is currently evaluating alternative methods for reducing some of the conservatism due the somewhat-unique variations in the water channel spacing anticipated in the UFTR LEU fuel assemblies. Measurements of the minimum channel spacing in each LEU fuel assembly are anticipated before August 10, 2006. Using the measured minimum channel spacing and less conservative, but more accurate, factors to represent more realistic coolant flow rate through these channels, it may be possible to reduce the coolant flow rate settings shown in Table 5.

9

Possible changes to the UFTR's Technical Specifications Based on the current analyses, the UFTR technical specifications could be modified as follows:

a) If the maximum tolerance for the water channel spacing is 10 mil, we propose the following specifications:

Maximum T (inlet) 110 F, and b) If the maximum tolerance for the water channel spacing is 15 mail we propose the following specifications:

Maximum T (inlet) = 100 F, and c) TrueM n.

If the maximum tolerance for the water channel spacing is 20 mil we propose the following specifications:

Maximum T (inlet) = 100 F, and 10

A 7TACHMENT III DISCUSSION AND ANALYSES OF FUEL STORAGE FACILITIES CRITICALITY AND SAFETY REQUIREMENTS FOR TECH SPEC AMENDMENT 26 CONSIDERATION

9. Auxiliary system 9.2 Handling and Storage of Reactor Fuel This section presents the systems for secure storage of unirradiated and irradiated LEU fuel. These systems prevent criticality (kerr not exceeding 0.80) under all conditions of moderation during storage.

9.2.1 Irradiated and Unirradiated Fuel Storage Descriptions Irradiated Fuel Storage Area Irradiated and uniriradiated reactor LEU fuel can be stored in an irradiated fuel storage area __________

4as illustrated in Figures 9-1 and 9-2.

Figure 9-1 Horizontal Projection of the Irradiated Storage Area Figure 9-2 Vertical Projection of Two Irradiated Storage Locations I

Unirradiated Fuel Storage Area Figure 9-3 XZ and YZ Projections of the Unirradiated Storage Area 9.2.3 Criticality Analyses In order to verify that the irradiated storage prevent criticality under normal and flooded conditions, it was analyzed for a single case whOrN bundle. For the unirradiated storage area, the following fuel arrangements are considered:

The kdr's for both systems were determined under two conditions of moderation using MCNP5. Table 9-1 summarizes the results of our Monte Carlo calculations.

2

Table 9-1 kIcrfor both S stems Under Normal and Flooded Conditions System kerr Normal Flooded IrradiatedStorageArea i)a locations containing fuel 0.13237 0.00011)' 0.36486 (0.00025)

UnirradiatedStorage Area i) 70 plates 0.00923 (0.00001) 0.38733 (0.00039)

Pitch (cm) x=1O.O, z-O.O0 ii) 70 plates 0.01095 (0.00001) 0.43479 (0.00041)

Pitch (cm) x=7.2263, z=0.0 iii) 70 plates/ r 0.01569(0.00001) 0.45756 (0.00041)

Pitch (cm) xl1Ocm iv) 98 plates 0.01110 (0.000001) 0.45590 (0.00044)

Pitch (cm) x-7.2263, z=0.0 v) 196 plates 0.01800 (0.00001) 0.73454 (0.00049)

Pitch (cm) x-7.2263, z=5.72516 11-a relative statistical uncertainty Based on the results presented in Table 9-1, the storage systems of the UFTR facility meet the criticality requirement of kfr < 0.8 for all conditions of moderation.

3

A TTACHMENT IV CONSISTENCY CHECKS OF CONTROL BLADE REACTIVITY WORTHS

UNIVERSITY OF FLORIDA TRAINING REACTOR SUPPLEMENTAL INFORMATION August 4, 2006 In the June 15th responses to the RAI (docket No. 50- 83), an inconsistency between the control blade worths given in Tables 4-1 and 4-13 was noticed. The current supplement addresses this issue by providing the correct set of values for both tables and the resulting modifications to the relevant pages of the submittal. These modifications were made to ensure that this inconsistency does not appear elsewhere in the submittal.

For convenience, the corrected values of Table 4-13 for the control blade integral reactivity worths are given below.

Table 4-13 Comparison of Control Blades Worth for the HEU and LEU Cores HEU TEU LEU-fresh LEU-depleted Control Blade (calculated) (measured) (calculated) (calculated)

Regulating 0.87% 0.82% 0.63% 0.66%

Safety 1 1.35% 1.21% 1.62% 1.65%

Safety 2 1.63% 1.36% 1.81%' 1.76%

Safety 3 2.03% 1 1.88% 1.42% 1.46%

The total integral reactivity worth of this blade was evaluated by positioning all the blades at their critical position and then rotating the blade of interest (see following section).

The corrected Table 4-1 is included in Appendix A of this supplement while the modified pages of the submittal are included in Appendix B.

APPENDIX A Table 4-1 Summary of Key Nominal Design Parameters of HEU (current) and LEU (expected) Cores HEU LIEVU DESIGN DATA Fuel Type U-Al alloy USi2-AI Fuel Meat Size Width (cm) 5.96 5.96 Thickness (cm) 0.102 0.051 Height (cm) 60.0 60.0 Fuel Plate Size Width (cm) 7.23 7.23 Thickness (cm) 0.178 0.127 Height (cm) 65.1 65.1 Cladding 1100 Al 6061 Al Cladding Thickness (cm) 0.038 0.038 Fuel Enrichment (nominal) 93.0% 19.75%

"Meat" Composition (wt% U) 14.05 62.98 Mass of 235U per Plate (nominal) 14.5 g 12.5 g Number of Plates per Fuel Bundle 11 14 Number of Full Fuel Bundles (current/expected) 21 22 Number of Partial Fuel Bundles 1 1 (5 fuel plates + 5 (10 fuel plates + 4 dummy plates) dummy plates)

Number of Dummy Bundles 2 I REACTOR PARAMETERS Fresh Core Excess Reactivity (% Ak/k) 1.09 0.925 Shutdown Margin (A/k) 3.11 3.17 Control blade worth, Regulating (% Ak/k) 0.87 0.63 Safety I (% Ak/k) 1.35 1.62.

Safety 2(% A0/Mk) 1.63 1.81 Safety 3 (% Ak/k) 2.03 1.42 Maximum Reactivity Insertion Rate (% A~k/s) 0.042 0.045 Ave. Coolant Void Coefficient, (% Ak/k/%void)

Fresh Core -0.148 -0.153 Depleted Core -0.146 Coolant Temp. Coefficient, (% Ak/k/0C)

Fresh Core -5.91E-03 -5.68E-03 Depleted Core -5.26E-03 Fuel Temp. Coefficient, (% Ak/k/ 0 C)

Fresh Core -2.91E-04 -1.65E-03 Depleted Core -1.49E-03 Effective Delayed Neutron Fraction Fresh Core 0.0079 0.0077 Depleted Core 0.00756

Neutron Lifetime (pis)

Fresh Core 187.4 177.5 Depleted Core 195.1 THERMAL-HYDRAULIC PARAMETERS (100kW. 43 irm, Tin=30 Q Max. Fuel Temperature' ( 0C) 66.5 64.5 Max. Clad Temperature' ( 0C) 66.5 64.4 Mixed Mean Coolant Outlet Temperature (QC) 40.8 40.5 Max. Coolant Channel Outlet Temp., (*C) 58.3 59.1 Minimum ONBR 1.98 2.09 Minimum DNBR 354 376 1At nominal operating conditions

APPENDIX B B.) Changes to page 28 ofthe December's Submittal Table 4-13 Comparison of Control Blades Worth for the HEU and LEU Cores lIEU HEU 1LEU-fresh LEU-depleted Control Blade (calculated) (measured)  !(calculated) (calculated)

Regulating 0.87% 0.82% 0.63% 0.66%

Safety 1 1.35% 1.21% 1.62% 1.65%

Safety 2 1.63% 1.36% 1.81%' 1.76%

Safety 3 2.03% 1 1.88% 1.42% 1.46%

The total integral reactivity worth of this blade was evaluated by positioning all the blades at their critical position and then rotating the blade of interest (see following section).

For the HEU core, the calculated and experimental data differ in a range of 6.1% to 19.9%. These differences can be attributed to experimental uncertainty and inconsistency between the experimental procedure to measure the blade worth and the modeling procedure.

Further, the two control blades on the south part of the reference LEU core (safety 1 and

2) have higher worths as compared to the HEU core, while the two control blades on the north part of the LEU core (safety 3 and regulating) have lower worth than in the HEU core. This finding is expected because of the observed power shift presented in Table 4-

10. This power shift is expected since more fuel is added to the south part of the core.

Maximum Reactivity Insertion Rate for HEU Core In addition to calculations of the total reactivity worth for the UFTR control blades, an analysis of the integral worth as a function of position was performed for the most reactive blade. In the prior calculations, Safety Blade 3 was determined to be the most reactive blade. An MCNP model of the UFTR fueled with 21.5 HEU fuel bundles was utilized. The calculations were performed by positioning the Safety 1, Safety 2, and Regulating Blades at a critical position for the core, and then moving Safety Blade 3 through its full range of motion (2.50 to 47.5C). Results are provided in Table 4-14 and Figure 4-16. The total blade worth calculated here is 2.03% Ak/k, which is almost the same as the prior calculation for the total blade worth (2.06 % Ak/k). In the prior calculations, the other blades were fully-withdrawn, while in the calculations presented in Table 4-14, the safety blades were inserted at 38.50 and the regulating blade was at 18.70.

Table 4-14 Intearal Reactivity Worth versus Position for Safety 3 in the HEU Core Time Blade Position kerr Reactivity Reactivity Insertion (s)" Degrees Units _ _ (%Ak/k) Rate (%Ak/k/s) 0.0 2.5 0 0.98747 k 0.04% !0.00% n/a 5.6 5 56 0.98936 - 0.03% 0.19% 0.034%

16.7 10 167 0.99330 - 0.04% 0 0.59% 0.036%

27.8 15 278 0.99789 - 0.04% 1.06% 0.042%

38.9 20 389 1.00158  : 0.02% 1.43% 0.034%

50.0 25 500 1.00423  : 0.02% 1.70% 0.024%

61.1 30 611 1.00576 . 0.02% 1.85% 0.014%

72.2 35 722 1.00664 . 0.02% 1.94% 0.008%

83.3 40 833 1.00704 . 0.02% 1.98% 0.004%

100.0 47.5 1000 1.00747 d 0.02% 2.03% 0.003%

'Assumes 100 seconds withdrawal time

B.2 Changes to page 29 ofthe December's Submittal Figure 4-16 shows calculated the Safety Blade 3 (most reactive blade for HEU core) worth as a function of position.

2.2%

2.0%

1.8%

1.6%

- 1.4%

• 1.2%

-> 1.0%

Q. 0.8%

0.6%

0.4%

0.2%

0.0%

0 200 400 600 800 1000 Blade Position (Units)

Figure 4-16 Integral Blade Worth versus Position for Safety 3 in the UFTR HEU Core.

The UFTR Technical Specifications require that the reactivity insertion rate from control blade withdrawal must be less than 0.06% Ak/kc/s when averaged over a 10 second interval. The rate of reactivity insertion resulting from withdrawal of the highest worth blade was approximated by assuming a 100 second (minimum allowed) blade withdrawal time. As shown in Table 4-14, the highest rate of reactivity insertion from withdrawal of Safety Blade 3 is 0.042% Ak/k/s, which meets the requisite UFTR Technical Specification.

Maximum Reactivity Insertion Rate for LEU Core The integral reactivity worth as a function of position was determined for Safety Blade 2 (most reactive blade for the LEU core) based on MCNP calculations in a manner similar to that employed for the HEU core calculations. The position of Safety Blade 1 and Safety Blade 3 was fixed at 26.3' and the Regulating Blade was positioned at 16.90, while Safety Blade 2 was rotated from 2.50 to 47.5'. Results are provided in Table 4-15 and

Figure 4-17. The total worth for Safety Blade 2 calculated in this manner is similar to that obtained in the prior calculations with the other blades fully-withdrawn (1.77%Al/k).