ML12020A219
| ML12020A219 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/17/2012 |
| From: | AREVA NP, Constellation Energy Nuclear Group |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| ANP-3043(NP), Rev. 1 | |
| Download: ML12020A219 (139) | |
Text
ATTACHMENT (3)
NON-PROPRIETARY - ANP-3043(NP), CALVERT CLIFFS RLBLOCA
SUMMARY
REPORT, REVISION 1, DECEMBER 2011 Calvert Cliffs Nuclear Power Plant, LLC January 17, 2012
ANP-3043(NP)
Revision 1 Calvert Cliffs RLBLOCA Summary Report December 2011 A
AREVA NP Inc. AREVA
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 2 Copyright © 2011 AREVA NP Inc.
All Rights Reserved AREVA NP Inc
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 3 Nature of Changes Item Page Description and Justification I All Revision 0: This is a new document.
2 30 Revision 1: Table 3-5. Consolidated the value for Total Core-Wide Oxidation to be applicable to Fresh Fuel and Once Burnt Fuel.
3 48 Revision 1: Figure 3-11: Replaced with correct figure.
4 24 Revision 1: Table 3-2, line item 1.0 (g): Deleted value under 'Operating Range.'
5 138 Revision 1: Inserted References 18 through 22 to make consistent with proprietary version.
AREVA NP Inc
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 4 Table of Contents 1.0 Introduction ................................................................................................................... 11 2.0 Sum m ary ........................................................................................................................ 14 3.0 Analysis ......................................................................................................................... 16 3.1 Description of the LBLOCA Event ................................................................... 17 3.2 Description of Analytical Models ...................................................................... 18 3.3 Plant Description and Summary of Analysis Parameters ................................ 20 3.4 SER Com pliance ............................................................................................ 22 3.5 Realistic Large Break LOCA Results .............................................................. 22 4.0 Generic Support for Transition Package ................................................................ 61 4.1 Reactor Power ................................................................................................. 61 4.2 Rod Quench ................................................................................................... 61 4.3 Rod-to-Rod Therm al Radiation ....................................................................... 61 4.4 Film Boiling Heat Transfer Lim it ..................................................................... 67 4.5 Downcomer Boiling ....................................................................................... 68 4.6 Break Size ...................................................................................................... 84 4.7 Detailed information for Containment Model (ICECON) .................................. 95 4.8 Cross-References to North Anna .................................................................. 99 4.9 GDC 35 - LOOP and No-LOOP Case Sets ...................................................... 100 4.10 Statem ent .......................................................................................................... 101 5.0 Conclusions ................................................................................................................. 102 6.0 Recent NRC Request for Additional Information (RAI) and AREVA NP Responses ................................................................................................................... 103 6.1 Thermal Conductivity Degradation - Once-Burnt Fuel ...................................... 103 6.2 Decay Heat Uncertainty Assum ption ................................................................. 115 6.3 Clad Swelling and Rupture ................................................................................ 122 6.4 Single-Sided Oxidation Model ........................................................................... 126 6.5 Limiting Condition - Single Failure ..................................................................... 127 6.6 Core Liquid Levels ............................................................................................. 135 6.7 Plant Input Selection and Technical Specifications ........................................... 136 7.0 References ................................................................................................................... 137 This document contains a total of 138 pages AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 5 List of Tables Table 2-1 Summary of Major Parameters for Limiting Transient ........................................... 15 Table 3-1 Sampled LBLOCA Parameters ............................................................................ 23 Table 3-2 Plant Operating Range Supported by the LOCA Analysis ..................................... 24 Table 3-3 Statistical Distributions Used for Process Parameters ......................................... 27 Table 3-4 SER Conditions and Limitations ............................................................................ 28 Table 3-5 Summary of Results for the Limiting PCT Case .................................................. 30 Table 3-6 Calculated Event Times for the Limiting PCT Case .............................................. 30 Table 3-7 Heat Transfer Parameters for the Limiting Case .................................................. 31 Table 3-8 Containment Initial and Boundary Conditions ....................................................... 33 Table 3-9 Passive Heat Sinks in Containment ..................................................................... 34 Table 3-10 Material Properties for Passive Heat Sinks in Containment ............................... 35 Table 4-1 Typical Measurement Uncertainties and Local Peaking Factors .......................... 64 Table 4-2 FLECHT-SEASET & 17x17 FA Geometry Parameters ......................................... 64 Table 4-3 FLECHT-SEASET Test Parameters ..................................................................... 66 Table 4-4 Minimum Break Area for Large Break LOCA Spectrum ........................................ 86 Table 4-5 Minimum PCT Temperature Difference - True Large and Intermediate B re a ks ............................................................................................................................. 88 AREVA NP Inc.
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Rev. 1 Page 6 List of Figures Figure 3-1 Primary System Noding for CCNPP ..................................................................... 36 Figure 3-2 Secondary System Noding ................................................................................... 37 Figure 3-3 Reactor Vessel Noding ........................................................................................ 38 Figure 3-4 Core Noding Detail .............................................................................................. 39 Figure 3-5 Upper Plenum Noding Detail .............................................................................. 40 Figure 3-6 Scatter Plot of Operational Parameters .............................................................. 41 Figure 3-7 Scatter Plot of PCT versus PCT Time ................................................................ 43 Figure 3-8 Scatter Plot of PCT versus Break Size ................................................................. 44 Figure 3-9 Scatter Plot of Maximum Transient Oxidation versus PCT ................................... 45 Figure 3-10 Scatter Plot of Total Oxidation versus PCT ....................................................... 46 Figure 3-11 Containment Volume versus PCT Scatter Plot from 59 LOOP C a lculatio ns ..................................................................................................................... 47 Figure 3-12 Peak Cladding Temperature (Independent of Elevation) for the Lim iting Case .................................................................................................................. 48 Figure 3-13 Break Flow for the Limiting Case ........................................................................ 49 Figure 3-14 Core Inlet Mass Flux for the Limiting Case ....................................................... 50 Figure 3-15 Core Outlet Mass Flux for the Limiting Case ..................................................... 51 Figure 3-16 Void Fraction at RCS Pumps for the Limiting Case ............................................ 52 Figure 3-17 ECCS Flows (Includes SIT, HPSI and LPSI) for the Limiting Case ................... 53 Figure 3-18 Upper Plenum Pressure for the Limiting Case .................................................. 54 Figure 3-19 Collapsed Liquid Level in the Downcomer for the Limiting Case ....................... 55 Figure 3-20 Collapsed Liquid Level in the Lower Plenum for the Limiting Case .................. 56 Figure 3-21 Collapsed Liquid Level in the Core for the Limiting Case ................................. 57 Figure 3-22 Containment and Loop Pressures for the Limiting Case .................................... 58 Figure 3-23 Reactor Vessel Liquid Mass for the Limiting Case ........................................... 59 Figure 3-24 GDC 35 LOOP versus No-LOOP Cases ............................................................ 60 Figure 4-1 R2RRAD 5x5 Rod Segment ................................................................................. 65 Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x1 7 FA .................................................................................................................................... 67 Figure 4-3 Reactor Vessel Downcomer Boiling Diagram ..................................................... 69 Figure 4-4 S-RELAP5 versus Closed Form Solution ............................................................ 72 AREVA NP Inc.
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Rev. 1 Page 7 Figure 4-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity ............................. 73 Figure 4-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity .............................. 74 Figure 4-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity ....................................... 75 Figure 4-8 Core Liquid Level - Wall Mesh Point Sensitivity .................................................. 76 Figure 4-9 Azimuthal Noding ................................................................................................ 78 Figure 4-10 Lower Compartment Pressure versus Time ....................................................... 79 Figure 4-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study ..................... 80 Figure 4-12 PCT Independent of Elevation - Axial Noding Sensitivity Study ....................... 81 Figure 4-13 Downcomer Liquid Level - Axial Noding Sensitivity Study ................................ 82 Figure 4-14 Core Liquid Level- Axial Noding Sensitivity Study ........................................... 83 Figure 4-15 Plant A - Westinghouse 3-Loop Design ............................................................ 89 Figure 4-16 Plant B - Westinghouse 3-Loop Design ............................................................ 90 Figure 4-17 Plant C - Westinghouse 3-Loop Design ............................................................ 91 Figure 4-18 Plant D - Combustion Engineering 2x4 Design ............................................... 92 Figure 4-19 Plant E - Combustion Engineering 2x4 Design ................................................ 93 Figure 4-20 Plant H - Westinghouse 4-Loop Design ............................................................ 94 Figure 4-21 PCT vs. Containment Volume ........................................................................... 96 Figure 4-22 PCT vs. Initial Containment Temperature ......................................................... 97 Figure 4-23 Containment Pressure for Limiting Case ............................................................ 98 Figure 6-1 Once-Burnt Fuel Power Ratios (2nd cycle) ............................................................ 108 Figure 6-2 Radial Temperature Profile for Hot Rod ................................................................ 109 Figure 6-3 Temperature versus Time for Fuel Centerline, Clad Surface, and Fuel Ave rag e ......................................................................................................................... 1 10 Figure 6-4 Fresh versus Once-Burnt UO2 Rod PCT Trace ...................................................... 111 Figure 6-5 Fractional Fuel Centerline Temperature Delta Between RODEX3A a n d Da ta ........................................................................................................................ 112 Figure 6-6 Fuel Centerline Temperature Delta of RODEX3A Calculations to Data (Original and Using the New Correlation) ..................................................................... 113 Figure 6-7 Correction Factor (as applied for temperatures in Kelvin) ...................................... 114 Figure 6-8 Decay Heat Comparisons, Infinite Operation U235, Finite Operation A ll Isotopes (0.1 - 10 sec) ............................................................................................. 118 Figure 6-9 Decay Heat Comparisons, Infinite Operation U235, Finite Operation A ll Isotopes (10 - 1000 sec) .......................................................................................... 119 AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. I Page 8 Figure 6-10 Decay Heat Ratios, Finite Operation over Infinite Operation for U235, All Isotopes (0 - 10 sec) ..................................................................................... 120 Figure 6-11 Decay Heat Ratios, Finite Operation over Infinite Operation for U235, All Isotopes (0 - 600 sec) ................................................................................... 121 Figure 6-12 Swell, Rupture, & Relocation Sensitivity ............................................................... 123 Figure 6-13 Clad Temperature Response from Single Failure Study ...................................... 129 Figure 6-14 Comparison of Maximum Clad Temperature Independent of Elevation for Maximum ECCS and Minimum ECCS (AOR) .......................................... 131 Figure 6-15 Comparison of Containment and System Pressure for Maximum ECCS and Minimum ECCS (AOR) ............................................................................... 132 Figure 6-16 Comparison of ECCS Flows for Maximum ECCS and Minimum EC C S (AO R ) ................................................................................................................. 133 Figure 6-17 Comparison of Downcomer Level for Maximum ECCS and Minimum E C C S (AO R ) ................................................................................................................. 134 AREVA NP Inc.
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Rev. 1 Page 9 Nomenclature AOR Analysis of Record ARO All Rods Out ASI Axial Shape Index CCTF Cylindrical Core Test Facility CCNPP Calvert Cliffs Nuclear Power Plant CE Combustion Engineering Inc.
CFR Code of Federal Regulations COLR Core Operating Limits Report CSAU Code Scaling, Applicability, and Uncertainty DC Downcomer DEGB Double-Ended Guillotine Break DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPH Effective Full Power Hours EM Evaluation Model FLECHT-SEASET Full-Length Emergency Core Heat Transfer Separate Effects and Systems Effects Tests F0 Total Peaking Factor Fr Nuclear Enthalpy Rise Factor HFP Hot Full Power HPSI High Pressure Safety Injection HTC Heat Transfer Coefficient LBLOCA Large Break Loss of Coolant Accident LANL Los Alamos National Laboratory LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LOFT Loss of Fluid Test LOOP Loss of Offsite Power LPSI Low Pressure Safety Injection MSIV Main Steam Isolation Valve MWt Mega-Watt thermal NRC U. S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System ORNL Oak Ridge National Laboratory AREVA NP Inc.
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Rev. 1 Page 10 Nomenclature (Continued)
PCT Peak Clad Temperature PDIL Power Dependent Insertion Limit PIRT Phenomena Identification and Ranking Table PLHGR Planar Linear Heat Generation Rate PWR Pressurized Water Reactor RAS Recirculation Actuation Signal RCP Reactor Coolant Pump RCS Reactor Coolant System RLBLOCA Realistic Large Break LOCA RV Reactor Vessel RWST Refueling Water Storage Tank SER Safety Evaluation Report SI Safety Injection SIAS Safety Injection Actuation Signal SIT Safety Injection Tank THTF Thermal-Hydraulic Test Facility TS Technical Specifications AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. I Page 11 1.0 Introduction This report describes and provides results from a RLBLOCA analysis for the two-unit Calvert Cliffs Nuclear Power Plant (CCNPP). The two-unit plant is a Combustion Engineering (CE) designed 2737 MWt plant with a large dry containment. The plant is a 2X4 loop design - two hot legs and four cold legs. The loops contain four reactor coolant pumps (RCPs), two U-tube steam generators and one pressurizer. The ECCS is provided by two independent safety injection trains and four safety injection tanks (SITs).
The analysis supports operation with AREVA NP's HTP 14X14 fuel design using standard U0 2 fuel and U0 2 fuel with 2%, 4%, 6% and 8% Gd 20 3 with M5 cladding, unless changes in the Technical Specifications, Core Operating Limits Report, core design, fuel design, plant hardware, or plant operation invalidate the results presented herein. The analysis was performed in compliance with the U.S. Nuclear Regulatory Commission (NRC) approved RLBLOCA EM (Reference 1) with exceptions noted below. Analysis results confirm that the 1 OCFR50.46 (b) acceptance criteria presented in Section 3.0 are met and serve as the basis for operation of the CCNPP with AREVA NP fuel.
This RLBLOCA analysis was originally performed for the first supply of AREVA fuel with parameters from CCNPP. The analysis is being updated to address NRC concerns on decay heat sampling and the inclusion of once-burnt fuel rods. The revised analysis is not a cycle specific analysis and is intended to be bounding for future AREVA fuel reloads at CCNPP.
The applicability of the analysis to follow-on fuel reloads will be evaluated on a cycle-to-cycle basis.
The non-parametric statistical methods inherent in the AREVA NP RLBLOCA methodology provide for the consideration of a full spectrum of break sizes, break configuration (guillotine or split break), axial shapes, and plant operational parameters. A conservative loss of an emergency diesel generator assumption is applied. The effect of this is the loss of one LPSI pump and one HPSI pump. The LPSI injects into the broken loop and one intact loop and HPSI injects into all four loops. Regardless of the single-failure assumption, all containment pressure-reducing systems are assumed fully functional. The effects of Gadolinia-bearing fuel rods and peak fuel rod exposures are considered.
The following are deviations from the approved RLBLOCA EM (Reference 1) that were requested by the NRC and are referred to as the "Transition Package." The "Transition Package" is fully described in Section 4.0.
The assumed reactor core power for the CCNPP realistic large break loss-of-coolant accident is 2754 MWt. This value represents the 100% primary power (2737 MWt) plus 17 MWt (0.62% of primary power) to account for the measurement uncertainty. The power was not sampled in the analysis. This is not expected to have a noticeable effect on the PCT results.
AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. I Page 12 The RLBLOCA analysis was performed with a version of S-RELAP5 that requires both the void fraction to be less than 0.95 and the clad temperature to be less than 900°F before the rod is allowed to quench. This may result in a slight increase in peak clad temperature (PCT) results when compared to an analysis not subject to these constraints.
The RLBLOCA analysis was performed with a version of S-RELAP5 that limits the contribution of the Forslund-Rohsenow model to no more than 15% of the total heat transfer at and above a void fraction of 0.9. This may result in a slight increase in PCT results when compared to previous analyses for similar plants.
The split versus double-ended guillotine break (DEGB) type is no longer related to break area.
In concurrence with Regulatory Guide 1.157, both the split and the double-ended guillotine break will range in area between the minimum break area (Amen) and an area of twice the cross-sectional area of the broken pipe. The determination of break configuration, split versus double-ended guillotine, will be made after the break area is selected based on a uniform probability for each occurrence. Amin was calculated to be 28.76% of the DEGB area (see Section 4.6 for further discussion). This is not expected to have an effect on PCT results.
In concurrence with the NRC's interpretation of GDC 35, a set of 59 cases was run with a loss of offsite power (LOOP) assumption and a second set with a No-LOOP assumption. The set of 59 cases that predicted the highest PCT is reported in Section 2.0 and Section 3.0, herein. The results from both case sets are shown in Figure 3-24.
During recent RLBLOCA EM modeling studies, it was noted that cold leg condensation efficiency may be under-predicted. Water entering the downcomer (DC) post-SIT injection remained sufficiently subcooled to absorb DC wall heat release without significant boiling.
However, tests (Reference 2) indicate that the steam and water entering the DC from the cold leg, subsequent to the end of SIT injection, reach near saturation resulting from the condensation efficiency ranging from 80 to 100 percent. To assure that cold leg condensation would not be under-predicted, a RLBLOCA EM update was made. Noting that saturated fluid entering the DC is the most conservative modeling scheme, steam and liquid multipliers were developed in order to approximately saturate the cold leg fluid before it enters the DC. The multipliers were developed through scoping studies using a number of plant configurations-Westinghouse-designed 3-loop and 4-loop plants, and CE-designed plants. The results of the scoping study indicated that [ ] were appropriate to produce saturated fluid entering the DC. This RLBLOCA EM departure was discussed recently with the NRC and the NRC agreed that the approach described immediately above was satisfactory in the interim. The modification is implemented post-SIT injection, 10 seconds after the vapor void fraction in the bottom of the SIT becomes greater than 90%. Thus, the SITs have injected all their water into the cold legs, and the nitrogen cover gas has entered the system and been mostly discharged through the break before [
]. Providing saturated fluid conditions at the DC entrance conservatively reduces both the DC driving head and the AREVA NP Inc.
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Rev. 1 Page 13 core flooding rate. Recall that test results indicate that fluid conditions entering the DC range from saturated to slightly subcooled. Hence, it is conservative to force an approximation of saturated conditions for fluid entering the DC.
The NRC raised the issue concerning fuel thermal conductivity degradation as a function of burnup in Information Notice 2009-23. In order to manage this issue, AREVA NP is modifying the way RODEX3A temperatures are compensated in the RLBLOCA Revision 0/Transition Package methodology. In the current process, the RLBLOCA computes PCTs at many different times during an operating cycle. For each specific time in cycle, the fuel conditions are computed using RODEX3A prior to starting the S-RELAP5 portion of the analysis. A steady-state condition for the given time in cycle using S-RELAP5 is established. A base fuel centerline temperature is established in this process. Then two-transformation adjustments to the base fuel centerline temperature are computed. The first transformation is a linear adjustment for an exposure of 10 GWd/MTU or higher. The second adjustment is performed in the S-RELAP5 initialization process for the transient case. In the new process, a polynomial transformation is used for the first transformation instead of a linear transformation. The rest of the RLBLOCA process for initializing the S-RELAP5 fuel rod temperature should not be altered and the rest of LOCA transient should also continue in the original fashion. Section 6.0 provides additional information on the adjustment and adding once-burnt fuel to the analysis. Note that these changes are also deviations required by the NRC that are departures from the approved RLBLOCA EM.
Recent NRC concerns raised in the form of RAIs are responded to in Section 6.0.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 14 2.0 Summary The limiting PCT analysis is based on the parameter specification given in Table 2-1 for the limiting case. U0 2 only and Gadolinia-bearing U0 2 rods of 2%, 4%, 6% and 8% Gd 20 3 were analyzed for fresh and once-burnt fuel in all cases.
The RLBLOCA result is based on a case set of 59 individual transient cases for offsite power not available (LOOP) and a second case set of 59 individual transient cases for offsite power available (No-LOOP) conditions. The core is composed of AREVA NP HTP 14x14 fuel, hydraulically compatible with 14X14 Westinghouse fuel. Only AREVA NP fuel is analyzed for PCT; a mixed core scenario is hydraulically modeled. The limiting PCT is 1620°F for a fresh fuel 8% Gd 20 3 rod (Case 47) with offsite power not available (LOOP) conditions. From the same case, the limiting PCT for all once-burnt fuel is 1545 0 F for a U0 2 rod.
The analysis assumed full core power operation at 2754 MWt. The value represents the 100%
primary power plus 17 MWt uncertainty. The analysis assumed a steam generator tube plugging level of 10% in each steam generator, a LHGR limit of 15.0 kW/ft per Technical Specification 3.2.1 and the COLR, which is equivalent to a total peaking factor (FQ) up to a value of 2.37, and a radial peaking factor (Fr) up to a value of 1.810 (including 6% uncertainty and 3.5% control rod insertion effect). This analysis bounds typical operational ranges or Technical Specification limits (whichever is applicable) with regard to pressurizer pressure and level; SIT pressure, temperature, and level; core inlet temperature; core flow; containment pressure and temperature; and RWST.
The AREVA NP RLBLOCA Transition Package methodology (on a forward fit basis) explicitly analyzes fresh and once-burnt fuel assemblies to respond to recent NRC RAls. The twice-burnt fuel assemblies are not considered in the analyses since burnups at this level or higher do not retain sufficient energy potential to achieve significant cladding temperatures or cladding oxidations during the transient. The analysis demonstrates that the 10 CFR 50.46(b) criteria listed in Section 3.0 are satisfied.
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Rev. 1 Page 15 Table 2-1 Summary of Major Parameters for Limiting Transient Fresh 8% Gd 2O 3 Fuel Once-Burnt U0 2 Fuel Core Average Burnup (EFPH) 88971 89232 Core Power (MWt) 2754 Hot Rod LHGR (kW/ft) / 14.3618 /
Total Peaking (FQ) 2.26884 Radial Peaking (Fr) 1.623 1.734 Axial Offset -0.0878 -0.0949 Break Type Guillotine Break Size (ft)/side 4.5832 Offsite Power Availability Not available Decay Heat Multiplier 1.0 1 This is - 15.3 GWd/MTU in assembly bumup for the fresh fuel.
2 This is - 36.5 GWd/MTU in assembly bumup for the once-burnt fuel.
3 [
I 4 [
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Rev. 1 Page 16 3.0 Analysis The purpose of the analysis is to verify the typical technical specification peaking factor limits and the adequacy of the ECCS and to demonstrate compliance to the I OCFR 50.46(b) criteria.
- 1. The calculated maximum fuel element cladding temperature shall not exceed 2200 0 F.
- 2. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
- 3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel excluding the cladding surrounding the plenum volume were to react.
- 4. The calculated changes in core geometry shall be such that the core remains amenable to cooling.
The analysis did not evaluate the 10CFR 50.46(b) long-term cooling criterion, as this is handled in a separate analysis.
The RLBLOCA analysis conservatively considers blockage effects due to clad swelling and rupture in the prediction of the hot fuel rod PCT. The effects of combined LOCA loads on the fuel assembly components have been evaluated by AREVA NP and the resulting loads are below the allowable stress limit for all the components. The combination of compliance with the 2200°F limit and the LOCA loads evaluation ensures no permanent deformation to the fuel assemblies; thereby demonstrating compliance with the criterion that the core remains amenable to cooling.
Section 3.1 of this report describes the postulated LBLOCA event. Section 3.2 describes the models used in the analysis. Section 3.3 describes the CE 2x4 PWR plant and summarizes the system parameters used in the analysis. Compliance to the SER is addressed in Section 3.4.
Section 3.5 summarizes the results of the RLBLOCA analysis.
Section 4.0 discusses the additional information provided under the "Transition Package" on EMF-2103. Section 5.0 provides the conclusions. Section 6.0 addresses recent NRC RAls on RLBLOCA submittals and Section 7.0 contains the list of references.
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Rev. 1 Page 17 3.1 Description of the LBLOCA Event An LBLOCA is initiated by a postulated large rupture of the RCS primary piping. Based on deterministic studies, the worst break location is in the cold leg piping between the reactor coolant pump and the reactor vessel for the RCS loop containing the pressurizer. The break initiates a rapid depressurization of the RCS. A reactor trip signal is initiated when the low pressurizer pressure trip setpoint is reached; however, reactor trip is conservatively neglected in the analysis. The reactor is shut down by coolant voiding in the core.
The plant is assumed to be operating normally at full power prior to the accident. The cold leg break is assumed to open instantaneously. For this break, a rapid depressurization occurs, along with a core flow stagnation and reversal. This causes the fuel rods to experience DNB.
Subsequently, the limiting fuel rods are cooled by film convection to steam. The coolant voiding creates a strong negative reactivity effect and core criticality ends. As heat transfer from the fuel rods is reduced, the cladding temperature increases.
Coolant in all regions of the RCS begins to flash. At the break plane, the loss of subcooling in the coolant results in substantially reduced break flow. This reduces the depressurization rate, and leads to a period of positive core flow or reduced downflow as the RCPs in the intact loops continue to supply water to the RV (in No-LOOP conditions). Cladding temperatures may be reduced and some portions of the core may rewet during this period. The positive core flow or reduced downflow period ends as two-phase conditions occur in the RCPs, reducing their effectiveness. Once again, the core flow reverses as most of the vessel mass flows out through the broken cold leg.
Mitigation of the LBLOCA begins when the SIAS is issued. This signal is initiated by either high containment pressure or low pressurizer pressure. Regulations require that a worst single-failure be considered. The AREVA NP RLBLOCA methodology assumes an on-time start and normal lineups of the containment spray and fan coolers to conservatively reduce containment pressure and increase break flow, regardless of the single failure assumed. This single-failure has been determined to be the loss of one emergency diesel generator, which takes one train of ECCS pumped injection out. LPSI injects into the broken loop and one intact loop, HPSI injects into all four loops, and the containment spray system is fully functional with both trains operating.
When the RCS pressure falls below the SIT pressure, fluid from the SITs is injected into the cold legs. In the early delivery of SIT water, high pressure and high break flow will drive some of this fluid to bypass the core. During this bypass period, core heat transfer remains poor and fuel rod cladding temperatures increase. As RCS and containment pressures equilibrate, ECCS water begins to fill the lower plenum and eventually the lower portions of the core; thus, core heat transfer improves and cladding temperatures decrease.
Eventually, the relatively large volume of SIT water is exhausted and core recovery continues relying solely on pumped ECCS injection. As the SITs empty, the nitrogen gas used to AREVA NP Inc.
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Rev. 1 Page 18 pressurize the SITs exits through the break. This gas release may result in a short period of improved core heat transfer as the nitrogen gas displaces water in the downcomer. After the nitrogen gas has been expelled, the ECCS temporarily may not be able to sustain full core cooling because of the core decay heat and the higher steam temperatures created by quenching in the lower portions of the core. Peak fuel rod cladding temperatures may increase for a short period until more energy is removed from the core by the fluid delivered by the HPSI and the LPSI, while the decay heat continues to fall. Steam generated from fuel rod rewet will entrain liquid and pass through the core, vessel upper plenum, the hot legs, the steam generators, and the reactor coolant pumps before it is vented out the break. The resistance of this flow path to the steam flow is balanced by the driving force of water filling the downcomer.
This resistance may act to retard the progression of the core reflood and postpone core-wide cooling. Eventually (within a few minutes of the accident), the core reflood will progress sufficiently to ensure core-wide cooling. Full core quench occurs within a few minutes after core-wide cooling. Long-term cooling is then sustained with the LPSI pumped injection system.
3.2 Description of Analytical Models The RLBLOCA methodology is documented in EMF-2103 Realistic Large Break LOCA Methodology (Reference 1). The methodology follows the CSAU evaluation methodology (Reference 3). This method outlines an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifies the uncertainties in a LOCA analysis.
The RLBLOCA methodology consists of the following computer codes:
- RODEX3A for computation of the initial fuel stored energy, fission gas release, and fuel-cladding gap conductance.
- S-RELAP5 for the system calculation (includes ICECON for containment response).
- AUTORLBLOCA for generation of ranged parameter values, transient input, transient runs, and general output documentation.
The governing two-fluid (plus non-condensibles) model with conservation equations for mass, energy, and momentum transfer is used. The reactor core is modeled in S-RELAP5 with heat generation rates determined from reactor kinetics equations (point kinetics) with reactivity feedback, and with actinide and decay heating.
The two-fluid formulation uses a separate set of conservation equations and constitutive relations for each phase. The effects of one phase on the other are accounted for by interfacial friction, and heat and mass transfer interaction terms in the equations. The conservation equations have the same form for each phase; only the constitutive relations and physical properties differ.
The modeling of plant components is performed by following guidelines developed to ensure accurate accounting for physical dimensions and that the dominant phenomena expected during AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 19 the LBLOCA event are captured. The basic building blocks for modeling are hydraulic volumes for fluid paths and heat structures for heat transfer. In addition, special purpose components exist to represent specific components such as the RCPs or the steam generator separators.
All geometries are modeled at the resolution necessary to best resolve the flow field and the phenomena being modeled within practical computational limitations.
System nodalization details are shown in Figures 3-1 through 3-5. Figure 3-1 shows the primary system nodalization and Figure 3-2 shows the secondary system nodalization. A point of clarification: in Figure 3-1, break modeling uses two junctions regardless of break type-split or guillotine; for guillotine breaks, Junction 151 is deleted, it is retained fully open for split breaks. Hence, total break area is the sum of the areas of both break junctions.
A typical calculation using S-RELAP5 begins with the establishment of a steady-state initial condition with all loops intact. The input parameters and initial conditions for this steady-state calculation are chosen to reflect plant operational characteristics or to match measured data.
Additionally, the RODEX3A code provides initial conditions for the S-RELAP5 fuel models.
Specific parameters are discussed in Section 3.3.
Following the establishment of an acceptable steady-state condition, the transient calculation is initiated by introducing a break into one of the loops (specifically, the loop with the pressurizer).
The evolution of the transient through blowdown, refill and reflood is computed continuously using S-RELAP5. Containment pressure is also calculated by S-RELAP5 using containment models derived from ICECON (Reference 4), which is based on the CONTEMPT-LT code (Reference 5) and has been updated for modeling ice condenser containments.
The methods used in the application of S-RELAP5 to the LBLOCA are described in Reference 1. A detailed assessment of this computer code was made through comparisons to experimental data, many benchmarks with cladding temperatures ranging from 1,700°F (or less) to above 2,200 0 F. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena in a PWR LBLOCA. Various models-for example, the core heat transfer, the decay heat model and the fuel cladding oxidation correlation-are defined based on code-to-data comparisons and are, hence, plant independent.
The RV internals are modeled in detail (Figure 3-3 through Figure 3-5) based on specific inputs supplied by Constellation Energy. Nodes and connectivity, flow areas, resistances and heat structures are all accurately modeled. The location of the hot assembly/hot pin(s) is unrestricted; however, the channel is always modeled to restrict appreciable upper plenum liquid fallback.
The final step of the best-estimate methodology is to combine all the uncertainties related to the code and plant parameters, and estimate the PCT at a high probability level. The steps taken to derive the PCT uncertainty estimate are summarized below:
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 20
- 1. Base Plant Input File Development First, base RODEX3A and S-RELAP5 input files for the plant (including the containment input file) are developed. Code input development guidelines are applied to ensure that model nodalization is consistent with the model nodalization used in the code validation.
- 2. Sampled Case Development The non-parametric statistical approach requires that many "sampled" cases be created and processed. For every set of input created, each "key LOCA parameter" is randomly sampled over a range established through code uncertainty assessment or expected operating limits (provided by plant Technical Specifications or data). Those parameters considered "key LOCA parameters" are listed in Table 3-1. This list includes both parameters related to LOCA phenomena (based on the PIRT provided in Reference 1) and to plant operating parameters.
- 3. Determination of Adequacy of ECCS The RLBLOCA methodology uses a non-parametric statistical approach to determine values of PCT at the 95% probability level. Total oxidation and total hydrogen are based on the limiting PCT case. The adequacy of the ECCS is demonstrated when these results satisfy the criteria set forth in Section 3.0.
3.3 Plant Descriptionand Summary of Analysis Parameters The plant analysis presented in this report is for a CE-designed PWR, which has 2X4-loop arrangement. There are two hot legs each with a U-tube steam generator and four cold legs each with an RCP 5 . The RCS includes one Pressurizer connected to a hot leg. The core contains 217 thermal-hydraulic compatible AREVA NP HTP 14X14 fuel assemblies with U0 2 rods and Gadolinia-bearing U0 2 rods of 2%, 4%, 6% and 8% Gd 20 3 pins. For both units of CCNPP, the cores contain co-resident Westinghouse and AREVA Advanced CE14 HTP fuel.
The two assembly types have different form loss coefficients for the grid spacers and the upper and lower tie plates. Adjustments were made to these losses in the basedeck to model the mixed core configuration. The break is modeled in the same loop as the pressurizer, as directed by the RLBLOCA methodology. The RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water (i.e., RAS) for ECCS pumped injection need not be considered The S-RELAP5 model explicitly describes the RCS, RV, Pressurizer, and ECCS. The ECCS includes one HPSI, one LPSI and one SIT injection path per RCS loop. The HPSI and LPSI feed into a common header that connects to each cold leg pipe downstream of the RCP s The RCPs are Byron-Jackson Type DFSS pumps as specified by Constellation Energy. The homologous pump performance curves were input to the S-RELAP5 plant model; the built-in S-RELAP5 curves were not used.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. I Page 21 discharge. The ECCS pumped injection is modeled as a table of flow versus backpressure.
This model also describes the secondary-side steam generator that is instantaneously isolated (closed MSIV and feedwater trip) at the time of the break. A symmetric steam generator tube plugging level of 10% per steam generator was assumed.
Plant input modeling parameters were provided by Constellation Energy specifically for the CCNPP. By procedure, Constellation Energy maintains plant documentation current, and directly communicates with AREVA NP on plant design and operational issues regarding reload cores. Constellation Energy and AREVA NP will continue to interact in that fashion regarding the use of AREVA NP fuel in the CCNPP. Both entities have ongoing processes that assure the ranges and values of input parameters for the CCNPP RLBLOCA analysis bound those of the as-operated plant.
As described in the AREVA NP RLBLOCA methodology, many parameters associated with LBLOCA phenomenological uncertainties and plant operation ranges are sampled. A summary of those parameters is given in Table 3-1. The LBLOCA phenomenological uncertainties are provided in Reference 1. Values for process or operational parameters, including ranges of sampled process parameters, and fuel design parameters used in the analysis are given in Table 3-2. Plant data are analyzed to develop uncertainties for the process parameters sampled in the analysis. Table 3-3 presents a summary of the statistical distributions and the uncertainty parameter ranges used in the analysis. The RWST temperature, containment spray and ECCS delay times, and diesel start time are set at conservative bounding values. Where applicable, the sampled parameter ranges are based on Technical Specifications limits or supporting plant calculations that provide more bounding values.
For the AREVA NP RLBLOCA EM, dominant containment parameters, as well as NSSS parameters, were established via a PIRT process. Other model inputs are generally taken as nominal or conservatively biased. The PIRT outcome yielded two important (relative to PCT) containment parameters--containment pressure and temperature. In many instances, the conservative guidance of CSB 6-2 (Reference 6) was used in setting the remainder of the containment model input parameters. As noted in Table 3-3, containment temperature is a sampled parameter. Containment pressure response is indirectly ranged by sampling the containment volume (Table 3-3). In accordance with Reference 1, the condensing heat transfer coefficient is intended to be closer to a best-estimate instead of a bounding high value. A [ ]
Uchida heat transfer coefficient multiplier was specifically validated for use in CCNPP through application of the process used in the RLBLOCA EM (Reference 1) sample problems.
The containment initial conditions and boundary conditions are given in Table 3-8. The building spray is modeled at maximum heat removal capacity. Passive containment heat sink parameters are listed in Table 3-9 and the heat sink material properties are listed in Table 3-10.
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Rev. 1 Page 22 3.4 SER Compliance A number of requirements on the methodology are stipulated in the conclusions section of the SER for the RLBLOCA methodology (Reference 1). Assessment of the 13 analysis-related SER restrictions concludes that all requirements are met for the RLBLOCA analysis for CCNPP for the current cycle and beyond unless future changes mandate further evaluation. A summary discussion of the assessments and conclusions is provided in Table 3-4. The heat transfer parameters for the limiting case are provided in Table 3-7.
3.5 Realistic Large Break LOCA Results Two case sets of 59 transient calculations were performed sampling the parameters listed in Table 3-1 and the results of the limiting PCT case are reported. For each case set, PCT was calculated for a U0 2 rod and for Gadolinia-bearing rods with concentrations of 2%, 4%, 6% and 8% Gd 20 3 . The limiting case set, that contained the PCT, was the set with no offsite power available. The limiting PCT (1620 0 F) occurred in Case 47 for a fresh 8% Gd 20 3 rod 6 . From the same case, the limiting PCT for a once-bumt rod is 1545 0 F and occurred for a U0 2 rod. The fraction of total hydrogen generated was not directly calculated; however, it is conservatively bounded by the calculated total percent oxidation, which is well below the 1% limit. The best-estimate (median) PCT case is Case 1, which corresponded to the median case out of the 59-case set with no offsite power available. The nominal PCT was 14240 F for a 8% once-burnt Gd 2O3 rod. This result can be used to quantify the relative conservatism in the limiting case result. In this analysis, it was 196 0 F.
The case results and event times for the limiting PCT case are shown in Table 3-5 and Table 3-6, respectively. The analysis plots for the limiting PCT case are shown in Figure 3-6 through Figure 3-23. The reference Level Zero for the liquid level in the reactor vessel is the bottom of the downcomer, which corresponds to the bottom of the lower support plate. Figure 3-6 through Figure 3-11 show linear scatter plots of the key parameters sampled for the 59 calculations.
Parameter labels appear to the left of each individual plot. These figures show the parameter ranges used in the analysis. Figure 3-6 shows the scatter plot of operational parameters.
Figure 3-7 and Figure 3-8 show scatter plots for the time of occurrence of PCT and break size versus PCT for the 59 calculations, respectively. Figure 3-9, Figure 3-10 and Figure 3-12 show the maximum oxidation, total oxidation and Containment Volume versus PCT scatter plots for the 59 calculations, respectively. Key parameters for the limiting PCT case are shown in Figure 3-12 through Figure 3-23. Figure 3-12 is the plot of PCT versus time, independent of elevation; this figure clearly indicates that the transient exhibits a sustained and stable quench. A comparison of PCT results from both case sets is shown in Figure 3-24. As seen in Figure 3-24, the peak PCT is from the LOOP case.
6 The PCT for fresh fuel for the U0 2 only rod (Case 7) was 1588 0F.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Paae 23 Table 3-1 Sampled LBLOCA Parameters Phenomenological Time in cycle (peaking factors, axial shape, rod properties, burnup)
Break type (guillotine versus split)
Critical flow discharge coefficients (break)
Decay heat 7 Critical flow discharge coefficients (surgeline)
Initial upper head temperature Film boiling heat transfer Dispersed film boiling heat transfer Critical heat flux Tmin (intersection of film and transition boiling)
Initial stored energy Downcomer hot wall effects Steam generator inlet plenum interfacial effects 8 Condensation interphase heat transfer coefficient8 Metal-water reaction Plant9 Offsite power availability10 Break size Pressurizer pressure Pressurizer liquid level SIT pressure SIT liquid level SIT temperature (based on containment temperature)
Containment temperature Containment volume Initial RCS flow rate Initial operating RCS temperature Diesel start (for loss of offsite power only) 7 Not sampled in analysis, multiplier set to 1.0.
8 Not sampled in analysis.
9 Uncertainties for plant parameters are based on typical plant-specific data.
10 Not sampled, see Section 4.9.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 24 Table 3-2 Plant Operating Range Supported by the LOCA Analysis Event Operating Range 1.0 Plant Physical Description 1.1 Fuel a) Cladding outside diameter 0.440 in.
I b) Cladding inside diameter 0.387 in.
c) Cladding thickness 0.0265 in.
d) Pellet outside diameter 0.3805 in.
e) Pellet density 96 percent of theoretical f) Active fuel length 136.7 in.
g) Resinter densification [ I h) Gd 2 0 3 concentrations 2%, 4%, 6%, 8%
1.2RC a) Flow resistance Analysis b) Pressurizer location Analysis assumes location giving most limiting PCT (broken loop) c) Hot assembly location Anywhere in core d) Hot assembly type 14X14 AREVA NP HTP fuel e) SG tube plugging 10%11 2.0 Plant Initial Operating Conditions 2.1 Reactor Power 12 a) Power Level 2754 MWt b) LHGR 15.0 kW/ft c) Fa 2.37 d) Fr 1.81013 2.2 Fluid Conditions a) Loop flow 370,000 gpm* M <422,250 gpm b) RCS Cold Leg temperature 546. 0F < T < 554.0°F c) Pressurizer pressure 2164 psia : P s 2336 psia d) Pressurizer level 32.2 percent < L* 67.2 percent e) SIT pressure 194.7 psia < P
- 264.7 psia f) SIT liquid volume 1090 ft3 < V < 1179 ft3 0
60°F 6O~_*T*<: 125 20 F g) SIT temperature g (Coupled with containment temperature)
In the RLBLOCA analysis, only the maximum 10% tube plugging in each steam generator was analyzed. By independently sampling the break loss discharge coefficients, any flow differences attributed to asymmetry in the SG tube plugging is covered by use of the RLBLOCA methodology.
12 Includes 17 MWt uncertainties 13 The radial power peaking for the hot rod is including 6% measurement uncertainty and 3.5% allowance for control rod insertion effect.
Friimit = Fr*(l+ uncertFr) * (l+uncert.crinsertion) = 1.65*(1.0+0.06)*(1+0.035)=1.810 AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Paae 25 Event Operating Range h) SIT resistance fL/D As-built piping configuration:
Line 11A: 5.80 Line 11B: 5.72 Line 12A: 5.19 Line 12B: 5.35 i) Minimum ECCS boron Žt 2300 ppm 3.0 Accident Boundary Conditions a) Break location Cold leg pump discharge piping b) Break type Double-ended guillotine or split c) Break size (each side, relative to cold 0.2876 _ A < 1.0 full pipe area (split) leg pipe area of 4.91 ft2) 0.2876*_ A _<1.0 full pipe area (guillotine) d) Worst single-failure Loss of one emergency diesel generator e) Offsite power On or Off f) ECCS pumped injection temperature 100°F 30.0 s (wi/offsite power) g) HPSI pump delay 30.0 s (w/o offsite power) 45.0 s (w/ offsite power) 45.0 s (w/ offsite power) h) LPSI pump delay 45.0 s (w/o off'site power) i) Containment pressure 13.7 psia, nominal value 14 0
j) Containment temperature 60OF _ T _ 125 F k) Containment Spray Delay 20 s I) HPSI flow RCS Cold Leg Broken loop flow Intact loop flow Pressure (psia) (gpm) (gpm) 14.7 164.96 155.5 215 164.96 155.5 615 125.25 116.8 900 87.04 79.5 1015 70.15 63.0 1100 50.87 44.2 1150 35.31 29.1 1180 21.81 15.9 1195 10.66 5.9 1195.1 0 0.0 14 Nominal containment pressure range is -1.0 psi to +1.8 psi. For RLBOCA, a reasonable value in this range is acceptable.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 26 m) LPSI flow RCS Cold Leg Broken loop flow Intact loop flow Pressure (psia) (gPm) (gpm) 14.7 1713.52 1659.62 64.7 1422.6 1377.41 114.7 1029.27 995.9 149.7 604.27 583.83 159.7 393.04 379.15 169.7 206.06 198.2 169.8 0 0 AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 27 Table 3-3 Statistical Distributions Used for Process Parameters1 5 Parameter Operational Uncertainty Parameter Range Parameter __Distribution Pressurizer Pressure (psia) Uniform 2164 - 2336 Pressurizer Liquid Level (percent) Uniform 32.2 - 67.2 SIT Liquid Volume (ft3) Uniform 1090.0 - 1179.0 SIT Pressure (psia) Uniform 194.7 - 264.7 Containment Temperature (OF) Uniform 60 - 125 Containment Volume (ft3) Uniform 1.989E+6 - 2.148E+6 Initial RCS Flow Rate (gpm) Uniform 370,000 - 422,250 Initial RCS Operating Temperature Uniform 546.0 - 554.0 (Tcod) (0F)
RWST Temperature for ECCS (OF) Point 100 Offsite Power Availability 16 Binary 0,1 Delay for Containment Spray (s) Point 20 LPSI Pump Delay (s) Point 30 (w/ offsite power) 30 (w/o offsite power)
Pump Delay (s) Point 45 (w/ offsite power)
IHPSI __ 1 145 (w/o offsite power) 15 Core power is not sampled, see Section 1.0 16 This is no longer a sampled parameter. One set of 59 cases is run with LOOP and one set of 59 cases is run with No-LOOP.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 28 Table 3-4 SER Conditions and Limitations SER Conditions and Limitations Response
- 1. A CCFL violation warning will be added to alert the analyst An occurrence-based scatter plot of downcomer velocities to CCFL violation in the downcomer should such occur. versus the applicable CCFL velocity (as Wallis parameters) was generated for all cases. These results were examined for gross CCFL violation (i.e., more than 50% of the data appearing above the Wallis model flooding line).
There was no significant occurrence of CCFL violation in the downcomer for this evaluation.
- 2. AREVA NP will not include the hot leg nozzle gaps in plant Hot leg nozzle gaps were not modeled.
models to which they apply the S-RELAP5 code.
- 3. If AREVA NP applies the RLBLOCA methodology to plants The PLHGR for CCNPP is lower than that used in the using a higher planar linear heat generation rate (PLHGR) development of the RLBLOCA EM (Reference 1). An end-than used in the current analysis, or if the methodology is to of-life calculation was not performed. However, an be applied to an end-of-life analysis for which the pin assessment for CCNPP against rupture criteria was pressure is significantly higher, then the need for a performed, which concluded that for the CCNPP RLBLOCA blowdown clad rupture model will be reevaluated. The analysis, cladding rupture prior to the initiation of reflood evaluation may be based on relevant engineering does not occur for either the first or second cycle fuel.
experience and should be documented in either the RLBLOCA guideline or plant specific calculation file.
- 4. Slot breaks on the top of the pipe have not been evaluated. The AREVA NP PWR analysis guidelines provide detailed These breaks could cause the loop seals to refill during late discussion on the generic treatment of top slot breaks. For reflood and the core to uncover again. These break CCNPP, the elevation of the top of active fuel is below the locations are an oxidation concern as opposed to a PCT elevation of the top of the crossover leg piping that would be concern since the top of the core can remain uncovered for susceptible to a top slot break. Thus no top slot break LOCA extended periods of time. Should an analysis be performed analysis is needed, and no additions to the calculation for a plant with spillunder (Top crossover pipe (ID) at the notebook or Design Report are required.
crossover pipes lowest elevation) that are below the top elevation of the core, AREVA NP will evaluate the effect of the deep loop seal on the slot breaks. The evaluation may be based on relevant engineering experience and should be documented in either the RLBLOCA guideline or plant-specific calculation file.
- 5. The model applies to 3 and 4 loop Westinghouse- and CCNPP is a CE-designed 2X4 loop plant and the RLBLOCA CE-designed nuclear steam systems. EM applies to this type of plant.
- 6. The model applies to bottom reflood plants only (cold side CCNPP is a bottom reflood plant and the RLBLOCA EM injection into the cold legs at the reactor coolant discharge applies to this type of plant.
piping).
- 7. The model is valid as long as blowdown quench does not The case set was examined and blowdown quench was not occur. If blowdown quench occurs, additional justification an issue in the CCNPP RLBLOCA uncertainty analysis.
for the blowdown heat transfer model and uncertainty are needed or the calculation is corrected. A blowdown quench is characterized by a temperature reduction of the peak cladding temperature (PCT) node to saturation temperature during the blowdown period.
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Rev. 1 Page 29 Table 3-4 SER Conditions and Limitations (continued)
- 8. The reflood model applies to bottom-up quench behavior. If The CCFL model is applied on all core exit junctions as a a top-down quench occurs, the model is to be justified or provision to prevent top-down quench.
corrected to remove top quench. Aovng top-down quench is characterized bychaactrizd te qenc b front the quench frnt moving romthetoptoNo from top-down the top to RLBLOCA quench analysis.
uncertainty effects are observed in the CCNPP the bottom of the hot assembly.
- 9. The model does not determine whether Criterion 5 of 10 Long-term cooling was not evaluated in this analysis.
CFR 50.46, long term cooling, has been satisfied. This will be determined by each applicant or licensee as part of its application of this methodology.
- 10. Specific guidelines must be used to develop the The nodalization in the plant model is similar with the plant-specific nodalization. Deviations from the reference Westinghouse 3-loop sample calculations that were plant must be addressed. submitted to the NRC for review with a slight deviation in the upper plenum nodalization. This deviation should not impact the acceptability of the nodalization.
Figure 3-1 shows the loop noding used in this analysis.
(Note only Loop 1 is shown in the figure; Loop 2 is identical to loop 1, except that only Loop 1 contains the pressurizer and the break.) Figure 3-2 shows the steam generator model. Figures 3-3, 3-4, and 3-5 show the reactor vessel noding diagrams.
- 11. A table that contains the plant-specific parameters and the Simulation of clad temperature response is a function of range of the values considered for the selected parameter phenomenological correlations that have been derived either during the topical report approval process must be analytically or experimentally. The important correlations provided. When plant-specific parameters are outside the have been validated for the RLBLOCA methodology and a range used in demonstrating acceptable code performance, statement of the range of applicability has been the licensee or applicant will submit sensitivity studies to documented. The correlations of interest are the set of heat show the effects of that deviation, transfer correlations as described in Reference 1. Table 3-3 presents the summary of the full range of applicability for the important heat transfer correlations, as well as the ranges calculated in the limiting case of this analysis. Calculated values for other parameters of interest are also provided.
As is evident, the plant-specific parameters fall within the methodology's range of applicability.
- 12. The licensee or applicant using the approved methodology The design report presents the results of the calculations in must submit the results of the plant-specific analyses, accordance with the sample problem.
including the calculated worst break size, PCT, and local and total oxidation.
- 13. The licensee or applicant wishing to apply AREVA NP CCNPP has 14x14 HTP fuel with M5 cladding. Constellation realistic large break loss-of-coolant accident (RLBLOCA) Energy has the permanent exemption approved for M5 methodology to M5 clad fuel must request an exemption for cladding for unrestricted use of AREVA fuel (Reference 7).
its use until the planned rulemaking to modify 10 CFR 50.46(a)(i) to include M5 cladding material has been completed.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 30 Table 3-5 Summary of Results for the Limiting PCT Case Case #47 Fresh Fuel Once-Burnt (Offsite power unavailable) 8% Gd O3 Rod U0 2 Rod PCT Temperature 1620°F 1545°F Time 8.52s 8.36s Elevation 7.859 ft 7.859 ft Metal-Water Reaction Pre-transient Local Oxidation (%) 1.214 1.997 Transient Local Oxidation Maximum (%) 0.543 0.463 Total Local Oxidation Maximum (%) 1.757 2.460 Total Core-Wide Oxidation (%) 0.0111 Table 3-6 Calculated Event Times for the Limiting PCT Case Event Time (s)
Break Opened 0.0 RCP Tnp N/A SIAS Issued 0.6 Start of Broken Loop SIT Injection 13.8 Start of Intact Loop SIT Injection 16.2, 16.1, and 16.1 (Loops 2, 3 and 4 respectively)
Broken Loop LPSI Delivery Began 45.6 Intact Loop LPSI Delivery Began 45.6, N/A and NIIA (Loops 2, 3 and 4 respectively)
Broken Loop HPSI Delivery Began 30.6 Intact Loop HPSI Delivery Began 30.6, 30.6 and 30.6 (Loops 2, 3 and 4 respectively)
Beginning of Core Recovery (Beginning of Reflood) 27.1 Broken Loop SIT Emptied 72.6 Intact Loop SITs Emptied 72.0, 68.2 and 68.5 (Loops 2, 3 and 4 respectively)
PCT Occurred 8.5 Transient Calculation Terminated 340.0 AREVA NP Inc.
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Rev. 1 Paqe 31 Table 3-7 Heat Transfer Parameters for the Limiting Case 17 AREVA NP Inc.
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Rev. 1 Pace 32 Table 3-7 Heat Transfer Parameters for the Limiting Case (continued)
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Rev. 1 Page 33 Table 3-8 Containment Initial and Boundary Conditions Containment Net Free Volume (ft3) 1,989,000- 2,148,000 Initial Conditions Containment Pressure (nominal) 13.7 psia Containment Temperature 60OF - 125 0 F Outside Temperature 100 F Humidity 0.9 Containment Spray Number of Pumps operating 2 Spray Flow Rate (Total, both pumps) 4600 gpm Minimum Spray Temperature 40OF Fastest Post-LOCA initiation of spray 20 sec Initial Time for:
a) Spray Flow (minimum) a) 20 sec b) Fans (minimum) b) 0 sec Containment Emergency Cooling Units Number of Emergency Cooling Units 4 Operating 4 Minimum Post Accident Initiation Time of Emergency Cooling Units 0 (sec)
Emergency Cooling Units Capacity (maximum per fan)
Containment (Vapor) Temperature Heat Removal Rate (OF) (BTU/sec) 60 0.0 120 5844 180 18654 275 49944 AREVA NP Inc.
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Rev. 1 Page 34 Table 3-9 Passive Heat Sinks in Containment Description Slab Material Material Area (ft2)
Thick. (ft)
Paint 2.50E-04 Shell and Dome C Steel 2.08E-02 73230 Concrete 3.OOE+00 Unlined Concrete Concrete 4.OOE+00 53000 Zinc 3.17E-04 Galvanized Steel C. cSteel S.17E-04 8.33E-03 100800 Painted Thin Steel Paint 2.50E-0470250 C. Steel 2.07E-02 Painted Steel Paint 2.50E-04 55000 C. Steel 5.25E-02 Painted Thick Steel Paint 2.50E-04 2966 C. Steel 2.01 E-01 Paint 2.50E-04 Containment Penetration Area C. Steel 6.25E-02 3000 Concrete 3.75E+00 S. Steel 1.56E-02 Stainless Steel Lined Concrete Cocrte Concrete 1.56E+02 4.00E+00 7925 Paint 2.50E-04 Containment Liner Plate Stiffeners C. Steel 6.67E-01 4000 Concrete 2.OOE+00 Base Slab Concrete 8.OOE+00 13300 Sump Strainer I S. Steel 1.31 E-02 308.774 Sump Strainer 2 S. Steel 1.97E-02 161.338 Sump Strainer 3 S. Steel 9.83E-03 3 Sump Strainer 4 S. Steel 4.08E-03 3433.5 Additional H/S 1 C. Steel 1.OOE-02 193.05 Paint 2.50E-04 Additional H/S 2 C.iSt C. Steel 2.0E-04 2.08E-02 42.79 Paint 2.50E-04 Additional HIS 3 Pan .0-456.54 C. Steel 4.17E-02 Improvised H/S S. Steel 8.33E-02 10000 AREVA NP Inc.
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Rev. I Page 35 Table 3-10 Material Properties for Passive Heat Sinks in Containment Thermal Volumetric Heat Heat Sink Conductivity Capacity (BTU/hr-ft-°F) (BTU/ft 3-°F)
Concrete 2.5 35 Carbon Steel 35 55 Stainless Steel 10 62 Zinc 70 45 Paint 1.5 32 AREVA NP Inc.
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Rev. 1 Page 36 Figure 3-1 Primary System Noding for CCNPP AREVA NP Inc.
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Rev. I Page 37 Figure 3-2 Secondary System Noding AREVA NP Inc.
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Rev. 1 Page 38 Figure 3-3 Reactor Vessel Noding AREVA NP Inc.
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Rev. 1 Page 39 Figure 3-4 Core Noding Detail AREVA NP Inc.
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Rev. 1 Paae 40 Figure 3-5 Upper Plenum Noding Detail AREVA NP Inc.
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Rev. 1 Page 41 One-Sided Break Area (ft2/side) m m moe
- m urn oumos j 1.0 2.0 3.0 4.0 5.0 Bum Time 1.0 2000.0 1000.0 45000.0 5.00 (hours) 0.0 5000.0 1000 15000.0 2000 )0.0 Core Power (MW)
I 2 I 2 2 I75. 2 275.2.0 2752.5 2753.0 2753.5 2754.0 2754.5 2755.0 LHGR (KW/ft) 12.0 rIe 13.0 mo 14.0 15.0 ASI mos nui Oum N O 000
-0.2 -0.1 0.0 0.1 0.2 Pressurizer Pressure (psia) 21 50.0 2200.0 2250.0 2300.0 23510.0 Pressurizer Liquid Level
(%)
RCS (Tcold) '
Temperature moo oe o o o e w oeo (OF) 546.0 548.0 550.0 552.0 554.0 Figure 3-6 Scatter Plot of Operational Parameters AREVA NP Inc.
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Rev. 1 Paqe 42 Total Loop Flow em M
- etmmin*ima-(MIb/hr) 13'0.0 135.0 140.0 145.0 150.0 155.0 160.0 SIT Liquid Volume r
- meu meem mm
- sea mimi. imimj 10180.0 1100.0 1120.0 1140.0 1160.0 1180.0 SIT Pressure (psia) 18 0.0 200.0 220.0 240.0 260.0 280 .0 Containment F Volume m mm inin m mm mm (ief) 1.95e+06 2.00e+06 2.05e+06 2.10e+06 2.1 5e+06
.SIT Temperature (6F) 60.0 Figure 3-6 Scatter Plot of Operational Parameters (Continued)
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Rev. 1 Page 43 PCT vs Time of PCT 2000 1800 1600 1400 0
LL.
1200 1000 800 600 400 0 100 200 300 400 500 Time of PCT (s)
Figure 3-7 Scatter Plot of PCT versus PCT Time AREVA NP Inc.
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Rev. 1 Page 44 PCT vs One-sided Break Area 2000 1800 F El 1600 F 1400 F Ell LIE]
U-I-- 1200 F 0
a.
1000 k LD*
800 F 600 F U Split Break El Guillotine Break 400 1.(
0 2.0 3.0 4.0 5.0 Break Area (ft2/side)
Figure 3-8 Scatter Plot of PCT versus Break Size AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. I Paae 45 Maximum Oxidation vs PCT 1.0 0 Split Break 0.9 L1 Guillotine Break 0.8 0.7 0.6 C El o0.5
~0 El 1!!
0.4 0.3 0.2 0.1 0.0 400 600 800 1000 1200 1400 1600 1800 2000 PCT (°F)
Figure 3-9 Scatter Plot of Maximum Transient Oxidation versus PCT AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. I Paae 46 Total Oxidation vs PCT 0.10 N Split Break El Guillotine Break 0.08 0.06 01 0
0 0.04 LI 0.02
,r Li I rn- I-rrn in 0.00 40 )0 600 800 1000 1200 1400 1600 1800 2000 PCT (0F)
Figure 3-10 Scatter Plot of Total Oxidation versus PCT AREVA NP Inc.
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Rev. 1 Page 47 PCT vs Containment Volume 2000 1800 I Li 1600 k EU U EP~ END* am 1400 F Ejl E1N W N 0 F]
I-- 1200 F 0
01.
1000 F 800 I 600 F E Split Break E] Guillotn ra 400 ' '
1.9500e+06 2.0500e+06 2.1500e+06 Containment Volume (ft3)
Figure 3-11 Containment Volume versus PCT Scatter Plot from 59 LOOP Calculations AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 48 PCT Trace for Case #47 PCT = 1619.6 OF, at Time = 8.52 s, on Fresh 8% Gad Rod 2000 1500 U.
a)
Q-4)
E 1000 a) 500 0 L 0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 3-12 Peak Cladding Temperature (Independent of Elevation) for the Limiting Case AREVA NP Inc.
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Rev. I Page 49 Break Flow 90 _ Vessel Side
Pump Side
-- - Total 70 50 0 30 M
10
-10
-30
-50 0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 3-13 Break Flow for the Limiting Case AREVA NP Inc.
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Rev. 1 Paqe 50 Core Inlet Mass Flux 1000 500 a,
LL 0
-500 0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 3-14 Core Inlet Mass Flux for the Limiting Case AREVA NP Inc.
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Rev. 1 Paae 51 Core Outlet Mass Flux 900 700 500 E 300
.S0 X
U, 100
-100
-300
-500 0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 3-15 Core Outlet Mass Flux for the Limiting Case AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Pane 52 Pump Void Fraction 1.0 0.8 0.6 o
03
(.U 0.4 Broken Loop 1 Intact Loop 2 Intact Loop 3 0.2 Intact Loop 4 0.0 0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 3-16 Void Fraction at RCS Pumps for the Limiting Case AREVA NP Inc.
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Rev. 1 Page 53 ECCS Flows 2000
_ Loop 1 (broken)
Loop 2 Loop 3 Loop 4 1500 E
1000 CC 0
UL 500 K
0-0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 3-17 ECCS Flows (Includes SIT, HPSI and LPSI) for the Limiting Case AREVA NP Inc.
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Rev. 1 Page 54 Upper Plenum Pressure 3000 2000
!D
()
1000 A
0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 3-18 Upper Plenum Pressure for the Limiting Case AREVA NP Inc.
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Rev. 1 Paae 55 Downcomer Liquid Level 30 20
-j 1.
10 0
0 100 200 300 400 Time (s) 11:39450 3Sep2O11 07:50:44 R5DMX Figure 3-19 Collapsed Liquid Level in the Downcomer for the Limiting Case AREVA NP Inc.
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Rev. 1 Paae 56 Lower Vessel Liquid Level 14 12 10 S8
-J 6
4 2
0 0 100 200 300 400 Time (s)
ID:39450 3Sep2O11 07:50:44 R5DMX Figure 3-20 Collapsed Liquid Level in the Lower Plenum for the Limiting Case AREVA NP Inc.
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Rev. 1 Page 57 Core Liquid Level 15 10 (D
.5 5
0 0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 3-21 Collapsed Liquid Level In the Core for the Limiting Case AREVA NP Inc.
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Rev. 1 Page 58 Containment and Loop Pressures 100 90 80 70 60 (n
0.
a) 50 ci, Co 0) 0~
40 30 20 10 A
0 0 100 200 300 400 Time (s)
ID:39450 3Sep2O11 07:50:44 R5DMX Figure 3-22 Containment and Loop Pressures for the Limiting Case AREVA NP Inc.
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Rev. 1 Paae 59 300000 250000 E 200000 C')
D 150000 0
t m 100000 50000 0 1 0 100 200 300 400 Time (s) 11:39450 3Sep2011 07:50:44 R5DMX:4 Figure 3-23 Reactor Vessel Liquid Mass for the Limiting Case AREVA NP Inc.
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Rev. 1 Paae 60 2200- 2200 S*LOOP O No LOOP 2000- -2000 1800 - 1800 1600-o :2 o *-- - -- -- -
cI------------------
'k - -- -
a-- - -
1600 F1400 *a a 00* 1400 aa 00 0 0 a0 0 0 a 0 a _ a 1200- - - - - -- - - - - - ---------------------------
--- a -------- 1200 a0
- 1000- V, 0 1000 a 9 a 0
19 9 a
800- -C- *800 600 600 0 10 20 30 40 50 60 Case Number Figure 3-24 GDC 35 LOOP versus No-LOOP Cases AREVA NP Inc.
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Rev. 1 Page 61 4.0 Generic Support for Transition Package The following sections are responses to typical RAI questions posed by the NRC on EMF-2103 Revision 0 plant applications, these responses and changes are known as the "Transition Package." In some instances, these requests cross-reference documentation provided on dockets other than those for which the request is made. AREVA NP discussed these and similar questions from the NRC (draft SER for Revision 1 of EMF-2103) in a meeting with the NRC on December 12, 2007. AREVA NP agreed to provide the following additional information within new submittals of a Realistic Large Break LOCA report.
4.1 Reactor Power Question: Reactor Power - Table 3-2, Item 2.1, and its associatedFootnote 1 indicate that the assumed reactorcore power "includes uncertainties." The use of a reactorpower assumption other than 102 %, regardless of BE or Appendix K methodology, is permitted by Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix K.l.A, "Required and Acceptable Features of The Evaluation Models, 'Sources of Heat During a LOCA." However, Appendix K.I.A also states: "... An assumed power level lower than the level specified in this paragraph
[1.02 times the licensed power level], (but not less than the licensed power level) may be used provided..." Please explain.
Response: As indicated in Item 2.1 of Table 3-2 herein, the assumed reactor core power for the CCNPP Realistic Large Break Loss-of-coolant Accident is 2754 MWt. This value represents the 100% primary power (2737 MWt) plus 17 MVWt (0.62%) measurement uncertainty.
4.2 Rod Quench Question: Does the version of S-RELAP5 used to perform the computer runs assure that the void fraction is less than 95 % and the fuel cladding temperature is less than 900 OF before it allows rod quench?
Response: Yes, the version of S-RELAP5 employed for the CCNPP LAR requires that both the void fraction is less than 0.95 and the clad temperature is less than the minimum temperature for film boiling heat transfer (Tmin) before the rod is allowed to quench. Tain is a sampled parameter in the RLBLOCA methodology. For the CCNPP case set, the mean value of Tmin, is 626 K with standard deviation of 33.6 K, making it very unlikely that Tmin would exceed 755 K (900°F). Therefore, Tmin was never sampled above 696 K (793.70 F). This is a change to the approved RLBLOCA EM (Reference 1).
4.3 Rod-to-Rod Thermal Radiation Question: Providejustification that the S-RELAP5 rod-to-rod thermal radiationmodel applies to the CCNPPUnits I and 2 cores.
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Rev. 1 Page 62 Response: The Realistic LBLOCA methodology, (Reference 1), does not provide modeling of rod-to-rod radiation. The fuel rod surface heat transfer processes included in the solution at high temperatures are: film boiling, convection to steam, rod-to-liquid radiation and rod-to-vapor radiation. This heat transfer package was assessed against various experimental data sets involving both moderate (1600°F - 2000°F) and high (2000°F to over 2200°F) peak cladding temperatures and shown to be conservative when applied nominally. The normal distribution of the experimental data was then determined. During the execution of an RLBLOCA evaluation, the heat transferred from a fuel rod is determined by the application of a multiplier to the nominal heat transfer model. This multiplier is determined by a random sampling of the normal distribution of the experimental data benchmarked. Because the data include the effects of rod-to-rod radiation, it is reasonable to conclude that the modeling implicitly includes an allocation for rod-to-rod radiation effects. As will be demonstrated, the approach is reasonable because the conditions within actual limiting fuel assemblies assure that the actual rod-to-rod radiation is larger than the allocation provided through normalization to the experiments.
The Full-Length Emergency Core Heat Transfer Separate Effects and Systems Effects Tests (FLECHT-SEASET) tests evaluated covered a range of PCTs from 1,651 OF to 2,239°F and the Thermal Hydraulic Test Facility (THTF) tests covered a range of PCTs from 1,000°F to 2,200 0F.
Since the test bundle in either FLECHT-SEASET or THTF is surrounded by a test vessel, which is relatively cool compared to the heater rods, substantial radiation from the periphery rods to the vessel wall can occur. The rods selected for assessing the RLBLOCA reflood heat transfer package were chosen from the interior of the test assemblies to minimize the impact of radiation heat transfer to the test vessel. The result was that the assessment rods comprise a set which is primarily isolated from cold wall effects by being surrounded by powered rods at reasonably high temperatures.
As a final assessment, three benchmarks independent of THTF and FLECHT-SEASET were performed. These benchmarks were selected from the Cylindrical Core Test Facility (CCTF),
Loss of Fluid Test (LOFT), and the Semiscale facilities. Because these facilities are more integral tests and together cover a wide range of scale, they also serve to show that scale effects are accommodated within the code calculations.
The results of these calculations are provided in Section 4.3.4, Evaluation of Code Biases, page 4-100, of Reference 1. The CCTF results are shown in Figures 4.180 through 4.192, the LOFT results in Figures 4.193 through 4.201, and the Semiscale results in Figures 4.202 through 4.207 (Reference 1). As expected, these figures demonstrate that the comparison between the code calculations and data is improved with the application of the derived biases. The CCTF, LOFT, and Semiscale benchmarks further indicate that, whatever consideration of rod-to-rod radiation is implicit in the S-RELAP5 reflood heat transfer modeling, it does not significantly affect code predictions under conditions where radiation is minimized. The measured PCTs in these assessments ranged from approximately 1,000°F to 1,540 0 F. At these temperatures, there is little rod-to-rod radiation. Given the good agreement between the biased code AREVA NP Inc.
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Rev. 1 Page 63 calculations and the CCTF, LOFT, and Semiscale data, it can be concluded that there is no significant over-prediction of the total heat transfer coefficient.
Notwithstanding any conservatism evidenced by experimental benchmarks, the application of the model to commercial nuclear power plants provides some additional margins due to limitations within the experiments. The benchmarked experiments, FLECHT-SEASET and ORNL Thermal Hydraulic Test Facility (THTF), used to assess the S-RELAP5 heat transfer model, Reference 1, incorporated constant rod powers across the experimental assembly.
Temperature differences that occurred were the result of guide tube, shroud or local heat transfer effects. In the operation of a pressurized water reactor (PWR) and in the RLBLOCA evaluation, a radial local peaking factor is present, creating power differences that tend to enhance the temperature differences between rods. In turn, these temperature differences lead to increases in net radiation heat transfer from the hotter rods. The expected rod-to-rod radiation will likely exceed that embodied within the experimental results.
4.3.1 Assessment of Rod-to-Rod Radiation Implicit in the RLBLOCA Methodology As discussed above, the FLECHT-SEASET and THTF tests were selected to assess and determine the S-RELAP5 code heat transfer bias and uncertainty. A uniform radial power distribution was used in these test bundles. Therefore, the rod-to-rod temperature variation in the rods away from the vessel wall is caused primarily by the variation in the sub-channel fluid conditions. In the real operating fuel bundle, on the other hand, there can be 5- to 10-percent rod-to-rod power variation. In addition, the methodology includes a provision to apply the uncertainty measurement to the hot pin. Table 4-1 provides the hot pin measurement uncertainty and a representative local pin peaking factor for several plants. These factors, however, relate the pin to the assembly average. To more properly assess the conditions under which rod-to-rod radiation heat transfer occurs, a more local peaking assessment is required.
Therefore, the plant rod-to-rod radiation assessments herein set the average pin power for those pins surrounding the hot pin at 96% of that of the peak pin. For pins further removed the average power is set to 94%.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Paae 64 Table 4-1 Typical Measurement Uncertainties and Local Peaking Factors FAH Measurement Local Pin Peaking Plant Uncertainty Factor P ek (percent) 1 4.0 1.068 2 4.0 1.050 3 6.0 1.149 4 4.0 1.113 5 4.25 1.135 6 4.0 1.058 4.3.2 Quantification of the Impact of Thermal Radiation using R2RRAD Code The R2RRAD radiative heat transfer model was developed by Los Alamos National Laboratory (LANL) to be incorporated in the BWR version of the TRAC code. The theoretical basis for this code is given in References 8 and 9 is similar to that developed in the HUXY rod heatup code (Reference 10, Section 2.1.2) used by AREVA NP for BWR LOCA applications. The version of R2RRAD used herein was obtained from the NRC to examine the rod-to-rod radiation characteristics of a 5x5 rod segment of the 161 rod FLECHT-SEASET bundle. The output provided by the R2RRAD code includes an estimate of the net radiation heat transfer from each rod in the defined array. The code allows the input of different temperatures for each rod as well as for a boundary surrounding the pin array. No geometry differences between pin locations are allowed. Even though this limitation affects the view factor calculations for guide tubes, R2RRAD is a reasonable tool to estimate rod-to-rod radiation heat transfer.
The FLECHT-SEASET test series was intended to simulate a 17x17 fuel assembly and there is a close similarity, Table 4-2, between the test bundle and a modem 17x17 assembly.
Table 4-2 FLECHT-SEASET & 17x17 FA Geometry Parameters Design Parameter FLECHT-SEASET 17x17 Fuel Assembly Rod Pitch (in) 0.496 0.496 Fuel Rod Diameter (in) 0.374 0.374 Guide Tube Diameter (in) 0.474 0.482 Five FLECHT-SEASET tests (Reference 11) were selected for evaluation and comparison with expected plant behavior. Table 4-3 characterizes the results of each test. The 5x5 selected rod array comprises the hot rod, 4 guide tubes and 20 near adjacent rods; the simulated hot rod is rod 7J in the tests (Figure 4-1).
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Rev. 1 Page 65 Guide Tube -~000 00000 00 @0 ao~ Hot Rod Adjacent Rods 0000 00000 Figure 4-1 R2RRAD 5x5 Rod Segment Two sets of runs were made simulating each of the five experiments and one set of cases was run to simulate the RLBLOCA evaluation of a limiting fuel assembly in an operating plant. For the simulation of Tests 31805, 31504, 31021, and 30817, the thimble tube (guide tube) temperatures were set to the measured values. For Test 34420, the thimble tube temperature was set equal to the measured vapor temperature. For the first experimental simulation set, the temperature of all 21 rods and the exterior boundary was set to the measured PCT of the simulated test. For the second experimental set, the hot rod temperature was set to the PCT value and the remaining 20 rods and the boundary were set to a temperature 25°F cooler providing a reasonable measure of the variation in surrounding temperatures. To estimate the rod-to-rod radiation in a real fuel assembly at LOCA conditions and compare it to the experimental results, each of the above cases was rerun with the hot rod PCT set to the experimental result and the remaining rods conservatively set to temperatures expected within the bundle. Because peak rod powers frequently occur at fuel assembly corners away from either guide tubes or instrument tubes and for added conservatism, the guide tubes (thimble tubes) were replaced by fuel rods in the input model described above. The surrounding 24 rods were set to a temperature estimated for rods of 4% lower power. The boundary temperature was estimated based an average power 6% below the hot rod power. For both of these, the temperature estimates were achieved using a ratio of pin power to the difference in temperature between the saturation temperature and the PCT.
AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Pace 66 "24 rod. = 0.96 * (PCT - Tt) + Tat and Trurrounding region = 0.94 - (PCT - Tat) + Tsat.
T,,t was taken as 270 0 F.
Figure 4-2 shows the hot rod thermal radiation heat transfer for the two FLECHT-SEASET sets and for the plant set. The figure shows that for PCTs greater than about 1700 0 F, the hot rod thermal radiation in the plant cases exceeds that of the same component within the experiments.
Table 4-3 FLECHT-SEASET Test Parameters 7j 22 PCT PCT HTC at PCT Steam Thimble Rod Test Time 2(Btulhr- Temperature at Temperature at 6-ft (OF) Time (s) ft .OF) 7122 (6-ft) (OF) at 6-ft (OF) 34420 2205 34 10 1850 1850*
31805 2150 110 10 1800 1800 31504 2033 100 10 1750 1750 31021 1684 29 9 1400 1350 30817 1440 70 13 900 750
- set to steam temp 22 Note: "7J"and "71" are the bundle locations for the Heater Rod and the Steam Probe thermocouples, respectively (Reference 11).
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Rev. 1 Paae 67 4.5-4-
32.5-3--2 1-2 0.5-0 1400 1500 1600 1700 1800 1900 2000 2100 2200 2300 2400 PCT (TF)
Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA 4.3.3 Rod-to-Rod Radiation Summary In summary, the conservatism of the heat transfer modeling established by benchmark can be reasonably extended to plant applications, and the plant local peaking provides a physical reason why rod-to-rod radiation should be more substantial within a plant environment than in the test environment. Therefore, the lack of an explicit rod-to-rod radiation model, in the version of S-RELAP5 applied for realistic LOCA calculations, does not invalidate the conclusion that the cladding temperature and local cladding oxidation have been demonstrated to meet the criteria of 10 CFR 50.46 with a high level of probability.
4.4 Film Boiling Heat TransferLimit Question: In the CCNPP calculations, is the Forslund-Rohsenow model contribution to the heat transfer coefficient limited to less than or equal to 15% when the void fraction is greater than or equal to 0.9?
Response: Yes, the version of S-RELAP5 employed for the CCNPP RLBLOCA analysis limits the contribution of the Forslund-Rohsenow model to no more than 15% of the total heat transfer AREVA NP Inc.
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Rev. 1 Page 68 at and above a void fraction of 0.9. Because the limit is applied at a void fraction of 0.9, the contribution of Forslund-Rohsenow within the 0.7 to 0.9 interpolation range is limited to 15% or less. This is a change to the approved RLBLOCA EM (Reference 1).
4.5 DowncomerBoiling Question: If the PCT is greaterthan 1800°F or the containment pressure is less than 30 psia, has the CCNPP downcomer model been rebenchmarked by performing sensitivity studies, assuming adequate downcomer noding in the water volume, vessel wall and other heat structures?
Response: The downcomer model for the CCNPP has been established generically as adequate for the computation of downcomer phenomena including the prediction of potential local boiling effects. The model was benchmarked against the UPTF tests and the LOFT facility in the RLBLOCA methodology, Revision 0 (Reference 1). Further, AREVA NP addressed the effects of boiling in the downcomer in a letter, from James Malay to U.S. NRC, April 4, 2003.
The letter cites the lack of direct experimental evidence but contains sensitivity studies on high and low pressure containments, the impact of additional azimuthal noding within the downcomer, and the influence of flow loss coefficients. Of these, the study on azimuthal noding is most germane to this question; indicating that additional azimuthal nodalization allows higher liquid buildup in portions of the downcomer away from the broken cold leg and increases the liquid driving head. Additionally, AREVA NP has conducted downcomer axial noding and wall heat release studies. Each of these studies supports the Revision 0 methodology and is documented later in this section.
This question is primarily concerned with the phenomena of downcomer boiling and the extension of the Revision 0 methodology and sensitivity studies to plants with low containment pressures and high cladding temperatures. Boiling, wherever it occurs, is a phenomenon that codes like S-RELAP5 have been developed to predict. Downcomer boiling is the result of the release of energy stored in vessel metal mass (Figure 4-3). Within S-RELAP5, downcomer boiling is simulated in the nucleate boiling regime with the Chen correlation. This modeling has been validated through the prediction of several assessments on boiling phenomenon provided in the S-RELAP5 Code Verification and Validation document (Reference 12).
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Rev. 1 Page 69 Figure 4-3 Reactor Vessel Downcomer Boiling Diagram Hot downcomer walls penalize PCT by two mechanisms: by reducing subcooling of coolant entering the core and through the reduction in downcomer hydraulic head which is the driving force for core reflood. Although boiling in the downcomer occurs during blowdown, the biggest potential for impact on clad temperatures is during late reflood following the end of SIT injection.
At this time, there is a large step reduction in coolant flow from the ECC systems. As a result, coolant entering the downcomer may be less subcooled. When the downcomer coolant approaches saturation, boiling on the walls initiates, reducing the downcomer hydraulic static level.
With the reduction of the downcomer level, the core inlet flow rate is reduced which, depending on the existing core inventory, may result in a cladding temperature excursion or a slowing of the core cooldown rate.
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Rev. I Page 70 While downcomer boiling may impact clad temperatures, it is somewhat of a self-limiting process. If cladding temperatures increase, less energy is transferred in the core boiling process and the loop steam flows are reduced. This reduces the required driving head to support continued core reflood and reduces the steam available to heat the ECCS water within the cold legs resulting in greater subcooling of the water entering the downcomer.
The impact of downcomer boiling is primarily dependent on the wall heat release rate and on the ability to slip steam up the downcomer and out of the break. The higher the downcomer wall heat release, the more steam is generated within the downcomer and the larger the impact on core reflooding. Similarly, the quicker the passage of steam up the downcomer, the less resident volume within the downcomer is occupied by steam and the lower the impact on the downcomer average density. Therefore, the ability to properly simulate downcomer boiling depends on both the heat release (boiling) model and on the ability to track steam rising through the downcomer. Consideration of both of these is provided in the following text. The heat release modeling in S-RELAP5 is validated by a sensitivity study on wall mesh point spacing and through benchmarking against a closed form solution. Steam tracking is validated through both an axial and an azimuthal fluid control volume sensitivity study done at low pressures. The results indicate that the modeling accuracy within the RLBLOCA methodology is sufficient to resolve the effects of downcomer boiling and that, to the extent that boiling occurs, the methodology properly resolves the impact on the cladding temperature and cladding oxidation rates.
4.5.1 Wall Heat Release Rate The downcomer wall heat release rate during reflood is conduction limited and depends on the vessel wall mesh spacing used in the S-RELAP5 model. The following two approaches are used to evaluate the adequacy of the downcomer vessel wall mesh spacing used in the S-RELAP5 model.
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Rev. 1 Page 71 4.5.1.1 Exact Solution In this benchmark, the downcomer wall is considered as a semi-infinite plate. Because the benchmark uses a closed form solution to verify the wall mesh spacing used in S-RELAP5, it is assumed that the material has constant thermal properties, is initially at temperature Ti, and, at time zero, has one surface, the surface simulating contact with the downcomer fluid, set to a constant temperature, To, representing the fluid temperature. Section 4.3 of Reference 13 gives the exact solution for the temperature profile as a function of time as (T(x,t) - To) / (T - To) = erf {x / (2o(a t)0 5 )}, (1) where, a is the thermal diffusivity of the material given by a = k/(p Cp),
k = thermal conductivity, p = density, Cp = specific heat, and erf(j is the Gauss error function (given in Table A-1 of Reference 13).
The conditions of the benchmark are Ti = 500OF and To = 3000 F. The mesh spacing in S-RELAP5 is the same as that used for the downcomer vessel wall in the RLBLOCA model.
Figure 4-4 shows the temperature distributions in the metal at 0.0, 100 and 300 seconds as calculated by using Equation 1 and S-RELAP5, respectively. The solutions are identical confirming the adequacy of the mesh spacing used in the downcomer wall.
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Rev. 1 Page 72 550 500 LL 450 I-E 400 E
S350 300 250 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Distance from Inner Wall, feet Figure 4-4 S-RELAP5 versus Closed Form Solution AREVA NP Inc.
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Rev. 1 Page 73 4.5.1.2 Plant Model Sensitivity Study As additional verification, a typical 4-loop plant case was used to evaluate the adequacy of the mesh spacing within the downcomer wall heat structure. Each mesh interval in the base case downcomer vessel wall was divided into two equal intervals. Thus, a new input model was created by increasing the number of mesh intervals from 9 to 18. The following four figures (Figure 4-5 through Figure 4-8) show the total downcomer metal heat release rate, PCT independent of elevation, downcomer liquid level, and the core liquid level, respectively, for the base case and the modified case. These results confirm the conclusion from the exact solution study that the mesh spacing used in the plant model for the downcomer vessel wall is adequate.
30 M0.00 -
[ . ...
~1
-Bass VSL Wall (9-mnhý 24000.00. 18-Mesh SLWaII FA 8M0000_
0Ci
- 4) 12000.00 w1W Time (sec)
Figure 4-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity AREVA NP Inc.
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Rev. 1 Page 74 0
L.
0*. 1200.00 E
4)
Time (sec)
Figure 4-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity AREVA NP Inc.
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Rev. 1 Page 75 30.00 4-a) a)
20.00
.4-4)
-J 0*
-J 400.0 Time (sec)
Figure 4-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity AREVA NP Inc.
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Rev. 1 Page 76 12.00 4) 0)
Time (sec)
Figure 4-8 Core Liquid Level - Wall Mesh Point Sensitivity AREVA NP Inc.
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Rev. 1 Page 77 4.5.2 Downcomer Fluid Distribution To justify the adequacy of the downcomer nodalization in calculating the fluid distribution in the downcomer, two studies varying separately the axial and the azimuthal resolution with which the downcomer is modeled have been conducted.
4.5.2.1 Azimuthal Nodalization In a letter to the NRC dated April 2003 (Reference 1), AREVA NP documented several studies on downcomer boiling. Of significance here is the study on further azimuthal break up of the downcomer noding. The study, based on a 3-loop plant with a containment pressure of approximately 30 psia during reflood, consisted of several calculations examining the affects on clad temperature and other parameters.
The base model, with 6 axial by 3 azimuthal regions, was expanded to 6 axial by 9 azimuthal regions (Figure 4-9). The base calculation simulated the limiting PCT calculation given in the EMF-2103 three-loop sample problem. This case was then repeated with the revised 6 x 9 downcomer noding.
The change resulted in an alteration of the blowdown evolution of the transient with little evidence of any affect during reflood. To isolate any possible reflood impact that might have an influence on downcomer boiling, the case was repeated with a slightly adjusted vessel-side break flow. Again, little evidence of impact on the reflood portion of the transient was observed.
The study concluded that blowdown or near blowdown events could be impacted by refining the azimuthal resolution in the downcomer but that reflood would not be impacted. Although the study was performed for a somewhat elevated system pressure, the flow regimes within the downcomer will not differ for pressures as low as atmospheric. Thus, the azimuthal downcomer modeling employed for the RLBLOCA methodology is reasonably converged in its ability to represent downcomer boiling phenomena.
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Rev. 1 Page 78 Base model
- -1 1 - L (r I' CHO CQHO i 'i a &
a Revised 9 Region Model IAII C-ACH 9 IJI IIZi CH 9 IL CHO Figure 4-9 Azimuthal Noding 4.5.2.2 Axial Nodalization The RLBLOCA methodology divides the downcomer into six nodes axially. In both 3-loop and 4-loop models, the downcomer segment at the active core elevation is represented by two equal length nodes. For most operating plants, the active core length is 12 feet and the downcomer segments at the active core elevation are each 6-feet high. (For a 14 foot core, these nodes would be 7-feet high.) The model for the sensitivity study presented here comprises a 4-loop plant with ice condenser containment and a 12 foot core. For the study, the two nodes spanning the active core height are divided in half, revising the model to include eight axial nodes. Further, the refined noding is located within the potential boiling region of the downcomer where, if there is an axial resolution influence, the sensitivity to that impact would be greatest.
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Rev. 1 Page 79 The results show that the axial noding used in the base methodology is sufficient for plants experiencing the very low system pressure characteristics of ice condenser containments.
Figure 4-10 provides the containment back pressure for the base modeling. Figure 4-11 through Figure 4-14 show the total downcomer metal heat release rate, PCT independent of elevation, downcomer liquid level, and the core liquid level, respectively, for the base case and the modified case.
The results demonstrate that the axial resolution provided in the base case, 6 axial downcomer node divisions with 2 divisions spanning the core active region, are sufficient to accurately resolve void distributions within the downcomer. Thus, this modeling is sufficient for the prediction of downcomer driving head and the resolution of downcomer boiling effects.
40,00-_
--- Base 6 32.00 24.00 -
0.
L.
00 00 w 16.00 I -.----- ~--.-----.-----.------ I -------------------
0~
8.00 .4- -~
0.001 0.0 80.0 180.0 240.0 320.0 400.0 Time (sec)
Figure 4-10 Lower Compartment Pressure versus Time AREVA NP Inc.
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0) 0)
w 0.0- °- 80.0---.o___ ,=-
145 "0.0 .o £o~
3200 ,___
40 Time (sec)
Figure 4-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study AREVA NP Inc.
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0)
I--
4000 Time (sec)
Figure 4-12 PCT Independent of Elevation - Axial Noding Sensitivity Study AREVA NP Inc.
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Rev. 1 Page 82 30.00 20.00 400.0 Time (sec)
Figure 4-13 Downcomer Liquid Level - Axial Noding Sensitivity Study AREVA NP Inc.
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0) 400.0 Time (sec)
Figure 4-14 Core Liquid Level - Axial Noding Sensitivity Study 4.5.3 Downcomer Boiling Conclusions To further justify the ability of the RLBLOCA methodology to predict the potential for and impact of downcomer boiling, studies were performed on the downcomer wall heat release modeling within the methodology and on the ability of S-RELAP5 to predict the migration of steam through the downcomer. Both azimuthal and axial noding sensitivity studies were performed. The axial noding study was based on an ice condenser plant that is near atmospheric pressure during reflood. These studies demonstrate that S-RELAP5 delivers energy to the downcomer liquid volumes at an appropriate rate and that the downcomer noding detail is sufficient to track the distribution of any steam formed. Thus, the required methodology for the prediction of downcomer boiling at system pressures approximating those achieved in plants with pressures as low as ice condenser containments has been demonstrated.
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Rev. I Page 84 4.6 Break Size Question: Were all break sizes assumed greaterthan or equal to 1.0 if2?
Response: Yes.
The NRC has requested that the break spectrum for the realistic LOCA evaluations be limited to accidents that evolve through a range of phenomena similar to those encountered for the larger break area accidents. This is a change to the approved RLBLOCA EM (Reference 1). The larger break area LOCAs are typically characterized by the occurrence of dispersed flow film boiling at the hot spot, which sets them apart from smaller break LOCAs. This occurs generally in the vicinity of 0.2 DEGB (double-ended guillotine break) size (i.e., 0.2 times the total flow area of the pipe on both sides of the break). However, this transitional break size varies from plant to plant and is verified only after the break spectrum has been executed. AREVA NP has sought to develop sufficient criteria for defining the minimum large break flow area prior to performing the break spectrum. The purpose for doing so is to assure a valid break spectrum is performed.
4.6.1 Break I Transient Phenomena In determining the AREVA NP criteria, the characteristics of larger break area LOCAs are examined. These LOCA characteristics involve a rapid and chaotic depressurization of the reactor coolant system (RCS) during which the three historical approximate states of the system can be identified.
Blowdown The blowdown phase is defined as the time period from initiation of the break until flow from the SIT begins. This definition is somewhat different from the traditional definition of blowdown, which extends the blowdown until the RCS pressure approaches containment pressure. The blowdown phase typically lasts about 12- to 25-seconds, depending on the break size.
Refill is that period that starts with the end of blowdown, whichever definition is used, and ends when water is first forced upward into the core. During this phase the core experiences a near adiabatic heatup.
Reflood is that portion of the transient that starts with the end of refill, follows through the refilling of the core with water and ends with the achievement of complete core quench.
Implicit in this break-down is that the core liquid inventory has been completely, or nearly so, expelled from the primary system leaving the core in a state of near core-wide dispersed flow film boiling and subsequent adiabatic heatup prior to the reflood phase. Although this break down served as the basis for the original deterministic LOCA evaluation approaches and is valid for most LOCAs that would classically be termed large breaks, as the break area decreases the depressurization rate decreases such that these three phases overlap substantially. During these smaller break events, the core liquid inventory is not reduced as much as that found in AREVA NP Inc.
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Rev. 1 Page 85 larger breaks. Also, the adiabatic core heatup is not as extensive as in the larger breaks which results in much lower cladding temperature excursions.
4.6.2 New Minimum Break Size Determination No determination of the lower limit can be exact. The values of critical phenomena that control the evolution of a LOCA transient will overlap and interplay. This is especially true in a statistical evaluation where parameter values are varied randomly with a strong expectation that the variations will affect results. In selecting the lower area of the RLBLOCA break spectrum, AREVA NP sought to preserve the generality of a complete or nearly complete core dry out accompanied by a substantially reduced lower plenum liquid inventory. It was reasoned that such conditions would be unlikely if the break flow rate was reduced to less than the reactor coolant pump flow. That is, if the reactor coolant pumps are capable of forcing more coolant toward the reactor vessel than the break can extract from the reactor vessel, the downcomer and core must maintain some degree of positive flow (positive in the normal operations sense).
The circumstance is, of course, transitory. Break flow is altered as the RCS blows down and the RC pump flow may decrease as the rotor and flywheel slow down if power is lost. However, if the core flow was reduced to zero or became negative immediately after the break initiation, then the event was quite likely to proceed with sufficient inertia to expel most of the reactor vessel liquid to the break. The criteria base, thus established, consists of comparing the break flow to the initial flow through all reactor coolant pumps and setting the minimum break area such that these flows match. This is done as follows:
Wbreak = Abreak
- Gbreak = Npump
- WRCP.
This gives Abreak =- (Npump
- WRcP)/Gbreak.
The break mass flux is determined from critical flow. Because the RCS pressure in the broken cold leg will decrease rapidly during the first few seconds of the transient, the critical mass flux is averaged between that appropriate for the initial operating conditions and that appropriate for the initial cold leg enthalpy and the saturation pressure of coolant at that enthalpy.
Gbreak = (Gbreak(Po, HCLO) + Gbreak(PCLsat, HCLO))I2 .
The estimated minimum LBLOCA break area, Amin, is 2.824 ft2 and the break area percentage, based on the full double-ended guillotine break total area, is 28.76 percent.
Table 4-4 provides a listing of the plant type, initial condition, and the fractional minimum RLBLOCA break area, for all the plant types presented as generic representations in the next section.
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Rev. 1 Page 86 Table 4-4 Minimum Break Area for Large Break LOCA Spectrum Saturated Spectrum Spectrum Plant System Cold Leg Subcoole Gbk RCP Minimum Minimum Pressure Enthalpy d Gbmk flow Break Break Depcrpaio (psa) (Btulbm) (bft-%aBuim(ibmft (Ibmt (HEM)2 s) (Ibmis) Area Area
________s _(ft 2
) (DEGB)
A 3-Loop W 2250 554.0 22198 6330 31558 2.21 0.27 Design B 3-Loop W 2250 544.5 23880 5450 28124 1.92 0.23 Design C Designn 3-Loop 2250 550.0 23540 5580 29743 2.04 0.25 D 9x4CE 2100 538.8 22860 5310 21522 1.53 0.24 Design E 2 CE 2060 531.0 22068 5694 38277 2.76 0.28 Design F 2 CE 2250 544 22930 5834 41230 2.87 0.29 Design G 2 CE 2250 548 22637 6091 42847 2.98 0.30 Design I I_ I H -iLoop n DesignIIII 2160 540.9 23290 5370 39500 2.76 0.33 The split versus double-ended guillotine break type is no longer related to break area. In concurrence with Regulatory Guide 1.157, both the split and the double-ended guillotine break will range in area between the minimum break area (Amin) and an area of twice the size of the broken pipe. The determination of break configuration, split versus double-ended guillotine, is made after the break area is selected based on a uniform probability for each occurrence.
4.6.3 Intermediate Break Size Disposition With the revision of the smaller break area for the RLBLOCA analysis, the break range for small breaks and large breaks are no longer contiguous. Typically the lower end of the large break spectrum occurs at between 0.2 to 0.3 times the total area of a 100% double-ended guillotine break (DEGB) and the upper end of the small break spectrum occurs at approximately 0.05 times the area of a 100% DEGB. This leaves a range of breaks that are not specifically analyzed during a LOCA licensing analysis. The premise for allowing this gap is that these breaks do not comprise accidents that develop high cladding temperature and thus do not comprise accidents that critically challenge the emergency core cooling systems (ECCS).
Breaks within this range remain large enough to blowdown to low pressures. Resolution is AREVA NP Inc.
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Rev. 1 Page 87 provided by the large break ECC systems and the pressure-dependent injection limitations that determine critical small break performance are avoided.
A variety of plant types for which analysis within the intermediate range have been completed were surveyed. Although statistical determinations are extracted from the consideration of breaks with areas above the intermediate range, the AREVA NP best-estimate methodology remains suitable to characterize the ECCS performance of breaks within the intermediate range.
Table 4-4 provides a listing of the plant type, initial condition, and the fractional minimum RLBLOCA break area. Figure 4-15 through Figure 4-20 provide the enlarged break spectrum results with the upper end of the small break spectrum and the lower end of the large break spectrum indicated by bars.
Table 4-5 provides differences between the true large break region and the intermediate break region (break areas between that of the largest SBLOCA and the smallest RLBLOCA). The minimum difference is 222 0 F; however, this case is not representative of the general trend shown by the other comparisons. Considering this point as an outlier, the table shows the minimum difference between the highest intermediate break spectrum PCT and large break spectrum PCT, for the eight plants, as at least 4630 F, and including this point would provide an average difference of 640°F for the CE 2x4 design plants and a maximum difference of 840°F for the 4-loop V plant design.
Thus, by both measures, the peak cladding temperatures within the intermediate break range will be several hundred degrees below those in the true large break range. Therefore, these breaks will not provide a limit or a critical measure of the ECCS performance. Given that the large break spectrum bounds the intermediate spectrum, the use of only the large break spectrum meets the requirements of 10CFR50.46 for breaks within the intermediate break LOCA spectrum, and the method demonstrates that the ECCS for a plant meets the criteria of 10CFR50.46 with high probability.
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Rev. 1 Page 88 Table 4-5 Minimum PCT Temperature Difference - True Large and Intermediate Breaks Generic Maximum Maximum Plant Plant PCT (OF) PCT (OF) Delta PCT Average Delta Description Label Intermediate Large Size (OF) PCT (OF)
(Table 4-4) Size Break Break A 1206 193023 724 3-Loop W B 1273 1951 678 622 Design C 1326 1789 463 D 984 1751 767 E 1049 1740 691 2x4 CE 640 Design F 791 1670 879 G 1464 1686 22224 4-Loop W H 1127 1967 840 840 Design 23 The analysis for this W 3-Loop plant was performed with the Transition methodology and no break sizes fell into the intermediate break range. The PCT value of 1206°F is the closest point to the maximum end of the intermediate break spectrum.
24 The analysis for this 2x4 CE plant was performed with the Transition methodology and no break sizes fell into the intermediate break range. The PCT value of 1464°F is the closest point to the maximum end of the intermediate break spectrum. From the trends of the other 2x4 CE analyses, breaks falling within the intermediate break spectrum would be significantly lower.
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Rev. 1 Paae 89 2000 -
Upper End of Large Break SBLOCA Spectrum Break Size Minimum Spectrum 1800 + Break Area 1600 -
S*
1400 +
a-0J 1200 + 4 4
1000 +/-
800 -
600 1 0.00 00 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1,0000 Break Area Normalized to Double Ended Guillotine Figure 4-15 Plant A- Westinghouse 3-Loop Design AREVA NP Inc.
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Rev. 1 Paae 90 Paae 90 2000 Upper End of Large Break SBLOCA Spectrum Break Size Minimum 1800 + Spectrum
- Break Area 1600-
- ~ **~
1400 -
1200-1000-800 -
6004-0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine Figure 4-16 Plant B - Westinghouse 3-Loop Design AREVA NP Inc.
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Rev. 1 Paae 91 2000 Upper End of Large Break SBLOCA Spectrum Break Size Minimum
- - Spectrum 1800 + Break Area 1600 1400
- ~
- 1-1200 1000 -
800 +
'nnqt a 000 0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine Figure 4-17 Plant C - Westinghouse 3-Loop Design AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. I Paae 92 2000 -
Upper End of Large Break SBLOCA Spectrum Break Size k' Minimum 1800 + --Spectrum Break Area 1600-
- 1 * *.*
1400 * #
- 0 1200 1000 800 +
P*
600 1__ I I-0.00OO00 0.1000 0.2000 0.3000 0.4000 0,5000 0.6000 0.7000 0).8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine Figure 4-18 Plant D - Combustion Engineering 2x4 Design AREVA NP Inc.
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Rev. 1 Paae 93 2000 T Upper End of Large Break l SBLOCA Spectrum Break Size - Minimum 1800 - Spectrum Break Area 1600 +
+
1400 +
- 't . ~.
A.-
1200 +
1000 +
.4 800 -
finn 4__ 4 0.2000 I 0.3000 0.4000 0.5000 0.6000 0.7000 0 0.0000 0. 1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0 .8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine Figure 4-19 Plant E - Combustion Engineering 2x4 Design AREVA NP Inc.
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Rev. 1 Pace 94 2200.0000 Large Break Spectrum Upper End of Minimum 2000.0000 - -
SBLOCA k" Break Area Break Size Spectrum 1800.0000 4
1600.0000 - -
4..ee
[,= 1400.0000 I
1200.0000 --
4 1000.0000 4 4
800.0000 -- 4 4
600.0000 -*1 , -.
0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine Figure 4-20 Plant H - Westinghouse 4-Loop Design AREVA NP Inc.
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Rev. I Page 95 4.7 Detailedinformation for ContainmentModel (ICECON)
Question: Verify that the ICECON model is that shown in Figure 5.1 of EMF-CC-39(P)
Revision 2, "ICECON:A Computer Program Used to Calculate Containment Back Pressure for L OCA Analysis (Including Ice CondenserPlants)."
The AREVA NP RLBLOCA Report shows that the containment parameterstreated statistically are: (1) upper compartment containment volume, (2) upper compartment containment temperature,and (3) lower compartment containment temperature. ANP-2903(P)states that "in many instances" the guidance of NRC Branch Technical Position CSB 6-2 was used in determiningthe other containmentparameters.
- a. How is the mixing of containment steam and ice melt modeled so as to minimize the containmentpressure?
- b. Verify that all containment spray and fan coolers are assumed operating at maximum heat removal capacity.
- c. Describe how the limits on the volume of the upper containmentwere determined.
d.How are the containment air return fans modeled and what is the effect of this modeling on the containmentpressure?
- e. Describe how passive heat sink areas and heat capacities are modeled so as to minimize containment pressure.
ResDonse: See Section 3.3 for discussion of questions (b), (c) and (e). Questions (a) and (d) are specific to ice condenser plants. Containment initial conditions and cooling system information are provided in Table 3-8 and Heat Sinks are provided in Table 3-9. For CCNPP, the scatter plots of PCT versus the sampled containment volumes and initial atmospheric temperature are shown in Figure 4-21 and Figure 4-22. Containment pressure as a function of time for limiting case is shown in Figure 4-23.
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Rev. 1 Page 96 PCT vs Containment Volume 2000 1800 LI 1600 0 E ENDI 11P LI WE D.
1400 *DD 0 El 0I 0
0I LI 800 600
- Split Break DI Guillotine Break 400 1 ,
1.9500e+06 2.0500e+06 2.1500e+06 Containment Volume (ft3)
Figure 4-21 PCT vs. Containment Volume AREVA NP Inc.
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Rev. I PaQe 97 2000 1800 1600 m
Ell UM E U
& FI U
1400 F-1 El 4 00 E El 0M
[]
U-1200 U
Lb 1000 El El El El 800 600 MSplit Break I LI Guillotine Breakj 400 60 70 80 90 100 110 120 130 140 SIT Temperature (OF)
Figure 4-22 PCT vs. Initial Containment Temperature AREVA NP Inc.
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Rev. 1 Paae 98 Containment Pressure 100 90 80 70 60 Co 0.
0 50 Co Co a) ci.
40 30 20 10 0
0 100 200 300 400 Time (s)
ID:39450 3Sep2011 07:50:44 R5DMX Figure 4-23 Containment Pressure for Limiting Case AREVA NP Inc.
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Rev. 1 Page 99 4.8 Cross-References to North Anna Question: In order to conduct its review of the CCNPP application of AREVA NP's realistic LBLOCA methods in an efficient manner, the NRC staff would like to make reference to the responses to NRC staff requests for additional information that were developed for the application of the AREVA NP methods to the North Anna Power Station, Units 1 and 2, and found acceptable during that review. The NRC Staff safety evaluation was issued on April 1, 2004 (Agency-wide Documentation and Management System (ADAMS) accession number ML040960040). The staff would like to make use of the information that was provided by the North Anna licensee that is not applicable only to North Anna or only to subatmospheric containments. This information is contained in letters to the NRC from the North Anna licensee dated September 26, 2003 (ADAMS accession number ML032790396) and November 10, 2003 (ADAMS accession number ML033240451). The specific responses that the staff would like to reference are:
September 26, 2003 letter: NRC Question 1 NRC Question 2 NRC Question 4 NRC Question 6 November 10, 2003 letter: NRC Question I Please verify that the information in these letters is applicable to the AREVA NP model applied to CCNPPexcept for that information related specifically to North Anna and to sub-atmospheric containments.
Response: The responses provided to questions 1, 2, 4, and 6 are generic and related to the ability of ICECON to calculate containment pressures. They are applicable to the CCNPP RLBLOCA submittal.
Question 1 - Completely Applicable Question 2 - Completely Applicable Question 4 - Completely Applicable (the reference to CSB 6-1 should now be to CSB Technical Position 6-2). The NRC altered the identification of this branch technical position in Revision 3 of NUREG-0800.
Question 6 - Completely applicable.
The supplemental request and response are applicable to CCNPP.
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Rev. 1 Page 100 4.9 GDC 35- LOOP and No-LOOP Case Sets Question: IOCFR50, Appendix A, GDC [General Design Criterion] 35 [Emergency core cooling] states that, "Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite electric power is not available) and for offsite electric power operation (assuming onsite power is not available) the system function can be accomplished,assuming a single failure."
The Staff interpretation is that two cases (loss of offsite power with onsite power available, and loss of onsite power with offsite power available) must be run independently to satisfy GDC 35.
Each of these cases is separate from the other in that each case is represented by a different statistical response spectrum. To accomplish the task of identifying the worst case would require more runs. However, for LBLOCA analyses (only), the high likelihood of loss of onsite power being the most limiting is so small that only loss of offsite power cases need be run. (This is unless a particularplant design, e.g., CE [Combustion Engineering] plant design, is also vulnerable to a loss of onsite power, in which situation the NRC may require that both cases be analyzed separately. This would require more case runs to satisfy the statistical requirement than forjust loss of offsite power.)
What is your basis for assuming a 50% probabilityof loss of offsite power? Your statisticalruns need to assume that offsite power is lost (in an independent set of runs). If, as stated above, it has been determined that Palisades,being of CE design, is also vulnerable to a loss of onsite power, this also should be addressed(with an independent set of runs).
ResDonse: In concurrence with the NRC's interpretation of GDC 35, a set of 59 cases each was run with a LOOP and No-LOOP assumption. The set of 59 cases that predicted the highest figure of merit, PCT, is reported in Section 2 and Section 3, herein. The results from both case sets are shown in Figure 3-24. This is a change to the approved RLBLOCA EM (Reference 1).
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Rev. 1 Page 101 4.10 Statement Question: Provide a statement confirming that Constellation Energy and its LBLOCA analyses vendor have ongoing processes that assure that the input variables and ranges of parameters for the CCNPP LBLOCA analyses conservatively bound the values and ranges of those parameters for the as operated CCNPP Units I and 2. This statement addresses certain programmaticrequirements of 10 CFR 50.46, Section (c).
Response: Constellation Energy and the LBLOCA Analysis Vendor have an ongoing process to ensure that all input variables and parameter ranges for the CCNPP realistic large break loss-of-coolant accident are verified as conservative with respect to plant operating and design conditions. In accordance with Constellation Energy Quality Assurance program requirements, this process involves
- 1. Definition of the required input variables and parameter ranges by the Analysis Vendor.
- 2. Compilation of the specific values from existing plant design input and output documents by Constellation Energy and Vendor personnel in a formal analysis input summary document issued by the Analysis Vendor and
- 3. Formal review and approval of the input document by Constellation Energy. Formal Constellation Energy approval of the input document serves as the release for the Vendor to perform the analysis.
Continuing review of the input document is performed by Constellation Energy as part of the plant design change process and cycle-specific core design process. Changes to the input summary required to support plant modifications or cycle-specific core alternations are formally communicated to the Analysis Vendor by Constellation Energy. Revisions and updates to the analysis parameters are documented and approved in accordance with the process described above for the initial analysis.
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Rev. 1 Page 102 5.0 Conclusions A RLBLOCA analysis was performed for the CCNPP using NRC - approved AREVA NP RLBLOCA methods (Reference 1). Analysis results show that the limiting case has a PCT of 1620°F for a fresh fuel 8% Gd 20 3 rod, and a maximum oxidation thickness and hydrogen generation that fall well within regulatory requirements.
The analysis supports operation at a nominal power level of 2754 MWt (including 17 MWt uncertainty), a steam generator tube plugging level of up to 10% in each steam generator, a total LHGR of 15.0 kW/ft, which is equivalent to a total peaking factor (FQ) up to a value of 2.37, and a radial peaking factor (Fr) up to a value of 1.810 (including 6% uncertainty and 3.5%
control rod insertion effect) with no axial or burnup dependent power peaking limit. The twice-burnt fuel assemblies are not considered in the analyses since burnups at this level or higher do not retain sufficient energy potential to achieve significant cladding temperatures or cladding oxidations during the transient. For large break LOCA, the four 10CFR50.46 (b) criteria presented in Section 3.0 are met and operation of CCNPP with AREVA NP-supplied 14x14 M5 clad fuel is justified.
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Rev. 1 Page 103 6.0 Recent NRC Request for Additional Information (RAI) and AREVA NP Responses The NRC staff has found that strict adherence to currently referenced, or proposed for referencing AREVA methodologies are inconsistent with the NRC's requirements and review guidance without appropriate justification. This section addresses the NRC staff's concerns for the AREVA RLBLOCA methodology.
6.1 Thermal Conductivity Degradation- Once-Burnt Fuel Question:
Please provide more information about the management of the fuel thermal conductivity degradation issue identified in NRC Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation." Specifically:
- a. "Foreach specific time in cycle, the fuel conditions are computed using RODEX3A prior to starting the S-RELAP5 portion of the analysis. A steady state condition for the given time in cycle using S-RELAP5 is established. A base fuel centerline temperature is establishedin this process. Then two-transformationadjustment to the base fuel centerline temperature is computed. The first transformation is a linear adjustment for an exposure of 10 MWdIMTU or higher. In the new process, a polynomial transformationis used in the first transformation insteadof a linear transformation." Pleaseclarify the following:
- i. Explain how the fuel pellet radialtemperature profile is computed.
ii. Explain which code is used to calculate this profile, both for initialconditions and through the postulated accident iii. Explain whether the polynomial transformationis applied merely to the centerline temperature,or to the entire pellet temperature
- b. Provide additionalinformation to describe the polynomial transformation. Summarize data used to develop the polynomial transformationand discuss considerationof applicable uncertainties.
Response
The NRC concern covers a wide range of specific items but can be paraphrased as: "How does the AREVA NP RLBLOCA analysis for CCNPP provide a licensing basis for fuel throughout its operational life with particular attention to the phenomena of thermal conductivity degradation with burnup?" In response, the following explanation of the methodology employed for CCNPP is provided and followed by specific responses to each of the particular questions.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 104 The AREVA transition package has been updated to specifically model both first and second cycle fuel rods. Third cycle fuel does not retain sufficient energy potential to achieve significant cladding temperatures nor cladding oxidation and is not included in the RLBLOCA individual pin calculations. The burnup for the individual first and second cycle rods analyzed is assigned according to the sampled time in cycle. The time in cycle is sampled once and is the same for both the fresh (first cycle) and once-burnt (second cycle) fuel. Bumup for the fresh and once-burnt rods is different in accordance with the cycle management. Likewise, pin pressure and thermal conductivity differ.
In addition to the thermal conductivity and fuel temperature adjustments for burnup, a burnup dependent reduction in allowed peaking is needed for the once-burnt fuel. For first cycle fuel, the RLBLOCA methodology increases the Fr to the Technical Specifications maximum (including uncertainty) for the first cycle hot rods in the model. Shortly into the cycle, once-burnt fuel has insufficient energy potential to achieve this peaking. A burnup dependent reduction in allowed peaking is therefore applied through an adjustment in the second cycle Fr. The modeling of the burnup dependent reduction in peaking is applied through an adjustment to the Fr based on the power ratio shown in Figure 6-1.
a.i The RODEX3 topical report, ANF-90-145(P)(A), Appendix B (Reference 14) details the calculation of the radial temperature distribution.
a.ii A portion of the RODEX3A fuel model was incorporated into the S-RELAP5 code to calculate fuel response for transient analyses. This coding, referred to as the S-RELAP5/RODEX3A model, deals only with transient predictions and does not calculate the bumup response of the fuel. Instead, fuel conditions at the burnup of interest are transferred via a binary data file from RODEX3A to S-RELAP5/RODEX3A, establishing the initial state of the fuel prior to the transient. The data transferred from RODEX3A describes the fuel at zero power.
A steady-state S-RELAP5/RODEX3A calculation is required to establish the fuel state at power. The transient fuel pellet radial temperature profile is computed by solving the conduction equation of S-RELAP5. Material properties are calculated in S-RELAP5/RODEX3A.
a.iii The adjustment is applied to the entire fuel pellet. The polynomial transformation provides a bias adjustment to the fuel centerline temperature. A sampled parameter provides a random assessment and adjustment of the centerline temperature uncertainty. These are combined and the total adjustment is achieved by iterating a multiplicative adjustment to the fuel thermal conductivity until the desired fuel centerline temperature is reached.
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A AR EVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 105 Paraphrased concern: Provide information on the treatment of thermal conductivity degradation.
Thermal conductivity degradation impacts the ability to transfer energy from within the pellet to the pellet surface and consequently through the cladding to the coolant. Both the initial pellet temperature and the transient release of energy from the pellet are affected. The impact of thermal conductivity changes with burnup are treated by applying a bias. This bias and a measure of the uncertainty in the data were determined by benchmarking the fuel performance code, RODEX3A, to a set of data that extends past the licensed burnup. The bias adjusts the initial fuel temperature to the mean of the benchmark results. The sampled uncertainty is used to provide for the variance of the benchmarks.
The database for the benchmarks is that used to qualify and approve the RODEX4 code (Reference 15). The data from three experimental rods (cases 432R2, 432R6, and 597R8) were not used in the benchmarks. Test 597R8 was not appropriate for this application. Cases 432R2 and 432R6 are rod studies that are not configured appropriately for these types of comparisons. Essentially, these fuel rods are not representative of commercial PWR fuel. Part of the benchmark activity was to incorporate a fractional representation of difference between the RODEX3A calculated results and the data. The fractional adjustment provides a better adjustment over a range of initial temperatures. Therefore, for each benchmark case the Tfr.cbon was determined.
T,.,3 - T ,,
Tfraction -data Trodex3A where:
Tfracbon = Delta fractional temperature of computed to data (K),
Trodex3A = Temperature computed by RODEX3A (K) and Tda=t = Temperature from the RODEX4 database (K).
Figure 6-5 shows the RODEX3A benchmark results along with a polynomial fitted to the results using the least squares method. The negative of this polynomial is the bias which is added to RODEX3A predictions to achieve agreement with the data. Figure 6-6 shows the results of applying this bias in comparison to the results of applying the original RLBLOCA methodology Revision 0 bias. It is evident from Figure 6-6 that the bias makes the adjustment for burnup effects in accordance with the data.
The application of the bias within the methodology proceeds as follows: The burnup for the case of hot rods, fresh and once burned, is determined by sampling the time in cycle and a RODEX3A calculation of the initial fuel centerline temperature performed.
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Rev. 1 Page 106 From the fit in Figure 6-5, an adjusted temperature is determined as per the following equation:
EI where:
Tn,, = Adjusted fuel centerline temperature (K),
B = Burnup (GWd/MTU or MWd/KgU) and Tongi,,ai = Unadjusted RODEX3A fuel centerline temperature (K).
Figure 6-7 provides the bias adjustment-, Toriginat as a function of burnup, using the above polynomial curve fit.
The uncertainty is determined from a Gaussian distribution characterized by a [ ]
standard deviation and added to Tew. The fuel temperature calculation is then repeated with a multiplier, fuel K, on the code calculated fuel thermal conductivity. The fuel centerline temperature is compared to 'T,,w + uncertainty' and the calculation is repeated with an adjusted fuel K as necessary. The process is continued until the calculated centerline fuel temperature matches 'Tnew + uncertainty'. Since the process applies an adjustment to the fuel thermal conductivity, the temperature throughout the pellet is adjusted appropriately. The final multiplier is applied to the thermal conductivity throughout the transient.
Because the data fitting covers the complete range of applicable bumup it is applied as such and the zero bias offset used in Revision 0 for the first 10 GWd/MTU burnup is eliminated.
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Rev. 1 Page 107 Follow-on Question:
Question:
The issue described in IN 2009-23 invalidates AREVA's generic disposition for analyzing fresh fuel only, which is based on sensitivity studies indicatingthat mid-second-cycle fuel had a PCT of 80°F lower than the limiting PCT. This work needs to be repeated accountingfor fuel thermal conductivity degradation. Please provide several cases run at various times-in-life for once-burnt fuel, with information similarto the above list provided; bumup for the limiting rod is only necessaryfor the most limiting second-cycle case analyzed.
Forthe PCT-limitingRLBLOCA case, please provide:
- a. Correctedand uncorrectedradialtemperature profile of the hot rod at the time and location of peak cladding temperature.
- b. Temperature vs. time for the limiting PCT case at the limiting location, including the fuel centerline, fuel average,and clad surface temperatures. Indicate the end of blowdown, start of refill, and start of reflood on this graph.
- c. Bumup for the limiting rod.
Response
Figure 6-2 shows the corrected and uncorrected radial temperature profile for the limiting PCT case hot rod at the initiation of the transient. Because the uncorrected radial profile is never used or recorded in the methodology, it cannot be provided. However, the uncorrected centerline temperature is available and shown on Figure 6-2. As the pellet power is not adjusted the radial temperature profile must follow the corrected profile closely and the two must converge at the surface of the pellet. Figure 6-3 shows the centerline, surface, and average fuel temperatures of the fresh 8% Gad rod at the PCT elevation for the limiting PCT case. In this case, all of the fresh rods have higher PCTs than the once-burnt rods. The most limiting once-burnt rod is the U0 2 Gad rod. With a cycle burnup of approximately 8,897 EFPH, the fresh 8%
Gad rod has a burnup of 15.3 MWd/kgU while the once-burnt U0 2 rod has a burnup of 36.5 MWd/kgU. A plot comparing the PCT of the fresh U0 2 with 8% Gad and the once-burnt U0 2 rods for this case is shown in Figure 6-4.
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Rev. 1 Paae 108 Paae 108 Figure 6-1 Once-Burnt Fuel Power Ratios (2nd cycle)
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Rev. 1 Paqe 109 Figure 6-2 Radial Temperature Profile for Hot Rod AREVA NP Inc.
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Rev. 1 Paae 110 Paae 110 Figure 6-3 Temperature versus Time for Fuel Centerline, Clad Surface, and Fuel Average AREVA NP Inc.
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Rev. 1 Page 111 Figure 6-4 Fresh versus Once-Burnt U0 2 Rod PCT Trace AREVA NP Inc.
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Rev. 1 Paae 112 Paae 112 Figure 6-5 Fractional Fuel Centerline Temperature Delta Between RODEX3A and Data AREVA NP Inc.
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Rev. 1 Paae 113 Figure 6-6 Fuel Centerline Temperature Delta of RODEX3A Calculations to Data (Original and Using the New Correlation)
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Rev. 1 Paae 114 Paae 114 Figure 6-7 Correction Factor (as applied for temperatures in Kelvin)
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Rev. 1 Page 115 6.2 Decay Heat UncertaintyAssumption Question:
Provide additionalinformation to justify the use of the selected analytic treatment for decay heat uncertainty in the RLBLOCA model.
- a. The NRC needs to understand the sensitivity that PCT has with respect to the decay heat uncertainty,please re-execute the limiting case with a 1.03 decay heat multiplierand report the results.
- b. The NRC requests to provide the following parametersfor the limiting LBLOCA with the 1.03 multiplier on the decay heat curve: core power, RCS pressure, hot assembly inlet mass flow rate, hot assembly outlet mass flow rate, break flow rate (pump side and vessel side), ECC injection mass flow rate, PCT, hot spot vapor temperature,hot spot heat transfercoefficient (total), downcomer liquid level, core liquid level and containment pressure.
Response
The RLBLOCA EM decay heat calculations are based on the 1979 ANSI/ANS standard (Reference 16). The standard is applicable to light water reactors containing low enriched uranium as the initial fissile material; all plants, to which the RLBLOCA EM is applicable, are such plants. The selected approach to simulate fission product decay assures a representative yet conservative treatment. The EM fission product decay heat simulation and the basis for the conservatism of the approach are outlined in the remainder of this response.
Non-Sampling Approach to Decay Heat The RLBLOCA methodology proposed herein utilizes the U235 decay curve from the 1979 ANSI/ANS standard for fully saturated decay chains as the decay for all fission products. The fully saturated chains result from an assumption of infinite operation. The total energy per fission is assumed to be 200 MeV (Reference 16). No bias or uncertainty is assigned to the fission product decay heat. Differing from the base EMF-2103 evaluation model approach, the uncertainty for the decay heat parameter is set to zero and no sampling is done on this parameter, resulting in the decay heat being used with a 1.0 multiplier. The decay heat in the analysis is always the 1979 ANS standard for decay heat from U235 with fully saturated decay chains, corresponding to infinite operation, assuming 200 MeV per fission.
Conservatism in the Approach In the approach used, the total energy per fission is assumed to be 200 MeV whereas a more accurate value for U235 would be greater than 202 MeV per fission. This imparts a direct 1%
conservatism.
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Rev. 1 Page 116 During irradiation, plutonium accumulates such that the ratio of plutonium-to-uranium fission-energy production rate is substantial and increasing. Because the decay energy resulting from plutonium fissions is less than that from U235, the decay energy is reduced from U235 fully saturated decay chains as the fuel is burnt. Thus, as bumup increases, the RLBLOCA decay heat modeling with U235 only accrues conservatism. This conservatism applies to all regions of the core according to the mix of burnups represented within each region.
The fresh fuel, hot pin and hot assembly, begin operation with no plutonium. Therefore, the reduction in decay heat due to plutonium build-up is not applicable to the low bumup fuel in the initial period of the cycle. However for fresh fuel, the concentrations of long decay term fission products will not have built up. The lack of long decay term sources comprises a reduction in decay heat rate of several percent over the first year of operation, making the infinite operation assumption conservative while the plutonium concentration is accumulating.
Calculations of these considerations based on the 1979 ANS standard have been performed to demonstrate the conservatism of the selected approach. Figure 6-8 and Figure 6-9 show the decay heat versus time for:
- 1) Infinite Operation of U235 (the AREVA NP decay heat model)
- 2) Finite Operation to 0.1 GWdlMTU of all fissionable isotopes with uncertainties added
- 3) Finite Operation to 1 GWd/MTU of all fissionable isotopes with uncertainties added
- 4) Finite Operation to 1 GWd/MTU of all fissionable isotopes without uncertainties
- 5) Finite Operation to 20 GWd/MTU of all fissionable isotopes with uncertainties added
- 6) Finite Operation to 40 GWd/MTU of all fissionable isotopes with uncertainties added
- 7) Finite Operation to 60 GWd/MTU of all fissionable isotopes with uncertainties added In order to treat the Plutonium buildup effect conservatively, the finite operations curves are based on cycle management and enrichment assumptions that minimize the build up of Plutonium. No uncertainty is included in the infinite operation curve. The uncertainties incorporated in the other curves are 2 sigma values for the individual isotopes as published in the 1979 ANS standard. This provides greater than a 95/95 confidence in each of the decay heat contributions. The contributions are added linearly according to the individual isotopes fractional occurrence of fission.
Because of the range of the decay heat parameter, the early comparison of the relationships is difficult to ascertain. Clearly the U235 infinite operation curve is conservative for all times after a few seconds (-2 seconds). To better demonstrate the relationships, Figure 6-10 and Figure 6-11 provide the ratios of the finite operation curves to the infinite operation curves. The curvature of the plotted ratios during the first 2 to 3 seconds is due to the increased uncertainties during this time phase. The 1979 ANS standard is based on measured data and the difficulty of measuring decay heat within a few seconds of shutdown is reflected in these AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 117 uncertainties. The highest combined finite operation decay heat curve with uncertainties exceeds the AREVA NP decay heat curve by only 2.5% at shutdown and falls below the AREVA NP curve in less than 2 seconds. Thus, there is only a 5% probability that the infinite operation curve of decay heat will be exceeded by up to 2.5% and that possibility exists for the first 2 seconds of the transient. The potential accumulated under-prediction is of short duration and of no consequence to the LOCA evaluation. The decay heat curve selected is suitable while somewhat conservative for the realistic evaluation of LOCA.
In conclusion, the choice of infinite operation with pure U235 fission product decay heat provides a base model that is conservative relative to the decay heat for finite operation. For RLBLOCA evaluation, the sampling of a decay heat multiplier has been removed such that the decay heat for all cases is 1.0 times the infinite operation U235 decay chain providing conservative treatment of the 1979 ANS standard with the assumption of 200 MeVlfission.
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Rev. 1 Page 118 Figure 6-8 Decay Heat Comparisons, Infinite Operation U235, Finite Operation All Isotopes (0.1 - 10 sec)
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Rev. 1 Page 119 Figure 6-9 Decay Heat Comparisons, Infinite Operation U235, Finite Operation All Isotopes (10 - 1000 sec)
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Rev. 1 Page 120 Figure 6-10 Decay Heat Ratios, Finite Operation over Infinite Operation for U236, All Isotopes (0 - 10 sec)
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Rev. 1 Page 121 Figure 6-11 Decay Heat Ratios, Finite Operation over Infinite Operation for U235, All Isotopes (0 - 600 sec)
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Rev. 1 Page 122 6.3 Clad Swelling and Rupture Question:
The following questions are based on a July 14, 2009, letter from Gardner,AREVA NP, to the USNRC, re: Informational Transmittal Regarding Requested White Papers on the Treatment of Exposure Dependent Fuel Thermal Conductivity Degradation in Legacy Fuel Performance Codes and Methods.
- a. AREVA postulates that clad swelling and rupture produces a benefit to PCT,and because of this, the realistic large break loss of coolant accident (RLBLOCA) model does not include a clad swelling and rupture model. Does this conjecture include consideration of test data, which has shown that following fuel rupture, the ballooned region fills with fuel fragments?
What analytic studies support this conclusion? How are they applicable to Calvert Cliffs?
Please also address the potentialfor co-planarblockage with the fuel relocationevaluation.
- b. Since blowdown ruptures can occur at end of life conditions, show that blowdown ruptures do not occurat the end of life for the postulated Calvert Cliffs large break LOCA.
Response
For CCNPP, the clad swell, rupture and relocation models were activated and the limiting set of 59 cases was re-executed as a sensitivity study. The calculation shows that no rupture occurred for CCNPP, thus there is no rupture or relocation issue for CCNPP.
The "White Paper" on swell, rupture, and relocation (SRR) was transmitted to the NRC via Reference 17 and the rupture models discussed in the paper were used in the sensitivity study.
L Additional discussion is provided following Figure 6-12 on the AREVA RLBLOCA provisions to not include swell, rupture, and fuel relocation.
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Rev. 1 Page 123 Figure 6-12 Swell, Rupture, & Relocation Sensitivity AREVA NP Inc.
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Rev. 1 Page 124 AREVA NP Inc.
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Rev. 1 Page 125 AREVA NP Inc.
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Rev. 1 Page 126 6.4 Single-Sided Oxidation Model Question:
Provide information to illustrate the conservative nature of the single-side only oxidation model and its applicationto the Calvert Cliffs RLBLOCA analysis.
Response
AREVA NP's NRC-approved RLBLOCA EM uses the maximum un-ruptured cladding oxidation as representative or bounding of the transient oxidation that would have been computed at a rupture location. The position is supported by three aspects of the performed oxidation calculation.
" The cladding is initialized with no initial corrosion layer. Because the oxidation rate is inversely proportional to the oxidation layer present, the use of clean cladding at the start of the accident leads to substantially higher reaction rates. For corrosions in the range of the first cycle of M5 cladding, the difference in rate is a minimum of a 50 %
increase and increases during the cycle. The increase applies to both exterior and post-rupture interior oxidation.
- The cladding temperature even in the presence of fuel relocation is reduced for the ruptured region of the cladding. In the KfK experiments (Reference 20) the temperature drop at rupture was between 50 and 75 K. Since the oxidation rate is exponentially proportional to the cladding temperature, a decrease in temperature of this magnitude would reduce the oxidation rate by approximately 50%.
" For ruptured cladding either the cladding interior oxidation rate is reduced by attached pellet fragments, moderate to highly burnt fuel, or the cladding temperature decrease at rupture is much more than the 50 to 75 K explained above. In either case, an additional mechanism exists to reduce the local oxidation at the rupture location.
In conclusion, insights into the EM oxidation process and those that will evolve after rupture clearly identify differences that will reduce the transient oxidation at the rupture location to less than that which the EM calculates at un-ruptured locations. Thus, the RLBLOCA Revision 0 EM approach of determining local transient oxidation is clearly appropriate to demonstrate compliance with the local oxidation criterion of 10CFR50.46, when combined with the pre-transient oxidation.
The initial corrosion layer was calculated to be 1.214-percent for the Fresh fuel rod (at a maximum bumup of 30.7 GWd/MTU) with M5 cladding and 1.997-percent for the once-burned fuel rod (at a maximum burnup of 54.2 GWd/MTU) with M5 cladding. The initial corrosion layer was added to the transient calculated value and the total is in Table 3-5.
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Rev. 1 Page 127 6.5 Limiting Condition - Single Failure Questions:
- 1. The current licensing basis, deterministic loss of coolant accident (LOCA) analysis concluded that the limiting condition did not involve a worst-case single failure, but rather that it depended on injected coolant delivered in such a condition that the resultant containment environment, specifically the lower containment pressure, contributed to the limiting peak cladding temperature (PCT). Please provide information describing how this potentially limiting scenario was evaluated using the proposed best-estimate methodology.
- 2. Please provide additional information summarizing the single-failure evaluation performed to establish compliance with General Design Criterion (GDC) 35 requirements. Identify which single failures were considered, discuss whether each failure was evaluated or explicitly analyzed, and for those failures which were explicitly analyzed, explain whether they were analyzed in a reference case or explicitly as a part of the statistical methodology. Also discuss the basis for the single failure evaluation. For example, were single failures considered as a matter of experience with CCNPP Units I and 2 specifically, or with a generic CE nuclear steam supply system design?
- a. The staff also needs to understandhow the limiting single failure for the CE 2x4 NSSS was determined, since the basis for the RAI response defers to NRC-approved methodology. Poring through EMF-2103, the staff only located sensitivity results on 3-loop W systems. In some cases, the limiting failure would be a single LPSI and in others it was a diesel. The staff could not locate a clear,generic disposition for the single failure at any place in EMF-2103.
- b. What was done under the auspices of EMF-2103 development to ensure that the containment analysis produced a sufficiently conservative prediction that a no failure, maximum SI spillage case, for a CE 2x4 NSSS, is bounded by the chosen single failure? The staff will need to see that work.
Response
- 1) The proposed licensing basis for CCNPP is AREVA's NRC-approved RLBLOCA evaluation model and the worst single failure considered is loss of diesel with fully functional containment sprays. The EM also conservatively prescribes:
(1) The use of full containment sprays with minimum time delay at the minimum technical specification temperature; (2) Pumped ECCS injection at the maximum technical specification temperature; and (3) Sampling of the containment volume (indirectly sampling containment pressure) from its nominal volume to its empty volume.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 128 Studies, comparing several failure assumptions, including a no-failure assumption (Reference 1, EMF-2103(P)(A) Revision 0, RAI response Numbers 26 and 111) validate that the ECCS and containment modeling of the AREVA methodology trends to the conservative. The containment pressure response is indirectly ranged by sampling the containment volume. The possible containment volume range to be sampled from was 1.99E+6 to 2.15E+06 ft3 for CCNPP.
Figure 4-21 shows that there is little sensitivity between containment volume (indirectly pressure) and PCT for a statistical application. Thus, the methodology is responsive to the goal of a realistic evaluation, yet slightly conservative.
- 2) Section 4.9 discusses GDC 35. The single failure prescribed by EMF-2103(P)(A) (AREVA NP's RLBLOCA EM) is a loss of one train of ECCS.
AREVA NP satisfies the GDC-35 criteria by running one set of 59 cases with offsite power available and one set of 59 cases with no offsite power available. The sampling seeds are held constant between these two case sets, with the only difference being the offsite power assumption. The case set that produces the most limiting PCT for CCNPP is reported was with offsite power not available. Figure 3-24 displays the results from the two case sets.
- a. The definition for loss of a diesel scenario by itself would mean that in addition to loss of one LPSI and one HPSI pump, one train of containment spray would not be available. The current method models all containment pressure-reducing systems as fully functional. Containment fans start at time zero and containment sprays have a 20 second delay (Table 3-8).
The response to RAI #111 for EMF-2103 (Reference 1, Attachment 1 page 185 -
189) was based on sensitivities to 3-loop W plants. The Base Case, which produced the most limiting results, is described in the RAI #111 response as the loss of one diesel with full containment spray. Figure 6-13 (recreated from RAI #111, Figure 111.2) shows that for the sample plant analysis, W 3-loop, the base case, AREVA NP ECCS failure assumptions, is 35°F higher in PCT than a fully consistent loss of diesel and over 170°F greater than the loss of one LPSI case.
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Rev. 1 Paae 129 Figure 6-13 Clad Temperature Response from Single Failure Study AREVA NP Inc.
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Rev. I Page 130
- b. A sensitivity study of the CCNPP limiting case (Case 47) from the Analysis of Record (AOR) was conducted with "maximum" ECCS flow conditions to demonstrate that the minimum ECCS single failure assumption is conservatively bounding. The study was performed for both the loss of offsite power (LOOP) and no loss of offsite power (No-LOOP) limiting cases, which were both Case 47.
The LOOP case for the maximum ECCS configuration had a PCT value of 1620°F, which is identical to the PCT of the AOR LOOP with minimum ECCS. The No-LOOP case for the maximum ECCS configuration had a PCT value of 1537 OF compared to the AOR No-LOOP case with minimum ECCS, which had a PCT of 1539 0 F. This demonstrates that the AREVA single failure assumption produces conservative results. Figure 6-14 through Figure 6-17 show the respective PCT trace, containment and system pressure, ECCS injection rates, and downcomer level for both the AOR and the maximum ECCS sensitivity.
Figure 6-15 demonstrates that the maximum ECCS flow (Figure 6-16) does not have a significant impact on the containment pressure up to about 70 seconds (approximately the time that the SIT empties); the maximum ECCS containment pressure overlaps the AOR containment pressure for the majority of the transient.
Figure 6-17 gives the downcomer level for both the AOR and the maximum ECCS case. It can be seen that the downcomer level in the maximum ECCS case is higher than the AOR, consequently providing more driving head for the reflood of the core.
The higher driving head in the maximum ECCS case is enough to compensate for small differences in containment pressure resulting in a faster post peak cooldown.
The AREVA NP RLBLOCA application, regardless of the loss of diesel assumption, models all containment pressure-reducing systems and conservatively assumes them to be fully functional. The AOR conservatively assumes an on-time start and normal lineups of the containment spray and Emergency Cooling Units to conservatively reduce containment pressure and increase break flow. The results of the study demonstrate that the AOR ECCS configuration is PCT-limiting.
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Rev. 1 Page 131 Figure 6-14 Comparison of Maximum Clad Temperature Independent of Elevation for Maximum ECCS and Minimum ECCS (AOR)
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Rev. 1 Page 132 Figure 6-15 Comparison of Containment and System Pressure for Maximum ECCS and Minimum ECCS (AOR)
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Rev. 1 Page 133 Figure 6-16 Comparison of ECCS Flows for Maximum ECCS and Minimum ECCS (AOR)
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Rev. 1 Page 134 Figure 6-17 Comparison of Downcomer Level for Maximum ECCS and Minimum ECCS (AOR)
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Rev. 1 Page 135 6.6 Core Liquid Levels Question:
Section 3.3 states, "the RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water (i.e., RAS) for ECCS pumped injection need not be considered." Is the SRELAP-5 model of the limiting case capable of generatingcredible results after 350s? If so, please provide results for a period of the transient sufficient to demonstrate that the core collapsedliquid levels are stable or increasing.
Response
In general the S-RELAP5 model does provide credible simulation of primary system responses at the end of the transient. In some cases, the reactor vessel and core levels do not show an increasing or stable trend toward the end of the transient run. However, the level will be sufficient to maintain core cooling and prevent further core heatup.
In the case of CCNPP, a very slight downward trend toward the end of the transient is observed in the core liquid level in Figure 3-21. Figure 3-21 shows an increase in core collapsed liquid level from about 80 seconds to about 120 seconds. After 120 seconds, it shows the core collapsed level slightly increasing and then it drops slightly by the time of PCT quenching (about 220 sec).
A better measure of a stable cooling inventory is the vessel mass, which shows that that water is being supplied to the vessel at the rate that boiling is occurring. The reactor vessel mass shown in Figure 3-23 confirms a stable reactor vessel mass from about 80 seconds until the end of the transient.
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Rev. 1 Page 136 6.7 PlantInput Selection and Technical Specifications Question:
- 1. Please provide information to enable comparison between Technical Specifications (TS) requirementsand analytic inputparametersfor PressurizerLevel.
- 2. Please provide discussion to confirm that the assumed containment temperature is an acceptableminimum without a TS requirement.
- 3. The TS minimum for the refueling water storage tank (RWST) temperature is 40°F.
Previous, deterministicanalyses demonstratedthat minimum safety injection temperatures resulted in a limiting PCT. In light of this information, please explain why a minimum RWST temperature case was not evaluated, or if a minimum RWST temperature case was evaluated,please summarize the evaluation and discuss its conclusions.
ResDonse:
Pressurizer water level > 133 inches and < 225 inches;"
The sampled range for the liquid level in the pressurizer is 32.2% to 67.2 % of span. The pressurizer liquid level % span translates into a liquid level range of 125.2 in to 251.2 in.
Therefore, the Technical Specification for CCNPP is bounded.
- 2. Technical Specification LCO 3.6.5 states "Containment average air temperature shall be <
120°F". The sampled range for the containment temperature was 60 to 125 0 F. The Technical Specification for CCNPP is bounded. The Technical Specification for CCNPP is bounded on the high side and the 60°F temperature for the low side was supported by a review of plant data.
- 3. The RWST borated water temperature Technical Specification requirement for CCNPP is
> 40°F and < 100I F (TS 3.5.4.1 and TS 3.5.4.2). The NRC-approved RLBLOCA EM, EMF 2103(P)(A), prescribes use of the maximum temperature for the ECCS pumped injection and use of the minimum temperature for containment sprays. The temperatures for the CCNPP analysis are 100°F for pumped injection and 40°F for the containment sprays.
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A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 137 7.0 References 1 EMF-2103(P)(A) Revision 0, Realistic Large Break LOCA Methodology, Framatome ANP, Inc., April 2003.
2 G.P. Liley and L.E. Hochreiter, "Mixing of Emergency Core Cooling Water with Steam: 1/3
- Scale Test and Summary," EPRI Report EPRI-2, June 1975.
3 Technical Program Group, Quantifying Reactor Safety Margins, NUREG/CR-5249, EGG-2552, October 1989.
4 XN-CC-39 (A) Revision 1, "ICECON: A Computer Program to Calculate Containment Back Pressure for LOCA Analysis (Including Ice Condenser Plants)," Exxon Nuclear Company, October 1978.
5 Wheat, Larry L., "CONTEMPT-LT A Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-Of-Coolant-Accident," Aerojet Nuclear Company, TID-4500, ANCR-1219, June 1975.
6 U. S. Nuclear Regulatory Commission, NUREG-0800, Revision 3, Standard Review Plan, March 2007.
7 Letter from D. V. Pickett to G.H. Gellrich, U.S. NRC to CENG, "Calvert Cliffs Nuclear Power Plant, Unit Nos 1 and 2 Exemption for use of M5 Cladding (TAC NOS. ME5154 AND ME5155)," January 13, 2011 (U.S. NRC ADAMS Accession # ML103070113).
8 NUREG/CR-0994, "A Radiative Heat Transfer Model for the TRAC Code,"
November 1979.
9 D. A. Mandell, "Geometric View Factors for Radiative Heat Transfer within Boiling Water Reactor Fuel Bundles," Nucl. Tech., Vol. 52, March 1981.
10 EMF-CC-1 30, "HUXY: A Generalized Multirod Heatup Code for BWR Appendix K LOCA Analysis Theory Manual," Framatome ANP, May 2001.
11 NUREG/CR-1532, EPRI NP-1459, WCAP-9699, "PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Report," June 1980.
12 EMF-2102(P)(A) Revision 0, S-RELAP5: Code Verification and Validation, Framatome ANP, Inc., August 2001.
13 J.P. Holman, Heat Transfer, 4th Edition, McGraw-Hill Book Company, 1976.
14 ANF-90-145(P)(A), "RODEX3 Fuel Rod Thermal-Mechanical Response Evaluation Model", April 1996.
15 EMF-2994(P) Rev. 4, "RODEX4: Thermal-Mechanical Fuel Rod Performance Code Theory Manual", December 2009.
16 ANSI/ANS-5.1-1979, American National Standard for Decay Heat Power in Light Water Reactors, approved August 29, 1979.
17 AREVA Letter NRC:11:101 from Pedro Salas to H.D. Cruz, September 20, 2011 "Presentation Materials for Pre-Submittal Meeting on PWR LOCA Supplements," enclosed "White Paper - Clad Swelling, Rupture, & Relocation."
AREVA NP Inc.
A AREVA Calvert Cliffs RLBLOCA Summary Report ANP-3043(NP)
Rev. 1 Page 138 18 AREVA Letter NRC:02:062, December 20, 2002, Responses to a Request for Additional Information on EMF-2103(P) Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors", (TAC No. MB2865).
19 BAW-10166(P)(A) - Revision 5, BEACH - Best Estimate Analysis Core Heat Transfer, AREVA, Lynchburg, Virginia, November 2003.
20 P. IhIe, Heat Transfer in Rod Bundles with Severe Clad Deformations, KfK 3607 B, April 1984.
21 M.J. Loftus, et al., PWR FLECHT SEASET 163-Rod Bundle Flow Blockage: Task Data Report, No. 13, NUREG/CR-3314, October 1983.
22 NEAICSNI/R(2004)19, SEGFSM Topical Meeting on LOCA Fuel Issues, Argonne National Laboratory, May 25-26 2004, Published by Organization for Economic Cooperation and Development Nuclear Energy Agency, Isy-les-Moulineaux, France, November 2004.
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