ML072260534

BAW-2501(NP), Rev. 2, Palisades Nuclear Plant Realistic Large Break LOCA Summary Report.

ML072260534
Person / Time
Site:	Palisades
Issue date:	06/30/2007
From:	Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation
References
TAC MD3492 BAW-2501(NP), Rev 2
Download: ML072260534 (50)

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REALISTIC LARGE BREAK LOCA AREVA NP NON-PROPRIETARY REPORT 49 Pages Follow

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Revision 2 Palisades Nuclear Plant Realistic Large Break LOCA Summary Report June 2007

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page i Customer Disclaimer Important Notice Regarding the Contents and Use of This Document Please Read Carefully AREVA NP Inc.'s warranties and representations concerning the subject matter of this document are those set forth in the agreement between AREVA NP Inc. and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither AREVA NP 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 in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or

b. assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document.

The information contained herein is for the sole use of the Customer.

In order to avoid impairment of rights of AREVA NP Inc. in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writing by AREVA NP Inc. or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document.

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Palisades Nuclear Plant Revision 2

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• O Nature of Changes Revision Page Description 0 All This is a new document 1 All Revision 1 supersedes revision 0 in its entirety.

2 Section 3.4 and Revision 2 adds justification for one Item 7 in Table 3.4 case of blowdown quench.

Revision 2 supersedes revision 1 in its entirety.

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page iii Contents 1 .0 In tro d u c tio n ..................................................................................................................... 1-1 2 .0 S u m m a ry ........................................................................................................................ 2-1 3 .0 A n a ly s is .......................................................................................................................... 3-1 3.1 Description of the LBLO CA Event ...................................................................... 3-1 3.2 Description of Analytical Models ......................................................................... 3-3 3.3 Plant Description and Sum mary of Analysis Parameters ................................... 3-5 3.4 SER Com pliance ................................................................................................ 3-7 3.5 Mixed-Core Considerations ................................................................................ 3-8 3.6 Realistic Large Break LOCA Results .................................................................. 3-8 4.0 Conclusions .................................................................................................................... 4-1 5.0 References ..................................................................................................................... 5-1

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page iv Tables Table 2.1 Summary of Major Parameters for the Limiting PCT Case ....................................... 2-1 Table 3.1 Sam pled LB LO CA Param eters ................................................................................. 3-9 Table 3.2 Plant Operating Range Supported by the LOCA Analysis ...................................... 3-10 Table 3.3 Statistical Distributions Used for Process Parameters ............................................ 3-12 Table 3.4 SE R Conditions and Lim itations .............................................................................. 3-13 Table 3.5 Summary of Hot Rod Limiting PCT Results ............................................................ 3-15 Table 3.6 Calculated Event Times for the Limiting PCT Case ................................................ 3-15 Table 3.7 Heat Transfer Parameters for the Limiting Case ..................................................... 3-16

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page v Figures Figure 3.1 Primary System Noding ......................................................................................... 3-17 Figure 3.2 Secondary System Noding .................................................................................... 3-18 Figure 3.3 Reactor Vessel Noding .......................................................................................... 3-19 Figure 3.4 Core Noding Detail ................................................................................................. 3-20 Figure 3.5 Upper Plenum Noding Detail ................................................................................. 3-21 Figure 3.6 S-RELAP5 Containment Pressure versus Best-Estimate R e s u lt ........................................................................................................................... 3-2 2 Figure 3.7 Scatter Plot of Operational Parameters ................................................................. 3-23 Figure 3.8 PCT versus PCT Time Scatter Plot from 59 Calculations ...................................... 3-24 Figure 3.9 PCT versus Break Size Scatter Plot from 59 Calculations .................................... 3-25 Plot from 59 Figure 3.10 Maximum Oxidation versus PCT Scatter C a lc u la tio n s .................................................................................................................. 3 -2 6 Figure 3.11 Peak Cladding Temperature (Independent of Elevation) for the Lim iting C a se .......................................................................................................... 3 -2 7 Figure 3.12 Break Flow for the Limiting Case ......................................................................... 3-28 Figure 3.13 Core Inlet Mass Flux for the Limiting Case ............................... ...................... 3-29 Figure 3.14 Core Outlet Mass Flux for the Limiting Case ....................................................... 3-30 Figure 3.15 Void Fraction at RCS Pumps for the Limiting Case ............................................. 3-31 Figure 3.16 ECCS Flows (Includes SIT, HPSI and LPSI) for the Limiting C a s e ............................................................................................................................. 3 -3 2 Figure 3.17 Upper Plenum Pressure for the Limiting Case ..................................................... 3-33 Figure 3.18 Collapsed Liquid Level in the Downcomer for the Limiting C a s e ............................................................................................................................. 3-3 4 Figure 3.19 Collapsed Liquid Level in the Lower Plenum for the Limiting C a s e ............................................................................................................................. 3-3 5 Figure 3.20 Collapsed Liquid Level in the Core for the Lim iting Case .................................... 3-36 Figure 3.21 Containment and Loop Pressures for the Limiting Case ..................................... 3-37

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Palisades Nuclear Plan Revision 2 D H In D.- I nr'A Q 1 U '0IIaL , alu I4 LI 0 O .. .l.I1111101 U.f V I V0LJIJ L r u VI Nomenclature ASI Axial Shape Index CCFL Counter Current Flow Limit CE Combustion Engineering, Inc.

CFR Code of Federal Regulations CHF Critical Heat Flux CL Cold Leg CSAU Code Scaling, Applicability and Uncertainty DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EM Evaluation Model FrT Total Radial Peaking Factor FSAR Final Safety Analysis Report HFP Hot Full Power HPSI High Pressure Safety Injection LBLOCA Large Break Loss-of-Coolant Accident LHR/LHGR Linear Heat Rate/Linear Heat Generation Rate LOCA Loss-of-Coolant Accident LPSI Low Pressure Safety Injection MOV Motor Operated Valve MSIV Main Steam Isolation Valve MTC Moderator Temperature Coefficient NMC Nuclear Management Company, LLC NRC U. S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PCS Primary Coolant System PCT Peak Clad Temperature PIRT Phenomena Identification and Ranking Table PWR Pressurized Water Reactor

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Reoort Page vii Nomenclature (Continued)

RCP Reactor Coolant Pump RLBLOCA Realistic Large Break LOCA RV Reactor Vessel SBLOCA Small Break Loss-of-Coolant Accident SER .Safety Evaluation Report SG Steam Generator SIAS Safety Injection Actuation Signal SIRWT Safety Injection and Refueling Water Tank SIT Safety Injection Tank

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 1-1 1.0 Introduction This report describes and provides results from a RLBLOCA analysis for the Palisades nuclear plant. The plant is a CE-designed 2,565.4 MWt PWR plant with a large dry containment. AREVA NP is the current fuel supplier. The plant is a 2x4-loop design-two hot legs and four cold legs. The loops contain four RCPs, two U-tube steam generators and a pressurizer. The ECCS is provided by two independent safety injection trains and four SITs.

The analysis herein supports operation for Cycle 18 and beyond with Zr-4 clad fuel, unless invalidated by changes in Technical Specifications, Core Operating Limits Report, core design, fuel design, plant hardware or plant operation. The reanalysis represents a large break LOCA methodology change (from deterministic to realistic), not a fuel design change. The core contains 204 AREVA NP 15x15 fuel assemblies with Zr-4 cladding. The analysis was performed in compliance with the NRC-approved AREVA NP RLBLOCA EM (Reference 1). Analysis results confirm that the 10CFR50.46(b) acceptance criteria presented in Section 3.0 are met and serve as the basis for operation of the Palisades Nuclear Plant with AREVA NP fuel.

The non-parametric statistical methods inherent to the AREVA NP RLBLOCA methodology provide for consideration of a full spectrum of break sizes, break configuration (guillotine or split break), axial power shapes, and plant operational parameters. A conservative single-failure assumption is applied in which the negative effects of the loss of a train of ECCS pumped injection is simulated.

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.

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Palisades Nuclear Plant Revision 2 Realistic Larqe Break LOCA Summary Report Page 2-1 2.0 Summary The limiting PCT is 1,751 OF; for a U0 2 rod. Gadolinia-bearing rods of 2 w/o and 6 w/o Gd 20 3 were also analyzed, but were not limiting. This RLBLOCA result is based on a case set comprised of 59 individual transient cases. The core is composed of only AREVA NP 15x15 fuel; hence, from the standpoint of LBLOCA analyses, no consideration of co-resident fuel (mixed core) is necessary.

Table 2.1 gives the analysis parameters for the limiting (95/95) PCT case.

The analysis assumed full-power operation at 2,565.4 MWt (plus uncertainties), a steam generator tube plugging level of 15 percent in both steam generators, a total LHR of 15.28 kW/ft (technical specification value including uncertainties, with no axial dependency), and an FrT of 2.04 (including uncertainty). The analysis addresses typical operational ranges or technical specification limits (whichever are applicable) with regard to pressurizer pressure and liquid level; SIT pressure, temperature (set to containment temperature) and liquid level; core inlet temperature; core flow; containment pressure and temperature; and SIRWT temperature.

The AREVA NP RLBLOCA methodology explicitly analyzes only fresh fuel assemblies (Reference 1, Appendix B). Previous analyses showed that once-and twice-burnt fuel is not limiting up to peak rod average exposures of 62,000 MWd/MTU. The analysis demonstrates that the 10CFR50.46(b) criteria listed in Section 3.0 are satisfied.

Table 2.1 Summary of Major Parameters for the Limiting PCT Case U0 2 Core Average Burnup (EFPH) 4,110.6 Core Power (MWt) 2,580.4 Hot Rod LHR, kW/ft 14.68 Total Hot Rod Radial Peak (FrT) 2.040 ASI -0.1611 Break Type Guillotine Break Size (ft2/side) 3.573 Offsite Power Availability Not Available Decay Heat Multiplier 1.0179

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 3-1 3.0 Analysis The purpose of the analysis is to verify the adequacy of the ECCS for the planned Cycle 18 plant configuration by demonstrating that the following criteria of 10CFR 50.46(b) are met:

The calculated maximum fuel element cladding temperature shall not exceed 2,200 '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.

Calculated changes in core geometry shall be such that the core remains amenable to cooling. The RLBLOCA analysis conservatively considers blockage effects due to clad swelling and rupture in the prediction of the hot fuel rod PCT. Therefore, compliance with Criterion 1, demonstrating that the PCT is less than 2,200 F, assures that the core remains amenable to cooling and satisfies Criterion 4.

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 2x4-loop PWR plant and summarizes the system parameters used in the analysis.

Compliance with the RLBLOCA evaluation model SER is addressed in Section 3.4. Section 3.5 addresses the mixed core. Section 3.6 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 PCS piping. Based on deterministic studies, the worst break location is in the cold leg piping between the RCP and the RV for the PCS loop containing the pressurizer. The break initiates a rapid depressurization of the PCS. 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

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 3-2 rapid primary system 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 fission ends. As heat transfer from the fuel rods is reduced, the cladding temperature rises.

Coolant in all regions of the PCS 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 RCPs 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 SIAS is tripped. This signal is initiated by either high containment pressure or low pressurizer pressure.

Regulations require that a worst active single-failure be considered for ECCS safety analysis. This worst active single failure was determined generically in the RLBLOCA evaluation model to be the loss of one ECCS train. The AREVA NP RLBLOCA methodology conservatively assumes a minimal time delay and a normal (no failure irrespective of the assumed worst single active failure) lineup of the containment sprays and fan coolers to reduce containment pressure and increase break flow. The analysis assumes that one HPSI pump, one LPSI pump, all containment spray pumps and all containment fan coolers are operational.

When the PCS 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 cause some of this fluid to bypass the core. During this bypass period, core heat transfer remains poor and fuel rod cladding temperatures increase. As PCS and containment pressures equilibrate, ECCS water begins to fill the lower plenum and eventually the lower portions of the core. This improves core heat transfer and cladding temperatures begin to decrease.

Eventually, the relatively large volume of SIT water is exhausted and core recovery relies solely on ECCS pumped injection. As the SITs empty, the nitrogen gas used to 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 is 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

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 3-3 quenching in the lower portions of the core. Peak fuel rod cladding temperatures may increase for a short period until additional 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 RCP before it is vented out the break. The resistance of this flow path to the steam flow (including steam binding effects) is balanced by the driving force of water filling the downcomer. This resistance (steam binding) may act to retard the progression of core reflooding and postpone core-wide cooling. Eventually (within a few minutes of the accident),

core reflooding 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 topical report EMF-2103, Realistic Large Break LOCA Methodology (Reference 1). The methodology follows the 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 LBLOCA analysis.

The RLBLOCA methodology uses 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, including the containment pressure response.

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 another 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 phenomenon expected during an 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

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 3-4 components exist to represent specific components such as the pumps or the steam generator separators. All geometries are modeled at a level of detail 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. 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 technical specifications 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.

Transient 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, many benchmarks with cladding temperatures ranging from 1,700 OF (or less) to above 2,200 OF. 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 (Figures 3.3 through 3.5) based on specific inputs supplied by NMC. 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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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 3-5

1. Base Plant Input File Development First, RODEX3A and S-RELAP5 base input files for the plant (including a containment input file) are developed. Code input development guidelines are followed to ensure that the model nodalization is consistent with that 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 percent probability level with 95 percent confidence (95/95). Total oxidation and total hydrogen generation are based on the 95/95 PCT case. The adequacy of the ECCS is demonstrated when these results satisfy the regulatory criteria set forth in Section 3.0.

3.3 Plant Descriptionand Summary of Analysis Parameters The plant analysis presented herein is for a CE-designed PWR, which has a 2x4-loop arrangement. There are two hot legs each with a U-tube steam generator and four cold legs each with a RCP 1 . The PCS also includes one pressurizer connected to a hot leg. The core contains 204 15x15 AREVA NP fuel assemblies. The ECCS includes four SIT lines, each connecting to a cold leg pipe downstream of the pump discharge. The HPSI and LPSI lines tee into the SIT lines prior to their connection to the cold legs. The ECCS HPSI pumps are cross-connected. The single failure assumption renders one LPSI pump, two LPSI injection MOVs, and a HPSI pump inoperable. This results in one LPSI pump injecting through two valves into cold legs 1A (leg containing the break) and 1B, and one HPSI pump injecting through four valves in all four of the cold legs. This models the break in the same loop as the pressurizer, as directed by the RLBLOCA methodology. The RLBLOCA transients are of sufficiently short The RCP are Byron-Jackson Type DFSS pumps as specified by NMC. 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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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Paqe 3-6 duration that the switchover to sump cooling water for ECCS pumped injection need not be considered.

The S-RELAP5 model explicitly describes the PCS, RV, pressurizer, and the ECCS. The model also describes the steam generator secondary side that is instantaneously isolated (closed MSIV and feedwater trip) at the time of the break. A steam generator tube plugging level of up to 15 percent per steam generator is assumed.

Plant input modeling parameters were provided by NMC specifically for Palisades. NMC maintains plant documentation, and directly communicates with AREVA NP on plant design and operational issues regarding reload cores. NMC and AREVA NP have ongoing processes that assure the ranges and values of input parameters for the Palisades RLBLOCA analysis bound those of the as-operated plant values.

As described in the AREVA NP RLBLOCA methodology, many parameters associated with LBLOCA phenomenological uncertainties and plant operation ranges are sampled. A list of the sampled 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 analyses. Table 3.3 presents a summary of the uncertainties used in the analyses. Two parameters, SIRWT temperature for ECCS pumped injection flows and diesel start time, are set at conservative bounding values for all calculations. Where applicable, the sampled parameter ranges are based on technical specification limits. Plant and design data are used to define range boundaries for some parameters, for example, loop flow and containment temperature.

For the AREVA NP RLBLOCA evaluation model, significant 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 6-1 (Reference 4) 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 is indirectly ranged by sampling the containment volume (Table 3.3). The containment-related technical specification minimum SIRWT temperature is used for the building sprays. A Palisades-specific [ ] Uchida heat transfer coefficient multiplier was established through application of the process used in the RLBLOCA EM (Reference 1) sample problems. The comparison, shown in Figure 3.6, is within the RLBLOCA guidelines acceptance criterion and validates the acceptability of the Palisades S-RELAP5 containment model using a [ ] Uchida multiplier.

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 3-7 3.4 SER Compliance The SER on the RLBLOCA evaluation model stipulates a number of requirements (Reference 1). The application reported herein complies with all SER requirements. The requirements are addressed in Table 3.4.

Since a non-limiting PCT case exhibited a blowdown quench (SER Item 7), a discussion and justification of this aspect is provided. Relevant parameters for the blowdown quench case along with the limiting case are presented in the table below.

PCT Tmin Break Peak Case PCT PCT time Sampled Break Size LHGR Offs

(°F) Elevation (sec) (K) Type (ft2/side) Sampled pow

______ _ _____(kw/ft)

____ ___

13 Node 36 Guillotine No (limiting) 1751 of 52 (in 27.2 650.6 (double 3.57 14.67 availe top half) sided) 32 890 Node 43 5.9 680.8 split 1.49 13.4 availE of 52 The applicable features for the case that exhibited a quench of the PCT node before the end of blowdown are:

- relatively small break area,

- relatively high Tmin,

- offsite power continues to be available to power Reactor Coolant Pumps, and

- relatively low Linear Heat Generation Rate (LHGR).

The case with blowdown quench is a split break with an area of 1.49 ft 2 (relative to a full cold leg piping flow area of 4.909 ft 2 for each of two sides). The limiting case (a double-ended guillotine break) did not exhibit a blowdown quench.

Mechanistically, the observed quench occurs because the small break area limits break flow. This reduces the rates at which pressure and flow decrease at the PCT location compared with the limiting case. The void fraction calculated at the PCT location indicates significantly more liquid is available for cooling in the case with blowdown quench. The continued operation of the Reactor Coolant Pumps also provides increased forced convection cooling. The resulting combination cools the cladding sufficiently to enable a return to nucleate boiling.

A factor contributing to the occurrence of blowdown quench in this case is the relatively high value of 680'K for Tmin (as a result of random sampling). When Tw < Tmin, the heat transfer mode is selected based on the larger of the heat

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Pame 3-8 fluxes from transition and film boiling and so a relatively high value of Tmin facilitates the transition to nucleate boiling.

It is therefore concluded that the predicted blowdown quench is appropriate for this non-limiting case and also that this behavior is not applicable in any way to the limiting case.

3.5 Mixed-Core Considerations The Palisades core model contains 204 15x15 AREVA NP fuel assemblies. All fuel assembly cages are similar in design. Hence, due to the homogenous nature of the core fuel assemblies, no mixed-core evaluation need be done and no mixed-core penalty need be applied to the LBLOCA analysis.

3.6 Realistic Large Break LOCA Results A case set comprising 59 transient calculations was performed sampling the parameters listed in Table 3.1. For each transient calculation, PCT was calculated for a U0 2 rod and for gadolinia-bearing rods with concentrations of 2 w/o and 6 w/o Gd 2 0 3 . The limiting PCT (1,751 OF) occurred in Case 13 for a U0 2 rod. The major parameters for the limiting transient are presented in Table 2.1. Table 3.5 lists the limiting PCT results for the hot fuel rod. The fraction of total hydrogen generated is conservatively bounded by the calculated total percent oxidation, which is well below the 1 percent limit. A nominal 50/50 PCT case, based on the U0 2 hot rod, was identified as Case 20. The nominal PCT is 1,369 OF. This result can be used to quantify the relative conservatism in the 95/95 result; in this analysis, it is 382 OF.

The hot fuel rod results are given in Table 3.5 and event times for the limiting PCT case are shown in Table 3.6, respectively. Figure 3.7 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 sample ranges used in the analysis. 2 Figures 3.8 and 3.9 are PCT scatter plots versus the time of PCT and versus break size 3 from the 59 calculations, respectively. Figure 3.10 shows the maximum oxidation versus PCT for the 59 calculations. Figures 3.11 through 3.21 present transient results for key parameters from the S-RELAP5 limiting case. Figure 3.11 is a PCT elevation-independent plot; this figure clearly indicates that the transient exhibits a sustained and stable quench.

2 Figure 3.7, also Figure 3.9, presents the break flow area for only one break flow junction; total break flow area is the sum of the break flow areas from both break flow junctions (see break modeling in Figure 3.1).

3 The RLBLOCA approval provides for break size ranging down to 10 percent of the pipe cross-sectional area. Case set results were examined for the occurrence of phenomena characteristic of small break LOCA (loop seals, periods of natural circulation cooling, no rapid DNB immediately after transient initiation, etc.). The smallest break in the case set showed complete core voiding during blowdown and core refilling after the start of SIT injection-all LBLOCA characteristics. No characteristics exclusive to SBLOCA were observed in the Palisades case set results.

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 3-9 Table 3.1 Sampled LBLOCA Parameters Phenomenological Time in cycle (peaking factors, axial shape, rod properties and burnup)

Break type (guillotine versus 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 Trin (intersection of film and transition boiling)

Initial stored energy Downcomer hot wall effects Steam generator interfacial drag Condensation interphase heat transfer Metal-water reaction 4

Plant Offsite power availability Core power and power distribution Pressurizer pressure Pressurizer liquid level SIT pressure SIT liquid level SIT temperature (based on containment temperature)

Containment temperature Containment volume Initial flow rate Initial operating temperature Diesel start (for loss of offsite power only) 4 Uncertainties for plant parameters are based on plant-specific values with the exception of "Offsite power availability," which is a binary result that is specified by the analysis methodology.

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Palisades Nuclear Plant Revision 2 Panlictir" I rr a Rrank I Or.A qi mmnr Parnnrt P~n a -1 n 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 0.417 in b) Cladding inside diameter 0.367 in c) Cladding thickness 0.025 in d) Pellet outside diameter 0.360 in e) Pellet density 95.85% of theoretical f) Active fuel length 132.6 in g) Resinter densification [ I h) Gd 20 3 concentrations 2 and 6 w/o 1.2 RCS a) Flow resistance Analysis considers plant-specific form and friction losses b) Pressurizer location Analysis assumes location giving most limiting PCT (broken loop) c) Hot assembly location Anywhere in core d) Hot assembly type 15x15 AREVA NP e) SG tube plugging 15%

2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Nominal reactor power 2,565.4 MWt 5

b) LHR < 15.28 kW/ft c) FrT *2.046 2.2 Fluid Conditions a) Loop flow 130 Mlbm/hr*_< M* 145 Mlbm/hr b) PCS inlet core temperature 537 < T < 544 OF7 c) Upper head temperature < core outlet temperature d) Pressurizer pressure 2,010*5 P *2,100 psia 8 e) Pressurizer liquid level 46.25% < L < 67.8%

f) SIT pressure 214.7 < P < 239.7 psia3 g) SIT liquid volume 1,040!< V!_ 1,176 ft h) SIT temperature 80 < T 5 140 OF (coupled to containment temperature) i) SIT fL/D As-built piping configuration j) Minimum ECCS boron Ž 1,720 ppm 5 Includes a 5% local LHR measurement uncertainty, a 3% engineering uncertainty and a 0.5925% thermal power measurement uncertainty.

Includes a 4.25% measurement uncertainty.

7 Sampled range of +7 OF includes both operational tolerance and measurement uncertainty.

8 Based on representative plant values, including measurement uncertainty.

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Palisades Nuclear Plant Revision 2 R*.*li.tic. 1 *rn*. Br*k LOflA Siimrnrv Rp~nnrt Dý ý1-1 1 Table 3.2 Plant Operating Range Supported by the LOCA Analysis (Continued)

Event Operating Range 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 CL 0.05!< A <ý0.5 full pipe area (split) pipe) 0.5 <ýA!< 1.0 full pipe area (guillotine) d) Worst single-failure Loss of one ECCS pumped injection train e) Offsite power On or Off f) LPSI flow Minimum flow g) HPSI flow Minimum flow h) ECCS pumped injection temperature 100 OF 30 (w/ offsite power) i) HPSI delay time 40 seconds (w/o offsite power) j) d30 (w/ offsite power)

)LPSI delay time 40 seconds (w/o offsite power) k) Containment pressure 14.7 psia, nominal value I) Containment temperature 80 < T < 140 OF m) Containment spray/fan cooler delays 0/0 seconds

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Palisades Nuclear Plant Revision 2 Pcliktir. I *rn* Rr~~k I NC)A Siimm~rx, R~nnrt P~n ."-19 Table 3.3 Statistical Distributions Used for Process Parameters Operational Measurement Standard Parameter Uncertainty Parameter Range Uncertainty Deviation Distribution Distribution Core Power Operation (%) Uniform 99.5- 100.5 Normal 0.5925 Pressurizer Pressure (psia) Uniform 2,010 - 2,100 N/A N/A Pressurizer Liquid Level (%) Uniform 46.25 - 67.8 N/A N/A SIT Liquid Volume (ft 3 ) Uniform 1,040 - 1,176 N/A N/A SIT Pressure (psia) Uniform 214.7 - 239.7 N/A N/A Containment/SIT Temperature ("F) Uniform 80- 140 N/A N/A Containment Volume' (xl0 ft3) Uniform 1.64-1.80 N/A N/A Initial Flow Rate (Mlbm/hr) Uniform 130- 145 N/A N/A Initial Operating Temperature (oF) Uniform 537 - 544 N/A N/A SIRWT Temperature (oF) Point 100 N/A N/A Offsite Power Availability10 Binary 0,1 N/A N/A Delay for Containment Sprays (s) Point 0 N/A N/A Delay for Containment Fan Point 0 N/A N/A Coolers (s)

HPSI Delay (s) Point 30 (w/ offsite power) N/A 40 (w/o offsite power)

LPSI Delay (s) Point 30 (w/ offsite power) N/A N/A

________________________ _________ 40 (w/o offsite power)N/

9 Uniform distribution for parameter with demonstrated PCT importance conservatively produces a wider variation of PCT results relative to a normal distribution. Treatment consistent with approved RLBLOCA evaluation model (Reference 1, Section 4.3.3.2.12).

10 No data are 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 (loss in coolant pump capacity), it is assumed that there is a 50 percent probability the offsite power is unavailable.

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Pane 3-13 Table 3.4 SER Conditions and Limitations SER Conditions and Limitations Response

1. A CCFL violation warning will be added to alert the There was no significant occurrence of CEL violations analyst to a CCFL violation in the downcomer in the downcomer for this analysis.

should such occur.

2. AREVA NP has agreed that it is not to use nodalization with hot leg to downcomer nozzle Hot leg nozzle gaps were not modeled.

gaps.

3. IfAREVA NP applies the RLBLOCA methodology to plants using a higher planar linear heat generation rate (PLHGR) than used in the current analysis, or if the methodology is to be applied to The PLHGR for Palisades is lower than the defined limit an end-of-life analysis for which the pin pressure is for the RLBLOCA EM (Reference 1). An end-of-life significantly higher, then the need for a blowdown calculation was not performed; thus, the need for a clad rupture model will be reevaluated. The evaluation may be based on relevant engineering blowdown cladding rupture model was not reevaluated.

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. These breaks could cause the loop seals 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 This is not applicable to the Palisades plant because it performed for a plant with loop seals with bottom does not have "deep loop seals."

elevations 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- The RLBLOCA evaluation model is applicable to the The

a. nPalisades plant since it is a CE-designed 2x4-loop and CE-designed nuclear steam systems. plant.

6. The model applies to bottom reflood plants only The RLBLOCA evaluation model is applicable to the (cold side injection into the cold legs at the reactor Palisades plant since it is a bottom reflood plant.

coolant discharge piping).

7. The model is valid as long as blowdown quench does not occur. Ifblowdown quench occurs, additional justification for the blowdown heat The limiting PCT case showed no evidence of transfer model and uncertainty are needed or the blowdown quench. Blowdown quench was observed in run corrected. A blowdown quench is one (single) other case. An explanation is provided in characterized by a temperature reduction of the Section 3.4.

peak cladding temperature (PCT) node to saturation temperature during the blowdown period.

8. The reflood model applies to bottom-up quench behavior. Ifa top-down quench occurs, the model Examination of the case set showed that core quench is to be justified or corrected to remove top quench. Examiatio of the case sethe auench core A top-down quench is characterized by the quench initiated at the bottom of the core and proceeded front moving from the top to the bottom of the hot upward.

assembly.

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Palisades Nuclear Plant Revision 2 Realistic Larqe Break LOCA Summary Report Page 3-14 Table 3.4 SER Conditions and Limitations (Continued)

SER Conditions and Limitations Response

9. The model does not determine whether Criterion 5 of 10CFR50.46, long-term cooling, has been satisfied. This will be determined by each Long-term cooling was not evaluated herein.

applicant or licensee as part of its application of this methodology.

The Palisades plant model nodalization is consistent

10. Specific guidelines must be used to develop the with the sample calculations given in the RLBLOCA plant-specific nodalization. Deviations from the evaluation model (Reference 1). Figure 3.1 shows the plaerenc-specificnodatinbadevtse. floop noding used in the analysis. Figure 3.2 shows the reference plant must be addressed. steam generator model. Figures 3.3, 3.4 and 3.5 show RV nodinq diagrams.

11. A table that contains the plant-specific parameters and the range of the values considered for the Table 3.7 presents the summary of the full range of selected parameter during the topical report applicability for the important heat transfer correlations, approval process must be provided. When as well as the ranges calculated in the limiting analysis plant-specific parameters are outside the range case. Calculated values for other parameters of used in demonstrating acceptable code interest are also provided. As is evident, the performance, the licensee or applicant will submit plant-specific parameters fall within the applicability sensitivity studies to show the effects of that range of the methodology.

deviation.

12. The licensee or applicant using the approved methodology must submit the results of the Analysis results are presented in Section 3.6.

plant-specific analyses, including the calculated worst break size, PCT and local and total oxidation.

13. Applicants or licensees wishing to apply the AREVA NP realistic large break loss-of-coolant accident (RLBLOCA) methodology to M5 clad fuel This is not applicable, the cladding material is Zr-4.

must request an exemption for its use until the planned rulemaking to modify 10CFR50.46(a)(i) to include M5 cladding material has been completed. I

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Palisades Nuclear Plant Revision 2 Realistic Larqe Break LOCA Summary Report Paae 3-15 Table 3.5 Summary of Hot Rod Limiting PCT Results 15 x 15 AREVA NP Fuel Type U0 2 Case Number 13 PCT Temperature 1,751 OF Time 27.2 s Elevation 7.748 ft Metal-Water Reaction Oxidation Maximum 0.87%

Total Oxidation 0.02%

Table 3.6 Calculated Event Times for the Limiting PCT Case Event Time (sec)

Break Opened 0 RCP Trip 0 SIAS Issued 0.6 Start of Broken Loop SIT Injection 13.9 Start of Intact Loop SIT Injection 16, 16, 16 Beginning of Core Recovery (Beginning of Reflood) 25.8 PCT Occurred 27.2 Start of HPSI 40.6 LPSI Available 40.6 Broken Loop LPSI Delivery Began 40.6 Intact Loop LPSI Delivery Began (loops 1B, 2A and 2B, respectively) 40.6, N/A, N/A Broken Loop HPSI Delivery Began 40.6 Intact Loop HPSI Delivery Began (loops 1B, 2A and 2B, respectively) 40.6, 40.6, 40.6 Broken Loop SIT Emptied 50.9 Intact Loop SIT Emptied (loops 1B, 2A and 2B, respectively) 50.9, 54.7, 53.3 Transient Calculation Terminated 300

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Report P;an 3-16 Table 3.7 Heat Transfer Parameters for the Limiting Case"1 Values in brackets show full range of applicability. Phasic data are provided regardless of the amount of that phase present during the respective period.

I

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Palisades Nuclear Plant Revision 2 Reaqlistic~ Larae Brek LOCA Su~mmary Re~nort Paae 3-17 Realistic Larne Break LOCA Summarv Report Paae 3-17 Figure 3.1 Primary System Noding

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Renort Pane 3-18 Figure 3.2 Secondary System Noding

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Reoort Paae 3-19 Paae 3-19 Figure 3.3 Reactor Vessel Noding

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Report Paae 3-20 Figure 3.4 Core Noding Detail

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Palisades Nuclear Plant Revision 2 Re.*lintic L~ran Rrn.*k LOCYA Siimm~rv Re~nnrt Realistic Larne Break LOCA Summarv Report F

• lJ*

Lj--f I

Figure 3.5 Upper Plenum Noding Detail

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Renort Do*n "1_09 Realistic Larce Break LOCA Summarv Report I

• --LL Figure 3.6 S-RELAP5 Containment Pressure versus Best-Estimate Result

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Palisades Nuclear Plant Revision 2 R*.Iii*tic. I nrn*. Brm.k l OCA Stimmnrv R*.nrtcr Prnp_ 3-23 Realistic Larne Break LOCA Surnmarv Renort Pane 3-23 One-Sied Break Area IeO91101 0 a Was M el oemmONS (ft /side) 0.0 1.0 2.0 3.0 4.0 5.0 Burn Time *oooo 000oo eoooooo eom (hrs) 0.0 5000.0 10000.0 15000.0 Core Power

• o Ini-i- i- m . * .

(MW) [

2520.0 2540.0 2560.0 2580.0 2600.0 2620.0 LHGR we o0 aeON 641111 /

(kW/ft) [

12.0 13.0 14.0 15.0 16.0 ASI [ mIO. mOO MI 01,1111e SIGN

-0.2 -0.1 0.0 0.1 0.2 Pressurizer Pressure Call mm. m O m.. .me I UM (psia) 2000.0 2020.0 2040.0 2060.0 2080.0 2100.0 Pressurizer '

Liquid Level mO m m Oc0.N Nm

(%) ~m 40.0 50.0 60.0 70.0 Temperature Mm a Mee I mmcm H 536.0 538.0 540.0 542.0 544.0 TotalI Loop Flow [ acm. e mIINOmI OeI *

(Mlb/hr) 130.0 135.0 140.0 145.0 SITLiquid F m mI m Volume (ft3)ý ý OC ýmI mm m

1000.0 1050.0 1100.0 1150.0 1200.0 SIT Pressure e ooI *IIe (psa)H 210.0 220.0 230.0 240.0 SIT Temperature 0110 4W MeND 04m111116 me aau

( °F) 80.0 100.0 120.0 140.0 Figure 3.7 Scatter Plot of Operational Parameters

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Palisades Nuclear Plant Revision 2 Re~alis*ti¢ L~ran Rre~k LQCA 5*imm~rv Rennrt Do*, '_)A Realistic Larne Break LOCA Summary Report 2000 1800 F 1600 1400 FE 1200 U

1000 U U

800 U

MSplit Break 1 El Guillotine Break 600 400 0 100 200 300 400 500 Time of PCT (s)

Figure 3.8 PCT versus PCT Time Scatter Plot from 59 Calculations

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Palisades Nuclear Plant Revision 2 R*.Ii.*tirc L rn*. Brmk l O'CA Siimmairv R*.nnrt Realistic Larne Break LOCA Summarv Report 2000 1800 F D

[]

El 1600 F DE2 Ell D

El DO 1400 k E2 qm DDE[

ED0 F-I- D]

1200 F 0O 0

1000 k U U

U U 800 ý U

U 600 F 0 Split Break EDGuillotine Break 400 0.0 1.0 2.0 3.0 4.0 5.0 Break Area (ft2/side)

Figure 3.9 PCT versus Break Size Scatter Plot from 59 Calculations

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Renort Pane 3-26 1.0 E Split Break I El Guillotine Breaki 1:

0 X DEl 0

x O E: El m-uEI 0.0 -

400 800 1200 1600 2000 PCT ('F)

Figure 3.10 Maximum Oxidation versus PCT Scatter Plot from 59 Calculations

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summarv Reoort Paae 3-27 2000.0 1500.0 E 1000.0

.5 0

500.0 0.0 L-0.0 100.0 200.0 300.0 Time (s)

Figure 3.11 Peak Cladding Temperature (Independent of Elevation) for the Limiting Case

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Palisades Nuclear Plant Revision 2 Re~Ii.*tic¢ L~ran Rr*.k LOCA Siimm~rv Ri~nnrt Dý,, . I_-,

Realistic Larne Break LOCA Summarv Renort I

• U*

• J--LLJ C) 80.0 60.0 40.0 0

u_ 20.0 0.0

-20.0 L 0.0 100.0 200.0 300.0 Time (s)

Figure 3.12 Break Flow for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Reoort Paae 3-29 1000.0 500.0 E

9 X

0.0

-500.0 0.0 100.0 200.0 300.0 Time (s)

Figure 3.13 Core Inlet Mass Flux for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Larme Break LOCA Summary Report Paae 3-30 900.0 700.0

.500.0 E 300.0 X

U-100.0

-100.0

-300.0

-500.0 L-0.0 100.0 200.0 300.0 Time (s)

Figure 3.14 Core Outlet Mass Flux for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Laree Break LOCA Summary Renort 1Don a~l-Realistic Larae Break LOCA Summarv ReDort 1.0 0.8 0.6 C

U-2 L0 O2 0

0.4 0.2 0.0 -

0.0 100.0 200.0 300.0 Time (s)

Figure 3.15 Void Fraction at RCS Pumps for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Reoort Paae 3-32 ECCS Flows 3000.0 2000.0 E

1000.0

.2 LL 0.0

-1000.0 0.0 100.0 200.0 300.0 Time (s)

Figure 3.16 ECCS Flows (Includes SIT, HPSI and LPSI) for the Limiting Case

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Palisades Nuclear Plant Revision 2 P~:np. 3-33 Realistic Larcqe Break LOCA Summary Report Pane 3-33 3000.0 2000.0 Cu Co 0.

Co Co 0,

1000.0 0.0 L 0.0 100.0 200.0 300.0 Time (s)

Figure 3.17 Upper Plenum Pressure for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Renort Pn a -_*/A Realistic Laroe Break LOCA Summarv Reoort 30.0 20.0 7)

-j 10.0 0.0 1 0.0 100.0 200.0 300.0 Time (s)

Figure 3.18 Collapsed Liquid Level in the Downcomer for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summary Report Page 3-35 10.0 8.0 6.0 4.0 2.0 0.0 1 0.0 100.0 200.0 300.0 Time (s)

Figure 3.19 Collapsed Liquid Level in the Lower Plenum for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Reoort Paae 3-36 15.0 10.0 0)

-J V

0~

-J 5.0 0.0 LI-L 0.0 100.0 200.0 300.0 Time (s)

Figure 3.20 Collapsed Liquid Level in the Core for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Larae Break LOCA Summarv Report Dnr ý 1-1 Realistic Larae Break LOCA Summarv Reoort 100.0 90.0 80.0 70.0 60.0 S-2i 50.0 Ci3 (n

40.0 30.0 20.0 10.0 0.0 -

0.0 100.0 200.0 300.0 Time (s)

Figure 3.21 Containment and Loop Pressures for the Limiting Case

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 4-1 4.0 Conclusions An RLBLOCA analysis was performed for the Palisades nuclear power plant using NRC-approved AREVA NP RLBLOCA methods (Reference 1). Analysis results show that the limiting AREVA NP fuel case has a PCT of 1,751 OF, and a maximum oxidation thickness and hydrogen generation that fall well within regulatory requirements. Mixed-core effects are a non-issue since the core is completely fueled with 15x15 AREVA NP fuel assemblies.

The analysis supports operation at a nominal power level of 2,565.4 MWt (plus uncertainty), a steam generator tube plugging level of up to 15 percent in both steam generators, a linear heat rate of 15.28 kW/ft, an FrT of 2.04 with no axially-dependent power peaking limit and peak rod average exposures of up to 62,000 MWd/MTU. For large break LOCA, all 10CFR50.46(b) criteria presented in Section 3.0 are met and operation of Palisades with AREVA NP-supplied 15xl 5 Zr-4 clad fuel is justified.

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Palisades Nuclear Plant Revision 2 Realistic Large Break LOCA Summary Report Page 5-1 5.0 References

1. AREVA NP Document, 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 Programfor Predicting Containment Pressure-TemperatureResponse to a Loss-Of-Coolant-Accident," Aerojet Nuclear Company, TID-4500, ANCR-1219, June 1975.

4. U. S. Nuclear Regulatory Commission, NUREG-0800, Revision 2, Standard Review Plan, July 1981.