ML100890593
| ML100890593 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 01/31/2010 |
| From: | AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| HNP-10-029 ANP-2853(NP), Rev 0 | |
| Download: ML100890593 (105) | |
Text
Enclosure 4 to SERIAL: HNP-10-029
.SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1.......DOCKET LICENSE NO.-NPF-63.:..
APPLICATION FOR REVISION TO TECHNICAL SPECIFICATION CORE OPERATING LIMITS REPORT (COLR) REFERENCES HARRIS NUCLEAR PLANT UNIT, REALISTIC LARGE BREAK LOCA
SUMMARY
REPORT;AREVA NP.Inc., ANP-2853(NP), Revision 000 (Noin-Proprietary)*..
(104 Pages)
AREVA NP Inc.ANP-2853(NP)
Revision 000 Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report January 2010 Harris Nuclear Plant Unit 1 ,I Realistic LarQe Break LOCA Summary Report.ANP-2853(NP)
Revision .000'Pame i Copyright
© 2010 AREVANP !nc ',..-All Rights Reserved.AREVA NP Inc.
Harris Nuclear Plant Unit 1 --, .Realistic Large Break LOCA Summary Report.ANP-2853(NP)
Revision 000...., .., * -iPage ii Nature of Changes Description and Jpustificatir, This is wnew dociment.i:
Item Page 1. All AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report'.ANP-2853(NP).
Revision 000, S-Page-iii Contents 1.0 Introduction
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1 1 2.0 Sum m ary ... ...... ..... .. ...... ...... ... 2 1 3.0 Ana rlys is ........ 3. 1..3.1.. Description of the LBLOCA, Event.. ....31..... ................................
3.2 Description
of Analytical Models...
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3-3 3.3 Plant Description and Summaryof Analysis ParmeOters..................
..... 3-6., 3.4 RLBLOCA.SER Compliance
...... ... ..... .L................3-8
3.5 Realistic
Large Break LOCA Results ....... ..... ....... ...................
3-8 4.0 -Generic Support for Transition Package.,.::..1:.:..............................
4-1 4.1 Reactor Power ..................................................
4-1 4.2 Rod Quench ........................................
...................
....... 4-1 4.3 Rod-to-Rod Thermal Radiation'
.:. .., ...............
4-2 4.4 Film Boiling Heat. Transfer Limit... .....................
4-.6 4.5 Downcomer Boiling...........
.. ..... .................
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4-6 4.6 Break Size .............
I ..... .......................
..... 4-11 4.7 Detailed Information.for Containment Model ...... .................
....4-15-4.8 Cross-References to'Generic Data on the North Anna Docket ........................
4-15 4.9 GDC 35 -LOOP and No-LOOP Case Sets. ...... ..... .... .............
4-18 4.10 Input Variables Statem ent ...............................................................................
4-18 5 .0 C o nclusio ns ..................................................................................................................
5-1 6 .0 R efe re nces ..................................................................................................................
6-1 AREVA NP Inc.
Harris Nuclear Plant ,ANP-2853(NP)
Unit 1 Revision 00Q Realistic Large Break LOCA Summary Report .Page. iv Tables-.Table 2-1 Summary of Major Parameters for Limiting Transient
.............................................
2-1 Table 3-1 Sampled LBLOCA Parameters..-.-.1.0.-:--.*.....
.... ...... .. ..... ... ... .. .. ...............
.... 3 Table 3-2 Plant Operating Range Supported bythe LOCA Analysis ..........................
3-1.1 Table 3-3 Statistical Distributions Used for Process Parameters
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..... 3-15.,-, ..... ..... .... ..... .... ... ...... ........ ........' " ." " 7 ' ..' .". , ., .Table 3-4. RLBLOCA EM SER Conditions and Limitations..;
......i......................
3-16 Table 3-5 Summary of Results for the.LiMiting P& Tas......
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3-18.: ....... ..... ..; .. :- .: -..: "; '
.". ....... Table 3-6 Calculated.
Event Times for the Limiting.
PCT Case,.;. ..... ... ............
.... 3-18 Table 3-7 'Heat Transfer Parameters for the Limiting Case.........
.... ................
3-19 Table 3-8.. Containment.
Initial and. Boundary .Conditions..:...:.'.I.
.......I.............,........
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3-20 Table 3-9 Passive Heat Sinks in Containment
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....... 3-21 Table 4-1, Typical, Measurement.
U ncertainties and Local Peaking Factors ...........
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4-20 Table 4-2 FLECHT-SEASET
& 17x17 FA Geometry Parameters......................
4-20 Table 4-3 FLECHT-SEASET Test Parameters
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20 Table .i~c .HT Test a~amete...................................
.............
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... ..... 4 2 Table 4-4 Minimum Break Area for Large Break LOCA Spectrum....':
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4-21 Table 4-5. Minimum. PCT Temperature Difference
-True Large and Intermediate.
B re a ks ....... .......................
........................
I .............
...............................
4 -2 1 AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)y Unit 1 :. Revision 000: Realistic Large Break LOCA Summary Report Pa'gev.Figures-
.Figure 3-1 Primary System Noding..:
..........................................................................
.3-23 Figure 3-2 Secondary System Noding ...................................................................................
3-24 Figure 3-3 Reactor Vessel Noding .... .. ...............
...... 3-25......
..Figure 3-4 C ore N oding D etail ................................................................................................
3-26 Figure'3-5 U pper Plen6u.m Noding Detail................................
3-27 Figure 3-6 Scatter Plot of Operational Parameters
...........
3-28 Figure 3-7 PCT versusPCT Time Scatter Plot from 59Calculatio'ns..................
........ 3-30 Figure 3-8 PCT versusBreak Size Scatter Plotfrom 59 Calculations
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3-31 Figure 3-9 Maximum Oxidation Versus PCT Scatter Plot' from 59 Calculation's
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...... 3-32 Figure 3-10 Total Oxidation versus PCT Scatter Plot from 59 Calculations
....... ..........
3-33 Figure 3-11 Peak Cladding Temperature (Inndependent of Elevation) fr the Limiting.
Case... ... ............................
,.. ..........
.... ...334 Figure 3-12 Break Flow.for the Limiting Case ...'3........
- ..3.35............,.....
.............................
3-35 Figure 3-13 Core Inlet Mass Flux for the Limiting Case .........................................................
3-36 Figure 3-14 Core Outlet Mass Flux for the Limiting Case ......................................................
3-37 Figure 3-15 Void Fraction at RCS Pumps for the Limiting Case ............................................
3-38 Figure 3-16 ECCS Flows (Includes Accumulator, Charging, Sl and RHR) for the L im iting C a se ..............................................................................................................
3-3 9 Figure 3-17 Upper Plenum Pressure for the Limiting Case ....................................................
3-40 Figure 3-18 Collapsed Liquid Level in the Downcomer for the Limiting Case .......................
3-41 Figure 3-19 Collapsed Liquid Level in the Lower Plenum for the Limiting Case ....................
3-42 Figure 3-20 Collapsed Liquid Level in the Core for the Limiting Case ..................................
3-43 Figure 3-21 Containment and Loop Pressures for the Limiting Case .....................................
3-44 Figure 3-22 GDC 35 LOOP versus No-LOOP Cases ............................................................
3-45 Figure 4-1 R2RRAD 5 x 5 Rod Segment ...............................................................................
4-22 Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 F A ..............................................................................................................................
4 -2 3 Figure 4-3 Reactor Vessel Downcomer Boiling Diagram .......................................................
4-24 Figure 4-4 S-RELAP5 versus Closed Form Solution .............................................................
4-25 Figure 4-5 Downcomer Wall Heat Release -Wall Mesh Point Sensitivity
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4-26 Figure 4-6 PCT Independent of Elevation
-Wall Mesh Point Sensitivity
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4-27 Figure 4-7 Downcomer Liquid Level -Wall Mesh Point Sensitivity
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4-28 Figure 4-8 Core Liquid Level -Wall Mesh Point Sensitivity
...................................................
4-29 Figure 4-9 A zim uthal N oding .................................................................................................
4-30 AREVA NP Inc.
Harris Plant Unit 1 Realistic Larcge Break LOCA Summary Report ANP-2853(NP)
Revision 000-ý Paqe vi v *Figure 4-10 Lower Compartment Pressure versus Time .......................................................
4-31 Figure 4-11 Downcomer Wall Heat Release -Axial Noding Sensitivity Study .......................
4-32:.' -... , .. ...... ..... ... .......... ..... .... ... ... ... .. .. ., : L ' i -: .' ." i .*.. , C 'Figure 4-12 PCT Independent of Elevation
-Axial Noding Sensitivity Study .................
4-33 Figure 4-13 Downcomer Liquid Level -Axial Noding Sensitivity Study ..................................
4-34 Figure 4-14 Core Liquid Level -Axial Noding Sensitivity Study ...........................................
4-35
~ ~ ~ ~~ ~~~~~~~~~.
...........
..... ... ... .......... : ., ." " " .. ." , Figure 4-15 Plant A -W estinghouse 3-Loop Design .............................................................
4-36 Figure 4-16 Plant B -Westinghouse 3-Loop Design ................
4-37 Figure 4-17 Plant C -W estinghouse 3-Loop Design ...................
....................
..............
4-38 Figure 4-18 Plant D -Combustion Engineering 2x4,Design
..............................................
4-39 Figure 4-19 Plant E -Combustion Engineering 2x4 Design ..............................................
4-40 Figure 4-20 Plant F -W estinghouse 4-loop Design ...........................
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4-41 Figure 4-21 PCT vs. Containment Volume .....................................
4-42 Figure 4-22 PCT vs. Initial Containment Temperature
....................
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........ 4-43 Figure 4-23 Containment Pressure as function of time for limiting case ...... ..........
4-44 This document contains a total of 104 pages.AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP):
Unit 1 Revision 000P Realistic Large Break LOCA Summary Report '.age vii-,Nomenclature:-:.-.:, AFD Axial Flux Difference CFR Code of Federal Regulations CCTF Cylindrical Core Test Facility..,;-!
CHF Critical Heat Flux CSAU Code Scaling, Applicability, and Urcehainhty
-CSIP Charging/Safety Injection Pump CSB Containment Systems Branch--
" DC Downcomer DEGB Double-Ended Guillotine Break DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPH Effective Full Power Hours EM Evaluation Model FQ Total Peaking Factor FAH Nuclear Enthalpy Rise Factor FSAR Final Safety Analysis Report HFP Hot Full Power HHSI High Head Safety injection HNP Harris Nuclear Plant LBLOCA Large Break Loss of Coolant Accident LANL Los Alamos National Laboratory LHSI Low Head Safety Injection LOCA Loss of Coolant Accident LOOP Loss of Offsite Power MSIV Main Steam Isolation Valve MTC Moderator Temperature Coefficient NRC U. S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PCT Peak Clad Temperature PIRT Phenomena Identification and Ranking Table PLHGR Planar Linear Heat Generation Rate PWR Pressurized Water Reactor RAS Recirculation Actuation Signal RCP Reactor Coolant Pump RCS Reactor Coolant System AREVA NP Inc.
Harris Nuclear Plant Unit1 :: Realistic Large Break LOCA Summary Report ANP.-2853(NP)
Revision 000 S,.. Pbge viii Nomenclature (Continued)
RHR RLBLOCA RV RWST SER SI SIAS Residual Heat Removal Realistic Large Break LOCA Reactor Vessel Refueling Water Storage Tank ',-Safety Evaluation Report,.: Safety Injection Safety Injection Actuation Signal AREVA NP Inc.
Harris Nuclear Plant -ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report .... '. Page 1-1 1.0
...: , -This report describes and provides results from a RLBLOCA analysis for the Harris Nuclear Plant, Unit 1 (HNP). The plant is a Westinghouse 3-loop design with a rated thermal power of 2900 MWt and dry, atmospheric containment.
The. loops contain, three RCPs, three U-tube steam generators and a pressurizer.
In the ECCS, there are two LHSI pumps -which are cross-connected to all three cold legs, two HHSI pumps (Charging/Safety Injection pumps) which are cross-connected to all three cold legs and one accumulator connected to each cold leg. The design includes an installed spare.,swing CSIPR.that isnormally but-;of service.The analysis supports operation for Cycle 16 and beyond with AREVA NP's 17X17 HTP fuel design using standard U0 2 fuel with 2%, 4%, 6% and 8% Gd9O 3.and Zr-4 cladding, unless changes in the Technical Specifications, Core Operating Limits Report, core design, fuel design, plant hardware,.or plant operation'm invalidate the results presented herein. .The analysis was performed in compliance with. the NRC-approved RLBLOCA EM (Reference
- 1) with exceptions noted:below.
Analysis results confirm the 10CFR50.46 (b) acceptance criteria.
presented, in Section 3.0 are. met and serve as.,the basis' for operation of the Harris: Nuclear' Plant with AREVA NP fuel.The non-parametric statistical methods inherent in the AREVA NP RLBLOCA methodology provide for the consideration of a full spectrum of break sizes, break configuration (guillotine or split, break), axial shapes, and.plant operational parameters.
.A conservative single-failure assumption is applied in which the loss of one train of.the pumped ECCS injection is simulated' Regardless of the single-failure assurmption,, all containment -pressure-reducing systems are assumed fully functional.
The effects of Gadolinia-bearing
'fuel rods.. and.:'peak fuel- rod exposures are considered.
The following are deviations from the approved RLBLOCA EM.. (Reference
- 1) that were requested and approved by the NRC on other implementation of the RLBLOCA EM on other.dockets. Further discussion of the origin of these deviations is contained in Section 4.0.The assumed reactor core power for the HNP realistic large break loss-of-coolant accident, is 2958 MWt. The value represents the plant rated thermal power of 2900 MWt with a maximum power measurement uncertainty of 2.0 percent (58 MWt) added to the rated thermal power. The AREVA NP Inc.
HarrisNuclear Plant ANP-2853(NP)
Unit 1 Revision 000 Realistic.Large Break LOCA Summary Report .....Page1,-2 power was not sampled in the analysis.
This is not expected to have an, effect on the PCT results. ., .'.The: RLBLOCA anaIysis wa perform ed .with a 'Viersion .of S-RELAP5 thaýt requires' both the void fraction to be less'than'0.95'a-n'theIcaiid be leiss th" 6'900before the rod is allowed to 'quench'-
This may resdRt in'a slight increase' ii PCT results when compared to an analysis not subject to these constraintsj
' "s" The RLBLOCA analysis .was -performed with a version of S-RELAP5 that limits the contribution of the Forslund-Rohsenow model to no more than 15 percent of the total heat transfer at and above a void fraction of 0.9. This may result in a slight increase in PCT'results When compared to previous analyses for similar plants.The split versus double-ended break type.is no-longer relatedto break area.: In concurrence with Regulatory Guide'.-1.157, both, the split and the double-ended break will .range in area between the, minimum ,.break area (Amin) and- an.area of twice the size: of the broken pipe. The determination of break configuration, split versus double-ended,'
will be made *after the break area is selected based on a uniform probability for each occurrence.
Amin was calculated to be 27 percent of the DEGB area (see Section 4.6 for further discussion).
This is not expected to have an effecton" PCT'results.
In concurrence with the. NRC's interpretation of GDC 35, a set of 59 cases was run with a LOOP assumption and a second set with a No-LOOP assumption.
The set of 59 cases that predicted the highest PCT is reported in Section 2 and Section 3, herein. The results from both case sets are shown in Figure 3-22., The effect on PCT results is expected to be minor.During recent RLBLOCA EM modeling studies, it was noted that cold leg condensation efficiency may be under-predicted.-
Water entering the DC post-accumulator injection remained sufficiently subcooled to absorb DC wall heat release without significant boiling., However, tests (Reference
- 7) indicate that the steam and water entering the DC from the cold leg, subsequent to the end of accumulator injection, reach near saturation resulting from the condensation efficiency ranging between 80 to 100 percent. To assure that cold leg condensation would not be under-predicted, a RLBLOCA EM update was made. Noting that saturated fluid entering the DC is the' most conservative modeling scheme, steam and liquid multipliers were developed so as to approximately saturate the cold leg fluid before it enters the DC. The multipliers were AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)., Unit 1 -Revision 000;Realistic Large Break LOCA Summary Report ______ --_ _Page 1-3 developed through scoping studies using a number of plant designed 3- and .4-loop plants, and CE-designed plants. The results of the.scoping study indicated that multip iers of 10 and 150 for liquid and steam, respectively, were appropriate.
to.produce saturated fluid entering the DC. This, RLBLOCA EM departure was recently discussed with the NRC and the NRC agreed that. the- approach described immediately., above was satisfactory in the interim. The modification:
is mplemented post-accumulation injection, 10 seconds after the vapor void fraction in the bottom of the accumulator becomes greater than 90 percent. Thus, the accumulators have injected all their water into the cold legs, and the nitrogen cover gas has entered the system and been mostly discharged through the break before the condensation*
efficiency -is : increased--
by, the: factors. of:* 1.0 and 150,. for .-liquid :-and vapor respectively:.,Providing saturated fluid..conditions at the-DC entrance conservatively:-reduces both the DC.driving head and the coreflooding~rate.
Recall that,-test-results indicate that fluid conditions entering the DC range from saturated to slightly subcooled.
Hence,;it is conservative to force an approximation of saturated conditions for, fluid entering the DC.------AREVA Inc. has acknowledged an issue concerning fuel thermal conductivity degradation as a function of burnup as raised by the NRC. In order to manage this issue, AREVA Inc. is modifying the way RODEX3A temperatures are compensated in the RLBLOCA Revision 0/Transition package methodology.
In the current process,,'the.
RLBLOCA computes PCTs at many different times during an-operating cycle.- For each -specific -time in cycle, the- fuel conditions are computed using RODEX3A prior to. starting the- S-RELAP5 portion of the analysis.
A steady state.condition for the. given time in.cycle using S-RELAP5 is established.
A base fuel centerline temperature is established in this process. Then two-transformation adjustment to the base fuel centerline temperature is computed.-The first transformation is a linear adjustment for-an exposure of 10 MWd/MTU-or.
higher. In the new. process, a polynomial transformation is used in the first- transformation instead of a linear transformation.
The rest of the RLBLOCA process for initializing the S-RELAP5 fuel rod ,temperature should not be altered and the rest of LOCA transient should also continue in the original fashion. This approach has been requested by the NRC.AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report 'Page 2A1 2.0..:. Summary , The limiting PCT analysis"is ba'sed 6n 'the parameter -specification given- in Table 2-1 f6r the limiting case. The limiting-PCT i19300F fo r an ;0 2 rod in-a cassewith LOOP, conditions.
Gadolinia-bearing rods of 2, 4, 6 ahd 8w/vi Gd 2 03' were also analyzed, blit were not limiiting' This RLBLOCA result is based on a case set of 59 indiVidual transient cases for LOOP and 59 individual triansient cases for No-LOOP conditions.
'Tlie'c~re is composed only of AREVA NP 17x17 thermal hydraulically' compa6tibil fuel designs; hece, there' is no mixed core conhsidersation.'o The' analyzed core power. is 2958 MWIt with a steam generator.-tube plugging level of:3,percent in.all steam generators, a-totalpeaking factor (F 0) up to a value,.of 2.52 (includinguncertainties; but ino axial dependency);,c.-and -a nuclear. enthalpy rise factor (FAH)'up' to.a .value of., 1.73 (including uncertainty).'
This analysis also addresses typical.:operatibnal ranges, or-technical specification limits (whichever is applicable) with regard to pressurizer pressure and level;accumulator pressure, temperature (based on containment temperature), and level; core average temperature; core flow; containment pressure and temperature; and RWST temperature.
The AREVA 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 rbd *average exposures.
of.62,000 MWd/MTU: *The analysis demonstratesthat 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) _ 11267 Analyzed Core Power (MWt) 2958 " Total Peaking (FQ) ' 2.47 Radial Peak (FAH), Tech Spec.. 1.66 " , Axial Offset -0.2028 Break Type Guillotine Break Size (ft 2/side) 1.5097 Offsite Power Availability Not available Decay Heat Multiplier 0.9896 AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1 Revision 000Q, Realistic Large Break LOCA Summary Report :.::Page 3-1 3.0 Analysis-
-- '."'.The purpose of the analysis -is to verify typical technical tspe6ification peaking4factor limit's and the adleq6uacy ofthe ECCS by demonstrating'that theý follo0wirg 1OC(FR 50.46(b) criteria are' met: (1) The calculated maximum fuel element cladding temperature shall not exceed 2200cF.(2) The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total claddin'gthickness' before oxidationl
- -" .,-(3) The calculated total amount of hydrogen generated from the chemical reaction of the""cladding with water Or~steam s'hal[ nof-exceed,'O0l times the hypothetical amount that would: be generated
- if all of the, metal.. in the, cladding -cylinders surrounding the fuel excluding the cladding surrounding the plenum volume were to react.(4) .T-,he. calculated changes in core geometry shall be such that the core remains amenable to cooling.(5) Long-term cooling is established and maintained after the LOCA.The analysis did not evaluate core coolability due to seismic events, nor did it consider the 100FR 50.46(b) long-term cooling criterion.
The analysis purpose does not change the LBLOCA licensing basis, therefore prior coolable geometry (LOCA-seismic loads) and long-term cooling licensing bases remain unaffected and valid. Thus, compliance with Criteria (4) and (5)is assured.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 RLBLOCA EM 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 rupture 'of the RCS primary piping. Based on deterministic'studies, the worst break location is in the cold leg piping between the reactor coolant pump and' the reactor vessel for the 'RCS loop containing the pressurizer.
The break initiates a rapid depressurization of the RCS. A reactor trip signal is initiated when the low pressurizer pressure trip setpoint is reached; however, reactor trip is conservatively neglected in the analysis.
The reactor is shut down by coolant voiding in the core.The plant is assumed to be operating normally at full power prior to the accident.
The cold leg break is assumed to open instantaneously.
For this break, a rapid depressurization occurs, AREVA NP Inc.
Harris Nuclear Plant ;ANP-2853(NP)
Unit 1 Revision 000 RealisticLarge Break LOCA Summary Report .. .;.. Page 3-2 along with a core flow stagnation and reversal.
This causes.the fuel rods to experience DNB.Subsequently,.the.
limiting fuel rodsare cooled by film convection to~steam.
T~he 2c[olant voiding creates a strong negative.
reactivity effect. and.core.
criticality ends. As. heat transfer from the fuel rods is reduced, the cladding temperature increases.
Coolant in all regions of the RCS begins to flash. At the break plane, the loss of subcooling in the coolant results in substantially reduced break flow.. This ,reduces the rate, and leads to a period of positive core flow or reduced downflow as the RCPs in the intact loops continue to supply .water to the,.RV (inNo-LOQP.
conditions)..
Cladding temperatures may be reduced and some, portions'of the coremay'rewet during this period.' The positive core flow or reduced downflow period ends as two-phase conditions occur in the RCPs, reducing their sis.ocu innheR6s reducingth theiroýtofh effectiveness.
Onde again, the core flow reverses as mostof the vesselmass flows" 6ut through the broken cold leg.Mitigation of the LBLOCA begins when the SIAS is issued. This signal is initiated by either high containment pressure or low pressurizer pressure.
Regulations require that a worst single-failure be considered.
This single-failure has been determined to be the loss of one ECCS pumped injection train. The AREVA RLBLOCA methodology conservatively assumes an on-time start and normal lineups of the containment spray to conservatively reduce containment pressure and increase break flow. Hence, the analysis assumes that one HHSI pump, one LHSI. pump, and all :containment spray .pumps are -operating..Seven fan coolers are assumed operating from time zero of the transient.:.
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.Event ually, the relatively large volume of accumulator'Vwater is exhausted and core' recovery continues relying solely on pumped ECCS injection.
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 ofimproved core heat transfer as the nitrogen gas displaces water in the downcomer.
After the nitrogen gas has been expelled, the ECCS temporarily may not be able to AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1 Revision 000, Realistic Large Break LOCA Summary Report .Page 33.sustain full core.codlingbecause .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 HHSI and IHsI whie:the decay h~eat: coiinues to fall." Stearhi generated from fuel reWet Will ent" liquid a6d pass throug h fhe core, vessel upper plehnmir, t Iehe 'i he- s,"th eI.e a` gen&bators,'
nd the reactor bCoolant pu bml 6fb e'itl isiv"entdediuftihe'birbk.The path:to the steam-flowi§-balribdd'by the deiving foribdf wiater filling the downcom-er.
'This resistance may .act-to retard tohe progressidni ofthe' core ref ood and postone' corew'ide cooling.( few minutes Of the acident), reflood will 'poge!ss'sufficiently to ensure core-wide cooling. Full core! quench occurs within a few minutes after 'or6-'wide cooling.Long-term cooling is then sustained with LHSI pumped injection system.3.2 Description of Analytical Model" 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). Thismethod:outlines an approach for defining and qualifying
- a."best-estimate.
thermal-hydraulic code. and :..quantifies the uncertainties in a LOCA analysis.The RLBLOCA methodology consists of the following computer codes: RODEX3A for. computation of the initial fuel stored energy, fission:.gas release, and fuel-cladding gap conductance.
.* S-RELAP5 for the system calculation (includes ICECON for containment response).
AUTORLBLOCA for generation of ranged parameter values, transient input, transient runs, and general output documentation.
The governing two-fluid (plus non-condensables) 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 kihetics equations (point kinetics) with reactivity feedback, and with actinide and decay heating.The two-fluid formulation uses a separate set of conservation equations and, constitutive relations for each phase. The effects of one phase on the other are accounted for by interfacial friction, and heat and mass transfer interaction terms, in the equations.
The conservation AREVA NP Inc.
Hari-is NuclearPlant
-.. ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report 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 phenomena expected during the LBLOCA event are captured.
The basic building blocks for modeling are hydraulic volumes for fluid paths-and heat structures for heat transfer.
In addition, special purpose components exist to represent specific components such as the RCPs or the steam generator separators.
All geometries are modeled, at the resolution necessary to.best resolve the flow field and the phenomena being, modeled within practical computational limitations.
.System nodalization details are shown in Figures 3-1 thrbugh 3-5.'- A pint 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 6f both break junctions.
.A typical calculation using S-RELAP5 begins with theestablishment 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 Sbetion 3'3.Following the establishment'of an acceptable steady-statecondition, 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 bl6wd6wn, refill and reflood is computed continuously using S-RELAP5.
Containment pressure is also calculated by .S-RELAP5 using containment models derived from ICECON (Reference 4), which is based on the CONTEMPT-LT code (Reference3).
The methods used in the application of S-RELAP5 Jto the LBLOCA are described in Reference
- 1. A detailed assessment of this computer code was made through comparisons to experimental data, many benchmarks with cladding temperatures ranging from 1,700'F (or less)to above 2,200cF. These assessments were -used to d evelop 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 AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1 i Revision 000: RealisticLarge Break LOCA Summary Report Page 3-5 correlation-are defined based -on. code-to-data., c-omparisonsý .and,- are;,. heiice,- plant independent.. , .., The RV internals are-modeled in detail (Figures 3-3 through 3-5) based on HNP spedific inputs S by Pr.g...s.Enr d anconnectivity, flow areas, resistances and heat supplibd'd by Progiress Energy. oIi~de`s :'in "" '" " ' ; .::... ..i.n° -. .. .: structures
-re all accurately modeled. The location of -the hot assembiy/hOt pin(s) is unrestricted; however, the chan'nel is always modeled to restrict appieciable ulppr 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: 1. Base. Plant Input File Development
-. , -First, base RODEX3A and S-RELAP5 input.files' for the plant'(including'the'containment input file) aredeveloped;
- Code input development guidelines are- applied:to-ensurethat model nodalization is consistent with the model nodalization used in the code validation.
- 2. Sampled Case Development:
The non-parametric statistical approach requires that many "sampled" cases be created and processed.
For every set of input created, each "key LOCA parameter" is randomly sampled over a range established -through code uncertainty assessment or expected operating limits (provided by plant technical specifications or data). Those parameters considered "key LOCA parameters" are listed in Table 3-1.:, This: list includes both--parameters, related to LOCA phernomena (based onthe PIRTprovided:in Reference, 1)and to plant operating parameters.
- 3. -Determination.of Adequacy of ECCS- -.-., -The RLBLOCA methodology uses: anon-parametric.
statistical appr6ach lto determine , values70f PCT at the 95 percent probability level. Total oxidation-and total hydrogen are based on the limiting PCT case. The adequacy of the ECCS is demonstrated when-these results satisfy the criteria set forth in Section 3.0..AREVA NP Inc.
<it..Harris Nuclear Plant .ANP-2853(NP)
Unit, 1 Revision 000 Realistic Large Break LOCA Summary Report .-Page 3-6 3.3. Plant Description and Summary-of Analysis Parameters The plant analysis presented in this report is for a Westinghouse-designed PWR, whhich has three loops, each with a hot leg, an U-tube steam generator, and a cold leg with a RCP'.-. The RCS also includes one pressurizer, connected to a hot leg. The core contains 157 C i. .: : : ; ? ? i .r : ' -: " ' .,' ; " :, : , , / , :;, " " , ' -, t, 1 , thermal-hydraulic compatible AREVA 17X17 HTP fuel assemblies with 2%, 4%, 6% and 8%gadolinia pins. The ECCS includes one HHSI, one LHSI and one accumulator injection path per.RCS loop. The break is modeled in the same loop as the pressurizer, as directed by the RLBLOCA methodology.
The RLBLOCA transients are of sufficiently short duration that the switchover to- sump cooling water (i.e., RAS)- for'-ECCS pumped injection-need not be considered.
., The S-RELAP5 model explicitly describes the RCS, RV, pressurizer, and accumulator lines.The ECCS includes an accumulator path and a LHSI/HHSl path per RCS loop., The HHSI and LHSI feed into a common header that connects to each cold leg pipe downstream of the RCP discharge.: -The ECCS. pumped injection
- iis modeled as -a table 'of flow versus backpressure.
This model also describes the secondary-side steam generator that is instantaneously isolated (closed MSIV and feedwater trip) at the time of the break. A symmetric steam generator tube plugging level of 3 percent per steam generator was assumed.As described in the.AREVA RLBLOCA methodology,, mahy 'parameters.
associated with iLBELOCA phenomenological uncertainties and plant operation ranges are sampled. A summary of those parameters is given in Table, 3-1:. -The LBLOCA phenomenological uncertainties are provided in Reference
- 1. Values for process or operational parameters, including ranges of sampled process parameters, and fuel design parameters used in the analysis are given in Table 3-2. Plant data are. analyzed to ;,develop..
uncertainties for the process parameters sampled in' the analysis.
Table :3-3 presents a summary of the* uncertainties Used in the analysis.
Two parameters (RWST temnperature for Si 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 or supporting plant calculations that provide more bounding values.For the AREVA NP RLBLOCA EM, dominant containment parameters, as well as NSSS parameters, were established via a PIRT process. Other model inputs are generally taken as AREVA NP Inc.
Harris Nuclear Plant ;ANP-2,853(NP)
Unit 1 .. Revision 000 Realistic Large Break LOCA Summary Report ,.. .-, ... Page 3-7 nominal or conservatively biased. The PIRT outcome yielded .two imprtqant%(Felative.'to POT)containment parameters--containment pressure and temperature.
In many instances, the conservative guidance of CSB 6-2 (Reference
- 5) was used in setting the remainder of the containment model input parameters.
.As noted in Table 3-3, containment temperature is a sampled parameter.
Containment pressure response is indirectly ranged by sampling the containment volume (Table 3-3). The minimum containment volurne value iscarried over from use in the :longrterm -containment-integrity.
analysis-of record for HN P., The maximum value is a: simplified value computed:
as, the available volume of,2.61 E6 ft. -This volume was:calculated as.the volume of the containment building, void of all interior walls-or other. structures.-." The containment initial conditions and boundary conditions aregiventin Table 3-8.-The building spray is modeled .at maximum -heat: removal capacity...All, spray*-flow is, -delivered to the containment.
Seven fan coolers are assumed:, operating from: time.zero of. the LBLOCA transient.
Containment heat sink data is 'given' in Table 3-9. In accordance with Reference 1, the condensing heat transfer coefficient is intended to be closer. to a best-estimate instead of a bounding high value. A [ ] Uchida heat transfer coefficient multiplier' Was specifically validated for use in HNP through application of the.. process, used *in the RLBLOCA EM (Reference
.1) sample problems.The RCPs are Westinghouse 93A type pumps. The homologous pump performance curves for this type of pump were input to the S-RELAP5 plant model.AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1. Revision 000 Realistic.Large Break. LOCA Report .. Page 3-8 3.4 RLBLOCA SER'Compliance-
.A number of requirements on the methodology are stipulated in the cohclusions section *of the SER for the RLBLOCA'methodology(Reference 1). These requirements have l11 bee n fulfilled during the application of the methodoiogy as addr-essed in Table 3-4".3.4.1. .Item 7: Blowdown Quench,..
.,..., , , -Fifteencbases were potential-candidates-for =blowdown, quench and were, closely inspected..For this set.of6al-culations, no evidernceeof blodwdown:
quench-was observed.:Therefore, compliance to the SER restriction has~been demonstrated." " ..3.4.2 Item 8: Top-down Quench Several provisions "have been implemented in the-S-RELAP5 model to prevent the top-down quench. The upper plenum nodalization features include: " the homogenous option is selected for the junction that connects the first axial level node above the hot channel to the second axial level node above the hot channel;" no cross-flow is allowed between the first axial level .Upper'Plenum nodes above the hot.channel to the average channel;, ..* the CCFL model is.applied on all core-exit junctions.
Four cases were closely examined for top-down quench. No evidence of top-down quench was'observed.
Therefore, compliance to the SER restriction has been demonstrated.
3.5 Realistic
Large Break LOCA Results Two case sets of 59 transient calculations were performed sampling the parameters listed in Table 3-1. For each case set, PCT was calculated for a U0 2 rod and for Gadolinia-bearing rods with concentrations of 2, 4, 6 and 8 w/o Gd 2 0 3.The limiting case set, that contained the PCT, was the set with no offsite power available.
The limiting PCT (1930 0 F) occurred in Case 5 for a U0 2 rod. The major parameters for the limiting transient are presented 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 percent limit. The best-estimate PCT case is Case 30, which corresponded to the median case out of the 59-case set with no offsite power available.
The nominal PCT was 1540°F for a U0 2 rod. This result can be used to quantify the relative conservatism in the limiting case result.. In this analysis, it was 390'F.AREVA NP Inc.
Harris:Nuclear Plant Unit 1 Realistic Larce Break LOCA Summary Report ANP-2853(NP)
Revision 000 Paqe 3-9 The case results, event times and. analysisIplots for, theý lirmiting, PCT case are shown in Table 3-5, Table 3-6, and in Figure 3-11 through Figure 3-21. Figure 3-6 shows linear scatter plots of the key parameters sampledfor the-59calculaitio'ns.ParaFiaete" lalbe appear tothe left of each individual.
plot. "hpese~figures show.. eparameterjranges used in the analysis.
Figure 3-7 and Figure 3-8 show the time of PCT anid'breaks."iiz§e-Vrsus PCT scatter plots for the 59 calculations with no offsite poweravaila-I able"" respect'iveiy:
F'igure 3-9 and Figure 3-10 show the maximum oxidation and total oxidation versus PCT scatter plots for the 59 calculations, respectively.
Key parameters for the .limiting.
PT case are shown in Figure 3-11 through Figure 3-21. Figure 3-11 is the plot of PCT! independent of. elevation -for the limiting case; this figure clearly indicates that the transient exhibits'a sustained and stable quench. A comparison of POT results from the LOOP and no-LOOP case sets is'shown in Figure 3-22.AREVA NP Inc.
Ha'rris 'Nuclear Plant Unit 11 Realistic Large.Break.l OCA. Suymmary Report ANP-2853(NP)
Revision 000-.-.. ....._Page 3-10..Table3.1Sampled LBLOCAParameters
- , :- .'. ", Tim -in cyc le :(peakirng fact0rs, axi ashpe- rod propertiesj,1burnujp)X.
'..Break type. (guillotine versus split),..
.-Critical flow discharge coefficients (break)DiCay heait C riti al flW " ischarg ciets sr eline) .. 6r46: .;Initial upper head, temperature Film boiling heattransfer 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 Plant'Offsite power availability 2 Break size Pressurizer pressure Pressurizer liquid level Accumulator pressure Accumulator liquid level Accumulator temperature (based on containment temperature)
Containment temperature Containment volume Initial RCS flow rate Initial operating RCS temperature Diesel start (for loss of offsite power only)Uncertainties for plant parameters are based on typical plant-specific data with the exception of"Offsite power availability," which is a binary result that is specified by the analysis methodology.
2 Not sampled, see Section 4.9.AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Larqe Break LOCA Summary Report ANP-2853(NP)
Revision 000 Paqe.3-1 1i Table 3-2 Plant Operating Range Suppbrted by4he LOCA Analysis Event Operating Range 11.0 Plant-Physical Description, .......' -" : ". .... ... 1.1 Fuel-. .. ___...._______________
____ ....___ .___, ____________
_ a) Cladding outside diameter 0.376 in b) Claddihginside diameter' "7 0.328'in:.
': c) Cladding thickness
.. ......_,_ 0.024 in.d) Pellet outside diameter, ___,_0.3215 in.e) Initial Pellet density!..
95 percent.of theoretical f) Active fuel length .'-' 144 in....... g).Resinterdensification
..._*_____..____]..
_"_.. __._.-.._ h) Gd 2 0 3 codn'centrations
.2, 4, 6, 8 w/o.. ... 1.2 R C S -_ _ _ _ __... i_..... ..a) low resistance
.. .. ....... Analysis b) Pressurizer location Analysis assumes.location giving... ..__ .most limiting PCT.(broken loop)c) Hot assembly location Anywhere.incore d) Hot assembly type: 17x17* e) SG tube pluggirig 3 percent.2.0 Plant Initial Operating Conditions 2.1 Reactor-Power-
___-,__....
a) Analyzed reactor power " 2958 MWt..b)F ..- 2.522 c) FAH 1 < 1.73.d) MTC 0 at HFP 2.2 Fluid Conditions a) Loop flow 109.2 Mlbm/hr < M < 117.8 Mlbm/hr b) RCS average temperature 582.0°F < T < 594.8 0 F c) Upper head temperature -Tcold Temperature 4 1 2 3 4 Includes 2% measurement uncertainties Ensures that a minimum 7 percent peaking margin is maintained to the FQ limits when operating at the positive or negative AFD limit Includes 4 percent measurement uncertainty Upper head temperature will change based on sampling of RCS temperature AREVA NP Inc.
Harris Nuclear Plant Unit 1.Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000_.-Page3-12 Table. 3-2 ;Plant Operating
- Range Supported bythe LOCA -Analysis (Continued through Page 3-14).. ... " ... ..-.d) Pressurizer-pressure 2200 psia
- P 2288 "ps'ia .e) Pressurizer level 53.25 percent.
L< 66.75 pe'rdrent
.. .f) Accumulator pressure..
.599.7.psia._<,P,<:679.7 psia-:>:: g) Accumulator liquid Yolume 994 6 ft 3<.V g, 1029.4 ft 3 h) Accumulator temperature 80°F T _ 130-F (It's Coupled with containment
.... ... .... .. ' ... ..... te m p e ra tu re ) ....... .... .. .i) Accurmhulator-resistance fL/D As-built piping configuration' j) Minimum ECCS boron > 2400 ppm 3.0 Accident Boundary Conditions a) Break location Any RCSipi cation b) Break type Double-end6d guillotine or, split c) Break size (each side, relative-to-cold 0.27 < A < 1.0 full pipe area (split).leg pipe area) .. 0.27 < A < 1.0 full pipe area (guillotine) d) Worst single-failure
-Loss of one train of ECCS_ e) Offsite power On or Off f)ECCS pumped injectiontemperature 125 0 F g) HHSI pump delay 17 s (w/ offsite power).. ... .. ...... ... 29 s- w/o offsite power)_-h) LHSI pump delay:..-
27 s (wI offsite power)37 s (w/o offsite power)i) Containment pressure 14.7 psia, nominal value.j) Containment temperature.
80°F <.T _ 1306F'k) Containment sprays delay 0 s 1) Containment spray water temperature 40'F .m) LHSI Flow BROKEN LOOP..RCS pressure LHSI'.fllow
psia gpm 0. 1832.0 15. 1832,0.0.20. 1791.1 30.. 1707.6 35 .1664.9 40. 1621.5 50. 1532.5 70. 1318.8 120. 546.2 125. 491.9 125.01 0.0 3000. 0.0 INTACT.LOOPI.
- RCS pressure LHSI flow----------
ps~ia gp, 0. 916.0 AREVA NP Inc.
Harris Nuclear Plant Unit 1, Realistic Large Break LOCA Summary Report:ANP-2853(NP)
Revision 000'Page 3-13 f" ! * ":' '/'. "' ": ' ' "L', i' ". , Table 3-2...(continued) 15.20.30.35.40.50.70.120.125.125.01 3000.916..0.895.6 853. 8 810.-8 766.3 659.3 273.1 246.0 0.0 0.0 INTACTLOOP2
- RCS pressure-----------
psia 0.15.20.30.35.40.50.70.120.125.125.01 3000.LHSI flow gpm 916.0 916.0 895.6 853.8 832.4 810.8 766.3 659.3 273.1 246.0 0.0 0.0 n) HHSI Flow BROKENLOOP
- RCS Pressure-----------
psia 10.15.20.30.40.50.70.120.500.1001.1150.1609.1775.2037.2141.2193.2246.2296.3000..INTACTLOOP1
- RCS Pressure-----------
psia 10.15.20.30.HHSI Flow gpm 206.3 206.3 206.1 205.7 205.3 204.9 204.1 202. I 186.3 161.9 154.0 124.4 114.5 91.2 72.7 60.8 35.1 0.0 0.0 HHSI Flow gpm 129.6 129.6 129.4 129.2 AREVA NP Inc.
Harris Nuclear Plant U nit 1,.... , Realistic Lame Break LOCA Summary Report ANP,-2853(NP)
Revision 000*Pale 3ý14 Table 3-2 (continued) 40.50, 70.120.500.1001.1150.1609.1775.2037.2141.2193.2246.2296.3000.INTACT LOOP2* RCS Pressure---------psia 10.15.20.30.40.50.70.120.500.1001.1150.1609.1775.2037.2141.2193.2246.2296.3000.128.9 128.7 128.2 ...126.9'..117... 0 101.7 96.8 78.3 72.4 58.7 49.2 44.6 28.6 0.0 0.0 HHSI Flow gpm 129.6 129.6 129.4 129.2 128.9 128.7 128.2 126.9 172. 0 101.7 96.8 78.3 72.4 58.7 49.2 44.6 28.6 0.0 0.0 AREVA NP Inc.
Harris Nuclear'Plant Unit 1 .-Realistic Large Break LOCA Summary Report ,ANP-2853(NP)
Revision 000, Page:3-15' Table 3 Statistical
- Distributions.Used for Process Parameters 1 Operational
[.Measurement
,- I Paraimeter Uncertainty
.Parameter Ranee .Uhcertainty S-tndard Ditiutioni
>'Distributioni evain Pressurizer Pressure (psia),- ..Uniformý .2200 -2288 N/A N/A Pressurizer LiquidLevel (percent)
Uniform,, 53.25-- N/A , .: N/A Accumulator Liquid Volume (ft 3) .Uniform .994.6 1029.4 N/A N/A Accumulator Pressure (psia) Uniform ' 5997 -!.679.7,'!":'ý,ý N/A N/A Containment Temperature
('F) -Uniform 80-130 N/A N/A ""_"__Containment Volume ( ft 3) Uniform 2.266E&6 -2.610E+6 N/A .' N/A Initial RCS Flow Rate (Mlbm/hr)
Uniform 109.2 -117:8. , N/A :N/A Initial RCS Operating Temperature Uniform 58M 594.8 N/A (Tavg) ('F) N/A RWST Temperature for ECCS (TF) Point 125 N/A " N/A Offsite Power Availabiiity , Binary, 01 .-N!A N/A Delay for Containment Quench Point' 0 .N/A N/A LHSpra CoopDlng (s) Pont 27 (w/ offsite power) N/A LHSI Pump Delay (s) 37 o offsite' .N/A N/A HHSI PumpDelay (s) Point 17 (w/.offsite power) N/A ;N/1 29 (w/o offsite power) _ _N/A______
_..., N/A._ ___1 2 3 Note that core power is not sampled, see Section 1.0 All measurement uncertainties were incorporated into the operational ranges This is no longer a sampled parameter.
One set of 59 cases is run with LOOP and one set of 59 cases is run with No-LOOP.AREVA NP Inc.
Harris Nuclear Plant Unit 1'Realistib Large Break LQ.CASumary.
Report ANP-2853(NP)
Revision 000 Page 3-16 Table 3-4 ;RLBLOCA EM SER'Conditions andLimitations:
.'-.,SER Conditions-ahd Limitations Response A CCFL!violation warn wilbe.added too a6erthe analyst There was no significant occurrence of CCFL violation in the-to CCFL violation in the downcomer should suc-roccur..----
downlomner for this. evaluation-:
-Violations of CCFL were ,._____.....
.....________ ....__,____ _. -noted in a statisticallyinsignificant number of time steps.2. AREVA NP: has agreed that it is not to use H,,Hot leg nozzle gaps were not modeled.:
.w ith hot leg to dow ncom er nozzle gaps.. ... .. ....3. If AREVA NP applies the RLBLOCA methodology to plants- :The HNP. analysis LHGR is consistent -with the 3-loop using a higher planailineaY heat generation rate (PLHGR). _sample-problem LHGR.-than used in' the current analysis, br if the methodology, is -...-to be 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 enbgineering:.
experience and should be documented in -either the -- ... -.. ..-RLBLOCA guideline or plant specific calculation file. --4. Slot breaks on the top of the pipe have-not been evaluated.
The evaluation of high. elevation slot.breaks is doc'umented These breaks could cause the loop seals to refill during late in the AREVA R.LBLOCA analysis guidelines.
reflood and the core to uncover again. These break j..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, AREVA NP will- ..evaluate the effect of the deep loop seal on the slot breaks.The evaluation may be based on relevant engineering experience and should be documented in either the RLBLOCA guideline or plant-specific calculation file.5. The model applies to 3 and 4 loop Westinghouse-and HNP is a Westinghouse 3-loop plant.CE-designed nuclear steam systems.6. The model applies to bottom reflood plants only (cold side HNP is a bottom reflood plant.injection into the cold legs at the reactor coolant discharge piping).7. The model is valid as long as blowdown quench does not The limiting case did not show any evidence of a blowdown occur. If blowdown quench occurs, additional justification quench.for the blowdown heat transfer model and uncertainty are needed or the calculation is corrected.
A blowdown quench is characterized by a temperature reduction of the peak cladding temperature (PCT) node to saturation temperature during the blowdown period.8. The reflood model applies to bottom-up quench behavior.
Core quench initiated at the bottom of the core and If a top-down quench occurs, the model is to be justified or proceeded upward.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.AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 31.-17 7".Table 3-4 -,RLBLOCA-SER Conditions-and Limitations (Continued)
SER Conditions and Limitations
.,Response
- 9. The model does not determine whether Long-term cooling was not evaluated in this analysis.
For the HNP's Criterion 5 of 10 CFR. 50.46, long term- -assessment-of-long- term cooling, please refer to Chapter 6.3.3 and cooling, has been satisfied..
This will be Chapter 15.6.5 of HNP FSAR;.determined by each applicant orlicensee as.part of its application of this methodology
........._.10. Specific guidelines must be used to develop The:nodalization in t he;plant model is consistent with the Westinghouse 3-the plant-specific nodalization. ,Deviations -loop sample calculation that was submittedto the NRC for review. Figure from the reference plant must-be addressed.
3-,1 showthelO6p.noding..used .in this.anaysis. (Note only Loop 1 is shown in the figure; Loops 2 and 3 are identical to loop 1, except that only Loop 1 contains the pressurizer and the break.) Figure 3-2 shows the steam generator model. Figures 3-3, 3-4, and 3-5 show the reactor vessel noding diagrams.11..A table that._ contains the .plant-specific Simulation
.of. .,clad .temperature,.
response J.is a function of parameters and the range of the values phenomenological correlations that have been derived either analytically considered for the selected-parameter during- or experimentally.-
The important correlations have been validated for the the topical report approval process must be RLBLOCA methodology and a statement ofthe range of applicability has provided.
When plant-specific parameters been documented.
The correlations of interest are the set of heat transfer are outside the range used in demonstrating correlations as described in Reference
- 1. Table 3-7 presents the acceptable code performance, the licensee or summary of the full range of applicability-for the important heat transfer applicant will submit sensitivity studies to correlations, as well as* the ranges calculated in the limiting case of this show the effects of that deviation..
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 a're '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., 13. The licensee or applicant wishing to apply The HNP plant will have 17x17 HTP fuel bundles with Zirc-4 clad.AREVA NP realistic large break loss-of- ......coolant accident (RLBLOCA) methodology to ." M5 clad fuel must request an exemption for ..its use until the planned rulemaking to modify 10 CFR 50.46(a)(i) to' include M5 cladding material has been completed.
____ __AREVA NP Inc.
Harris Nuclear.Plant Unit 1. " .Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page3-18 Table 3-5 Summary of Results for the Limiting PCT Case Case # 5 U0 2 Rod S P C .T. .. ...... ............ .. ... ......................emperature
.. 130°F:-Time ~ i'~' '126 Elevation:
7 .':. ', ., 10.043,ft Metal-W ater Reaction .........._____.Percent:Oxidation*Maxi-mum;, , 1.9498-.7 .Percent I otal Oxldation
..u.ub65 I -Table 3-6 Calculated Event Times for the Limiting PCT Case Event' "'Tinie (s)Break Opened, ... 0.0 RCP"Trip " N/A SIAS Issued:..
0.6 Start
ofBroken Loop Accumýulator Injection 19.0-Start of Intact Loop Accumulator Injection2 (Loops 2 and 3 respectively)
Broken Loop HHSI Delivery Began -29.6 Intact Loop HHSI Delivery'Began 29.6, 29.6 (Loops 2 and 3 respectively)
Broken Loop LHSI Delivery Began '37.6 IntactLoop LHSI Delivery Began 37.6, 37.6 (Loops 2 and 3 respectively)
Beginning of Core Recovery (Beginning of Reflood) ..41.6 Broken Loop Accumulator Emptied 44.9 Intact Loop Accumulators Emptied 45.6, 45.4 (Loops 2 and 3 respectively)
PCT Occurred 132.6 Transient Calculation Terminated 610.1 AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
- Unit.! '-" Revision 000-Realistic Large Break LOCA Summary Report Page'3-19 ,:Table,3-7 ,Heat,,Transfer-Parameters for the Limiting Case 4 4~'AREVA NP Inc.
- Harris Nuclear Plant-'U nit 1 Realistic Larqe Break LOCA Summary Report ANP-2853(NP)
Revision 000 S". Paqe.e3-20
..Table 3-8 Containment-Initial and Bou ndaryConditions Containment Net Free Volume (ft 3) 2,266,000
-2,610,000 Initial Conditions Containment Pressure (nominal) 14.7 psia Containment Temperature 80°F -130°F RWST Temperature 125 0 F Outside Temperature 40°F Humidity 1.0 Containment Spray (only Quench System Sprays are considered)
Number of Pumps operating 2 Quench System Total Spray Flow 5,000 gpm Minimum Spray Temperature 40°F Fastest Post-LOCA initiation of spray 0 sec AREVA NP Inc.
Harris Nuclear Plant Unit 1 ......Realistic Larae Break LOCA Summary Report ANP.-2853(NP)
Revision 000 Paae 3-21 Table 3-9 Passive Heat Sinks in Containment Structure name Surface Area(ft 2) Slab Material Th'ickness (in)-Containment Dome Paint-2 ,:'0.005 .-26546.0 Carbon Steel 0.5 Concrete 30.0 External Cylinder Wall Paint-2 0.005 63065.0 Carbon Steel 0.375 Concrete 54:0 1 In. Steel'Liner Concrete' Paint-2 0.005 2280.0 Carbon Steel 1.0 Concrete 54.0 Concrete 82525.0 Concrete 45.0 Stainless Steel Liner Conrrete ' Stainless Steel 0.1872 6756.0 Concrete 0.6 Sump 29320.0 Concrete -45.0 Piping , Paint-3 0.005... ... .. 5703.0 0.. .. ... ...005 ..5703.0.. Carbon. Steel ... 0:1966 Piping ' ...... .. .. .:. ,~~Paint-3
..- ... ...0 ..Piping .3870.0 0.005 3870.0 .Carbon Steel 0.4181.Structural Heat Sink. Paint-2 0.005 5Carbon Steel .0.312 Electrical
-. Galvanizing (Zinc) -0,.0015 33066.0 ".Carbon Steel 0.1745 Embedded Stainless Stainless.
Steel -0.3902.1030.0 Concrete 3.2244 Effective Stainless (Not Embedded, Steel Pipe, Structural 9143.0 Stainless Steel 0.22397 Steel, and Strainer Screen)Structural Heat Sink Paint-2 0.005 30300.0 Carbon Steel 1.0 Not Embedded Structural Paint-2 0.005 119467.0 Carbon Steel 0.1738 AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 , Page 3-22 Table 3-9 Passive Heat Sinks in Containment (continued)
Structure name I Surface Area(ft) -Slab Material~~
Thickness (in)Structural Heat Sink .Paint-2.
0.005--66753.0 Carbon Steel- 0.5004.-Embedded StructUiral.
Paint-2 0.005 1. 34;72.0 .Carbon-Steel 7, 0.3405" Concrete ..3.2244 Embedded Structural Paint-2 0.005 13899.0 Carbon Steel 1.444 .Concrete 3.2244 Ductwork 5430.0.Paint-4 0.008.. ..... .....5430 .0 .Carbon Steel 0.1248 Ductwork Galvanizing Zinc 0.0015 39672. 0:;1-, 396..0 .Carbon Steel' 0.029 Seismic Hangers 84386.6 Paint-2 0.005 Carbon Steel 0.1876 Material Properties.
Thermal conductivity Volumetric heat capacity Material f (Btulhr-ft-°F) (Btuihr~ft 3--F)Carbon Steel 26.0. 53.9 Paint-2' 0.23 42.6.Paint-3 .,0.23 147.0 Paint-4 0.23' 42.6 Galvanizing (Zinc) 64.0 40.6 Concrete, _.-0.92 22.62 Stainless Steel 9.4 53.9 AREVA NP Inc.
Hatris Nuclear Plant Unit 1 .Realistic Large Break LOCA Summary Report.ANP-21853(NP),, Revision 000... Page 3-23 Figure 3-1 Primary System Noding AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic-Large Break LOCA Summary Report ANPR2853(NP)
Revision 000 3-24.Figure 3-2 Secondary System Noding AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP), Revision 000.;Page 3-25: Figure 3-3 Reactor Vessel Noding AREVA NP Inc.
Harris Nuclear-Plant Unit 1 Realistic.
Large Break LOCA Summary Report ANP-2853(NP-)
Revision 000 Page 3-26 Figure 3-4 Core Noding Detail AREVA NP Inc.
Harris Nuclear Plant Urnit 1 .,. -Realistic Large Break LOCA Summary Report ANP-2853(NP)*
Revision 000,--,Page 3-27 Figure 3-5 Upper Plenum Noding Detail AREVA NP Inc.
Harris Nuclear Plant Unit 1, : Realistic Large Break LOCA Summary eport ANP-2853(NP)
Revision 000 Paqe 3-28 0 B)ne-Sided
-reak Area ommmmo o oo oom ooo (ft 2/side)1.0 2.0 3.0 4.0 Burn Time F m N e OOOOOOOO ge
- g m (hours)0.0 5000.0 10000.0 15 C o r e ' F 'Power (MW)2957.0 2957.5 2958.0 2958.5 2 Fq m m m mo Peaking 1.8 1.9 2.0 2.1 2.2 2.3 2.4 2.5 AO-04 -3 ,- 2 -I 0 0.2 0 3-0.4 -0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 5.0 000.0 959.0 2.6 0.4 Pressurizer Pressure (psia)Pressurizer Liquid Level (%)RCS (Tavg)Temperature
('F).0 2220.0 2240.0 2260.0 2280.0 230 0 55.0 60.0 65.0 70 50.('.0 580.0 585.0 590.0 595.0 Figure 3-6 Scatter Plot of Operational Parameters AREVA NP Inc.
Harris NuclearPlant Unit 1 ,.. .-RealisticLarcme Break LOCA Summary Report ANP-2853(NP)
Revision 000.Pacqe 3-29 Total .Loop Flow -O0 m i I mI inmm ý6miOO (Mlb/hr) k 108.0 110.0 112.0 114.0 116.0 118.0 Accumulator , I I Liquid Volume em i llON m eenlmm e 990.0 1000.0 1010.0 1020.0 1030.0 Accumulator
'Pressure -N m mmmcme me mmcO NOe (psia)600.0 620.0 640.0 660.0 680.0 Containment I Volume ONee l i~ g 00 i m (ft3)2.20e+06 2.30e+06 2.40e+06 2.50e+06 2.60e+06 2.70e+06 Containment F .. " (Accumulator) mmmc m Ome l Oil N a OO mO-O Temperature
('F), , 80.0 90.0 100.0 110.0 120.0 130.0 Figure 3-6 Scatter Plot of Operational Parameters (Continued)
AREVA NP Inc.
Harris Nuclear Plant1 Realistic Larcie Break LOCA Summary Report ANP-2853(NP)
Revision 000 Pacie 330: PCT vs Time of PCT 2000 1800 1600 1400 1[mE]m lpli Break 0L G n U Split Break El Guillotine Break E-0-1200 1000 800 600 400 0 100 200 300 Time of PCT (s)400 500 Figure 3-7 PCT versus PCT Time Scatter Plot from 59 Calculations AREVA NP Inc.
HarrisNuclear Plant U n it 1 -Realistic Larce Break LOCA Summary Report ANP-2853(NP).
Revision 000 Paqe 3-31 PCT vs One-sided Break Area 2000 1800 1600 1400 1200 D E 012000 800 600 400 1.0 0I E u E*F F1i~ LI LI El* Split Break EL Guillotine Break 2.0 3.0 " Break Area (ft 2/side)4.0 5.0 Figure 3-8 PCT versus Break Size Scatter Plot from 59 Calculations AREVA NP Inc.
Harris Nuclear Plant Unit 1 RealisticLarge Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 3-%32 Maximum Oxidation vs PCT 3.0 2.8 2.6 2.4 2.2 2.0 1.8" 1.6 0-g 1.4 X 0 1.2 1.0 0.8 0.6 0.4 0.2 0.0 40 a Split Break El Guillotine Break!El[] 0 DD 10E[] 0 0] [.... 91)0 600 800 1000 1200 1400 1600 PCT (°F)1800 2000 Figure 3-9 Maximum Oxidation versus PCT Scatter Plot from 59 Calculations AREVA NP Inc.
Harris Nuclear Plant Unit 1 RealisticLarge Break LOCA Summary Report S.ANP,-2853(NP)
Revision 000.. .PaQe 3-33 Total Oxidation vs PCT 0.10 N Split Break ED Guillotine Break El 0.08 k 0.06 0 0-X 0 0.04 0.02 F nI D1 D IýEl]LI U]0.00 L 400 600 800 1000 1200_ PCT (0 F)1400 1600 1800 2000 Figure 3-10 Total Oxidation versus PCT Scatter Plot from 59 Calculations AREVA NP Inc.
Harris Nuclear'Plant Unit 1 i ýRealistic Large Break LOCA-Summary-Report ANP-2853(NP)
Revision 000'.Page.3-34 PCT Trace for Case #5 PCT = 1929.4 OF, at Time = 132.63 s, on Hot U02 Rod 2000 1500 E 1000 0-U5 500 0 0 800 Time (s)ID:05622 290ct2009 17:56:48 R5DMX Figure 3-11 Peak Cladding Temperature (Independent of Elevation) for the Limiting Case AREVA NP Inc.
Harris:Nuclear Plant Unit 1 I Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 3-35-Break-ý Fl ow, 60 40 0 E ai)M 0 U-20 0-20 L 0 200,-ý- 400 600 800 Time (s)ID:05622 290ct2009 17:56:48 R5DMX Figure 3-12 Break Flow for the Limiting Case AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ,ANP-2853(NP)
Revision 000'F- Page 3-36 Core I nlet Ma ,s;'Flux 800 600 C )E-D C,)('3 400 200 0-200 0 200 400 600. -Time (s)800 ID:05622 290ct2009 17:56:48 R5DMX Figure 3-13 Core Inlet Mass Flux for the Limiting Case AREVA NP Inc.
Harris Nuclear Plant Unit 1 , , .Realistic Larae Break LOCA Summary Report ANP-2853(NPP)
Revision 000.Paae.3-37-.
Core Outlet Mass Flux- -900 700 500 E U)U)300 100-100-300-500 0 200 400,, 600 800 Time (s)ID:05622 290ct2009 17:56:48 R5DMX Figure 3-14. Core Outlet Mass Flux for the Limiting Case AREVA NP Inc.
Harris Nuclear 'Plant Unit 1 -.: Realistic Large Break LOCA Summary Report..-ANP-2853(NP)
Revision 000 Page 3-38 Pumnp Void 'Fraction 1.0 0.8 0.6 0.4 C: 0 cc U-0_I -__ Broken Loop 1 Intact Loop .2 : Intact Loop 3 0.2 0.0 0 200 400 600 800 Time (s)ID:05622 290ct2009 17:56:48 R5DMX Figure 3-15 Void Fraction at RCS Pumps forthe Limiting Case AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Summary Report-ANP-2853(NP)
Revision 000: , Pace 3-39.ECCS Flowsi".4000 3000 E 2000 0 LL 1000 0 0 200 .400- :. 600 Time (s)800 ID:05622 290ct2009 17:56:48 R5DMX Figure 3-16 ECCS Flows (Includes Accumulator, Charging, SI and RHR) for the Limiting Case AREVA NP Inc.
Harris Nuclear Plant Unit 1 :...Realistic Large Break LOCA Summar Report ANP-2853(NP)
Revision 000.Page 3-40 Upper Plenum Pressure 3000 2000 0~0)U)U)0)1000 0 0 200 400.. 600 .800, Time (s)-ID:05622 290ct2009 17:56:48 R5DMX Figure 3-17. Upper Plenum Pressure for the Limiting Case AREVA NP Inc.
Harris Nuclear Plant Unit 1.: -.-Realistic Large Break LOCA Summary Report ANP-2853(NP), Revision 000-, Page 3-41-Downcomer Liquid Leveli,,,, 30 20 U,-j._10 0 0 200 400 600 Time (s)800 ID:05622 290ct2009 17:56:48 R5DMX Figure 3-18 Collapsed Liquid Level in the Downcomer for the Limiting Case AREVA NP Inc.
Harris, Nluclear Plant Unit 1 Rdalistic Larae Break LOCA Summary Report*ANP-2853(NP)
Revision 000 Pacqe 42:-..............
-LowebrzF" ssel, Liquid Level-10 8 6-J 4 2 0 0 200 " 400 600 Tim6 (s)800 ID:05622 290ct2009 17:56:48 R5DMX Figure 3-19 Collapsed Liquid, Level in the Lower Plenum for the Limiting.
Case AREVA NP Inc.
Harris Nuclear Plant U nit 1.Realistic.
LarQe Break LOCA Summary Report ANP-2853(NP)
Revision 000 ,.... .-/.Pa ie,3-.43 Core Liquid: Level..-.15 10 (D-j V 5 0 0 200 400 600 Time,(s): 800 ID:05622 290ct2009 17:56:48 R5DMX Figure 3-20 Collapsed Liquid Levelin the Core for the Limiting Case AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large.Break LOCA Summary _Repprt ANP-2853(NP)
Revision 000 Page..3-44 Containment" and LOop Pressures (U 0~0):3 Cd)U, C)0~100 90 80 70 60 50 40 30 20 10 0 0 200 400 600 Time (s)800 -*ID:05622 290ct2009 17:56:48 R5DMX Figure 3-21 Containment and Loop Pressures for the Limiting Case AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000.Page 3-45 2000 T------------I ----------#LOOP -NoLOOP.1800 +/-b 0 1* *03 W: *3 0 .0 03 w 1600 +/-0 0*1 0-i 0[0 **0 13 03:13~-7 -~~ 0 I~j El~~ ~ -* --0-*3 2000 1800 1600 1400 1200 1000 1400 +0-0.0[][]0 0 1200 +0:[0 '0 3-13 0. 0 ..[] +/ .+1000 +P 800 800 0 M0'20..30 Case Number 40-U /, 50 60 Figure:3-22 GDC 35 LOOP Versus No-LOOP Cases AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report Page 4-1 4.0 Generic Support for Transition Package The following sections are responses to typical RAI questions posed by the NRC on EMF-2103 Revision 0 plant applications.
In some ihstances,-
these requests, cross-referenced documentation provided on dockets other than those for which the request is-made. AREVA discussed these and similar questions from the NRC draft SER for Revision 1 of EMF-2103 in a meeting with. the NRC on December 12, 2007. AREVA agreed to provide the following" additional information within new submittals of a Realistic Large!Break LOCA report. The NRC questions have been modified to fit the context of the pre-emptive consideration of the NRC's feedback provided in the review of RLBLOCA applications on other dockets.4.1 Reactor Power Question:
It is indicated in the RLBLOCA analyses that the assumed reactor core power"includes uncertainties." The use of a reactor power assumption other than 102 percent, regardless of BE or Appendix K methodology, is permitted by Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix K.I.A, "Required and Acceptable Features of The Evaluation Models, 'Sources of Heat During a LOCA." However, Appendix K.LA also states: "...Ah assumed power level lower than the level specified in this paragraph
[1.02 times the licensed power level], (but not less than the licensed power level) may be used provided.
.." Please explain.Response:
As indicated in Item 2.1 of Table 3-2 herein, the analyzed reactor core power for the HNP Realistic Large Break Loss-of-coolant Accident is 2958 MWt. The value represents the maximum power measurement uncertainty of 2% measurement uncertainty to the current rated'thermal power, (2900 MWt).4.2 Rod Quench Question:
Does the version of S-RELAP5 used to perform the computer runs assure that the void fraction is less than 95 percent and the fuel cladding temperature is less than 900'F before it allows rod quench?Response:
Yes, the version of S-RELAP5 employed for the HNP requires that both the void fraction is less than 0.95 and the clad temperature is less than the minimum temperature for film boiling heat transfer (Tmin) before the rod is allowed to quench. Tmin is a sampled parameter in AREVA NP Inc.
Harris Nuclear Plant -ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report .Page 4z2'the RLBLOCA -methodology.
that typically does: not -exceed,.755 0 K (900'F). Thisis.!a .change to the approved.
RLBLOCA EM -(Reference 1)..This feature'is -carried forward into.the UAPRO9.version of S-RELAP5.4.3Rod-to-Rod:
Thermal. Radiation
., .*., , .*Ques-tion:Provide justification that the S-RELAP5 rod-to-rod thermal radiation model applies to the HNP care....... .....,. :: : ..: ..:- -. ..: Response:, The'Realistic.LBLOCA methodology, (Reference
'1), does not,.provide modeling :of rod-to-rod radiation.
The, fuel rod surfaceheat.transfer, processes:.included, in the-:solution, at.high temperaturesare:::
film boiling,;
convection-to steam,, rod:to liquid, radiation.and rod to vapor radiation.
- This heat :transfer
ýpackage was..assessed against'various' experimental, data. sets involving both.moderate (1600'F -2000YF) and high, (2000cF=to over.220QT) peak cladding temperatures and. shown to be conservative when applied nominally. -The. normal distribution of the experimental data was then determined.
During the execution of an:.RLBLOCA evaluation, the heat transferred from a fuel rod is determined .by. the application of.-a. multiplier.to the nominal heat transfer model. This multiplier is determined by a random sampling of the normal distribution"of the experimental'data benchmarked.
Because the data include the effects of rod-to-rod radiation, it is reasonable to conclude that the modeling 'implicitly includes an allocation for rod-to-rod radiation effects. As Will be'demonstrated, the approach:
is reasonabie because the *conditions within actual limiting fuel'assemblies assure that the actual rod-to-rod'radiation is larger than the allocation provided through normalization to the experiments.
The FLECHT-SEASET tests. evaluated covered a range of PCTs from 1,651 to 2,239cF and: the THTF tests covered a range of PCTs from 1,000 to 2,2007., Since the. test bundle in either, FLECHT-SEASET or THTF is surrounded by a test.vessel, which is relatively cool compared to, the. heater rods, substantial radiation from.the periphery rods to the vessel wall can occur. The rods selected for assessing the RLBLOCA reflood heat transfer package were chosen from the.interior of the test assemblies to minimize the impact of radiation heat transfer to the test vessel.The result was that the assessment rods comprise a set which is primarily isolated from cold wall effects by being surrounded by powered rods at reasonably high temperatures.'
As a final assessment, threebenchmarks independent of THTF and FLECHT-SEASET were performed.
These benchmarks.
were.selected from the Cylindrical Core Test Facility (CCTF), AREVA NP Inc.
Harris NuclearPlant ,ANP-2853(NP)
Unit 1. " Revision 000 Realistic Large Break LOCA Summary Report .,.. Page 4-3 LOFT, and, the .Semiscare
,;facilities.-
- Because-theseý facilities
'are more integral.
tests; and together.
cover a rwide, range .of -scale, they, also -serve,.to show that_ scale .effects are accommodated within the code calculations.
-The results of these calculations are provided in Section 4.3.4, Evaluation)of Code, Biases, page 4-100, of Reference 1, The CCTF results are, shown in Figures 4.180 through 4.192, the LOFT results in Figures 4.193 through 4.201, and the Semiscale results in Figures 4.202 through 4.207. As expected, these figures demonstrate that the comparison between the code calculations.
arnd. data lis- improved with the applicatic n'of the deriveddbiases:.The CCTF,. LOFT, and Semiscale benchmarks.
further indicate that, whatever consideration of.rod-to-rodý radiation is implicit' in the S-RELAP5 reflood heat transferimodeling, it',does not significantly effect code predictions' under conditions -where 'radiation is minimized.
- The.measured PCTs in these assessments rangedfrom approximately 1,000 to 1,540cF.
- -At these temperatures,.there is little-rod-to-rod
'adiation.'
Given the good' agreementbetween ,the biased code calculations and the CCTF:,: LOF.T,: and Semiscale data,'-it can be0`c6ncluded that there is -no significant over prediction Of the-total heat transfer coefficient.
.Notwithstanding.
any conservatism evidenced by experimental benchmarks, the application of the model to commercial nuclear power plants provides some additional margins due to limitations within the experiments.
The benchmarked experiments, -FLECHET SEASET and ORNL Thermal Hydraulic Test Facility (THTF), used to assess the S-RELAP5 heat transfer model, Reference 1, incorporated constant, rod powers across the, experimental assembly.Temperature differences that occurred were the result of guide tube, shroud or local heat transfer-effects: In the operation' f a pressurized water 'reactor (PWR)-and in the RLBLOCA evaluation;a
- radial local peaking factor is present, ;creating power differences that tend to enhance'the temperature differences between rods. '-In turn, these temperature differences lead to increases
'in net radiation heat transfer 'from. the hotter rods. :.The expected rod-to-rod radiation will likely exceed that embodied'within the experimental:results:
'... " 4.3.1 Assessment of Rod-to-Rod Raediation Implicit in the RLBLOCA Methodoloqy As discussed above, the FLECHT-SEASET and THTF tests were selected to assess and determine the S-RELAP5 code heat transfer bias and uncertainty.
Uniform radial power distribution was -used in these test-bundles.
Therefore, the rod-to-rod temperature-variation in the rods away-from the vessel wall is caused primarily by the variation in the sub-channel fluid AREVA NP Inc.
Harris Nuclear Plant .*ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report ...' Page.4-4 conditions:',.
Inthet real:operating fuel.bundle, on.,thO other can-be-.5:to'i10..percent rod-to-vrod,.power variation..'
In-addition', the methodology.; -includes-, a. provision to. apply the uncertaintyrneasurement to,,the hot pin.- " .. ... .. :-. ... -," -Table 4-1 provides the hot pin measurement uncertainty and a representative local pin peaking factor for several plants. These factors, however, relate the pin to the assembly average. To more properly assess the conditions under which rod-to-rod radiation heat transfer occurs, a more local peaking assessment is required.
Therefore, .the plant rod-to-rod, radiation assessments, herein set the average pin power for those pins surrounding the hot pin at 96 percent of that of the peak pin. For pins further removed the average power is set to 94 percent.4.3.2 Quantification of the Impact of Thermal' Radiation using R2RRAD1Code
..,i The R2RRAD radiative heat'transfer mnode Was de'veloped by Los-Alamos National Laboratory (LANL) to be incorporated in the BW/R version of the- TRAC 'code.' The'theoretical basis for this code is given'in References 8 'and 1'and is similarr-to that developed'-in the HUXY rod heatup code (Reference 10, Setion 2.1.2) used by AREVA for BWR LOCA applications.
The-version of R2RRAD used herein was obtained' from' the NRC to examine the" rod- tb'-rod :radiation characteristics of a 5x5 rod segment of the 161 rod FLECHT-SEASET bundle : The output provided by the R2RRAD code includes an estimate of the net radiation heat transfer from each rod in the defined array. The code allows the input of different temperatures for each rod as well as for a boundary surrounding the pin array. No geometry differences between pin locations are allowed. Even though this limitation affects the view factor calculations:, for .guide,. tubes*, R2RRAD is a reasonable tool to estimate rod-to-rod radiation heat transfer.The FLECHT-SEASET test series, was intended to simulate a .17x17 fuel assemblyand there is a 'close similarity, as shown in. Table 4-2, between the test bundle Iand a modern 17xl,7 assembly.Five FLECHT-SEASET tests (Reference
- 6) were selected -for evaiuation-and
'comparison with expected plant behavior.
Table 4-3 characterizes the results of each test. The 5x5 selected rod array comprises the. hot'rod, 4 guide tubes and. 20 near adjacent rods. The simulated hot rod is rod:7J in the tests.Two sets of runs were made simulating each of the five experiments and one set of cases was run to simulate the RLBLOCA evaluation of a limiting fuel assembly in an operating plant. For AREVA NP Inc.
Harris Nuclear.Plant
.ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report_ ___ :' ____ ' Page 4-5 the simulation of. Tests 3.31805;" 31504;' 31021, ahd 30817,1-,the thimble, tube.. (guide tube)temperatures were.set to thermeasured .values;::
For, Test:34420,:
the thimble tube-temperature was set equal to the measured vapor temperature.
For the first experimental.simulation set, the temperature of all 21 rods and the exterior boundary was set to the measured PCT of the si'mulated'test.
For the s~6o~d experirmie'ntal set, th6lhot rod tempia~tuie was set to the PCT value and the remaining 20 rodsa'6nd the boundary were set- toa temperature 251F cooler providing a reasonable measure of the'variation in surrounding temperatures.
To e stimate the rod-to-rod iatio in 'a real fu6el assenbly'at aLOCA conditions and:compare it to the experimental results, eachý of the"abh6ve cases' was 'rerun with the' hot rod PT set to the eremaining rods 0 ,tily set to tehiperatures "expected within experimental res ult 'and the'm a irgrdscon'rserVaifvey``t o the bundle. The guide tubes (thimble.
tubes.) werelremoved for conservatism and because peak rod powers -frequently occur at fuel assembly corners. away. from, either guide tubes or instrument tubes. In line.with the discussion in Section 4.3.1;.,:the surrounding 24 rods were set to a temperature estimated for rods of 4 percent lower power. The boundary temperature was estimated based an average power 6 percent below, the hot rod power. For- both of these, the temperature -estimates were achieved using a ratio of pin power to the difference in temperature between the saturation temperature and the PCT.: "T 2 4 rods = 0.96 (PCT -Tsat) + Tsat ::and ..Tsurround'ing region '= 0.-94 (PCT"- Tsat- ..Tsat was taken as 270 0 F :" Figure 4-2 shows the hot rod thermal radiation heat transfer for the two FLECHT-SEASET sets and for the plant set., The figure shows that-.for-PCTs greater than about 1700'F,: the hot rod thermal ra-diation in the plant .cases exceeds that -of the same. component
- within the experiments.
4.3.3. Rod-to-Rod Radiation Summary In summary, the conservatism of the heat transfer modeling' established by benchmark can be reasonably extended to plant applications, and the- plant local peaking provides a physical reason why rod-to-rod radiation should be more substantial within a plant environment than in the test environment.
Therefore, the lack of an explicit rod-to-rod radiation model, in the version of S-RELAP5 applied for realistic LOCA calculations, does not invalidate the conclusion that the AREVA NP Inc.
Harris Nuclear Plant Unit 1-Realistic Large Break LOCA Summary Report"ANP-2853(NP)
Revision 000 Page 4-6 cladding temperature and ldcal:cladding oxidation have been demonstrated tomeet the criteria of.10*CFR 50.46 with a high level: ofu.probability: , -.','- .-4.4 Film Boiling Heat Transfer Limit .Question:
4s,the, Forslund-Rohsenow model ;contribution to: theheat transfer coefficient-limited to. less than or.:equal to 15 percent when the,.void fraction.is greater than or0equa! to 0.97 -Response:
Yes, the version of S-RELAP5 eemployed for the HNP RLBLOCA analysis limits the contribution of the Forslund-Rohsenow model to no more than 15 percent of the total heat transfer at and abovea Because.the.
limit is applied~ata voidfraction of 0.9, the contribution of Forslund-Rohsenow within the, 0.7 to. 0.9 interpolation range. is, limited to -15 percent or less. This is a change to the approvedRLBLOCA EM (Referencei
).This feature is carried forward into the UARR09. version of S-RELAP5..
- 4.5 Downcomer Boiling Question:
If the PCT is greater than 1800'F or the-containmen t pressure isless than 30,psia, has the HNP downcomer model been rebenchmarked by performing sensitivity studies, assuming adequate downcomer noding in the water volume, vessel wall and other heat structures?.
...-,", ....Response:
The downcomer model for HNP has been established generically as adequate for the computation of downcomer phenomena including the prediction of potential local boiling effects. The" model .was 'benchmarked against the UPTF tests and the LOFT facility in thle RLBLOCA methodology, Revision 0 (Reference 1). FurtherI AREVAk addressed'the effects- of boiling in the downComer ina letter, fromrn3ames Malay" to U.S. NRC,,April 4', 203:' The letter cites'the lack of direct experimental evidence but contains studies: on' high and low pressure containments, the impact-of additional a-zimuthal n6ding within'the downcomer,-and the, influence of flow loss coefficieints.
Of these, the study on azimuthal noding is most germane to this question; indicating that additional azimuthal nodalization allows higher liquid buildup in portions.
of the downcomer away from the broken cold leg and increases the liquid driving head.Additionally, AREVA has conducted downcomer axial noding and wall heat release studies.Each of these studies supports the Revision 0 methodology and is documented later in this section.AREVA NP Inc.
Harris Nuclear Plant :ANP-2853(NP)
Unit 1 Revision Q00 Realistic Large Break LOCA Summary Report JPage.47 This question -is,. primarily:
concerned, with- the. phenomena'of.,downcomer.:
boiling and,.-the extension of the Revision 0 methodology and sensitivity studies to. plants with low.cohtainment pressures and high cladding temperatures.
Boiling, wherever it occurs, is a phenomenon that codes like S-RELAP5 have been developed to predict. Dwnhcomer'boiling is the'result"of the release of energy, stored* in"-vess_!&
m-nietal, mass.,.- Within ZS-RELAP5,.:downcomer, boiling is simulated, in the: nucleate:-
boiling:,regirnima&'ith' theI. Chenr correlation.
-This ' modeling has .been validated through the prediction of several assessments on boiling phenomenon provided in the S-RELAP5 Code Ver'ification and Validation document (Reference 12).Hot downcorrier walls penalize PCT by ltwo .,mechanisms:
by reducing : subc6oling of coolant entering the core and through the reductin., in 'down comer, hydraulic head' which is the driving force for core reflood. :AlthoughlbOiling in the downcomer.
1 occurs during blowdown, the biggest potential for impact on clad temperatures is during late reflood following the end of accumulator injection.
At this time, there is a large step reduction in coolant flow from the ECC systems. As a result, coolant entering the downcomer may be less subcooled.
When the downcomer coolant approaches saturation, boiling on the walls initiates, reducing the downcomer hydraulic staticlevel.'
' '.With the reduction of the downcomer level, the core inlet flow rate is reduced which,,depending on the existing core inventory, may result in a cladding temperature excursion or a slowing of the core cooldown rae.- '- .' -While downcomer boiling may impact. clad, temperatures, it .is. somewhat of a self-limiting process. If, cladding.
temperatures increase, less energy is transferred, in the core boiling process -and the loop.steam.
flows are reduced. This reduces the required driving head. to support continued core ,reflood and reduces the steam available to heat the ECCS water within the, cold legs resulting in.greater subcooling of the water entering the downcomer..
The impact of downcomer boiling is primarily dependent on the wall heat release'rate and on the ability to slip steam up the downcomer and out of the break. The higher the downcomer wall heat release, the more steam is generated within the downcomer and the larger the impact on core reflooding.
Similarly, the quicker the passage of steam up the downcomer, the less resident volume within the downcomer is occupied by steam and the lower the impact on the downcomer average density. Therefore, the ability to properly simulate downcomer boiling depends on both the heat release (boiling) model and on the ability to track steam rising through AREVA NP Inc.
Harris Nuclear'Plant ANP-2853(NP)
Unit 1 " -: Revision 000 Realistic Large Break LOCA Summary Report .Page'4-8 the downcomer.:, Considerationh of both-of: theseb.is-pitovided:
in the following Aext., The heat release modeling.in;S-RELAP5&s validated by, ai:sensitivity:
study-:ýon wall.: mesh. point spacirig andthrough bierichmarking aigainst .a closed form:-solutioni
- Steam- tracking is validated, through.both an axial and an azimuthal fluid control v0lumeiseh'snitivity-study done at lowIpressures.
- .The.
- results indicate that the-modeling .accuracy, withinijthe,-RLBLOCA.
methodology.is.isufficient, to'resolve the effects of downcomer boiling and that, to the extent; that -boiling., occurs; .the methodology properly resolves the impact on the cladding temperature and cladding oxidation rates.4.5.1 Wall Heat Release Rate The downcomer wall-heat release rate during reflood is conduction limited. and.depends on the vessel wall meshspacingused in the S-RELAP5 model..The followingtwo approaches are used to.. evaluate the adequacy ofthe downcomer vessel wall mesh spacing used in the S-RELAP5 model. .4.5.1.1 Exact Solution In this benchmark, the downcomer wall is considered as a semi-infinite plate'." Because the benchmark uses a closed form solution to verify the wall mesh spacing used in.S-RELAP5, -it is-assumed that the material has constant thermal properties, is-initially at :temperature Ti,. and, at time zero, has one surface, the surface simulating contact with the downcomer fluid, set~to.a constant temperature, T,, representing the fluid temperature.
Section 4.3 ofReference 9 gives the exact solution for the temperature profile as a function of time as (T(x.,t) TO) (TI -'T,)) erf {x (2-(at) )}. (1) ' .where, a is the thermal diffusivity of the material given by ..." " ' * .a =kI(p Op), k = thermal conductivity, p = density," Cp = specific heat, and erf{} is the Gauss error function (given in Table A-1 of Reference 9).AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1 Revision 000'Realistic'Large Break LOCA SummarReport
____-__._'
_ Page 4-9.The. conditions of, the b6nchmark:.are Ti, =.50 0 F and'T, -300 0 F: -The meshi-spacing -in 'S-RELAPI5.is the same' as 1hat ,used -for. the wall:, in the"!RLBLOCA model.Figure 4-4 -shows the 'temperaturef distributionsin; the:.metalat'.O.0, .100 and .300 seconds.:as calculated by ;usirig Equatio.r: -J:1.and :S-RELAP5; respectivbly.
.:,The solutions,:
are, identical; confirming the adequacy.of the binmsh!.slpacing used in the downcomer wall. ..415'.1:2 Plant Model Sensitivity'Siudy
' ' :" ' ; ' : As additional verification, a typical 4-loop plant case was used to evaluate the adequacy of the mesh spacing within the downcomer wall heat structure.
Each mesh interval in the base case downcomer vessel wall was divided into two equal intervals.
Thus, a new input- model was created by increasing the number of mesh intervals from 9 to 18.- Figures'4-5 through 4"-8 show the total downcomer metal heat release rate, PCT independent of. elevation,;.downcomer .liquid level, and the core liquid level,'respectively,,:for,.the' base case and the modified case. These results confirm the conclusion from the exact solution study that the mesh spacing used. in the.plant model for the downcomer vessel wall is adequate.4.5.2 Downcomer Fluid Distribution........
, To justify the adequacy of the downcomer nodalization in calculating the fluid distribution in the downcomer, two~studies varying separately the axial and.the azimuthal resolution with which the downcomer is modeled have been conducted., '.4.5.2.1 Azimuthal Nodalization In a letter to the NRC dated April, 2003 (Reference 1), AREVA documented several studies on downcomer boiling. Of significance here. is the study..on further azimuthal break up of the downcomer noding. The study, based on a 3-loop plant with a containment pressure of approximately 30 psia during reflood, consisted of several calculations examining the affects on clad temperature and other parameters.:', The base model, with 6 axial by 3 azimuthal regions, was expanded to 6 axial by 9 azimuthal regions (Figure 4-9). The base calculation simulated the limiting PCT calculation given in the EMF-2103 three-loop sample problem. This case was then repeated with the revised 6 x 9 downcomer noding.The change resulted in an alteration of the blowdown evolution of the transient with little evidence of any affect during reflood. To isolate any possible reflood impact that might have an AREVA NP Inc.
Harris NuclearPlant ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report .-: Page 4-10 influence on downcomer boiling, the case was repeated with: a-.-_slightly
- adjusted vessel-side break flow. Again, little evidence of impact on the reflood portionrof the transient was observed.The study conclud6d that blowdown or near blowdown events could be byrefiningtthe"azimuthal rsolutibn in the doWncomer but that' refl6odWould not be impacted.
Although the study :was performred dfor a som6what elevated systemi pressure, the fldwv regimes within fhe downcomer will not differ for pressu66s
'as ow as atmosphenric Thu's,'thzeazimuthalI downcoiner modeling employed fortlhe RLBLOCA methdoloogy is reasonably converged in'-its ability to represent downcomrer boiling'phenomena. " ".4.5.2.2 Axial Nodalization The RLBLOCA methodology divides the-downcomer into six nodes axially. In-both 3-loop and 4-loop models, the downcomer segment at the active core elevation is represented by two equal length nodes. For most operating plants, the active core length is 12 feet and the downcomer segments at the active c6re elevation are, each 6-feet high: (For:.a 14 foot core,' these nodes would be 7-feet high.) The model for the sensitivity study presented here comprises a 3-loop plant with an ice condenser containment and a .12 foot core. For the.study,.,the two nodes spanning the active core height are ,divided in half, revising.
the model to include eight axial nodes. Further, the refined noding is located within the potential boiling region of the downcomer where, if there is, an axial resolution influence, the sensitivity to that impact would be greatest.The results show that the axial noding used in the base metHodolo"gy':is'sufficient for plants experiencing the very' low system pressures ,characteristic of"ice conclenser containments.
Figure 4-10 provides the containment back'pressure for the base modeling.
Figure 4-1 1 through Figure 4-14'show the total downcomer metal hea't release 'rate, PCT independent of'elevation, dow "omer'liquid level, and 'the core liquid level,; respectively, for the base case and' the modified case.The results demonstrate that the axial resolution provided in the base case, 6 axial downcomer node divisions with 2 divisions spanning the core active region, are sufficient to accurately resolve void distributions within the downcomer.'
Thus, this modeling is sufficient for the prediction of downcomer driving head and the resolution of downcomer boiling effects.AREVA NP Inc.
Harris Nuclear Plant ., ANP-2853(NP) unit 1 Revision 000 Realistic Large Break LOCA Summary Report ...Page 4-11 4.5.3 .:Downcomer Boiling;ConclIusionsI'.
To further iustify the ability of the RLBLOCA" methodology'to predict thepotential for and impact of downcomer boiling, studies were performed on the downcomer wall heat release modeling within the methodology and op the ability of S-RELAP5 to predict the migration of steam through the downcomer.
Both azimuthal and axial noding sensitivity studies were performed.
The axial noding study was based on an iqe condenser plant that is near atmospheric pressure during reflood. These studies demonstrate that SRELAP5 delivers energy to the downcomer liquid volumes at an appropriate rate and that the downcomer noding detail is sufficient to track the distribution of any steam formed. Thus, the required methodology for the prediction of downcomer boiling at system pressures approximating those achieved inpla'ntsvwith pressures as low as ic6econdenser'containments has beendemonstrated.
.4.6 Break Size ....Question:
Were.all break sizes assumed greater than or'equal to 1.0. ft 2?Response:
Yes,: The NRC has requested that the break spectrum for the realistic LOCA evaluations be limited to accidents'that evolve through a range of phenomena similar to those encountered for the'largeý break'area accidents.
This' is'a change to the approved RLBLOCA EM (Reference 1). The larger' br-ea k area' LoCAs are typically characterized by the occurrence of dispersed .flow film boiling at the hot spot, which sets them apart from smaller break LOCAs. This occurs generally inthe vicinity of, 0.2 DEGB (double-ended guillotine break) size (i.e., 0.2 times the total flow area of the pipe on both sides of the break). However, this transitional break size varies from plant to plant and is verified only after the break spectrum has been executed.
AREVA NPhas sought to develop sufficient criteria.
for defining the minimum large break flow area prior to performing the break spectrum.
The purpose .for doing so is to assure a valid break spectrum is performed.
4.6.1 Break
/ Transient Phenomena In determining the AREVA NP criteria, the characteristics of larger break area LOCAs are examined.
These LOCA characteristics involve a rapid.and chaotic. depressurization of the reactor coolant system (RCS) during which the three historical approximate states of the system can be identified.
...AREVA NP Inc.
Harris'Nuclear Plant .ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report __._..,,,.
_ ..Page.4-12 Blowdown phase"is.
defined'as:.the:time; from initiationr of the breaký until flow-from the. accumulators
begins:.,.:This,, definition
.is,-somewhat-;different from .,the traditional, definition.of bloWdown Which extends. the .blowdown until -the RCS. pressure approaches Uý_". -..,containment pressure.
!The. blowdown phase.typically lasts about .12. to-.25- seconds,%depending on the-break size;. ....: ... ... .::. -.Refill is that period that starts with the end of blowdown,.
whichever definition is used, and ends when water is first forced upward into the core. During this phase the core experiences a near adiabatic.heatup.
, Reflood is that portion of the transient that starts with the. end of refill, follows through the filling of the core with water and ends with the achievement of complete core quench.Implicit in this break-down is that the core liquid';inventdry hasý beenhcompletely, or hearly so, expelled from the primary -systern' leaving the core ina state 6f near core-wide'dispersed flow film boiling and subsequent adiabatic heatup prior to the reflood phase'. Altho6igh'this break down served as the basis for the original deterministic LOCA evaluati6n app:roachies and is valid for most LOCAs that would classically be termed large breaks, as the break area decreases the depressurization rate decreases suclh that these three phases overlap substantially.
During these smaller break events, the core liquid inventory is not reduced as much as that found in larger breaks. Also, the adiabatic core heatup is not as extensive as in the larger breaks which results in much lower cladding temperature excursions.
4.6.2 New Mihimum Break size Determ'ination
, No determination of the lower limit can be exact. The values of critical phenomena that control the evolution of a LOCA transient will overlap and interplay.
This is especially true in a statistical evaluation where parameter.
values are varied randomly with a strong expectation that the variations will affect results..
In selecting..the lower area of the RLBLOCA. break spectrum, AREVA. sought to preserve the generality of a complete or nearly complete core, dry out accompanied by a substantially.reduced lower plenum liquid inventory.
It was reasoned that such conditions would be unlikely if the break flow rate was. reduced to less than- the reactor coolant pump flow. That is, if the reactor coolant pumps are capable of forcing more coolant toward the reactor vessel than the break can extract from the reactor vessel, the downcomer and core must maintain some degree of positive flow (positive in the normal operations sense).AREVA NP Inc.
Harris Nuclear Plant .:.ANP 7 2853(NP)Unit 1 " Revision 000 Realistic Large Break LOCA Summary_ Report __......___
.____ -Page 4-13 The circumstance.
is, :of Break flow! is alteredas the. RCS blows down and the RC- pump. flow. may, decrease'as, the rotor and flywheel slow.-down if-power is lost.: However, if the core flow: was reduced tobzeroor.becare .negative immediately, after the break initiation, then the..evdnt.w-as
- quite likely tor.proceed with:sufficient inertia to-expel:
most: of the reactor vessel liquid to the break. The criteria base, thus established, consists:of comparing the break flow to the initial flow through all reactor coolant pumps and setting the minimum break area such'that'these fl6ws" match. This is.dne Wbreak = Abreak
- Gbreak -"Npump
- WRCP-This gives ,., ..-Abreak = (Npump WRcP)/Gbreak.
The break mass flux is.determined, from critical flow.. Because the RCS pressure in the broken cold leg will. decrease rapidly during~the first few seconds of the transient, the.critical mass flux is averaged between that appropriate for the initial operating conditions and that appropriate for.the initial cold;.Ieg enthalpy and the saturation pressure of coolant at that enthalpy, Gbreak = (Gbreak(PO,-HcLo)
+ Gbreak(PCLsat, HcLo))/2., The estimated minimum LBLOCA break area, Amin, is 2.21 ft 2 and the break areapercentage, based on the full double-'ended-guillotine break total area, 'is 27.0 percent.Table 4-4 provides a listing of the plant type,, initial condition, and the fractional minimum RLBLOCA break area, for all the plant types presented as generic representations in the next section.The split versus double-ended break type is no longer related to break area. In concurrence with Regulatory Guide 1.157, both the split and the double-ended break will range in area between the minimum break area (Amin) and an area of twice the size of the broken pipe. The determination of break configuration, split versus doub le-ended, is made-after the break area is selected based on a uniform probability for each occurrence.
AREVA NP Inc.
Har'ris Nuclear Plant .ANP-2853(NP)
Unit 1: Revision 000'Realistic Large Break LOCA Summary Report .,. Page 4-14 4.6:3 Intermediate Break-Size Disposition With the revision ofthe smaller break area for th'e-RLBLOCAa[ysiasl,, the reak range for'sm'al breaks and large breaks are no longer contiguous.
Typically th8 l6We'enid f 'the large break spectrum occurs at between 0.2 to 0.3 times the total. area of.. a 100 percent double-ended guillotine break (DEGB) and the upper end of the small break spectrum occurs at approximately 0.05 times:the (a1rea o'f"fa :10( percent DEGB.I This 16' a V ara' of' br" k t-that ar not specifically analyzed during a LOcA licensing analysis.
The premise for allownig this ggap'is that these breakss-do not compri§6 sccidents that develop hig Cladding temperature and thus'do not compris-e accidents that critically' challenge the emebig-ehccy core coolinI systems (ECCS).Breaks within this range remain large ,enotugh to blowdown.
to low pressures.., Resolution is provided by the large break ECC systems and the pressure-dependent injection limitations that determine critical small break performance are avoided. Further, these accidents develop relatively slowly, assuring maximum effectiveness of those ECG systems.A variety of plant types for which analysis within the intermediate range have-been completed were surveyed.
Although statisticai determinations are extracted fr6mn the consideration.
of breaks with areas above the intermediate range, the AREVA best-estimate.
rmiethodology remains suitable to characterize'the ECCS performance of breaks within the intermediate range'-Table 4-4, provides alisting ýof the plant.type, initial condition, and the fractional minimum RLBLOCA break area.Figure 4-15 through Figure 4-20 provide the enlarged break spectrum results with the upper end of the small break spectrum and the lower end of the large break spectrum indicated by bars.Table 4-5 provides differences between the true large break region and the intermediate break region (break areas between that of the largest SBLOCA and the smallest RLBLOCA).
The minimum difference is 463TF. The table shows the minimum difference between the highest intermediate break spectrum PCT and large break spectrum PCT, for the six plants, as at least 463'F, and including this point would provide an av erage difference of 694'F and a maximum difference of 840cF.Thus, by both measures, the peak cladding temperatures within the intermediate break range will be several hundred degrees below those in the true large break range. Therefore, these breaks will not provide a limit or a critical measure of the ECCS performance.
Given that the large break spectrum bounds the intermediate spectrum, the use of only the large break AREVA NP Inc.
Harris Nuclear'Plant .ANP-2853(NP).
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report .Page 4-15 spectrum meets the requirements of .10CFR50.46 for breaks within 1the intermediate break LQCA.,spectrum, and the. methcd Odemonstrates,.that.
the ,ECCSfor.
a plant. meets" the criteria of.10CFR50 46 with high probability.
" 4.7 Detailed Information for'Contaihment Model Containment initial conditions and coolingr system. information are,.provided in Table 3-8 and Heat Sinks are provided in Table 3-9.- For, HNP, the scatter plots of PCT versus the, sampled containment volumes and initial atmospheric temperature are. shown in Figure 4-21 and Figure 4-22. Containment pressure as a function of time for limiting case is shown in Figure 4-23.4.8 Cross-References-to6 Generic Data on the North: AnnaDocket Question:
In order to conduct its review of the HNP application of AREVA's realistic LBLOCA methods in an efficient manner, the NJRC staff would like to make reference to the responses to NRC staff requests for additional information that were developed for the application of the AREVA methods :to the North.Anna Power, Station, Units I and 2; and found acceptable during that review._ The NRC Staff .safety -evaluation .was -issued on April 1, 2004 (Agency-wide Documentation
'and Management System (ADAMS) accession number ML040960040)..
The staff would- like to, make-use, ofthe information that was provided ,by the North Anna licensee that .is not applicable-:;only to. -North Anna or only to subatmospheric containments.
This information is contained in letters to the NRC from the North Anna licensee dated September 26, 2003 (ADAMS accession number ML032790396) and November 10, 2003 (ADAMS accession number ML03324045 1). The specific respo/nses that the staff'would like to reference are: ,September 26, 2003 letter: NRC Question 1.NRC Question 2.NRc Question 4 NRC Question 6 November 10, 2003 letter: NRC Question 1 AREVA NP Inc.
Harris NudclearPlant ANP-2853(Np), Unit 1 Revision 000.Realistic Large Break LOCA Summary Report ... ..Page 4-16 Please* verify that. the information, in'.these letters is applicable.-to the AREVA model applied to other, analyses except for that information
- ,-related
NorthkAnna .and. to. sub-.atmospheric containments.
Response:
The responses provided to questions 1 2, 4, and 6 are generic and related to the ability -of ICECON to calculate containment pressures.
They are applicable to the HNP RLBLOCA submittal.
Please note that HNP is a dry atmospheric containment.
Question" 1 :ComlleteyApplicable", ... ,:, Question 2 -Completely Applicable Question 4- Comp!etely, Appl.icable, (the reference to CSB 6-1 should, now be, to CSB Technical Position 6-2). The. NRC altered the identificationof this branchtechnical position in Revision 3 of NUREG-0800.
.., ...* , Question 6- Completely applicable.
November 10, 2003 letter: NRC Question 1 The supplementalrequest and response are applicable to HNP.Both part a and. b of this question areaddressed in this response.
Consistent with the Code Scaling, Applicability and. Uncertainty.. (CsAU), methodology, containment modeling in the Framatome ANP (FANP) Realistic Large Break Loss of Coolant Accident (RLBLOCA) analysis emphasizes the important processes influencing large break LOCA: initial conditionsý, active heat sinks, and passive heat sinks., The FANP RLBLOCA Phenomena Identification and Ranking Table.- (PIRT) identifies "containment pressure as 'the ronily containment-"related phenomenon
- 'directly
- influencing
'-clad temperature".
resp.onse.
Accordingly, containment processes directly influencing containment pressure response are the focus of the modeling effort.The FANP RLBLOCA methodology was approved as a "Best-Estimate" methodology, which conforms to the guidance of Regulatory Guide 1.157. As applied to North Anna-Units 1 and 2, the RLBLOCA methodology involves a realistic simulation of containment backpressure.
Containment modeling in the RLBLOCA methodology includes both statistical and non-AREVA NP Inc.
Harris Nuclear Plant -:ANP-2853(NP)
Unit 1 Revision 000 Realistic.Large Break LOCA Summary Report ._ Page'4-17 statistical
- treatmentrof significart',tniode!lrinputs: ,The6 general.iobjective is to obtain containment backpressure resuilts,'that acbommodatee;pected modeling, :uncertainties..
The most:dominant phenomenological influences (ignoring active systems) on containment pressure -are: heat transfer to internal structures, break size and effluent modelin'g, and initial pressure and volume.Consistent with the CSAU methodology, parameters with lesser influence on containment pressure are modeled by assuming nominal "or conservatively biased values. "Thet net effect of treating certain parIame teris with conservatively biased values and others asstattical values is to produce a conservative backpressure result that accommodates ,expected modeling uncertainties.
Branch Technical Position CSB 6-1 was established to provide guidance to plant licensees and vendors a's* to how c6ntainment system- sre' to be modeled Jfo Appendix K-based LOCA e(aluations'The'itent'0fCSB 6-1 is to'prvide gui dance for the performance of a minimum containment backpressure analysis.
The RLBLOCA methodology was approved as a "Best-Estimate" methodology that conforms to Regulatory Guide 1.157. With regard to containment pressure, the Regulatory Guide states (Section 3.12.1):."The containment pressure used for evaluating cooling effectiveness during the post-blowdown phase of a loss-of-coolant
=accident, should,.:be calculated -in a .best-estimate manner and should include the effects of operation of all pressure reducing equipment assumed to be available.
Best-estimate models will be 6onsid red acceptable prbvided their technical basis is demonstrated With appropriate-data and'analyses." The containment pressure response used in the North Anna. RLBLOCA analyses is. a realistic calculation,"applying both best-estimate and conservative modeling, assumptions..
The CSAU methodology requires., that the treatment -of important; phenomena accommodate anticipated uncertainties.
As previously stated, the dominant phenomena influencing containmentresponse are: heat transfer to internal -structures, -break size and effluent modeling, and initial., pressure, and volume. A discussion of these dominant influences is presented below. In the North Anna RLBLOCA analyses, other parameters are, with only a few exceptions, modeled by applying CSB 6-1 recommendations.
Generally'for Framatome ANP RLBLOCA analyses, active systems are assumed to operate at maximurmi efficiency (as was done for the North Anna analyses and is consistent with CSB 6-1) unless data are available to treat their operation in a best-estimate manner.AREVA NP Inc.
Harris Nuclear Plant ... ANP-2853(NP)
Unit 1 .Revision 000 Realistic Large Break LOCA Summary Report ,. -Page 4-18 4.9 GDC 35- LOOP and No-LOOP.Case Sets. '..>.: Question:
IOCFR50, Appendix A, GDC [General Design Criterion]
35 [Emergency core cooling].-
states, that, "Suitable redundancy in components
,-and -features .and suitable interconnections, leak, detection, isolation, and containment capabilities shall .be provided to assure that for onsite electric power system operation (assuming offsite electricpower is, not available) .and for offsite electric power operation (assuming oQsite pwr is not available) the system function can be accomplished, assuming a single failure." The Staff interpretation is that two cases (loss of offsite power with onsite power available, and loss ofonsite' powerwith o'ffsite power available) must be'run independently'toi satisfy'GDC 35.-Each of these cases is separate from the bthei' 'in thatl each case is',reprie~ehtd by a different statistical response spectrum.i' To'6 acbcomlish'the task of identifying.
the"worst case would require more runs. However, for LBLOCA analyses (only), the high likelihood'of loss of onsite power being the most limiting is so small that only loss of offsite power cases need be run. (This is unless a particular plant design, e.g., CE [Combustion Engineering]
plant design, is also vulnerable to a loss of onsite power, in which situation the NRC may require that both cases be analyzed separately.
This would require more case runs to satisfy the statistical requirement than, forjust loss of offsite power.):, ..What is your basis for assuming a 50% probability of loss of offsite power? Your statistical runs need to assume that offsite power is lost (in an independent set of runs). If, as stated above, it has been determined that Palisades, being of CE design, is also vulnerable to a loss of onsite power, this also should be addressed (with an independent set of runs).Response:
In concurrence with the NRC's interpretation of GDC 35, a set of 59 cases each was run with a LOOP and No-LOOP assumption.
The set of 59 cases that predicted the highest figure of merit, PCT, is reported in Section 2 and Section 3, herein. The results from both case sets are shown in Figure 3-22. This is a change to the approved RLBLOCA EM (Reference 1).4.10 Input Variables Statement Question:
Provide a statement confirming that Progress Energy and its LBLOCA analyses vendor have ongoing processes that assure that the input variables and ranges of parameters for the LBLOCA analyses conservatively bound the values and ranges of those parameters for AREVA NP Inc.
Harris NuclearPlant
-ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summa yReport ...... _,p_"- Page 4-19 the as-operated HNP. This statement addresses.
certain programmatic requirements of 10 CFR 50.46, Section (c), Repon'se:
Pr6grebssEnergy a-d: te LBILOCA Analysis Vendo-r have an ongoing process to ensure that al input vriable-s'and*
parameter ranges for thHNP rea Istciarge break 0ossof-a as conservatve with respect to' plant" operatingAnd' design conditions:
'l In "acicrdarVce with Energy Quality Assurance" prdg -am requirements, this process involves 1) Definition of the required input yariables and parameter ranges by the Analysis Vendor.2)-Compilation of the specific values from existing p!ant.design input.and output documents by Progress Energy. and Vendor personnel in.a formal analyiss input summary document issued by the Analysis Vendor and, ...3) Formal review and approval of the input summary document by Progress Energy. -Formal Progress Energy approval of the input document serves as the release for the Vendor to perform the analysis.Continuing review of the input summary document is performed byProgress Energy as part of the plant design change process and cycle-specific core design process. Changes to the input summary required to support plant modifications or cycle-specific core alternations areformally communicated to the Analysis Vendor by Progress Energy. Revisions and updates to the analysis parameters are documented and approved in accordance with the process described above for the initial analysis.AREVA NP Inc.
Harris Nuclear Plant Unit 1. : .Realistic Larqe Break LOCA Summary Report ANP-2853(NP).
Revision 000 Page 4-20 Table 4-1 Typical Measurement Uncertainties-and Local: Peaking Factors FAH Measurement" c-I Local'Pin Peaking Plant_ Uncertainty.'..-.
Ft (-).... ___'_ _ ' "(percent)
, S1 .4_0 1.068 2 ...... .. ... .. 4 .0 .-.050 .3 .6.0 1.149" _"_-_4 4.0 1.113 5_. 4.25 .'.135 .6 4.0 .1.058 Table 4-2 FLECHT-SEASET
& 17x17 FA Geometry Parameters Design Parameter FLECHT-SEASET 17x1 7 Fuel Assembly Rod Pitch (in) 0.496 0.496 Fuel Rod Diameter (in) 0.374 .0374 Guide Tube Diameter (in) 0.474 0.482 Table 4-3 FLECHT-SEASET Test Parameters htc at PCT Steam .Thimble, Test Rod 7J PCT PCT T(es) Time(s) time Temperature -at Temperature at -ft .(Btu/hr-ft 2-'F) 7.1 (6-ft) (IF) at 6-ft'(CF)34420 2205 34 .101 1850 ..1850*31805 2150 110 10 1800 -1800 31504 2033 -100 10 1750 1750: 31021 1684 -29 9 1400 -1350 30817 1440 70 13 900 750* set to steam temp.AREVA NP Inc.
T A"i N T t Harris Nuclea Plant Unit 1 " Realistic Larqe Break LOCA Summary Report ANP-2853(NP)'
Revision 000 Page 4-21 Table 44 ý Minim m' Break'Area'for Break LOCA Spectrum Spectrum Spectrum Plant System Cold Leg Subcooled Saturated ,RCP flow. Minimum Minimum Prio issure Enthalpy 'bGa -Gbreak (HEM)Description brak break. BraMre)rakAe (psia). .(Btu/Ibm) (Ibm/ft 2-s) (!bm/ft -s),,?, (Ibm/s) Break fArea BreakeGB)a A 3-LoopW 2250 554.0 22198 6330 "31558 2.21 0.27 Design ., ... _ .........B D3-Ligp i.... 2250 ....-5&44.5 .... ....... 23880 ..........-.
-5450,I -,28124 ..... 1.92' 0.23 S3-Loop- 5 C Design 2250
- 550.0 23540 .5580 29743 2.04 0.25 D 2x4 CE 2100 -. .-538.8 .22860........5310 21522 .1:53 0.24 DDesign E2x4 CE .. 2060- 531 0 22068 5694 38277 2.76 0.28 FDesign 210 509 2205370 j39500 2.76 0.33 Table 4-5 Minimum PCT Temperature Difference
-True Large and Intermediate Breaks ,Generic Maximum Maximum Plant Plant PCT (iF) PCT (7) Delta PCT Average Delta Description Label Intermediate Large Size (IF) PCT (IF)(Table 4-4) _Size Break -Break.-A 1206 1930 724 3-LOOD: W ... .. .. ; "... ." 3-LeopW B 1273 1951 678;6 622 Design _____C 1326 1789 463..CE .D .984 .. 1751 767.729'Design E 1049 -1740 ..691 4-eigopW F 1127 1967 840 840 Design.AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1 Revision 000 Realistic'Large Break LOCA Summary Report Page 4-22 Guide Tube "- Hot Rodf-0 ,0Adjacent Rods 00000 Figure 4-11 R2RRAD 5 x 5 Rod Segment AREVA NP Inc.
ti t...Harris Nuclear Plant Unit 1 :. .Realistic Larae Break LOCA Summary Report ANP-2853(NP)
Revision.
000 Paae.4-23 4.5 4 3.5 3 2 M 0 1.5 I 1 0.5 0-1400 1500 1600 .. 1700 ,-1800 , 1900 ' 2000! 2100- 2200 2300 PCT (TF)2400 Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Summary Report ANP-2853(NP)
Revision 000 Paae 4-24 Figure 4-3 Reactor Vessel Downcomer Boiling Diagram AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-25 550 500 U. 450 CL 400 E 4)1-350 300 250 C closed Form, 0 s Form, 100 s-4"--Closed Form, 300 s--0 -S-RELAP5, 0 s--- S-RELAP5, 100 si--0 -S-RELAP5, 300 s j-r-0.0 0.1 1 0.2 0.3 -.-.4 -0.5 .Distance from Inner Wall, feet:10.6 0,7 .' 0.8 Figure 4-4 S-RELAP5 versus Closed Form Solution AREVA NP Inc.
Harris Nuclear Plant Ur~it.1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000: Page 4-26;* " .': .. ....30000.00 Ca a)LU Timne (sec)Figure 4-5 Downcomer Wall Heat Release -Wall Mesh Point Sensitivity AREVA NP Inc.
Harris NuclearPlant Unit 1 .Realistic Large Break LOCA Summary Report.ANP-2853(NP)
Revision 000 I -Page.4-27 24000 =1000.00 U-a)CL E a)1200M00 00000 -0.00 L"0.0 240.0 320.0 400.0 Time (sec)Figure 4-6 PCT Independent of Elevation
-Wall Mesh Point Sensitivity AREVA NP Inc.
Harris NuclearPlant Unit1 : -Realistic Larae Break LOCA Summary Report ANP-2853(NP)ý Revision 000 , Page 4-28, 30.00'-i 160.0 240,0 Time (sec)400.0 Figure 4-7 Downcomer Liquid Level'- Wall Mesh Point Sensitivity AREVA NP Inc.
Harris NuclearPlant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP), Revision 000 Page 4-29, 12.00 0)4-0)-J-o 0~-J Time:(se.c)
Figure 4-8 Core Liquid Level -Wall Mesh Point Sensitivity AREVA NP Inc.
Harris Nuclear Plant unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-30 Base model Revised 9 Region Model IILIIJJII2IIIL2IIII1 Q 9 QH9 Figure 4-9 Azimuthal Noding AREVA NP Inc.
Harris Nuclear Plant Unit 1 .., Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page,4-3.1 40.00r- .1 .-. Base 6x6 32.00* 4t- -- ---4 --(Vý70 W0 P0 IL 24.00 16.00--5---4---O--@ -O---5--*-- I 0.00 0.0 0 00.0 .160.0 .240.0 320.0.Time (sec)..Figure 4-10 Lower Compartment Pressure versus Time 4M0.0 AREVA NP Inc.
Harris-Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-32 30000,00 (3 aJ Cu U)400-0 1 :-: Time .(sec)Figure 4-11 Downcomer Wall Heat Release -Axial Noding Sensitivity Study AREVA NP Inc.
Harris:Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report."ANP-2853(NP), Revision 000 Page 4-33 U-E E I--Time (sec)Figure 4-12 PCT Independent Of Elevation -Axial Noding Sensitivity Study AREVA NP Inc.
,Harris-Nuclear Plant Unit' 1 Realistic Lame Break LOCA Summary Report ANP-2853(NP)
Revision 000.-.... Page 4-34 30.00 20.00._J 27"j 10.00 0.00 Time (sec)Figure 4-13 Downcomer Liquid Level,- Axial Noding Sensitivity Study AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report A.P, ANP-2853(NP)
Revision 000 Page 4-35 27 0--J 12.00 10,00 8.00 6.00 4.00 2.00-.Base 6wx6 1 asýZe 1ý1 I L 0.00 0.0 400,0 Time (sec)Figure 4-14 Core Liquid Level -Axial Noding. Sensitivity Study, AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Summary ReDort ANP-2853(NP)
Revision 000 Pacie 4-36 2000 -1800 +Upper End of SBLOCA Break Size Spectrum Large Break ,,;..-.Sp-ectrum A..Minimum-Break Area 1600 1400 1200 T '=.... ~.I*1*--I i T1 , 4 ---*-V.4 4 4 1000**1 41 I,--~ -I -I ---V -*800 -600o,-0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 Break Area Normalized to Double Ended Guillotine Note: This plot represents a "Transition" application.
Figure 4-15 Plant A -Westinghouse 3-Loop Design 0.8000 0.9000 1.0000 AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-37 2000 1800 1600 1400 +C o..[-Upper End of SBLOCA Break Size-_ Spectrum F.I.Large Break: .*Spectrum Minimum.Break Area--- ----------
i F F. I i --F--s---- -i i F i i I F*1200 -1000 -F.I.I.I--I.F I F F .I ..I .T- -t-,,t .i " t Ft 4
- FI F F F F 4 -.- --t- ---i I F F F_
- F 4 4*F F F F F* F F--I* .Fi 800 600 0.0000 0.1000 0.2 000 0.3000 0.4000 " 0,5000 0.6000 0.7000 Break Area Normalized to Double Ended Guillotine Figure 4-16 Plant B -Westinghouse 3-Loop Design 0.8000 0.9000 1.0000 AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-38 2000 T 1800 +1600 +1400 +---Z, oJ ii..Upper End of SBLOCA Break Size-Spectrum ---1--i i... .I--*-- ----i -------Large Break Spectrum Minimum -Break Area-F F I" , I -.I.I .I-I ---I. *** 4--f -1200-** i*1000 +----*-4'I I I I I I I -4. i-i- -i.-'L 800 -----I....600____ -t -1 0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000'Break Area Normalized to Double Ended Guillotine Figure 4-17 Plant C -Westinghouse 3-Loop Design AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-39 2000 -1800 -.-.-1600 -1400 +--o Upper End of SBLOCA Break Size--Spectrum I .I I-i.I.1*Large Break -Spectrum V1 Minimum Break Ar~a.5-I-S--.*I S I1 F* F II I F F3 F " F F F-I*:. *: IF i I I.*%* F* F--F* F*F-F 1200 -1000 +800 +600 0.0000 0.1000 0.2000, 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000' Break Area Normalized to Double Ended Guillotine Figure 4-18 Plant D -Combustion Engineering 2x4 Design AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-40 2000 T ---1800 -g-1600 +1400 iz Upper End of SBLOCA Break Size-Spectrum II " Ii*-- --L Large Break Spectrum Minimum Break Aiea** I I.-.1 --~2~ --~ -~-. V ------ *1- --I*---L --.1---I* I-I-* I* I* I.I.** I 2. I-' I-- I 1200 +-I 1000 ----I 800 +600 0.0000 o.1060 .' 0.2000 0.3000 -.0.4000 0.5000 0.6000 0.7000 Break Area Normalized to Double Ended Guillotine 0,8000 0.9000 1.0000 Figure 4-19 Plant E -Combustion Engineering 2x4 Design AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-41 2200 -F 2000 4-1800 + ...1600 +Z 0 U A.Upper End of SBLOCA Break Size Spectrum I,'i-I.-------F-*i1 I i-------I*-I' Large Break Spectrum Minimum"' Break Area*I T i;4-**I.* -.I.* 'I -.1 1400 .---I V L*I* I I.I--i 1.0000 1200 -1000 .---800 0nn0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000-Break Area Normalized to Double Ended Guillotine Figure 4-20 Plant F -Westinghouse 4-loop Design AREVA NP Inc.
Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Summary Report ANP-2853(NP)
Revision 000 Page 4-42 PCT vs Containment Volume 2000 1800 1600 1400 1200 1000 800 600 400 2.20e+06 F-El F1 m 0 LI U U U El EL Im[ L ]IE LEl El 0C[,.:: 1 -i rl LI*EL 0I LI 0 El LI 0 El[L nI LI 0 Split Break LI Guillotine Break I ...2.30e+06 2.40e+06 '2.50e+06 2.60e+06 2.70e+06 Containment Volume (ft 3)Figure 4-21 PCT vs. Containment Volume AREVA NP Inc.
Harris Nuclear Plant Unit 1 -Realistic Larme Break LOCA Summary Report , .ANP-2853(NP)
Revision 000 Pace 4-43 PCT vs Containment Temperature 2000 1800 1600 1400 1200 C-El El Ell El E E ED E El El [I l ]0 El El PEl U El E: E Split Break El Guillotine Breaki r 1000 800 600 400 I 80 90 100 110 120 130 Containment Temperature
()F)Figure 4-22 PCT vs. Initial Containment Temperature AREVA NP Inc.
Harris.Nuclear Plant unit 1 Realistic Large Break LOCA Summary Report.ANP-2853(NP)
Revision 000:- :. Page 4-44 100 90 80 70 60 Cl.50 Cl)40 30 20 10 0 0 Containment Pressures Containment 200 400 Time (s)600 800 ID:05622 290ct2009 17:56:48 R5DMX Figure 4-23 Containment Pressure as function of time for limiting case AREVA NP Inc.
Harris Nuclear Plant ANP-2853,(NP.)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report Page 5-1.5.0 Conclusions A RLBLOCA analysis was performed for the Harris Nuclear Plant, Unit 1 (HNP) using NRC -approved AREVA NP RLBLOCA methods (Reference 1). Analysis results show that the limiting LOOP case has a PCT of 1930'F, and a maximum oxidation thickness and hydrogen generation that fall well within regulatory requirements.
The analysis supports operation at a Rated Thermal Power of 2900 MWt and includes a measurement uncertainty of 2%, a steam generator tube plugging level of. up to 3 percent in all steam generators, a total peaking factor. (FQ) of 2.52 (including uncertainty) and a nuclear enthalpy rise factor (FAH) of 1.73 (including 4% uncertainty) with no axial or burnup dependent power peaking limit and peak rod average exposures of up to 62,000 MWd/MTU. For large break LOCA, the three 10CFR50.46(b) criteria presented in Section 3.0 are met and operation of HNP with AREVA NP-supplied 17X17 Zr-4 clad fuel is justified.
AREVA NP Inc.
Harris Nuclear Plant ANP-2853(NP)
Unit 1 Revision 000 Realistic Large Break LOCA Summary Report Page 6-1 6.0 References
- 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. XN-CC-39 (A) Revision 1, "ICECON: A Computer Program to Calculate Containment Back Pressure for LOCA Analysis (Including Ice Condenser Plants)," Exxon Nuclear Company, October 1978.5. U. S. Nuclear Regulatory Commission, NUREG-0800, Revision 3, Standard Review Plan, March 2007.6. NUREG/CR-1532, EPRI NP-1459, WCAP-9699, "PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Report," June 1980.7. G.P. Liley and L.E. Hochreiter, "Mixing of Emergency Core Cooling Water with Steam: 1/3 -Scale Test and Summary," EPRI Report EPRI-2, June 1975.8. NUREG/CR-0994, "A Radiative Heat Transfer Model for the TRAC Code" November 1979.9. J.P. Holman, Heat Transfer, 4 th Edition, McGraw-Hill Book Company, 1976.10. EMF-CC-1 30, "HUXY: A Generalized Multirod Heatup Code for BWR Appendix K LOCA Analysis Theory Manual," Framatome ANP, May 2001.11. D. A. Mandell, "Geometric View Factors for Radiative Heat Transfer within Boiling Water Reactor Fuel Bundles," Nucl. Tech., Vol. 52, March 1981.12. EMF-2102(P)(A)
Revision 0, S-RELAP5:
Code Verification and Validation, Framatome ANP, Inc., August 2001.AREVA NP Inc.