ML050700408

EMF-3030(NP), Revision 0, Robinson Nuclear Plant, Realistic Large Break LOCA Analysis, February 2004, Non-Proprietary Version

ML050700408
Person / Time
Site:	Robinson
Issue date:	02/20/2004
From:	Cudlin J, Holm J, Shaw R Framatome ANP
To:	Office of Nuclear Reactor Regulation
References
EMF-3030(NP), Rev 0
Download: ML050700408 (52)

Text

EMF-3030(NP)

Revision 0 Robinson Nuclear Plant Realistic Large Break LOCA Analysis I

February 2004

ISSUED INFRAANP ON-UNE DOCULENTSYSTEM DATE: P- o704X Frarmatom e ANP, Inc.

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Revision 0 Robinson Nuclear Plant Realistic Large Break LOCA Analysis Prepared: 5 4tc c4o R. A. Shaw, Engineer Date PWR Safety Analysis Approved: : , _ Z/7 Loty J. Cudlin, Manager Date Analysis Services I PWR Safety Analysis Approved: 7/Da t 4e J. S. Oicmanager Date' ProddctYLicensing

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Customer Disclaimer Important Notice Regarding the Contents and Use of This Document Please Read Carefully Framatome ANP, Inc.'s warranties and representations concerning the subject matter of this document are those set forth in the agreement between Framatome ANP, Inc. and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided In such agreement, neither Framatome ANP, Inc. nor any person acting on Its behalf:

a. makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained Inthis document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or

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Robinson Nuclear Plant Revision 0 D : I ---- M -1. I BO.A -^1-t-: Date:-

r-WalI51C Large D[UdK LULtk anadlysis b raget a Nature of Changes Item Page Description and Justification

1. All This is a new document.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page ii Contents 1.0 Introduction ......................................... 1-1 2.0 Summary ......................................... 2-1 3.0 Analysis .......................................... 3-1 3.1 Description of the LBLOCA Event ........................................ 3-1 3.2 Description of Analytical Models ........................................ 3-3 3.3 Plant Description and Summary of Analysis Parameters ................................... 3-5 3.4 SER Compliance ........................................ 3-6 3.5 Realistic Large Break LOCA Results ........................................ 3-6 4.0 Conclusions ......................................... 4-1 5.0 References ......................................... 5-1 Tables 2.1 Summary of Major Parameters for Limiting Transient ......................................... 2-1 3.1 Sampled LBLOCA Parameters ......................................... 3-7 3.2 Plant Operating Range Supported by the LOCA Analysis ......................................... 3-8 3.3 Statistical Distributions Used for Process Parameters ......................................... 3-10 3.4 SER Conditions and Limitations ......................................... 3-11 3.5 Summary of Results for the Limiting PCT Case ......................................... 3-13 3.6 Calculated Event Times for the Limiting PCT Case ......................................... 3-13 3.7 Heat Transfer Parameters for the Limiting Case ......................................... 3-14 Figures 3.1 Primary System Noding ......................................... 3-15 3.2 Secondary System Noding......................................... 3-16 3.3 Reactor Vessel Noding ......................................... 3-17 3.4 Core Noding Detail ......................................... 3-18 3.5 Upper Plenum Noding Detail ......................................... 3-19 Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page iii 3.6 Scatterplot of Operational Parameters ...................................................... 3-20 3.7 PCT vs. PCT Time Scatterplot From 59 Calculations .................................................. 3-22 3.8 PCT v. Break Size Scatterplot From 59 Calculations ................................................... 3-23 3.9 Maximum Oxidation vs. PCT Scatterplot From 59 Calculations .................................. 3-24 3.10 Peak Cladding Temperature (Independent of Elevation) for Limiting Break ................ 3-25 3.11 Break Flow for the Limiting Break (Early) ...................................................... 3-26 3.12 Core Inlet Mass Flux for Limiting Break (Early) ...................................................... 3-27 3.13 Core Outlet Mass Flux for Limiting Break (Early) ...................................................... 3-28 3.14 Void Fraction at RCS Pumps for Limiting Break . 3-29 3.15 ECCS Flows (Includes Accumulator, HPSI, and LPSI) for Limiting Break .................. 3-30 3.16 Upper Plenum Pressure for Limiting Break (Early) .................................... 3-31 3.17 Collapsed Liquid Level in the Downcomer for Limiting Break ..................................... 3-32 3.18 Collapsed Liquid Level in the Lower Plenum for the Limiting Break ........................... 3-33 3.19 Collapsed Liquid Level in the Core for the Limiting Break .......................................... 3-34 3.20 Containment and Loop Pressures for Limiting Break ................................................. 3-35 This document contains a total of 47 pages.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page iv Nomenclature BOC Beginning of Cycle CFR Code of Federal Regulations DEGB Double-Ended Guillotine Break DESB Double-Ended Split Break ECCS Emergency Core Cooling System EOC End of Cycle FT Total Peaking Factor Q

FAH Nuclear Enthalpy Rise Factor FCTF Fuel Cooling Test Facility FANP Framatome Advanced Nuclear Power HFP Hot Full Power HHSI I HPSI High Head / Pressure Safety Injection LBLOCA Large Break Loss of Coolant Accident LHSI / LPSI Low Head / Pressure Safety Injection MOC Middle of Cycle MTC Moderator Temperature Coefficient NRC U. S. Nuclear Regulatory Commission PCT Peak Clad Temperature RCP Reactor Coolant Pump RCS Reactor Coolant System RLBLOCA Realistic Large Break LOCA RV Reactor Vessel Si Safety Injection Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 1-1 1.0 Introduction This report describes a RLBLOCA analysis of a postulated large-break loss-of-coolant accident (LBLOCA) for the Robinson Nuclear Plant, which is a 3-loop PWR operating with Framatome ANP, Inc. (FANP) fuel.

The analysis supports operation during Cycle 23 and future cycles, 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 non-parametric statistical methods inherent in the FANP 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 single-failure assumption is applied in which the negative effects of both the loss of a LPSI pump and the loss of a diesel generator are simulated. (Containment coolers and sprays are assumed to be fully functional in this analysis.) The effects of Gadolinia-bearing fuel rods and peak fuel rod exposures are also considered.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 2-1 2.0 Summary The limiting peak cladding temperature (PCT) analysis is based on the parameter specification given in Table 2.1, based on a set of 59 calculations. The limiting PCT, occurring on a 4 w/o Gd2O3 bearing rod, was calculated to be 19520F.

The analysis assumes full-power operation at 2339 MWt (plus uncertainties), a steam generator tube plugging level of up to 10% in any generator, a total peaking factor (F0 ) of 2.62 (including uncertainties, but no axial dependency), and a nuclear enthalpy rise factor (FAH) of 1.80 (including uncertainty). This analysis also addresses typical operational ranges or technical specification limits (whichever is applicable) with regard to pressurizer pressure and level; accumulator pressure, temperature (containment temperature), and level; core average temperature; core flow; containment pressure and temperature; and refueling water storage tank temperature.

The FANP RLBLOCA methodology explicitly analyzes only fresh fuel assemblies. (See Reference 1, Appendix B.) Previous analyses have shown that once- and twice-burnt fuel will not be limiting up to peak rod average exposures of 62,000 MWd/MTU, which corresponds to a maximum assembly burnup of 57,000 MWd/MTU. The analysis demonstrates that the 10 CFR 50.46(b) criteria listed in Section 3.0 are satisfied.

Table 2.1 Summary of Major Parameters for Limiting Transient Core Average Burnup (EFPH) 5821 Core Power (MW) 2336 Total Peaking (Fo) 2.594 Radial Peak (F~h) 1.8 Axial Offset +0.20 Break Type DEGB Break Size (ft2/side) 3.64 (-88%)

Offsite Power Availability Yes Decay Heat Multiplier 1.02 Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-1 3.0 Analysis The purpose of the analysis is to verify typical technical specification peaking factor limits and the adequacy of the emergency core cooling system (ECCS) by demonstrating that the following 10 CFR 50.46(b) criteria are met:

• The calculated maximum fuel element cladding temperature shall not exceed 2200'F.

• The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

• 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.

• The calculated changes in core geometry shall be such that the core remains amenable to cooling.

Note that this analysis does not evaluate for changes in core coolability due to seismic events, nor does it evaluate the 10 CFR 50.46(b) long-term cooling criterion.

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 3-loop 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.

3.1 Description of the LBLOCA Event A LBLOCA is initiated by a postulated large rupture of the Reactor Coolant System (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 large 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 Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-2 departure from nucleate boiling (DNB). Subsequently, the limiting fuel rods are cooled by film convection to steam. The coolant voiding creates a strong negative reactivity effect, and core fission ends. As heat transfer from the rods is reduced, the cladding temperature rises.

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 may also lead to a period of positive core flow or reduced downflow as the reactor coolant pumps in the intact loops continue to supply water to the vessel. Cladding temperatures may be reduced, and some portions of the core may rewet during this period.

This positive core flow or reduced downflow period ends as two-phase conditions occur in the reactor coolant pumps, 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 safety injection actuation signal (SIAS) is tripped.

This signal is initiated by either high containment pressure or low pressurizer pressure.

Regulations require that a worst single-failure be considered for ECCS safety analysis. This single-failure has been determined to be the loss of one ECCS train, including one high-pressure safety-injection (HPSI) pump and one low-pressure safety injection pump (LPSI) pump. The FANP RLBLOCA methodology conservatively assumes an on-time start and normal lineups of the containment spray and fan coolers to conservatively reduce containment pressure and increase break flow. Hence, the analysis assumes that one HPSI pump, one LPSI pump, two containment spray pumps, and four fan coolers are operating.

When the RCS pressure falls below the accumulator pressure, fluid from the accumulators is injected into the cold legs. In the early delivery of accumulator 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 accumulator water is exhausted and core recovery must rely on SI coolant delivery alone. As the accumulators empty, the nitrogen gas used to pressurize the accumulators 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.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-3 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 LPSI and 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 generator, and the reactor coolant pump 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.

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 Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology (Reference 2). 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.

The governing two-fluid (plus noncondensibles) 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 another are accounted for by interfacial friction and heat and mass transfer interaction terms in the equations. The conservation Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-4 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 phenomenon expected during the LBLOCA event are captured. The basic building block for modeling is the hydraulic volume for fluid paths and the heat structure for a heat transfer surface. In addition, special purpose components exist to represent specific components such as the pumps 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.

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 technical specifications or to match measured data.

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 the CONTEMPT-LT code (Reference 3).

The methods used in the application of S-RELAP5 to the large break LOCA are described in Reference 1. A detailed assessment of this computer code was made through comparisons to experimental data. These assessments were used to develop quantitative estimates of the code's ability to predict key physical phenomena in a PWR large break LOCA. 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 95% probability. The steps taken to derive the PCT uncertainty estimate are summarized below:

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 the model nodalization is consistent with'the model nodalization used in the code validation.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-5

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 with 95% confidence. Total oxidation and total hydrogen are based on the 95/95 PCT case. The adequacy of the ECCS is demonstrated when these results satisfy the criteria set forth in Section 3.0.

3.3 Plant Description and Summary of Analysis Parameters The plant analysis presented in this report is for a Westinghouse designed pressurized water reactor (PWR), which has three loops, each with a hot leg, a U-tube steam generator, and a cold leg with a RCP. The RCS also includes one pressurizer. The ECCS includes one accumulator/LPSI and one HPSI injection path per RCS loop. The HPSI and LPSI feed into common headers which are connected to the accumulator lines.

The S-RELAP5 model explicitly describes the RCS, reactor vessel, pressurizer, and ECCS back to the common LPSI header and accumulators. 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 up to 10% per steam generator was assumed.

As described in the FANP RLBLOCA methodology, many parameters associated with LBLOCA phenomenological uncertainties and plant operation ranges are sampled. A summary of those parameters sampled 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 is analyzed to develop uncertainties for the process parameters sampled in the analyses. Table 3.3 presents a summary of the uncertainties used in the analyses. Two parameters (refueling water storage tank (RWST) temperature for Si flows and diesel start time) are set at conservative bounding values for all calculations. Where applicable, the sampled Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-6 parameter ranges are based on technical specification limits. Plant data are used to define range boundaries for some parameters, e.g., loop flow (high end) and containment temperature (low end).

The containment response was modeled conservatively relative to an Appendix K response.

Specifically, Technical Specification minimum RWST temperature was used for the containment sprays along with an Uchida heat transfer coefficient multiplier of [ ]. The containment model initial pressure was set to a nominal value.

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). These requirements have all been complied with during the application of the methodology as addressed in Table 3.4.

3.5 Realistic Large Break LOCA Results A set of 59 transient calculations was performed sampling the parameters listed in Table 3.1.

For each transient calculation, PCT was calculated for a U02 rod and for gadolinia bearing rods with concentrations of 2 w/o, 4 w/o, and 8 w/o Gd2O3. The limiting PCT (19520F) occurred in Case I for a 4 w/o Gd2O3 rod. The major parameters for the limiting transient are characterized in Table 2.1. Table 3.5 lists the results of the limiting case. 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. A nominal 50/50 PCT case was identified as Case 46. The nominal PCT is 1521'F. This result can be used to quantify the relative conservatism in the 95/95 result. In this analysis, it is 431"F.

The hot fuel rod results, event times and analysis plots for the limiting PCT case are shown in Table 3.5, Table 3.6, and in Figures 3.10 through 3.20. Figure 3.6 shows 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. Figures 3.7 and 3.8 show PCT scatter plots vs. the time of PCT and vs. break size from the 59 calculations, respectively. Figure 3.9 shows the maximum oxidation vs. PCT for the 59 calculations.

Figures 3.10 through 3.20 show key parameters from the S-RELAP5 calculation. Figure 3.10 is the plot of PCT independent of elevation.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-7 Table 3.1 Sampled LBLOCA Parameters Phenomenological Time in cycle (peaking factors, axial shape, rod properties, burnup)

Break type (guillotine vs. split)

Break size Critical flow discharge coefficients (break)

Decay heat 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 interfacial drag Condensation interphase heat transfer Metal-water reaction Planta Offsite power availability Core power Pressurizer pressure Pressurizer level Accumulator pressure Accumulator level Accumulator temperature (based on containment temperature)

Containment temperature Containment volume Initial flow rate Initial operating temperature Diesel start (for loss of offsite power only)

Uncertainties for plant parameters are based on typical plant-specific data with the exception of

'Offsite power availability" which Is a binary result that Isspecified by the analysis methodology.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Paae 3-8 Table 3.2 Plant Operating Range Supported by the LOCA Analysis J Event Operating Range 1.0 Plant Physical Description 1.1 Fuel a) Cladding outside diameter 0.424 in.

b) Cladding inside diameter 0.364 in.

c) Cladding thickness 0.030 in.

d) Pellet outside diameter 0.357 in.

e) Pellet density 95% of theoretical f) Active fuel length 144 in.

g) Resinter densification I I h) Gd2O3 concentrations 0, 2, 4, 8 w/o 1.2 RCS 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 15x15 e) SG tube plugging

• 10%

2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Nominal reactor power 2339 MWt b) F0

• 2.62a c) FAH 1. 8 b d) MTC

• 0 at HFP 2.2 Fluid Conditions a) Loop flow 97.3 Mlbm/hr S M

• 113 Mlbm/hr b) RCS average temperature 569.9

• T

• 581.90Fc c) Upper head temperature < Core Outlet Temperature Includes 5% measurement uncertainty and 3% manufacturing uncertainty.

b Include 4% measurement uncertainty.

c Sampled range of +/-61F includes both operational tolerance and measurement uncertainty.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-9 I ~'  ; i Table 3.2 Plant Operating Range Supported by the LOCA Analysis (Continued) d) Pressurizer pressure 2200

• P

• 2300 psia3 e) Pressurizer level 43.3%

• L

• 63.3%

f) Accumulator pressure 615

• P

• 675 psia g) Accumulator liquid volume 825 5 V

• 841 ft3 h) Accumulator temperature 80

• T

• 130OF (coupled to containment temperature) i) Accumulator fUD As-Built piping configuration j) Minimum ECCS boron 2 1950 ppm 3.0 Accident Boundary Conditions a) Break location Any RCS piping location b) Break type Double-ended guillotine or split c) Break size (each side, relative to 0.05

• A

• 0.5 full pipe area (split) cold leg pipe) 0.5

• A

• 1.0 full pipe area (guillotine) d) Worst single-failure Loss of one LPSI and one HPSI e) Offsite power On or Off f) LPSI flow Bounding minimum of current pump deliveryb g) HPSI flow Bounding minimum of current pump deliveryc h) Safety injection temperature

• 100OF i) HPSI delay

• 20.5 (wI offsite power)

• 40 (w/o offsite power) j) LPSI delay

• 29 (wI offsite power)

• 44 (w/o offsite power) k) Containment pressure Bounding current configuration I) Containment temperature 80

• T

• 1300 F m) Containment sprays delay 2 0 seconds a Based on representative plant data, including measurement uncertainty.

b Per the direction given in Reference 4, the LPSI pump delivery curve provided in Reference 5 was degraded by 5% and used in the subject analysis. Flow spilts were calculated by S-RELAP5.

C Total HPSI pump delivery the one-pump HPSI curve provided in Reference 5, with uniform delivery to each loop.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-1 0 Table 3.3 Statistical Distributions Used for Process Parameters Operational Measurement Uncertainty Uncertainty Standard Parameter Distribution Parameter Range Distribution Deviation Core Power (%) Uniform 99.5 - 100.5 Normal 0.3 Pressurizer Pressure (psia) Uniform 2220 - 2300 N/A NIA Pressurizer Level (%) Uniform 43.3 - 63.3 N/A N/A Accumulator Liquid Volume (ft3) Uniform 825 - 841 N/A N/A Accumulator Pressure (psia) Uniform 615 - 675 N/A N/A Containment/Accumulator Uniform 80- 130 N/A N/A Temperature (TF)

Containment Volume (x106 ft3) Uniform 1.95 - 2.23 N/A N/A Initial Flow Rate (Mlbm/hr) Uniform 97.3 - 113 N/A N/A Initial Operating Temperature (TF) Uniform 569.9 - 581.9 N/A N/A RWST Temperature (OF) Point 100 N/A N/A Offsite Power Availability' Binary 0,1 N/A N/A Delay for Containment Cooling (s) Point 0.0 N/A N/A HHSI Delay (s) Point 20.5 (w/ offsite power) N/A N/A LHSI___Delay__(s) __Point___2940 (w/o offsite power) N/A N/A LHSl Delay (s) Point 29 (w/ offsite power) N/A N/A

_____ _____ __ _ _____ ____ 44 (w/o offsite pow er) _ _ _ _ _ _ _ __ _ _ _ _ _ _ _

a No data is available to quantify the availability of offsite power. During normal operation, offsite power is available. Since the loss of offsite power is typically more conservative (some loss in coolant pump capacity), it is assumed that there is a 50% probability the offsite power is unavailable.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-11 Table 3.4 SER Conditions and Limitations SER Conditions and Limitations I *Response

1. A CCFL violation warning will be added to alert the There was no significant occurrence of CCFL violation in analyst to CCFL violation in the downcomer should the downcomer for this analysis. Occurrences of CCFL such occur. were noted in a statistically insignificant number of opportunities.

2. Framatome ANP has agreed that it is not to use Hot leg nozzle gaps were not modeled.

nodalization with hot leg to downcomer nozzle gaps.

3. If Framatome ANP applies the RLBLOCA The PLHGR is consistent with the 3-loop sample methodology to plants using a higher planar linear problem for the Reference I methodology. An end-of-life heat generation rate (PLHGR) than used In the calculation was not explicitly performed thus the need for current analysis, or if the methodology is to be a blowdown clad rupture model was not reevaluated.

applied to an end-of-life analysis for which the pin pressure is significantly higher, then the need for a blowdown clad rupture model will be reevaluated.

The evaluation may be based on relevant engineering 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 The evaluation of slot breaks Is documented in the FANP evaluated. These breaks could cause the loop seals RLBLOCA analysis guidelines.

to refill during late reflood and the core to uncover again. These break locations are an oxidation concern as opposed to a PCT concern since the top of the core can remain uncovered for extended periods of time. Should an analysis be performed for a plant with loop seals with bottom elevations that are below the top elevation of the core, Framatome ANP 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- The plant is a Westinghouse 3-loop plant.

and CE-designed nuclear steam systems.

6. The model applies to bottom reflood plants only The plant is a bottom reflood plant.

(cold side injection into the cold legs at the reactor coolant discharge piping).

7. The model is valid as long as blowdown quench Examination of the cases showed no evidence of does not occur. If blowdown quench occurs, blowdown quench.

additional justification for the blowdown heat transfer model and uncertainty are needed if 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.

8. The reflood model applies to bottom-up quench Core quench initiated at the bottom of the core and behavior. If a top-down quench occurs, the model is proceeded upward.

to be justified or corrected to remove top quench. A top-down quench is characterized by the quench front moving from the top to the bottom of the hot assembly.

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Robinson Nuclear Plant Revision 0 Realistic Lame Break LOCA Analysis Page 3-12 Table 3.4 SER Conditions and Limitations (Continued)

SER Conditions and Limitations [ Response

9. The model does not determine whether Criterion 5 Long term cooling was not evaluated in this analysis.

of 10 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 consistent with the plant-specific nodalization. Deviations from the Westinghouse 3-loop sample calculation that was reference plant must be addressed. submitted to the NRC for review. Figure 3.1 shows the loop noding used in this analysis. (Note only Loop I is shown in the figure; Loops 2 and 3 are identical to loop 1, except that only Loop I contains the pressurizer and the break.) Figure 3.2 shows the steam generator model.

Figures 3.3, 3.4, and 3-5 show RV noding diagrams.

11. A table that contains the plant-specific parameters Simulation of clad temperature response is a function of and the range of the values considered for the phenomenological correlations that have been derived selected parameter during the topical report either analytically or experimentally. The important approval process must be provided. When plant- correlations have been validated for the RLBLOCA specific parameters are outside the range used In methodology and a statement of the range of applicability demonstrating acceptable code performance, the has been documented. The correlations of interest are licensee or applicant will submit sensitivity studies to the set of heat transfer correlations as described in show the effects of that deviation. Reference 1. Table 3.7 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 Analysis results are discussed in Section 3.5.

methodology must submit the results of the plant-specific analyses, including the calculated worst break size, PCT, and local and total oxidation.

Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-13 Table 3.5 Summary of Results for the Limiting PCT Case Case # I 1 PCT Temperature 19520 F Time 29.6 seconds Elevation -9.4 ft Metal-Water Reaction

% Oxidation Maximum 1.82%

% Total Oxidation 0.04%

Total Hydrogen 0.68 lb Table 3.6 Calculated Event Times for the Limiting PCT Case Event Time (sec)

Break Opened 0.0 RCP Trip N/A SIAS Issued 0.6 Start of Broken Loop Accumulator Injection 6.5 Start of Intact Loop Accumulator Injection 10.4, 10.4 Start of HHSI 21.1 End of Bypass / Beginning of Refill 23 Beginning of Core Recovery (Beginning of Reflood) 29.1 LHSI Available 29.6 PCT Occurred 29.6 LHSI Delivery Began (to common header) 37 Broken Loop Accumulator Emptied 47.5 Intact Loop Accumulators Emptied 50.2, 47.9 (Loop 2 and 3 respectively)

Transient Calculation Terminated 414.8 Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Paae 3-14 Table 3.7 Heat Transfer Parameters for the Limiting Case' r

J a Values in brackets show full range of applicability. Phasic data is provided regardless of the amount of that phase present during the respective period.

[

]

Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-15 r

J Figure 3.1 Primary System Noding Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-16 r

i Figure 3.2 Secondary System Noding Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-17 F

J Figure 3.3 Reactor Vessel Noding Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-18 r

i Figure 3.4 Core Noding Detail Framatome ANP, Inc.

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Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-19 r

.1 Figure 3.5 Upper Plenum Noding Detail Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-20 Break  :  :

(f1) :1 Area .i .. Om 0.0 1.0 2.0 3.0 4.0 5.0 Bum  :

Time ,ggg * *-......... @ @ @

• soo (hrs)  :

0.0 5000.0 10000.0 15000.0 Core F '

Power

• 11e s *

(MW) [

2300.0 2310.0 2320.0 2330.0 2340.0 2350.0 2360.0 2370.0 2380.0 Fq - Ho FFI e oo_

ein m _a Peaking 2.0 2.2 2.4 2.6 2.8 ASI _-_S M We eeisew=*qSeMi

-0.4 -0.3 -0.2 -0.1 0.0 0.1 Pressurizer .

Pressure -em emee_

• _ e m,.0 23s0.0 (psia)l 2220.0 2240.0 2260.0 2280.0 2300.0 j

Pressurizer Level M seo _

• me_ *o -

(%)  :

40.0 45.0 50.0 55.0 60.0 65.0 70.0 RCS . .

Temperature

• m 01Me" em_

N*m . uERSOn (F) 570.0 575.0 580.0 585.0 Figure 3.6 Scatterplot of Operational Parameters Framatome ANP, Inc.

EM F-3030(N P)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Paqe 3-21 Total  :  :

LoopFlow - m *owe_ _- _

_

• 1 (Mlb/hr)  :

9.50e+07 1.OOe+08 1.05e+08 1.1 Oe+08 1.15e+08 Accumulator Volume _ mew.. _

825.0 830.0 835.0 840.0 845.0 Accumulator Pressure (psia) [ ***-a msso oo *"emme 600.0 620.0 640.0 660.0 680.0 Accumulator  :

Temperature _ _ _ ma_ *e. *09 _* ,

(F) 80.0 90.0 100.0 110.0 120.0 130.0 Figure 3.6 Scatterplot of Operational Parameters (Continued)

Framatome ANP. Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Larae Break LOCA Analysis Page 3-22 2000 0 . .P . . .]

1800 - El

- 4 11 1600

- w aa 1400 *1so N C A: M 1- 1200

• U 0 0

0 0L 1000 800

• Split Break II Il O Guillotine Break 600 I I I I 400 0 100 200 300 400 500 Time of PCT (s)

Figure 3.7 PCT vs. PCT Time Scatterplot From 59 Calculations Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Lame Break LOCA Analysis Page 3-23 2000 0

1800 ° 0 0

0 00 0 1600 D *0 1400 100

• Uf i- 1200 0

0c 1000 800 600 l *Split Break l 0 Guillotine Break 400

0. 0 1.0 2.0 3.0 4.0 5.0 Break Area (ft2)

Figure 3.8 PCT v. Break Size Scatterplot From 59 Calculations Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-24 2.0 . . . . . . . . . . . . . . .

X Split Break o o Guillotine Break 13 D

C 0

1.0 13

0 To

_ A__

0.0 _ ,_

. ~ 1 , ,

• 400 800 1200 1600 2000 PCT (F)

Figure 3.9 Maximum Oxidation vs. PCT Scatterplot From 59 Calculations Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-25 2200 2000 1800 1600 1400 El

. 1200 0L 1000 E

0) 800 600 400 200 Time (s)

Figure 3.10 Peak Cladding Temperature (Independent of Elevation) for Limiting Break Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-26 80000 70000 60000 l _Vesel-sde I Pumfp-side lll-_Total W 50000 I L1L II ID 40000 ai:

0 300001 C)

Ca) 20000 10000 0I

-10000 10 20 30 40 s0 60 Time (s)

Figure 3.11 Break Flow for the Limiting Break (Early)

Framatome ANP, Inc.

EM F-3030(N P)

Robinson Nuclear Plant Revision 0 Realistic Lame Break LOCA Analysis Page 3-27 300 250 200 150 100 50 a,

x Z, 0 en -50

-1 00

-150

-200

-250

-300 0 10 20 30 40 50 60 Time (s)

Figure 3.12 Core Inlet Mass Flux for Limiting Break (Early)

Framatome ANP, Inc.

EM F-3030(N P)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Paqe 3-28 300 250 200 r --- 0 Hot Asseffidy

..... a Con., Core 150 - -* Average Core -.

I *Wei Core 100 50 h

0 4z, f -:

-50

-100 j \ I I .11

-150

-200 _-

-250

-300t . . . . . I . . . I . . . . I . . . I I . I . . . .

o 10 20 30 40 50 60 Time (s)

Figure 3.13 Core Outlet Mass Flux for Limiting Break (Early)

Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Larae Break LOCA Analysis Page 3-29 C

0

-o IL 500 Time (s)

Figure 3.14 Void Fraction at RCS Pumps for Limiting Break Framatome ANP. Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-30 2000 I_ Loop I Broken)

.. ELoop 2(Intart)

} 1 -_ Loop 3 (ntaM) 1500

-p 0

cc 1000 5

0s (n . tl. . . . . I . . . . . . . . . I . . . . . . . . . I . . . . . .' . . I . .

500 0 100 200 300 400 500 Time (s)

Figure 3.15 ECCS Flows (Includes Accumulator, HPSI, and LPSI) for Limiting Break Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-31 2400 -

2200 -

2000 -

1800 1600 co 1400 2 1200 -

C,,

2- 1000 800 600 _

400 -

200 -

0O 0

Figure 3.16 Upper Plenum Pressure for Limiting Break (Early)

Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-32 30 25 20

-i 15

.5 10 5

0 Time (s)

Figure 3.17 Collapsed Liquid Level in the Downcomer for Limiting Break Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-33 14 12 10

-J

- 8 64o 2

0~

Time (s)

Figure 3.18 Collapsed Liquid Level in the Lower Plenum for the Limiting Break Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-34 16 14 12 In S-i 62

-6 4

2 0

0 100 200 300 400 500 Time (s)

Figure 3.19 Collapsed Liquid Level in the Core for the Limiting Break Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 3-35 80 l I _ Contalnmentl 70 I- SG-I Oullet I--4Upper PenumI

.nle.

.o.re 60 50 In hi 40 40 Vn V) 2 30 20 10 0

25 50 75 100 125 150 175 200 Time (s)

Figure 3.20 Containment and Loop Pressures for Limiting Break Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 4-1 4.0 Conclusions The results of the RLBLOCA analysis show that the limiting case has a PCT of 19520F and a maximum oxidation thickness and hydrogen generation that fall well within regulatory requirements.

The analysis supports operation at a nominal power level of 2339 MWt (plus uncertainty), a steam generator tube plugging level of up to 10% in any generator, a total peaking factor (FT) of 2.62 and a nuclear enthalpy rise factor (FAH) of 1.80 with no axially-dependent power peaking limit.

Framatome ANP, Inc.

EMF-3030(NP)

Robinson Nuclear Plant Revision 0 Realistic Large Break LOCA Analysis Page 5-1 5.0 References I

1. EMF-2103(P)(A) Revision 0, Realistic Large Break LOCA Methodology, Framatome ANP, Inc., April 2003.

2. Technical Program Group, Quantifying Reactor Safety Margins, NUREG/CR-5249, EGG-2552, October 1989.

3. 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.

4. E. J. Geyer (Progress Energy) letter to P. G. Newby (Framatome-ANP), "RNP Cycle 23 and RLBLOCA RHR Pump Performance, and RLBLOCA HPSI Delay," File NF-401.2604, Serial: NF-03A-0123, FANP Document: 38-5030411-00, July 14, 2003.

5. E. J. Geyer (Progress Energy) letter to P. G. Newby (Framatome-ANP), "Transmittal of Robinson Cycle 23 Plant Parameters Document," File NF-401.2604, Serial:

NF-03A-0106, FANP Document: 38-5027076-01, July 25, 2003.

Framatome ANP, Inc.

Robinson Nuclear Plant EMF-3030(NP)

Realistic Large Break LOCA Analysis Revision 0 Distribution N. F. Fausz, RC36 T. R. Lindquist, RC36 R. P. Martin, OF53 P. G. Newby, OF11 R. A. Shaw, OF53 S. M. Sloan, OF53 C. T. Stebbings, OF53 Framatome ANP, Inc.

United States Nuclear Regulatory Commission Attachment VI to Serial: RNP-RA/05-0006 48 pages including cover page H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING REVISION TO CORE OPERATING LIMITS REPORT (COLR) REFERENCES EMF-3030(P)

Robinson Nuclear Plant Realistic Large Break LOCA Analysis February 2004 Proprietary Version

United States Nuclear Regulatory Commission Attachment VII to Serial: RNP-RA/05-0006 4 pages including cover page H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING REVISION TO CORE OPERATING LIMITS REPORT (COLR) REFERENCES Framatome ANP Affidavit Regarding Proprietary Attachment VI

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing in Regulatory Affairs, for FANP, and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by FANP to determine whether certain FANP information is proprietary. I am familiar with the policies established by FANP to ensure the proper application of these criteria.

3. I am familiar with the reports EMF-3030, Revision 0, entitled "Robinson Nuclear Plant Realistic Large Break LOCA Analysis" and EMF-3031, Revision 1, entitled "Harris Nuclear Plant Realistic Large Break LOCA Analysis" and referred to herein as "Documents."

Information contained in these Documents have been classified by FANP as proprietary in accordance with the policies established by FANP for the control and protection of proprietary and confidential information.

4. These Documents contain information of a proprietary and confidential nature and is of the type customarily held in confidence by FANP and not made available to the public.

Based on my experience, I am aware that other companies regard information of the kind contained in these Documents as proprietary and confidential.

5. These Documents have been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in these Documents be withheld from public disclosure.

6. The following criteria are customarily applied by FANP to determine whether information should be classified as proprietary:

(a) The information reveals details of FANP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for FANP.,

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for FANP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by FANP, would be helpful to competitors to FANP, and would likely cause substantial harm to the competitive position of FANP.

7. In accordance with FANP's policies governing the protection and control of information, proprietary information contained in these Documents have been made available, on a limited basis, to others outside FANP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. FANP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this 4 day of J5 2004.

Ella F. Carr-Payne NOTARY PUBLIC, STATE OF VIRGINIA MY COMMISSION EXPIRES: 8/31/05