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Univ. of Maryland - Revision 0, Document No. 911206, Support Calculations for Mutr'S Maximum Inlet Temperature
	ML16008A073
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Site:	University of Maryland
Issue date:	08/01/2011
From:	Crozier J; General Atomics
To:	; Office of Nuclear Reactor Regulation, US Dept of Energy (DOE)
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Apvd ISSUED 2011108101 Revision 0 UNIVERSITY OF MARYLAND REACTOR ANALYSIS AND SUPPORT Support Calculations for MUTR's Maximum Inlet Ternperatu re Prepared by: TRIGA Reactor Division of General Atom ics Address: P0 Box 85608 San Diego, CA 92186-5608 Prepared under Contract No. 00096970 for the U.S. Department of Energy GA PROJECT 39364

+GENERAL ATOMIC5

+GENERAL GA 1485 (REV. 08106E)

ATOMiCS ISSUE/RELEASE

SUMMARY

O] R &D APPVL DISC QA LEVEL sYs DCC, TYPE PROJECT DOCUMIENT NO. REV O] DV&S LEVEL 0] DESIGN O] T&E 2 N lB DCR 39384 911208 (C)

[] NA TITLE:

Support Calculations for MUTR's Maximum Inlet Temperature

,,,,iAPPROVAL(,S) _____

REVISION CM APPROVAL! PREPARED DESCRIPTIONI DATE REV BY ENGINEERING QA PROJECT W.O. NO.

E D0 J. Crozier J. Balin K. Partain V Initial Issue

~\UG 1 2~'A39364 -0410 CONTINUE ON GA FORM 1485-1 NEXT INDENTURED DOCUMENT(S)

COMPUTER PROGRAM PIN(S)

El GA PROPRIETARY INFORMATION THIS DOCUMENT IS THE PROPERTY OF GENERAL ATOMICS. ANY TRANSMITTAL OF THIS DOCUMENT OUTSIDE GA WILL BE IN CONFIDENCE. EXCEPT WITH THE WRITTrEN CONSENT OF GA, (1) THIS DOCUMENT MAY NOT BE COPIED IN WHOLE DR IN PART AND WILL BE RETURNED UPON REQUEST OR WHEN NO LONGER NEEDED BY RECIPIENT AND (2) INFORMATION CONTAINED HEREIN MAY NOT BE COMMUNICATED TO OTHERS AND MAY BE USED BY RECIPIENT ONLY FOR THE PURPOSE FOR WHICH IT WAS TRANSMITTED.

0] NO GA PROPRIETARY INFORMATION PAE IOF 14

Support Calculations for MUTR's Maximum inlet Temperature 911206/0 TABLE OF CONTENTS AC RONYMS/ABBREVIATIONS ...................................................................... iii I PURPOSE ........................................................................................

2 METHODS - RELAP5 IMOD3.3-Patch03 CODE ............................................. I 3 RESULTS -RELAP5 ........................................................................ I..

4 CONCLUSIONS ................................................................................. 3 5 REFERENCES .................................................................................. 4 6 INDEPENDENT REVIEW ...................................................................... 4 APPENDIX A - INDEPENDENT REVIEW REPORT............................................... A-I APPENDIX B - CALCULATION FILE LISTINGS.................................................. B-I LIST OF FIGURES Figure 3-1 RELAP Natural Convection (No Primary Coolant Flow) Block Diagram ............... 2 Figure 3-2: MUTR DNBR as a Function of Pool Inlet Temperature.................................. 3 iii

Support Calculations for MUTR's Maximum Inlet Temperature 911206/0 910/

ACRONYMSIABB REVIATIONS AcronymlAbbreviation Description DNBR Departure from Nucleate Boiling Ratio GA General Atom ics MUTR University of Maryland TRIGA Reactor PRNC Puerto Rico TRIGA reactor RA! Request for Additional Information SAR Safety Analysis Report TAMU Texas A&M University TRIGA Training Research Isotope General Atomics iii

Support Calculations for MUTR's Maximum Inlet Temperature 911206/0 9101 I PURPOSE The purpose of this calculation file is to document the thermal hydraulic analyses that were performed by General Atomics (GA) in support of the activities necessary to answer the University of Maryland TRIGA Reactor (MUTR) Safety Analysis Report (SAR) submittal Request for Additional Information (RAI) question regarding the maximum inlet temperature. The nuclear analyses used in these analyses are reported in "Support Calculations for MUTR's Request for Additional Information" (Ref. 1).

Additional explanation and calculations were requested for the question: 'For a DNBR of 2.0, what is the inlet (bulk fluid) temperature under natural circulation conditions? This temperature would be used as the Lim iting Condition for Operation."

2 METHODS - RELAP5 /MOD3.3-PATCH03 CODE RELAP has been an industry standard code for the analysis of power reactors. Its development and improvements have occurred over at least 25 years. The code performs steady state and transient reactor neutronics, thermal hydraulics and fuel rod thermal analysis. It allows for very general modeling - multiple rod and channel configurations and heat structures for rod thermal performance. The code has sophisticated single and two phase flow modeling both in its continuity, momentum and energy formulations, as well as for wall and interfacial friction and heat transfer correlations. The version RELAP5/MOD3.3-Patch03 is the latest available validated version of the code (Ref. 2) (Ref.3).

3 RESULTS -RELAP5 The latest version of the RELAP Code (RELAP5/MOD3.3-Patch03) was used to evaluate the MUTR thermal/hydraulic performance for this RAI.

The RELAP model for this request contains an average powered rod representing the entire core and a hot rod representing the maximum powered rod. The MUTR TRIGA reactor uses four rod fuel element clusters as used in the Neutron Radiography Facility at Idaho National Laboratory (Ref. 4), Puerto Rico Nuclear Center (PRNC) (Ref. 5), and Texas A&M University (TAMU) TRIGA reactors (Ref. 6); As such, the RELAP modeling of the average and hot rods used in the RELAP model for these previous HEU-to-LEU conversions is used for the MUTR RELAP model, see Figure 3-1.

1

Support Calculations for MUTR's Maximum Inlet Temperature 911206/0 910/

Figure 3-1 RELAP Natural Convection (No Primary Coolant Flow) Block Diagram Some of the important conditions in the MUTR RELAP model include:

The average powered rod represents 93 rods fuel rods.

Hot rod power factor taken as 1.597.

Departure from Nucleate Boiling Ratio (DNBR) calculated using Bernath correlation, (Ref. 7)

Natural circulation core cooling.

Tank diameter - 7 ft., water depth from surface to reactor top - 17.1 ft.

14.7 psia pressure above tank water.

2

Support Calculations for MUTR's Maximum Inlet Temperature 910/

911206/0 Using these conditions, a RELAP run was performed at ,a reactor power of 600 kW for various pool temperatures. The RELAP results were used to construct a plot of pool temperature versus DNBR, Figure 3-2. The plot shows that the inlet pool temperature can reach 92 0C and only yield a DNBR of 2.96, at 600 kW. This is approximately equal to a DNBR of 5.92 for a reactor power level of 300 kW. At a pool inlet temperature of 95 00, the RELAP5 code fails to generate a stable solution due to the large vapor volume. The results demonstrate significant vapor formation. Additionally, the hot channel flow rate, quality, and bulk temperature vary irregularly along the flow channel. This RELAP5 behavior may be indicative of parametric fluctuations in the actual core at this power level and pool inlet temperature. Based on the RELAP model, the maximum theoretical achievable inlet temperature of the MUTR reactor would be 92 0C, at which point DNB would be predicted to occur. Associated RELAP run files and MathCad calculations are listed in Appendix B and included in a local subdirectory.

DNBR vs Pool Temp - 600 kW 6.0 5.5 5.0 4.5 4.0 zS3.5 3.0 2.5 2.0 1.5 1.0 65 90 95 Pool Temp, 'C Figure 3-2: MUTR? DNBR as a Function of Pool Inlet Temperature 4 CONCLUSIONS The analyses herein demonstrates that the MUTR TRIGA reactor reaches a DNBR of 2.96 when operating at a maximum pool inlet temperature of 92 00 and a power level of 600 kW.

Hence, for normal operations at a power level of 300 kW and a maximum pool inlet temperature 3

Support Calculations for MUTR's Maximum Inlet Temperature 910/

911206/0 of 92 °C, the DNBR is greater than 2.0. Based on the data from Figure 3-2, with a maximum pooi inlet temperature of 92 °C, a DNBR of 2.0 could not occur.

5 REFERENCES

1. Ellis, C. "Support Calculations for MUTR's Request for Additional Information," GA 911199, December 2010.

2. "RELAP5/MOD3.3 Code Manual Volume 3: Developmental Assessment Problems,"

NUREG-CR-5535/Rev. P3 - Volume Ill, 2006.

3. Tuey, J. "RELAP5/MOD 3.3 Code Verification Report," General Atomics, 09490R00001, Rev. A, July 2011

4. Crozier, J. and Ellis, C., "LEU Upgrade of the NRAD Reactor Final Report," GA Document No. 911193 Revi sion 0, February 2011

5. Safeguards Summary Report for the TRIGA Reactor at the Puerto Rico Nuclear Center, Mayaquez, Puerto Rico, PRNC 123, April 1969.

6. "Safety and Accident Analyses Report Texas A&M University (TAMU) Conversion from HEU to LEU Fuel," Texas A&M University, Docket 50-1 28, December 2005.

7. Bernath, L.B., "A Theory of Local Boiling Burnout and Its Application to Existing Data",

Chem. Eng. Progress Symposium Series No. 30, Vol. 56, pp 95-116., 1960 6 INDEPENDENT REVIEW John Bolin -

Review results (Thermal Hydraulics):

No comments on the thermal hydraulic analyses.

4

Support Calculations for MUTR's Maximum Inlet Temperature 911206/0 APPENDIX A - INDEPENDENT REVIE*W REPORT A-1

Support Calculations for MUTR's Maximum Inlet Temperature 911206/0 CALCULATION REVIEW REPORT TITLE DOCUMENT NUMBER REV Suppord C,oCulaI~ons for MUTR's MaximnummIndTempornluro 11112060 0 INDEPENDENT REVIE*WE R:

NAME: John Bolin ORGANIZATION: 480 MODULAR HELUUM REACTOR REVIER,SELCTION APPROVAL.

MANAGER: Tony Veco NAME I TURE' REVIEW CRITERIA YES NO COMMENT REVIEWCALCULATION METHOD 0] 0 COMPARE* CALCULATION METHODWITH SIMILAR APROVED M.E~hOD0 03 PREPARE ALTERNATE CALCULATION USING DIFFERENT METHOD 0.- 0]

REVIEW#ASSUMPTIONS:

UNIQUELY IDENTIFIED 0] 0

- ADEQUATELY DESCRJBED 0] 0

- REASONABLE 0] 0

- UNCONFIRMED SPECIFICALLY IDENTIFIED 0] 0 REVIEW INPUTS 00 [

REVIEW RESULT.SJCONCLUSIONS [] 0 VERIFY COMPUTER CODE INFORMATION CORRECT 0 0 VERIFY OUAUITYASSURANCE LEVE*L 0 0]

RECORD OR ATTIACH0] LIST OF DOCUMENTS USED IN REVIEW (ANOIOR) 0] ALTERNATE CALCULATIONS OTHER REMARKS:

SCALCULATION FOUND TO BE VALIDAND CONCLusIONS TO BE CORRECT (ORI 0] EXCEPT AS NOTED ABOVE.

INDEPENDENT REVIEWER *L* SGiGNATURE p->DATE "** l FINAL REVIEW BEFORE IESUEIRELEASE:

INDEPENDENT REVIEWER

NTUR 1 ---. AE 1""'*-***"/"

'/i I A-2

Support Calculations for MUTR's Maximum Inlet Temperature 911206/0 APPENDIX B - CALCULATION FILE LISTINGS B-I

Support Calculations for MUTR's Maximum Inlet Temperature 910/

911206/0 Calculation files in support of Section 3.

File Description File Type DVD Directory Location MUTR RELAP DNBR input file RELAP5 input MUTR_dnbr_600kW_40C.inp 600kW, 40C MUTR RELAP DNBR output file RELAP5 output MUTR_dnbr_600kW_.40C.out 600kW, 40C MUT RLAPDNR npu fle RELAP5 input MUTR_dnbr 600kW_80C.inp 600kW, 80C _________________

MUTR RELAP DNBR output file RELAP5 output MUTR_dnbr_600kW_80C.out 600kW, 80C__________________________

MUT RLAPDNR npu fle RELAP5 input MUTR_dnbr_600kW_90C.inp 600kW, 90C MUTR RELAP DNBR output file RELAP5 output MUTR_dnbr_600kW_90C.out 600kW, 90C MUR ELP NB ipu fle RELAP5 input M UTR_dnbr_600kW_92C.inp 600kW, 92C_

MUTR RELAP DNBR output file RELAP5 output MUTR_d nbr_600kW_92C.out 600kW, 92C ________

MUTR MathCAD file MathCAD Bernath dnbr.mncd B-2

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