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{{#Wiki_filter:Jmj I TEXAS A&M ENGINEERINGEXPERIMENT STATIONNUCLEAR SCIENCE CENTERNovember 13, 2014 2014-0069Document Control DeskATTN: Geoffrey WertzU.S. Nuclear Regulatory CommissionWashington, DC 20555- | {{#Wiki_filter:Jmj I TEXAS A&M ENGINEERINGEXPERIMENT STATIONNUCLEAR SCIENCE CENTERNovember 13, 2014 2014-0069Document Control DeskATTN: Geoffrey WertzU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001 | ||
==Subject:== | |||
Response to U.S. NRC Request for Additional Information, Review of The Fuel Pool Temperatureon Fuel Temperature for the License Renewal for the Nuclear Science Center TRIGA Reactor(TAC No. ME 1584), from the Texas A&M University System, Texas Engineering ExperimentStation, Nuclear Science Center Reactor (NSCR, License No. R-83, Docket 50-128)To Whom It May Concern:The Texas A&M University System, Texas Engineering Experiment Station (TEES), Nuclear Science Center (NSC,License No. R-83) operates a LEU, 1MW, TRIGA reactor under timely renewal. In December, 2003 the NSCsubmitted a Safety Analysis Report (SAR) as part of the license renewal process. In December, 2005 a conversionSAR (Chapter 18) was submitted resulting in an order to convert from the USNRC. In July 2009, the NSCsubmitted an updated SAR, dated June 2009, to the USNRC. This updated 2009 version of our SAR incorporatedthe information from the conversion SAR and the startup of the new LEU reactor core. On July 15, 2014 the U.S.NRC submitted a Request for Additional Information re: Review of The Fuel Pool Temperature on FuelTemperature for the License Renewal for the Nuclear Science Center TRIGA Reactor (TAC No. ME 1584) as a partof the review process. Attached is our reply to this question.It is interesting to note that after the initial license application was made in 2009, a study was commissioned fromGeneral Atomics in 2010 to investigate the implications of a power uprate at NSC. Many of the issues regarding thisRAI were carefully analyzed as part of that study and have been used as the basis for part of our response. Thisresponse was reviewed internally by Dr. Karen Vierow, Associate Professor of the Texas A&M UniversityDepartment of Nuclear Engineering. We believe the contents of this response will satisfy the concerns raised by thereviewers.If you have any questions, please contact Dr. Sean McDeavitt or Mr. Jan Vermaak at 979-845-7551.TEL. 979.845.7551 I FAX 979.862.2667nsc.tamu.edu1095 Nuclear Science Rd. 1 3575 TAMU I College Station, TX 77843-3575 I declare under penalty of perjury that the foregoing is true and correct. Executed on November 13, 2014.Dr. Sean M. McDeavittInterim Director, Nuclear Science CenterAssociate Professor of Nuclear EngineeringDepartment of Nuclear EngineeringDwight Look College of EngineeringTexas A&M Universitymcdeavitt@tamu.eduXc: 2.1 1/Central FileDuane Hardesty, USNRC Project Manager In response to the four items in the request for additional information (RAI) by letter dated July 15,2014, the Nuclear Science Center (NSC) would like to withdraw the reference report "Evaluation of PoolTemperature on Fuel Temperature and MDNBR [Minimum Departure from Nucleate Boiling Ratio]"[1]and replace it with "Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade"[2], prepared for the Texas A&M Engineering Experiment Station by General Atomics underResearch Service Agreement C13-00345 and dated.November 26, 2013.JustificationThe methodology used for the analysis presented in[1]involved the application of a single phase flowsolver combined with a predictive correlation to determine the critical heat flux (Bernath correlation).Upon further investigation, it was determined that the implementation was not subjected to extensivereview to determine the validity of its application. This is of particular concern when regarding the twophase flow regime existing during an approach to the Critical Heat Flux (CHF) which hints at thefundamental need for the flow solver to be able to calculate two phase flow. Stemming from thisconcern, the NSC will not continue to support the analysis done with this methodology and as acorrective measure will replace this supporting analysis with the safety analysis performed by GeneralAtomics (GA) in November 2013 [2].The GA safety analysisis contained in a report [2] which was originally requested in order to support theoperation of the NSC reactor at 1.25 MW, but with a nominal operating power of 1.0 MW. Additionally,the thermal-hydraulic analyses were performed for a pool inlet temperature of 60°C which was requiredto support a proposed limiting safety system setting (LSSS) on pool inlet temperature.The GA safety analysis indicated an acceptable minimum DNBR of greater than 2.0 at a reactor power of1.25 MW and a pool inlet temperature of 60°C.Summary of the relevant information in the GA safety analysis:In section 4.7 of[2], the analysis of the steady operation of the NSC reactor is presented. In this safetyanalysis General Atomics has diverted from using their proprietary STAT code (as used in [3]) andreplaced it with RELAP5/Mod 3.3 [4]for which a model description is provided[2]. The steady stateanalysis was performed for a pool inlet temperature of 60°C which differs that of [3] which was for apoolinlet temperature of 30°C. The 60°C inlet temperature, as well as an additional analysis at 1.25 MW,are parameters that were requested by the NSC as part of the service agreement to GA to allow for amore flexible basis for the reactor safety case (i.e. SCRAM limits). The following changes in operatingparameters (between [2] and [3]) were observed:* Maximum clad outer temperature at 1.0 MW increased to 137.2°C from 135.9°C." Maximum fuel temperature at 1.0 MW increased to 526°C from 368.1°C.* Hottest channel outlet temperature at 1.0 MW increased to 101.0°C from 76.8°C.* Minimum DNBR increased to 2.53 at 60°C inlet temperature from 2.42 at 30°C inlettemperature.The minimal increase in cladding outer temperature is expected due to the nature of the heat transferproperties at increased coolant temperature (i.e. dynamic viscosity decreases, thermal conductivityincreases) as well as the higher thermal expansion coefficient of water at higher temperature (i.e. morebuoyancy driven mass flow).The change in outlet temperature is directly related to the difference ininlet temperature (60°C vs 30'C) as well as minor differences in the total heat generation. | |||
The only major change is that of the maximum fuel centerline temperature which increased by 158°C.This large step is counter-intuitive especially since the two models essentially should have the samesolid conduction models, however, it was found that the step change resulted from the assumptionsmade for the gap between the cladding and the fuel pellets. In [3] it was assumed constant at 3micrometers but in [2] it was specified as 12 micrometers at a reactor power of 1.0 MW which is 4 timeslarger. This increase in gap size dramatically increases the thermal resistance of the gap and translatesto the associated increase in centerline temperature.The change in minimum DNBR is largely due to the non-conventional method used in [3] to establish theDNBR, where the reactor power was increased to determine the point of unity in terms of DNBR.However, such a method unnecessarily penalizes the operational conditions and after internalinvestigations it was determined that, if [3] applied the conventional method for determining DNBR, thevalue would approximately be 3.32 at 30°C inlet temperature. Also, the DNBR value at 60°C inlettemperature was also validated.References[1] General Atomics, "Evaluation of Pool Temperature on Fuel Temperature and MDNBR [MinimumDeparture from Nucleate Boiling Ratio]", GA Project 30398, 30398R00001 Rev. 0, Nov. 2013.[2] "Safety Analysis Report for the Texas A&M University Nuclear Science Center Power Upgrade",General Atomics, Project 30398, November 2013.[3] "Safety Analysis Report -Texas A&M University System -Nuclear Science Center Reactor",Docket number 50-128, License number R-83, May 2011.[4] "RELAP5/Mod 3.3 Code Manual, Vols. 1-8", US Nuclear Regulatory Commission, NUREG/CR-5535, Rev. P3, March 2006. | The only major change is that of the maximum fuel centerline temperature which increased by 158°C.This large step is counter-intuitive especially since the two models essentially should have the samesolid conduction models, however, it was found that the step change resulted from the assumptionsmade for the gap between the cladding and the fuel pellets. In [3] it was assumed constant at 3micrometers but in [2] it was specified as 12 micrometers at a reactor power of 1.0 MW which is 4 timeslarger. This increase in gap size dramatically increases the thermal resistance of the gap and translatesto the associated increase in centerline temperature.The change in minimum DNBR is largely due to the non-conventional method used in [3] to establish theDNBR, where the reactor power was increased to determine the point of unity in terms of DNBR.However, such a method unnecessarily penalizes the operational conditions and after internalinvestigations it was determined that, if [3] applied the conventional method for determining DNBR, thevalue would approximately be 3.32 at 30°C inlet temperature. Also, the DNBR value at 60°C inlettemperature was also validated.References[1] General Atomics, "Evaluation of Pool Temperature on Fuel Temperature and MDNBR [MinimumDeparture from Nucleate Boiling Ratio]", GA Project 30398, 30398R00001 Rev. 0, Nov. 2013.[2] "Safety Analysis Report for the Texas A&M University Nuclear Science Center Power Upgrade",General Atomics, Project 30398, November 2013.[3] "Safety Analysis Report -Texas A&M University System -Nuclear Science Center Reactor",Docket number 50-128, License number R-83, May 2011.[4] "RELAP5/Mod 3.3 Code Manual, Vols. 1-8", US Nuclear Regulatory Commission, NUREG/CR-5535, Rev. P3, March 2006. | ||
Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgrade4.7 Thermal Hydraulic Analysis -NSC LEU4.7.1 Analysis of Steady State Operation -NSCThe following evaluation has been made for the TRIGA LEU 30/20 fuel system with four-rodconfiguration operating with cooling from natural colvection water flow through four-rodclusters of fuel. The steady state thermal-hydraulic performance of the NSC LEU TRIGA corewas determined for operation at 1.0 and 1.25 MW with a water inlet temperature of 60 'C.The RELAP5 mod 3.3 computer code4 was used to determine the natural convection flow rate,the coolant and clad axial temperature profiles, and the clad wall heat flux axial profile. TheRELAP5 code was also used to determine the clad wall maxi -mum heat flux for departure fromnucleate boiling. RELAP5 was also used to determine the fuel average and fuel maximumtemperatures.An average powered fuel rod and the maximum powered fuel rod were analyzed. Four differentflow geometries were used to analyze the maximum powered fuel rod so that the most limitingthermal hydraulic condition could be determined. The arrangement of the four flow geometriesaround the maximum power fuel element (5D3) is depicted in Figure 4-2 1. Flow channel 221 isformed by the maximum rod 5D3, the transient rod 5D4, and fuel rods 5D I and 5D2. It has thesmallest hydraulic diameter and smallest flow area. Flow channel 222 is formed by themaximum rod 5D3, the transient rod 5D4, and fuel rods 4D1 and 4D2. Flow channel 223 isformed by the maximum rod 5D3 and fuel rods 5E4, 4D2, and 4El. It has the largest hydraulicdiameter. Flow channel 224 is formed by the maximum rod 5D3 and fuel rods 5D2, SEl, and5E4. The RELAP5 model shown in Figure 4-22 also includes flow channels for theinstrumented fuel element (lFE). The fuel element with the maximum intra-rod power density isalso shown in Figure 4-22 but is only used in the model for reactivity insertion transients.41 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398R00001/0U,tJUAllt~i'Uii;Figure 4-21: Flow Chnnnels That Include Maximum Fuel Element 5D342 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398REJ0001/0Figure 4-22: RELAP5 Model of Texas A&M NSC TRIGA4.7.2 RELAPS Model Input -NSCThe'RELAP5 analysis is a more modern and validated method than the method outlined inSection 4.6. The general reactor geometry and hydraulic data for the RELAP5 model are given inTable 4-16. The natural convection system for the NSC TRIGA is based on the four-rod clusterof fuel elements. The representation used herein establishes multiple flow channels eachtypically bounded by four fuel elements.A RELAP5 steady state thermal hydraulic analysis was done for an average flow channel, an IFEchannel and four flow channels containing the maximum powered fuel rod. The average flowchannel has a flow area associated with the 90 fuel rods in the NSC LEU core. The IFE has thesame hydraulic characteristics as the average flow channel except its flow area is for only onefuel rod. The hydraulic characteristics of the flow channels and power factorsof the associatedfuel rods are presented in Table 4-17.The gap between the.fuel and cladding is an important parameter for predicting fueltemperatures. Differences in thermal expansion between the fuel and cladding at operatingtemperatures reduce the initial manufactured gap and can result in gap closure at locations of43 Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgradehigh heat generation. The average radial gap in the NSC LEU fuel is 2.15 rail (54.6 pul). Theradial gap for the IFE in location 5E4 is 2.75 mil (69.9 pm). The radial gap for the maximumpower fuel rod in location 5D3 is 2.66 mil (67.6 r.m). Table 4-18 presents the effective radialgap at various reactor operating power levels. For the maximum and IFE fuel rods, the gap sizereduction is based on the peak axial power profile. For the average fuel element, the gap sizereduction is based on an axial power factor of 1.0 so that the gap represents the average gapwithin the average fuel element.Table 4-16: RELAP5 Input for Reactor and Core Geometry and Hydraulic Data, NSC: Parameter ValueCore and Reactor GeometryUnrteated grid plate length, in 2.00Unheated lower refleclor length, in 3.95Unheated top reflector length, in 4.06Length from top of unheated length to top of pool surface, ft 23.685Fuel element hetited length (in) 15Fuel element diameter (in) 1.4126Hydraulic DataSurface RoughnessGrid plate K lossAmbient pressure at elevation 320 ft, psiaTable 4-17: Hydraulic Flow Parameters and Power Factors, NSC..ara.eter .. :. ... !.' ig : I Channel Channel .Channel:. Channel..22.1 222 223. 224.Flow area (in2) 76.7196 0.852440 0.732495 0.929865 0.929581 0.735436WVetted perimeter (in/element) 399.403 4.437814 4.494677 4.494677 4.437814 4.437814Hydraulic diameter (in) 0.768342 0.768342 0.651877 0.827525 0.837873 0.662881Fuel rod entrance K loss 0.236 0.236 0.174 0.281 0.281 0.176Fuel rod exitX loss 1.289 1.289 0.952 1.534 1.533 0.960Ma.ximum rod power factor 1.00 1.484 1.648 1.648 1.648 1.648Power factor of other rods -- -- 1.584 1.432 1.400 1.493Peak axial power factor 1.318 1.318 1.294 1.294 1.294 1.29444 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398RO0001/0Table 4-18: Radial Gap Between Fuel and Cladding*Radial Cap in 1111 (11mr)Fuel ElementCold 0.5 o 1MW 1.0 MW .1.25 nVAverage 2.15 (54.6) 1.55 (39.4) 0.95 (24.1) 0.77 (19.6)Instrumented Fuel Element 2.75 (69.9) 1.16 (29.5) 0.55 (14.0) 0.39 (9.91)Maximum Power Fuel Element 2.66 (67.6) 1.04 (26.4) 0.47 (I1.9) 0.33 (8.38)4.7.3 Steady State Analysis Results -NSCTables 4-19 and 4-20 list the pertinent heat transfer and hydraulic steady state results for the 1.0and 1.25 MW NSC TRIGA reactor, respectively. Results are presented for an average chatneland the four maximum powered flow channels (hot channels) at initial core conditions. Theaverage core temperature at 1.0 MW is 354.9 'C. It increases to 373.4 'C at 1.25 MW. The IFElocated at 5E4 has a maximum temperature of 497.3 °C at 1.0 MW. It increases to 527.1 °C at1.25 MW.Table 4-19: Steady State Results for NSC, 1.00 MWChannel Chann el Channel .'Channel*ý.lInitial core (90 Elemients) Aerage 21222. :223 .224Mass flow rate, kg/sec 0.0860 0.0875 0.0912 0.0984 0.0989Exit coolant temperature, 'C 90.8 96.4 92.8 99.3 101.0Coolant saturation temperature, °C 1 15.2 115.2 115.2 115.2 115.2Maximum wall heat flux, WV/cm2 34.09 55.16 55.16 55.16 55.16Maximum flow velocity, cm/sec 16.20 19.26 15.76 17.18 21.78Max. clad outer temperature, OC 131.9 136.7 137.2 137.2 136.6Exit clad outer temperature, C 124.6 127.1 127.5 127.7 127.3Peak fuel temperature, *C 478.7 525.3 525.7 525.6 525.2Minimum DNBR -Bemath -- 2.58 3.05 2.92 2.53Minimum DNBR- Groeneveld '86 -- 4.79 4.88 4.75 4.75Minimum DNBR -Groeneveld '06 -- 4.24 4.04 3.94 4.1845 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398R00001/0Table 4-20: Steady State Results for NSC, 1.25 MWChannel Channel Channel :ChannelInitial Cot e (90 Elements) 'Average .1 222 223 224Mass flow rate, kg/sec 0.0931 0.0998 0.1043 0.1170 0.1179Exit coolant temperature, 0C 95.5 99.9 95.7 101.3 103.0Coolant saturation temperature, °C 115.2 115.2 115.2 115.2 115.2Maximum wall heat flux, \V/cnr2 42.62 68.95 68.95 68.95 68.95Maximum flow velocity, cmn/sec 17.61 22.19 18.22 20.72 26.38Max. clad outer temperature, 0C 134.1 139.3 139.8 139.8 139.2Exit clad outer temperature, 0C 126.4 129.0 129.2 129.5 129.1Peak fuel temperature, 0C 516.5 561.6 562.0 562.0 561.5Minimum DNBR -Bernath -- 2.02 2.39 2.31 2.01Minimum DNBR -Groeneveld '86 3.83 3.91 3.85 3.89Minimum DNBR-Groeneveld '06 -- 3.57 3.22 3.15 3.36Tables 4-19 and 4-20 include results for the minimum departure from nucleate boiling (DNB)ratio using three correlations. Historically, TRIGA reactors have used the Bernath correlation5to predict DNB. RELAP5 Mod 3.3 uses the 1986 Critical Heat Flux Lookup Table byGroeneveld6.This Lookup Table was updated in 20067 to include more experimental data and toaddress concerns with DNB at low pressure. The Bernath correlation is considered to be quiteconservative and the Groeneveld 2006 correlation is considered to be more accurate. Allcorrelations have a minimum DNBR greater than the requirement of 2.0 given in NUREG-1537.4.8 Thermal Neutron Flux Valtes -NSC LEU Core4.8.1 Thermal Neutron Flux Values in LEU Core, Calculated[No change]4.8.2 Thermal Neutron Fltx Valtes in LEU Core, Measured[No change]46 Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgrade1 Lawrence, R.D., "The DJF3D Nodal Neutronics Option for Two-and-Three-Dimensional Diffusion TheoryCalculations in Hexagonal Geometry," Doc. No. ANL-83-1, Argonne National Laboratory, March 1983.2 Derstine, K.L., "DIF3D: A Code to Solve One-, Two-, and Three-Dimensional Finite Difference Diffusion TheoryProblems," Doc. No. ANL-82-643 Los Alamos X-5 Monte Carlo Team, "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5,"LA-UR-03-1987, April 24, 2003.4, "RELAPS/Mod 3.3 Code Manual, Vols. 1- 8," US Nuclear Regulatory Commission, NUREG/CR-5535, Rev. P3,March 2006.s Bernath, L. B., "A Theory of Local Boiling Burnout and Its Application to Existing Data," Chem. Eng. ProgressSymposium Series No. 30, Vol. 56, pp 95-116.6 Groeneveld, D. C., et al., "1986 AECL-UO Critical Heat Flux Lookup Table," Heat Transfer Engineering, Vol. 7, Nos.1-2, pp 48-62, 1986.7 Groeneveld, D. C., et al., "The 2006 CHF Look-up Table," Nuclear Engineering and Design 237, pp 1909-1922,2007.47 | Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgrade4.7 Thermal Hydraulic Analysis -NSC LEU4.7.1 Analysis of Steady State Operation -NSCThe following evaluation has been made for the TRIGA LEU 30/20 fuel system with four-rodconfiguration operating with cooling from natural colvection water flow through four-rodclusters of fuel. The steady state thermal-hydraulic performance of the NSC LEU TRIGA corewas determined for operation at 1.0 and 1.25 MW with a water inlet temperature of 60 'C.The RELAP5 mod 3.3 computer code4 was used to determine the natural convection flow rate,the coolant and clad axial temperature profiles, and the clad wall heat flux axial profile. TheRELAP5 code was also used to determine the clad wall maxi -mum heat flux for departure fromnucleate boiling. RELAP5 was also used to determine the fuel average and fuel maximumtemperatures.An average powered fuel rod and the maximum powered fuel rod were analyzed. Four differentflow geometries were used to analyze the maximum powered fuel rod so that the most limitingthermal hydraulic condition could be determined. The arrangement of the four flow geometriesaround the maximum power fuel element (5D3) is depicted in Figure 4-2 1. Flow channel 221 isformed by the maximum rod 5D3, the transient rod 5D4, and fuel rods 5D I and 5D2. It has thesmallest hydraulic diameter and smallest flow area. Flow channel 222 is formed by themaximum rod 5D3, the transient rod 5D4, and fuel rods 4D1 and 4D2. Flow channel 223 isformed by the maximum rod 5D3 and fuel rods 5E4, 4D2, and 4El. It has the largest hydraulicdiameter. Flow channel 224 is formed by the maximum rod 5D3 and fuel rods 5D2, SEl, and5E4. The RELAP5 model shown in Figure 4-22 also includes flow channels for theinstrumented fuel element (lFE). The fuel element with the maximum intra-rod power density isalso shown in Figure 4-22 but is only used in the model for reactivity insertion transients.41 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398R00001/0U,tJUAllt~i'Uii;Figure 4-21: Flow Chnnnels That Include Maximum Fuel Element 5D342 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398REJ0001/0Figure 4-22: RELAP5 Model of Texas A&M NSC TRIGA4.7.2 RELAPS Model Input -NSCThe'RELAP5 analysis is a more modern and validated method than the method outlined inSection 4.6. The general reactor geometry and hydraulic data for the RELAP5 model are given inTable 4-16. The natural convection system for the NSC TRIGA is based on the four-rod clusterof fuel elements. The representation used herein establishes multiple flow channels eachtypically bounded by four fuel elements.A RELAP5 steady state thermal hydraulic analysis was done for an average flow channel, an IFEchannel and four flow channels containing the maximum powered fuel rod. The average flowchannel has a flow area associated with the 90 fuel rods in the NSC LEU core. The IFE has thesame hydraulic characteristics as the average flow channel except its flow area is for only onefuel rod. The hydraulic characteristics of the flow channels and power factorsof the associatedfuel rods are presented in Table 4-17.The gap between the.fuel and cladding is an important parameter for predicting fueltemperatures. Differences in thermal expansion between the fuel and cladding at operatingtemperatures reduce the initial manufactured gap and can result in gap closure at locations of43 Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgradehigh heat generation. The average radial gap in the NSC LEU fuel is 2.15 rail (54.6 pul). Theradial gap for the IFE in location 5E4 is 2.75 mil (69.9 pm). The radial gap for the maximumpower fuel rod in location 5D3 is 2.66 mil (67.6 r.m). Table 4-18 presents the effective radialgap at various reactor operating power levels. For the maximum and IFE fuel rods, the gap sizereduction is based on the peak axial power profile. For the average fuel element, the gap sizereduction is based on an axial power factor of 1.0 so that the gap represents the average gapwithin the average fuel element.Table 4-16: RELAP5 Input for Reactor and Core Geometry and Hydraulic Data, NSC: Parameter ValueCore and Reactor GeometryUnrteated grid plate length, in 2.00Unheated lower refleclor length, in 3.95Unheated top reflector length, in 4.06Length from top of unheated length to top of pool surface, ft 23.685Fuel element hetited length (in) 15Fuel element diameter (in) 1.4126Hydraulic DataSurface RoughnessGrid plate K lossAmbient pressure at elevation 320 ft, psiaTable 4-17: Hydraulic Flow Parameters and Power Factors, NSC..ara.eter .. :. ... !.' ig : I Channel Channel .Channel:. Channel..22.1 222 223. 224.Flow area (in2) 76.7196 0.852440 0.732495 0.929865 0.929581 0.735436WVetted perimeter (in/element) 399.403 4.437814 4.494677 4.494677 4.437814 4.437814Hydraulic diameter (in) 0.768342 0.768342 0.651877 0.827525 0.837873 0.662881Fuel rod entrance K loss 0.236 0.236 0.174 0.281 0.281 0.176Fuel rod exitX loss 1.289 1.289 0.952 1.534 1.533 0.960Ma.ximum rod power factor 1.00 1.484 1.648 1.648 1.648 1.648Power factor of other rods -- -- 1.584 1.432 1.400 1.493Peak axial power factor 1.318 1.318 1.294 1.294 1.294 1.29444 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398RO0001/0Table 4-18: Radial Gap Between Fuel and Cladding*Radial Cap in 1111 (11mr)Fuel ElementCold 0.5 o 1MW 1.0 MW .1.25 nVAverage 2.15 (54.6) 1.55 (39.4) 0.95 (24.1) 0.77 (19.6)Instrumented Fuel Element 2.75 (69.9) 1.16 (29.5) 0.55 (14.0) 0.39 (9.91)Maximum Power Fuel Element 2.66 (67.6) 1.04 (26.4) 0.47 (I1.9) 0.33 (8.38)4.7.3 Steady State Analysis Results -NSCTables 4-19 and 4-20 list the pertinent heat transfer and hydraulic steady state results for the 1.0and 1.25 MW NSC TRIGA reactor, respectively. Results are presented for an average chatneland the four maximum powered flow channels (hot channels) at initial core conditions. Theaverage core temperature at 1.0 MW is 354.9 'C. It increases to 373.4 'C at 1.25 MW. The IFElocated at 5E4 has a maximum temperature of 497.3 °C at 1.0 MW. It increases to 527.1 °C at1.25 MW.Table 4-19: Steady State Results for NSC, 1.00 MWChannel Chann el Channel .'Channel*ý.lInitial core (90 Elemients) Aerage 21222. :223 .224Mass flow rate, kg/sec 0.0860 0.0875 0.0912 0.0984 0.0989Exit coolant temperature, 'C 90.8 96.4 92.8 99.3 101.0Coolant saturation temperature, °C 1 15.2 115.2 115.2 115.2 115.2Maximum wall heat flux, WV/cm2 34.09 55.16 55.16 55.16 55.16Maximum flow velocity, cm/sec 16.20 19.26 15.76 17.18 21.78Max. clad outer temperature, OC 131.9 136.7 137.2 137.2 136.6Exit clad outer temperature, C 124.6 127.1 127.5 127.7 127.3Peak fuel temperature, *C 478.7 525.3 525.7 525.6 525.2Minimum DNBR -Bemath -- 2.58 3.05 2.92 2.53Minimum DNBR- Groeneveld '86 -- 4.79 4.88 4.75 4.75Minimum DNBR -Groeneveld '06 -- 4.24 4.04 3.94 4.1845 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398R00001/0Table 4-20: Steady State Results for NSC, 1.25 MWChannel Channel Channel :ChannelInitial Cot e (90 Elements) 'Average .1 222 223 224Mass flow rate, kg/sec 0.0931 0.0998 0.1043 0.1170 0.1179Exit coolant temperature, 0C 95.5 99.9 95.7 101.3 103.0Coolant saturation temperature, °C 115.2 115.2 115.2 115.2 115.2Maximum wall heat flux, \V/cnr2 42.62 68.95 68.95 68.95 68.95Maximum flow velocity, cmn/sec 17.61 22.19 18.22 20.72 26.38Max. clad outer temperature, 0C 134.1 139.3 139.8 139.8 139.2Exit clad outer temperature, 0C 126.4 129.0 129.2 129.5 129.1Peak fuel temperature, 0C 516.5 561.6 562.0 562.0 561.5Minimum DNBR -Bernath -- 2.02 2.39 2.31 2.01Minimum DNBR -Groeneveld '86 3.83 3.91 3.85 3.89Minimum DNBR-Groeneveld '06 -- 3.57 3.22 3.15 3.36Tables 4-19 and 4-20 include results for the minimum departure from nucleate boiling (DNB)ratio using three correlations. Historically, TRIGA reactors have used the Bernath correlation5to predict DNB. RELAP5 Mod 3.3 uses the 1986 Critical Heat Flux Lookup Table byGroeneveld6.This Lookup Table was updated in 20067 to include more experimental data and toaddress concerns with DNB at low pressure. The Bernath correlation is considered to be quiteconservative and the Groeneveld 2006 correlation is considered to be more accurate. Allcorrelations have a minimum DNBR greater than the requirement of 2.0 given in NUREG-1537.4.8 Thermal Neutron Flux Valtes -NSC LEU Core4.8.1 Thermal Neutron Flux Values in LEU Core, Calculated[No change]4.8.2 Thermal Neutron Fltx Valtes in LEU Core, Measured[No change]46 Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgrade1 Lawrence, R.D., "The DJF3D Nodal Neutronics Option for Two-and-Three-Dimensional Diffusion TheoryCalculations in Hexagonal Geometry," Doc. No. ANL-83-1, Argonne National Laboratory, March 1983.2 Derstine, K.L., "DIF3D: A Code to Solve One-, Two-, and Three-Dimensional Finite Difference Diffusion TheoryProblems," Doc. No. ANL-82-643 Los Alamos X-5 Monte Carlo Team, "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5,"LA-UR-03-1987, April 24, 2003.4, "RELAPS/Mod 3.3 Code Manual, Vols. 1- 8," US Nuclear Regulatory Commission, NUREG/CR-5535, Rev. P3,March 2006.s Bernath, L. B., "A Theory of Local Boiling Burnout and Its Application to Existing Data," Chem. Eng. ProgressSymposium Series No. 30, Vol. 56, pp 95-116.6 Groeneveld, D. C., et al., "1986 AECL-UO Critical Heat Flux Lookup Table," Heat Transfer Engineering, Vol. 7, Nos.1-2, pp 48-62, 1986.7 Groeneveld, D. C., et al., "The 2006 CHF Look-up Table," Nuclear Engineering and Design 237, pp 1909-1922,2007.47 | ||
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ML15009A279 | |
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Site: | 05000128 |
Issue date: | 11/13/2014 |
From: | McDeavitt S M Texas A&M Univ |
To: | Wertz G A Document Control Desk, Office of Nuclear Reactor Regulation |
References | |
2014-0069, TAC ME1584 | |
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Jmj I TEXAS A&M ENGINEERINGEXPERIMENT STATIONNUCLEAR SCIENCE CENTERNovember 13, 2014 2014-0069Document Control DeskATTN: Geoffrey WertzU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001
Subject:
Response to U.S. NRC Request for Additional Information, Review of The Fuel Pool Temperatureon Fuel Temperature for the License Renewal for the Nuclear Science Center TRIGA Reactor(TAC No. ME 1584), from the Texas A&M University System, Texas Engineering ExperimentStation, Nuclear Science Center Reactor (NSCR, License No. R-83, Docket 50-128)To Whom It May Concern:The Texas A&M University System, Texas Engineering Experiment Station (TEES), Nuclear Science Center (NSC,License No. R-83) operates a LEU, 1MW, TRIGA reactor under timely renewal. In December, 2003 the NSCsubmitted a Safety Analysis Report (SAR) as part of the license renewal process. In December, 2005 a conversionSAR (Chapter 18) was submitted resulting in an order to convert from the USNRC. In July 2009, the NSCsubmitted an updated SAR, dated June 2009, to the USNRC. This updated 2009 version of our SAR incorporatedthe information from the conversion SAR and the startup of the new LEU reactor core. On July 15, 2014 the U.S.NRC submitted a Request for Additional Information re: Review of The Fuel Pool Temperature on FuelTemperature for the License Renewal for the Nuclear Science Center TRIGA Reactor (TAC No. ME 1584) as a partof the review process. Attached is our reply to this question.It is interesting to note that after the initial license application was made in 2009, a study was commissioned fromGeneral Atomics in 2010 to investigate the implications of a power uprate at NSC. Many of the issues regarding thisRAI were carefully analyzed as part of that study and have been used as the basis for part of our response. Thisresponse was reviewed internally by Dr. Karen Vierow, Associate Professor of the Texas A&M UniversityDepartment of Nuclear Engineering. We believe the contents of this response will satisfy the concerns raised by thereviewers.If you have any questions, please contact Dr. Sean McDeavitt or Mr. Jan Vermaak at 979-845-7551.TEL. 979.845.7551 I FAX 979.862.2667nsc.tamu.edu1095 Nuclear Science Rd. 1 3575 TAMU I College Station, TX 77843-3575 I declare under penalty of perjury that the foregoing is true and correct. Executed on November 13, 2014.Dr. Sean M. McDeavittInterim Director, Nuclear Science CenterAssociate Professor of Nuclear EngineeringDepartment of Nuclear EngineeringDwight Look College of EngineeringTexas A&M Universitymcdeavitt@tamu.eduXc: 2.1 1/Central FileDuane Hardesty, USNRC Project Manager In response to the four items in the request for additional information (RAI) by letter dated July 15,2014, the Nuclear Science Center (NSC) would like to withdraw the reference report "Evaluation of PoolTemperature on Fuel Temperature and MDNBR [Minimum Departure from Nucleate Boiling Ratio]"[1]and replace it with "Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade"[2], prepared for the Texas A&M Engineering Experiment Station by General Atomics underResearch Service Agreement C13-00345 and dated.November 26, 2013.JustificationThe methodology used for the analysis presented in[1]involved the application of a single phase flowsolver combined with a predictive correlation to determine the critical heat flux (Bernath correlation).Upon further investigation, it was determined that the implementation was not subjected to extensivereview to determine the validity of its application. This is of particular concern when regarding the twophase flow regime existing during an approach to the Critical Heat Flux (CHF) which hints at thefundamental need for the flow solver to be able to calculate two phase flow. Stemming from thisconcern, the NSC will not continue to support the analysis done with this methodology and as acorrective measure will replace this supporting analysis with the safety analysis performed by GeneralAtomics (GA) in November 2013 [2].The GA safety analysisis contained in a report [2] which was originally requested in order to support theoperation of the NSC reactor at 1.25 MW, but with a nominal operating power of 1.0 MW. Additionally,the thermal-hydraulic analyses were performed for a pool inlet temperature of 60°C which was requiredto support a proposed limiting safety system setting (LSSS) on pool inlet temperature.The GA safety analysis indicated an acceptable minimum DNBR of greater than 2.0 at a reactor power of1.25 MW and a pool inlet temperature of 60°C.Summary of the relevant information in the GA safety analysis:In section 4.7 of[2], the analysis of the steady operation of the NSC reactor is presented. In this safetyanalysis General Atomics has diverted from using their proprietary STAT code (as used in [3]) andreplaced it with RELAP5/Mod 3.3 [4]for which a model description is provided[2]. The steady stateanalysis was performed for a pool inlet temperature of 60°C which differs that of [3] which was for apoolinlet temperature of 30°C. The 60°C inlet temperature, as well as an additional analysis at 1.25 MW,are parameters that were requested by the NSC as part of the service agreement to GA to allow for amore flexible basis for the reactor safety case (i.e. SCRAM limits). The following changes in operatingparameters (between [2] and [3]) were observed:* Maximum clad outer temperature at 1.0 MW increased to 137.2°C from 135.9°C." Maximum fuel temperature at 1.0 MW increased to 526°C from 368.1°C.* Hottest channel outlet temperature at 1.0 MW increased to 101.0°C from 76.8°C.* Minimum DNBR increased to 2.53 at 60°C inlet temperature from 2.42 at 30°C inlettemperature.The minimal increase in cladding outer temperature is expected due to the nature of the heat transferproperties at increased coolant temperature (i.e. dynamic viscosity decreases, thermal conductivityincreases) as well as the higher thermal expansion coefficient of water at higher temperature (i.e. morebuoyancy driven mass flow).The change in outlet temperature is directly related to the difference ininlet temperature (60°C vs 30'C) as well as minor differences in the total heat generation.
The only major change is that of the maximum fuel centerline temperature which increased by 158°C.This large step is counter-intuitive especially since the two models essentially should have the samesolid conduction models, however, it was found that the step change resulted from the assumptionsmade for the gap between the cladding and the fuel pellets. In [3] it was assumed constant at 3micrometers but in [2] it was specified as 12 micrometers at a reactor power of 1.0 MW which is 4 timeslarger. This increase in gap size dramatically increases the thermal resistance of the gap and translatesto the associated increase in centerline temperature.The change in minimum DNBR is largely due to the non-conventional method used in [3] to establish theDNBR, where the reactor power was increased to determine the point of unity in terms of DNBR.However, such a method unnecessarily penalizes the operational conditions and after internalinvestigations it was determined that, if [3] applied the conventional method for determining DNBR, thevalue would approximately be 3.32 at 30°C inlet temperature. Also, the DNBR value at 60°C inlettemperature was also validated.References[1] General Atomics, "Evaluation of Pool Temperature on Fuel Temperature and MDNBR [MinimumDeparture from Nucleate Boiling Ratio]", GA Project 30398, 30398R00001 Rev. 0, Nov. 2013.[2] "Safety Analysis Report for the Texas A&M University Nuclear Science Center Power Upgrade",General Atomics, Project 30398, November 2013.[3] "Safety Analysis Report -Texas A&M University System -Nuclear Science Center Reactor",Docket number 50-128, License number R-83, May 2011.[4] "RELAP5/Mod 3.3 Code Manual, Vols. 1-8", US Nuclear Regulatory Commission, NUREG/CR-5535, Rev. P3, March 2006.
Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgrade4.7 Thermal Hydraulic Analysis -NSC LEU4.7.1 Analysis of Steady State Operation -NSCThe following evaluation has been made for the TRIGA LEU 30/20 fuel system with four-rodconfiguration operating with cooling from natural colvection water flow through four-rodclusters of fuel. The steady state thermal-hydraulic performance of the NSC LEU TRIGA corewas determined for operation at 1.0 and 1.25 MW with a water inlet temperature of 60 'C.The RELAP5 mod 3.3 computer code4 was used to determine the natural convection flow rate,the coolant and clad axial temperature profiles, and the clad wall heat flux axial profile. TheRELAP5 code was also used to determine the clad wall maxi -mum heat flux for departure fromnucleate boiling. RELAP5 was also used to determine the fuel average and fuel maximumtemperatures.An average powered fuel rod and the maximum powered fuel rod were analyzed. Four differentflow geometries were used to analyze the maximum powered fuel rod so that the most limitingthermal hydraulic condition could be determined. The arrangement of the four flow geometriesaround the maximum power fuel element (5D3) is depicted in Figure 4-2 1. Flow channel 221 isformed by the maximum rod 5D3, the transient rod 5D4, and fuel rods 5D I and 5D2. It has thesmallest hydraulic diameter and smallest flow area. Flow channel 222 is formed by themaximum rod 5D3, the transient rod 5D4, and fuel rods 4D1 and 4D2. Flow channel 223 isformed by the maximum rod 5D3 and fuel rods 5E4, 4D2, and 4El. It has the largest hydraulicdiameter. Flow channel 224 is formed by the maximum rod 5D3 and fuel rods 5D2, SEl, and5E4. The RELAP5 model shown in Figure 4-22 also includes flow channels for theinstrumented fuel element (lFE). The fuel element with the maximum intra-rod power density isalso shown in Figure 4-22 but is only used in the model for reactivity insertion transients.41 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398R00001/0U,tJUAllt~i'Uii;Figure 4-21: Flow Chnnnels That Include Maximum Fuel Element 5D342 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398REJ0001/0Figure 4-22: RELAP5 Model of Texas A&M NSC TRIGA4.7.2 RELAPS Model Input -NSCThe'RELAP5 analysis is a more modern and validated method than the method outlined inSection 4.6. The general reactor geometry and hydraulic data for the RELAP5 model are given inTable 4-16. The natural convection system for the NSC TRIGA is based on the four-rod clusterof fuel elements. The representation used herein establishes multiple flow channels eachtypically bounded by four fuel elements.A RELAP5 steady state thermal hydraulic analysis was done for an average flow channel, an IFEchannel and four flow channels containing the maximum powered fuel rod. The average flowchannel has a flow area associated with the 90 fuel rods in the NSC LEU core. The IFE has thesame hydraulic characteristics as the average flow channel except its flow area is for only onefuel rod. The hydraulic characteristics of the flow channels and power factorsof the associatedfuel rods are presented in Table 4-17.The gap between the.fuel and cladding is an important parameter for predicting fueltemperatures. Differences in thermal expansion between the fuel and cladding at operatingtemperatures reduce the initial manufactured gap and can result in gap closure at locations of43 Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgradehigh heat generation. The average radial gap in the NSC LEU fuel is 2.15 rail (54.6 pul). Theradial gap for the IFE in location 5E4 is 2.75 mil (69.9 pm). The radial gap for the maximumpower fuel rod in location 5D3 is 2.66 mil (67.6 r.m). Table 4-18 presents the effective radialgap at various reactor operating power levels. For the maximum and IFE fuel rods, the gap sizereduction is based on the peak axial power profile. For the average fuel element, the gap sizereduction is based on an axial power factor of 1.0 so that the gap represents the average gapwithin the average fuel element.Table 4-16: RELAP5 Input for Reactor and Core Geometry and Hydraulic Data, NSC: Parameter ValueCore and Reactor GeometryUnrteated grid plate length, in 2.00Unheated lower refleclor length, in 3.95Unheated top reflector length, in 4.06Length from top of unheated length to top of pool surface, ft 23.685Fuel element hetited length (in) 15Fuel element diameter (in) 1.4126Hydraulic DataSurface RoughnessGrid plate K lossAmbient pressure at elevation 320 ft, psiaTable 4-17: Hydraulic Flow Parameters and Power Factors, NSC..ara.eter .. :. ... !.' ig : I Channel Channel .Channel:. Channel..22.1 222 223. 224.Flow area (in2) 76.7196 0.852440 0.732495 0.929865 0.929581 0.735436WVetted perimeter (in/element) 399.403 4.437814 4.494677 4.494677 4.437814 4.437814Hydraulic diameter (in) 0.768342 0.768342 0.651877 0.827525 0.837873 0.662881Fuel rod entrance K loss 0.236 0.236 0.174 0.281 0.281 0.176Fuel rod exitX loss 1.289 1.289 0.952 1.534 1.533 0.960Ma.ximum rod power factor 1.00 1.484 1.648 1.648 1.648 1.648Power factor of other rods -- -- 1.584 1.432 1.400 1.493Peak axial power factor 1.318 1.318 1.294 1.294 1.294 1.29444 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398RO0001/0Table 4-18: Radial Gap Between Fuel and Cladding*Radial Cap in 1111 (11mr)Fuel ElementCold 0.5 o 1MW 1.0 MW .1.25 nVAverage 2.15 (54.6) 1.55 (39.4) 0.95 (24.1) 0.77 (19.6)Instrumented Fuel Element 2.75 (69.9) 1.16 (29.5) 0.55 (14.0) 0.39 (9.91)Maximum Power Fuel Element 2.66 (67.6) 1.04 (26.4) 0.47 (I1.9) 0.33 (8.38)4.7.3 Steady State Analysis Results -NSCTables 4-19 and 4-20 list the pertinent heat transfer and hydraulic steady state results for the 1.0and 1.25 MW NSC TRIGA reactor, respectively. Results are presented for an average chatneland the four maximum powered flow channels (hot channels) at initial core conditions. Theaverage core temperature at 1.0 MW is 354.9 'C. It increases to 373.4 'C at 1.25 MW. The IFElocated at 5E4 has a maximum temperature of 497.3 °C at 1.0 MW. It increases to 527.1 °C at1.25 MW.Table 4-19: Steady State Results for NSC, 1.00 MWChannel Chann el Channel .'Channel*ý.lInitial core (90 Elemients) Aerage 21222. :223 .224Mass flow rate, kg/sec 0.0860 0.0875 0.0912 0.0984 0.0989Exit coolant temperature, 'C 90.8 96.4 92.8 99.3 101.0Coolant saturation temperature, °C 1 15.2 115.2 115.2 115.2 115.2Maximum wall heat flux, WV/cm2 34.09 55.16 55.16 55.16 55.16Maximum flow velocity, cm/sec 16.20 19.26 15.76 17.18 21.78Max. clad outer temperature, OC 131.9 136.7 137.2 137.2 136.6Exit clad outer temperature, C 124.6 127.1 127.5 127.7 127.3Peak fuel temperature, *C 478.7 525.3 525.7 525.6 525.2Minimum DNBR -Bemath -- 2.58 3.05 2.92 2.53Minimum DNBR- Groeneveld '86 -- 4.79 4.88 4.75 4.75Minimum DNBR -Groeneveld '06 -- 4.24 4.04 3.94 4.1845 Safety Analysis Report for the Texas A&M University Nuclear Science Center PowerUpgrade30398R00001/0Table 4-20: Steady State Results for NSC, 1.25 MWChannel Channel Channel :ChannelInitial Cot e (90 Elements) 'Average .1 222 223 224Mass flow rate, kg/sec 0.0931 0.0998 0.1043 0.1170 0.1179Exit coolant temperature, 0C 95.5 99.9 95.7 101.3 103.0Coolant saturation temperature, °C 115.2 115.2 115.2 115.2 115.2Maximum wall heat flux, \V/cnr2 42.62 68.95 68.95 68.95 68.95Maximum flow velocity, cmn/sec 17.61 22.19 18.22 20.72 26.38Max. clad outer temperature, 0C 134.1 139.3 139.8 139.8 139.2Exit clad outer temperature, 0C 126.4 129.0 129.2 129.5 129.1Peak fuel temperature, 0C 516.5 561.6 562.0 562.0 561.5Minimum DNBR -Bernath -- 2.02 2.39 2.31 2.01Minimum DNBR -Groeneveld '86 3.83 3.91 3.85 3.89Minimum DNBR-Groeneveld '06 -- 3.57 3.22 3.15 3.36Tables 4-19 and 4-20 include results for the minimum departure from nucleate boiling (DNB)ratio using three correlations. Historically, TRIGA reactors have used the Bernath correlation5to predict DNB. RELAP5 Mod 3.3 uses the 1986 Critical Heat Flux Lookup Table byGroeneveld6.This Lookup Table was updated in 20067 to include more experimental data and toaddress concerns with DNB at low pressure. The Bernath correlation is considered to be quiteconservative and the Groeneveld 2006 correlation is considered to be more accurate. Allcorrelations have a minimum DNBR greater than the requirement of 2.0 given in NUREG-1537.4.8 Thermal Neutron Flux Valtes -NSC LEU Core4.8.1 Thermal Neutron Flux Values in LEU Core, Calculated[No change]4.8.2 Thermal Neutron Fltx Valtes in LEU Core, Measured[No change]46 Safety Analysis Report for the Texas A&M University Nuclear Science Center Power 30398R00001/0Upgrade1 Lawrence, R.D., "The DJF3D Nodal Neutronics Option for Two-and-Three-Dimensional Diffusion TheoryCalculations in Hexagonal Geometry," Doc. No. ANL-83-1, Argonne National Laboratory, March 1983.2 Derstine, K.L., "DIF3D: A Code to Solve One-, Two-, and Three-Dimensional Finite Difference Diffusion TheoryProblems," Doc. No. ANL-82-643 Los Alamos X-5 Monte Carlo Team, "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5,"LA-UR-03-1987, April 24, 2003.4, "RELAPS/Mod 3.3 Code Manual, Vols. 1- 8," US Nuclear Regulatory Commission, NUREG/CR-5535, Rev. P3,March 2006.s Bernath, L. B., "A Theory of Local Boiling Burnout and Its Application to Existing Data," Chem. Eng. ProgressSymposium Series No. 30, Vol. 56, pp 95-116.6 Groeneveld, D. C., et al., "1986 AECL-UO Critical Heat Flux Lookup Table," Heat Transfer Engineering, Vol. 7, Nos.1-2, pp 48-62, 1986.7 Groeneveld, D. C., et al., "The 2006 CHF Look-up Table," Nuclear Engineering and Design 237, pp 1909-1922,2007.47