ML15253A206

ANP-3315NP, Revision 0, Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensing Report

ML15253A206
Person / Time
Site:	Millstone  (DPR-065)
Issue date:	09/01/2015
From:	AREVA
To:	Dominion Nuclear Connecticut, Office of Nuclear Reactor Regulation
References
15-411 ANP-3315NP, Rev 0
Download: ML15253A206 (60)

Text

Serial No.15-411 Docket No. 50-336 ANP-3315NP, Revision 0 MILLSTONE UNIT 2 M5 UPGRADE, SMALL BREAK LOCA ANALYSIS LICENSING REPORT (NON-PROPRIETARY)

DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

A ARE VA Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensing Report ANP-3315NP Revision 0 April 2015 AREVA Inc.

(c) April 2015 AREVA Inc.

Copyright © April 2015 ARE VA Inc.

All Rights Reserved

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Paqei Nature of Changes Section(s)

Item or Page(s)

Description and Justification 1

All Initial Issue

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinq Report Pacie ii Contents

1.0 INTRODUCTION

..................................................................... 1-1 2.0

SUMMARY

OF RESULTS........................................................... 2-1

3.0 DESCRIPTION

OF ANALYSIS..................................................... 3-1 3.1 Description of SBLOCA Event.............................................. 3-1 3.2 Analytical Methods........................................................... 3-3 3.3 Plant Description and Summary of Analysis Parameters.................. 3-8 3.4 SER Compliance............................................................. 3-9 4.0 ANALYTICAL RESULTS............................................................ 4-1 4.1 Results for Break Spectrum................................................. 4-1 4.2 Discussion of Transient for Limiting Break................................. 4-1 4.3 RCP Trip Sensitivity Study................................................... 4-4 4.3.1 Cold Leg Breaks RCP Trip Sensitivity Study....................... 4-5 4.3.2 Hot Leg Breaks RCP Trip Sensitivity Study........................ 4-5 4.4 Attached Piping Break Sensitivity Study.................................... 4-6 4.5 Safety Injection Low Fluid Temperature Sensitivity Study................ 4-6

5.0 REFERENCES

....................................................................... 5-1

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ANP-331 5NP Revision 0 Millstone Unit 2 M5© Upgrade, Small Break LOCA Analysis Licensingq Report Paoqe iii List of Tables Table 3-1 System Parameters and Initial Conditions................................... 3-1 1 Table 3-2 High Pressure Safety Injection Flow Rates for Cold Leg Breaks........... 3-12 Table 4-1 Summary of SBLOCA Break Spectrum Results.............................. 4-8 Table 4-2 Sequence of Events for Break Spectrum (seconds).......................... 4-9

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PRevso iv Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Page iv List of Figures Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization................ 3-5 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization...................... 3-6 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization.......................... 3-7 Figure 3-4 Axial Power Distribution Comparison........................................ 3-13 Figure 4-1 Peak Cladding Temperature versus Break Size (SBLOCA Break Spectrum)................................................................... 4-10 Figure 4-2 Reactor Power - 3.78-inch Break........................................... 4-11 Figure 4-3 Primary and Secondary System Pressures - 3.78-inch Break............ 4-12 Figure 4-4 Break Mass Flow Rate - 3.78-inch Break................................... 4-13 Figure 4-5 Break Vapor Void Fraction - 3.78-inch Break.............................. 4-14 Figure 4-6 Loop Seal Void Fraction - 3.78-inch Break................................. 4-15 Figure 4-7 Total Core Inlet Mass Flow Rate - 3.78-inch Break........................ 4-16 Figure 4-8 Downcomer Collapsed Liquid Level - 3.78-inch Break.................... 4-17 Figure 4-9 Inner and Outer Core Collapsed Liquid Level - 3.78-inch Break.......... 4-18 Figure 4-10 Reactor Vessel Mass - 3.78-inch Break................................... 4-19 Figure 4-11 RCS Loop Mass Flow Rates - 3.78-inch Break........................... 4-20 Figure 4-12 Steam Generator Main Feedwater Mass Flow Rates - 3.78-inch Break........................................................................ 4-21 Figure 4-13 Steam Generator Auxiliary Feedwater Mass Flow Rates - 3.78-inch Break........................................................................ 4-22 Figure 4-14 Steam Generator Total Mass - 3.78-inch Break.......................... 4-23 Figure 4-15 Steam Generator Narrow Range Level % - 3.78-inch Break............ 4-24 Figure 4-16 High Pressure Safety Injection Mass Flow Rates - 3.78-inch Break....4-25 Figure 4-17 Low Pressure Safety Injection Mass Flow Rates - 3.78-inch Break....

4-26 Figure 4-18 Safety Injection Tank Mass Flow Rates - 3.78-inch Break............... 4-27 Figure 4-19 Integrated Break Flow and ECCS Flow - 3.78-inch Break............... 4-28 Figure 4-20 Hot Assembly Collapsed Liquid Level - 3.78-inch Break................. 4-29 Figure 4-21 Hot Assembly Mixture Level - 3.78-inch Break............................ 4-30 Figure 4-22 Peak Cladding Temperature at PCT Location (11.02 ft) - 3.78-inch Break........................................................................ 4-31

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ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Paqe v Nomenclature Acronym Definition AFAS Auxiliary Feedwater Actuation Signal AFW Auxiliary Feedwater AREVA AREVA Inc.

BOG Beginning-of-Cycle CE Combustion Engineering CEA Control Element Assembly CFR Code of Federal Regulations DC Downcomer ECCS Emergency Core Cooling System EDG-Emergency Diesel Generator EOC End-of-Cycle HMPTM High Mechanical Performance Spacer Grid HTPTM High Thermal Performance Spacer Grid HPSI High Pressure Safety Injection LHR Linear Heat Rate LOCA Loss-of-Coolant Accident LPSI Low Pressure Safety Injection MFW Main Feedwater MSIV Main Steam Isolation Valve MSSV Main Steam Safety Valve NRC Nuclear Regulatory Commission RCP Reactor Coolant Pu~mp.........

RCS Reactor Coolant System RV Reactor Vessel PCT Peak Cladding Temperature PWR Pressurized Water Reactor PZR Pressurizer SBLOCA Small Break Loss-of-Coolant-Accident SER Safety Evaluation Report SG Steam Generator SI Safety Injection SIAS Safety Injection actuation signal SIT Safety Injection tank

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensing R~eport Pagqe 1-1

1.0 INTRODUCTION

This report summarizes the small break loss-of-coolant accident (SBLOCA) analysis for Millstone Nuclear Plant Unit 2. The purpose of the SBLOCA analysis is to support the upgrade to the AREVA Advanced Combustion Engineering (CE) 14 x 14 Fuel Design with M5@l cladding for the Millstone Nuclear Plant Unit 2 core design. This analysis was performed in accordance with the Nuclear Regulatory Commission (NRC)-approved S-RELAP5 methodology described in Reference 1 and as modified by Reference 2.

Millstone Nuclear Plant Unit 2 is a 2x4-loop, Combustion Engineering (CE)-designed pressurized water reactor (PWR). The AREVA Advanced CEI4 Fuel Design with M5 cladding for Millstone Nuclear Plant Unit 2 consists of a 14x14 CE array with HTPTM2 intermediate grids and a lower HMPTM2 grid. The fuel assembly will include a Zirc-4 MONOBLOCTM2 guide tube design, M5 fuel rod design and FUELGUARDTM2 debris-resistant lower tie-plate design.

A complete spectrum of cold leg break sizes was considered, ranging from 2.0 inch diameter to 9.49 inch diameter.

In addition, sensitivity studies were performed to consider delayed reactor coolant pump (RCP) trip sensitivity, attached piping break sensitivity and safety injection low fluid temperature sensitivity.

The analysis supports plant operation, at a core power level of 2754 MWt (including measurement uncertainty), a peak linear hear rate (LHR) of 15.1 kW/ft, a radial peaking factor of 1.854 (including uncertainty), and 5.87% steam generator tube plugging.

1 M5 is a registered trademark of AREVA.

2 HTP, HMP, MONOBLOC and FUELGUARD are trademarks of AREVA.

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ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Page 2-1 2.0

SUMMARY

OF RESULTS A SBLOCA break spectrum analysis was performed for Millstone Nuclear Plant Unit 2 using the NRC-approved AREVA method (Reference 1) as modified by Reference 2.

The analysis results demonstrate that the following acceptance criteria for Emergency Core Cooling Systems (ECCS), as stated in 10 Code of Federal Regulations (CFR) 50.46(b)(1 -4) (Reference 4), have been met.

1. The calculated maximum fuel element cladding temperature shall not exceed 2200°F.

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

3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

The limiting PCT is 1707°F for a 3.78-inch diameter cold leg pump discharge break.

The total maximum local oxidation isiless than_6_°/, ncludinga pre-transient oxidation of 2.3% and transient maximum oxidation of 3.6%. The maximum core-wide oxidation is less than 0.04%. The results of the analysis demonstrate the adequacy of the ECCS to support the 10 CFR 50.46(b) (1-4) criteria (Reference 4).

In addition to the cold leg pump discharge break spectrum analysis, three sensitivity studies were performed to consider a delayed RCP trip, break in an attached pipe and low fluid temperature safety injection. The results of the delayed RCP trip sensitivity demonstrated that there is at least 2 minutes for operators to trip all four RCPs after subcooling margin is lost in the cold leg pump suction in order to meet the 10 CFR 50.46(b)(1 -4) criteria (Reference 4). The break in an attached piping sensitivity study

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensincj Report Paae 2-2 was performed with a 10.5-inch diameter break in the safety injection tank (SIT) line.

The result of the SIT break was a PCT of 1239°F, which is less limiting than the cold leg pump discharge limiting break. The last sensitivity study performed was on reducing the SIT and safety injection (SI) fluid temperatures to approximate nominal temperatures as opposed to the maximum temperatures used in the spectrum analysis.

The result of the sensitivity study confirmed that the SIT and SI fluid temperatures used in the break spectrum are conservative.

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinq Report Panqe 3-1

3.0 DESCRIPTION

OF ANALYSIS Section 3.1 of this report provides a brief description of the postulated SBLOCA event.

Section 3.2 describes the analytical models used in the analysis. Section 3.3 presents a description of the Millstone Nuclear Plant Unit 2 plant parameters and outlines the system parameters used in the SBLOCA analysis.

Section 3.4 describes the Safety Evaluation Report (SER) compliance.

3.1 Description of SBLOCA Event The postulated SBLOCA is defined as a break in the Reactor Coolant System (RCS) pressure boundary for which the area is up to approximately 10% of a cold leg pipe area. The most limiting break location is in the cold leg pipe on the discharge side of the RCP. This break location results in the largest amount of RCS inventory loss and the largest fraction of ECOS fluid ejected out through the break. This produces the greatest degree of core uncovery, the longest fuel rod heatup time, and consequently, the greatest challenge to the 10 CFR 50.46(b)(1-4) criteria (Reference 4).

The SBLOCA event progression develops in the following distinct phases: (1) subcooled depressurization (also known as blowdown), (2) natural circulation, (3) loop seal

-clearing, (4) core boil-off (5) core recovery and long-term cooling. The duration of each of these phases is break size and system dependent.

Following the break, the RCS rapidly depressurizes to the saturation pressure of the hot leg fluid. During the initial depressurization phase, a reactor trip is generated on low pressurizer pressure; the turbine is tripped on the reactor trip. The assumption of a Ioss-of-offsite-power concurrent with the reactor SCRAM results in reactor coolant pump trip.

In the second phase of the transient, the RCS transitions to a quasi-equilibrium condition in which the core decay heat, leak flow, steam generator heat removal, and

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break L0CA Analysis Licensincq Report Paqe 3-2 system hydrostatic head balance combine to control the core inventory.

During this period, the reactor coolant pumps are coasting down and the system drains top down with voids beginning to form at the top of the steam generator (SG) tubes and continuing to form in the reactor vessel upper head and at the top of the reactor vessel upper plenum region.

Also, loop seals remain plugged during this phase, trapping vapor generated by the core in the RCS, and resulting in a low quality flow at the break.

The third phase in the transient is characterized by loop seal clearing. During this phase, the loop seal, which is liquid trapped in the reactor coolant pump suction piping, can prevent steam from venting via the break. When the maximum pressure difference between the reactor vessel upper head and downcomer is reached, loop seal upflow is pushed, clearing the loop seal, and the trapped steam can be vented to the break. For a small break, the transient develops slowly, and liquid level in the reactor coolant system may drop to the loop seal level prior to establishing a steam vent. The core can become temporarily uncovered in this loop seal clearing process. Following loop seal clearing, the break flow transitions to primarily steam and the core recovers to approximately the cold leg elevation, as pressure imbalances throughout the RCS are relieved.

The fourth phase is characterized as core boil-off.

With the loop seal cleared, the venting of steam through the break causes a rapid RCS depressurization below the secondary pressure. As boiling increases in the core, the core mixture level decreases.

The core mixture level will reach a minimum, in some cases resulting in deep core uncovery. The boil-off period of the transient ends when the core liquid level reaches this minimum. At this time, the RCS has depressurized to the point where ECCS flow into the reactor vessel matches the rate of boil-off from the core.

The last phase of the transient is characterized as core recovery and long-term cooling.

The core recovery period extends from the time at which the core mixture level reaches a minimum in the core boil-off phase, until all parts of the core are quenched and covered by a low quality mixture. During this time, the PCT occurs. Core recovery is

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break L0CA Analysis Licensingq Report Paqe 3-3 provided by pumped injection and passive SIT injection when the RCS pressure decreases below the SIT pressure.

The SBLOCA transient progression is dependent on the size of break and is typically broken into three different break size ranges. For break sizes towards the larger end of the break spectrum, significant primary system inventory loss results in larger primary system depressurization to the SITs, which provides sufficient inventory in time to limit the core uncovery and clad heatup, thus hot rod heatup is typically not limiting. For break sizes in the middle of the spectrum, the rate of inventory loss from the primary system is such that the HPSI pumps cannot preclude significant core uncovery. The primary system depressurization rate is slow, extending the time required to reach the SIT injection pressure or to recover core liquid level on HPSI flow.

This tends to maximize the heatup time of the hot rod which produces the maximum PCT and local cladding oxidation.

The limiting break case will either exhibit core recovery with the HPSI pumped injection alone while the primary system pressure remains barely above the SIT injection setpoint, or core recovery from SIT injection after an extended period of uncovery. For very small break sizes, the primary system pressure does not reach the SIT injection pressure; however, primary system inventory loss is not significant and typically within the means of HPSI makeup capacity such that core uncovery is minimal if not precluded.

3.2 Analytical Methods The AREVA S-RELAP5 SBLOCA evaluation model for event response of the primary and secondary systems and the hot fuel rod used in this analysis is based on the use of two computer codes. The appropriate conservatisms, as prescribed by Appendix K of 10 CFR 50 (Reference 7), are incorporated.

This analysis was performed in accordance with the NRC-approved S-RELAP5 methodology described in Reference I and as modified by Reference 2.

The two AREVA computer codes used in this analysis are:

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinq Report Paaqe 3-4

1. The ROD EX2-2A code (References 5 and 6) was used to determine the burnup-dependent initial fuel rod conditions for the system calculations.

2. The S-RELAP5 code was used to predict the thermal-hydraulic response of the primary and secondary sides of the reactor system and the hot rod response.

Representative system nodalization figures are shown in Figure 3-1 (RCS), Figure 3-2 (Secondary System), and Figure 3-3 (Reactor Vessel).

Since these figures are representative, they may not contain all of the model details specific to Millstone Unit 2.

For example, the charging system is not simulated in the SBLOCA analysis; therefore, the charging system noding diagram shown in Figure 3-1 is not used.

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinq Report Paqe 3-5 Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensincq Report Paqe 3-6 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization

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ANP-331 5NP Revision 0 Millstone Unit 2 M5~ Upgrade, Small Break LOCA Analysis Licensinaq Report Panqe 3-7 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Paqe 3-8 3.3 Plant Description and Summary of Analysis Parameters Millstone Nuclear Plant Unit 2 is a CE-designed PWR with two hot legs, four cold legs, and two vertical U-tube SGs. The reactor has a core power of 2754 MWt (including measurement uncertainty). The reactor vessel contains a downcomer, upper and lower plenums, and a reactor core containing 217 fuel assemblies. The hot legs connect to the reactor vessel with the vertical U-tube steam generators. Main feedwater (MEW) is injected into the downcomer of each SG. There are three AFW pumps, two motor-driven and one turbine (steam)-driven.

The ECCS contains two HPSI pumps, two LPSI pumps, and four SITs.

The RCS was nodalized in the S-RELAP5 model with control volumes interconnected by flow paths or "junctions." The model includes four SITs, a pressurizer, and two SGs with both primary and secondary sides modeled. All of the loops were modeled explicitly to provide an accurate representation of the plant. A SG tube plugging level of 5.87%

was modeled in each SG. Important system parameters and initial conditions used in the analysis are given in Table 3-1. The heat generation rate in the S-RELAP5 reactor core model was determined from reactor kinetics equations with actinide and decay heating as prescribed by Appendix K.

The analysis assumed a Ioss-of-offsite power concurrent with reactor SCRAM, which is based-on the low pressurizer pressure reactor trip and includes delays for Reactor Protection System (RPS) circulation and Control Element Assembly (CEA) coil delay.

The assumption of Ioss-of-offsite concurrent with reactor SCRAM results in RCP trip.

Tripping the reactor coolant pumps at the time of SCRAM instead of time zero is reasonable since a small delay relative to the time of loop seal uncovery for the limiting case is considered to be conservative due to the additional loss of primary system inventory through the break.

The single failure criterion required by 10 CFR 50 Appendix K (Reference 7) was satisfied by assuming the loss of one emergency diesel generator (EDG). Thus, this

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinqi Report Paqe 3-9 results in the loss of one HPSI pump, one LPSI pump and one motor-driven AFW pump.

The initiation of the HPSI and LPSI systems were delayed by 25 and 45 seconds, respectively, following safety injection actuation system (SIAS) activation.

Table 3-2 and Table 3-3 show the minimum EGGS flow rates with diesel generator failure for HPSI and LPSI, respectively. The HPSI system was modeled to deliver the highest SI flow to both legs of the broken loop (Loop 2A and Loop 2B (broken loop)).

The LPSI system was modeled to deliver the highest SI flow to the broken loop (Loop 2B) and the lower SI flow to one of the intact legs in the intact loop (Loop 1A). Although the charging system is considered safety grade, it was not modeled in the analysis.

The disabling of a motor-driven AFW pump leaves one motor-driven pump and the turbine-driven pump available. The initiation of the motor-driven pump was delayed 240 seconds beyond the time of the auxiliary feedwater actuation signal (AFAS) indicating low SG level (0.0% narrow range). The turbine-driven AFW pump was not credited in the analysis.

The input model included details of both main steam lines from the SGs to the turbine control valve, including the MSSV inlet piping connected to the main steam lines. The MSSVs were set to open at their nominal setpoints plus 3% tolerance.

The axial power shapes for this analysis are shown in Figure 3-4. Figure 3-4 compares the axial power shape at mid-node elevation for hot rod, hot assembly, inner and outer core used in the analysis.

3.4 SER Compliance A spectrum of cold leg break sizes from 0.02182 ft2 (2.0-inch diameter) to 0.49120 ft2 (9.49-inch diameter, 10% of cold leg pipe area) was analyzed. This satisfies the limitation placed on EMF-2328 (Reference 1), that the methodology is acceptable for modeling transients where the break flow area is less than or equal to 10% of the cold

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Pagqe 3-10 leg flow area.

There is no other SER requirement or restriction on EMF-2328 (Reference 1 ).

In addition, the reactor coolant pump trip sensitivity study, attached piping break study and safety injection low liquid temperature study have been performed to support the operation of Millstone Nuclear Plant Unit 2 with M5 cladding.

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensing Renort Paae 3-11

,1 r

Table 3-1 System Parameters and Initial Conditions Reactor Power, MWt 27541 Axial Power Shape Figure 3-4 Peak LHR, kW/ft 15.1 Radial Peaking Factor (1.69 plus uncertainties) 1.854 RCS Flow Rate, gpm 360,000 Pressurizer Pressure, psia 2250 Core Inlet Coolant Temperature, 0F 549 SIT Pressure, psia 214.7 SIT Fluid Temperature, °F 120 SIT Water Volume, ft3 1135 Maximum SG Tube Plugging Level per SG, %

5.87 SG Secondary Pressure, psia 880 MFW Temperature, 0F 435 AFW Flow Rate per SG, gpm 72 AFW Temperature, °F 70 Low-Low SG Level Setpoint, % Narrow Range Span 0

AFW Delay, sec 240 HPSI and LPSI Fluid Temperature, 0F 140 Pressurizer Pressure - Low Reactor Trip Setpoint (RPS), psia 1700 Reactor Trip Delay Time on Low Pressurizer Pressure, sec 0.9 SCRAM CEA Holding Coil Release Delay Time, sec 0.5 SIAS Activation Pressurizer Pressure Setpoint (Harsh 1500 Environment Conditions), psia HPSI Pump Delay Time on SIAS, sec 25 LPSI Pump Delay Time on SIAS, sec 45 MSSV Lift Pressure and Tolerance Nominal + 3%

1 Includes 2.0% measurement uncertainty

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Report Paae 3-12 Table 3-2 High Pressure Safety Injection Flow Rates for Cold Leg Breaks RCS Cold Leg Loop IA Loop lB Loop 2A Loop 2B Pressure (lbmls)

(Ibmls)

(lbmls)

(lbmls)

(psia)__ _

14.7 19.555 19.418 19.828 19.828 50 19.281 19.144 19.418 19.418 100 18.871 18.871 19.008 19.008 150 18.461 18.461 18.461 18.461 200 18.050 18.050 18.050 18.050 300 17.093 17.093 17.093 17.093 500 14.905 14.905 14.905 14.905 700 12.581 12.581 12.581 12.581 900 9.709 9.709 9.709 9.709 1000 8.068 8.068 8.068 8.068 1050 7.111 7.111 7.111 7.111 1100 5.8.80 5.880 5.880 5.880 1150 4.102 4.102 4.102 4.102 1190 2.051 2.051 2.051 2.051 1204 0.0 0.0 0.0 0.0 Table 3-3 Low Pressure Injection Flow Rates for ColdLegBreaks RCS Cold Leg Intact Loop Intact Loop Intact Loop Broken Loop Pressure IA lB 2A 2B (psia)

(Ibmls)

(Ibmls)

(Ibmls)

(Ibm/s) 14.7 179.683 0.0 0.0 187.204 50 159.171 0.0 0.0 166.008 100 123.617 0.0 0.0 129.224 150 70.971 0.0 0.0 74.663 200 0.0 0.0 0.0 0.0

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Paqe 3-13 Figure 3-4 Axial Power Distribution Comparison

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Lic~ensincq Report Paqe 4-1 4.0 ANALYTICAL RESULTS The analysis results demonstrate the adequacy of the EGOS to support the criteria given in 10 CFR 50.46(b)(1 -4) for Millstone Nuclear Plant Unit 2 operating with AREVA supplied Advanced CE14 Fuel with M5 cladding.

Section 4.1 describes the SBLOOA break spectrum for the cold leg break. Section 4.2 describes the event for the limiting break size. Section 4.3 discusses the delayed RCP trip sensitivity study. Section 4.4 discusses the attached piping break sensitivity study.

Section 4.5 discusses the safety injection low fluid temperature sensitivity study.

4.1 Results for Break Spectrum The Millstone Nuclear Plant Unit 2 break spectrum analysis for SBLOOA inciudes breaks of varying diameter up to 10% of the flow area for the cold leg. The spectrum includes a wide enough range of break sizes from 2.0 inch diameter to 9.49 inch diameters to establish a POT trend. Additional break sizes are performed with a smaller break interval once the potential limiting break size is determined to confirm the limiting break size.

Figure 4-1 shows the calculated POTs for these breaks. For the break spectrum analysis, RCP trip is assumed to occur on reactor S GRAM.

The results of the cold leg SBLOCA break spectrum analysis are presented in Table 4-1. The predicted event times for the break spectrum are provided in Table 4-2. The limiting break size was determined to be 3.78-inch diameter (0.07793 ft2), resulting in a PCT of 1707°F.

4.2 Discussion of Transient for Limiting Break The break opens at t=0 seconds and initiates a subcooled depressurization of the primary system. The low pressurizer pressure trip setpoint is reached at 19 seconds

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ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Pane 4-2 and within 2 seconds the reactor is tripped then scrammed, the offsite power is lost, coincident with the turbine trip, RCP trip, and MEW pump trip (Figure 4-2, Figure 4-1 1, Figure 4-12 and Table 4-2). The SIAS is issued at 27 seconds on low pressurizer pressure. As MEW to the SGs is ramped down, the pressure in the SGs increase for approximately 30 seconds until MSSV inlet reaches the lowest opening pressure setpoint. This provides core heat removal in the early stages of the transient.

The primary system depressurization continues at a relatively fast rate for the first 125 seconds as fluid rushes out of the break (Figure 4-3).

The primary side pressure continues to decrease, reaching that of the secondary side at approximately 300 seconds, thus ending SG secondary side inventory reduction and producing SG secondary side condensation (Figure 4-14). Shortly thereafter, the broken leg loop seal clears (356 seconds, Figure 4-6), and the primary side pressure reaches the saturation pressure of the fluid in the hot legs and reactor vessel upper plenum (366 seconds),

which results in the fluid beginning to flash to steam as demonstrated in the horizontal loop seals and break void fractions in Figure 4-6 and Figure 4-5, respectively.

Prior to loop seal clearing in the broken leg, the core uncovers about 4 feet below the top of the active fuel (Figure 4-9, Figure 4-20 and Figure 4-21). As there is no loop flow, a large amount of steam is generated and accumulated in the core by the decay heat power until enough pressure is built to blow the upflow leg of the loop seal in the broken leg around 356-seconds Into the transient. This causes an abrupt level drop in the downcomer region (Figure 4-8) with a simultaneous core recovery (Figure 4-9). As the broken leg clears, the plant then enters a fairly slow boil-off phase where mass is lost out the break, and the primary system continues to empty. All intact loops remained plugged for the duration of the transient.

As liquid drains out of the loop piping, the break flow transitions from liquid to two phase flow, and then to steam. The break flow becomes primarily steam around 366 seconds resulting in a reduced mass flow rate out of the break (Figure 4-4) and an increase in the depressurization rate of the primary system (Figure 4-3). The liquid level in the

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ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensin~q Report Paqe 4-3 reactor vessel continues to drop until the reactor vessel reaches a minimum level at 1386 seconds (Figure 4-10).

Although HPSI flow to the primary system cold legs began at approximately 62 seconds into the transient (Figure 4-16), it does not provide sufficient inventory at this time to offset the large amounts lost out the break at this time. As effective cooling is lost in the core, the fuel rods begin to heat up at approximately 800 seconds (Figure 4-22). The fuel continues to heat up until the maximum POT of 1707°F is reached at 1824 seconds.

Fuel rod rupture does occur for the hot rod, but the calculated blockage factor indicates that the channel around the hot rod is not completely blocked and that all other channels in the core are also not completely blocked. Therefore, the hot rod and all other channels in the core are amenable to cooling.

For this break size, the HPSI flow is eventually sufficient to compensate for the rate of inventory loss out of the break.

At the time of the POT, the primary system has depressurized to a pressure slightly above the SIT pressure and the LPSI shut-off head.

SIT injection begins at 4580 seconds (Figure 4-18), followed by LPSI injection 56 seconds later (Figure 4-17), resulting in no influence on POT turnaround.

The downcomer level (Figure 4-8) and the reactor vessel inventory (Figure 4-10) start slowly increasing at approximately 1400 seconds. The onset of SIT and LPSI injection helps the reactor vessel levels to step up, ensuring core recovery and long term core cooling.

In conclusion, the limiting POT break spectrum case is a 3.78-inch diameter cold leg break. The POT of this case is 1707°F. The maximum local oxidation is 3.6% and the maximum core-wide oxidation is less than 0.04%. The total maximum local oxidation is less than 6%, including a pre-transient oxidation of 2.3%.

The hot rod resulted in rupture, but remained amenable to cooling. The results of the analysis demonstrate the adequacy of the ECOS to support the 10 CFR 50.46(b) (1-4) criteria (Reference 4).

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Paqe 4-4 4.3 RCP Trip Sensitivity Study For plants such as Millstone Nuclear Plant Unit 2 that do not have an automatic RCP trip, a delayed RCP trip can potentially result in a more limiting condition than tripping the RCPs at reactor SCRAM. Continued operation of the RCPs can result in earlier loop seal clearing with associated two-phase flow out the break, which would result in less inventory loss out the break early in the transient, but in the longer term could result in more overall inventory loss out the break. It has been postulated that tripping the pumps when the minimum RCS inventory occurs could cause a collapse of voids in the core, thus depressing the core level and provoking a deeper core uncovery, and a potentially higher PCT. Therefore, an RCP trip sensitivity for both the cold and hot leg breaks was performed with delayed time following loss of subcooling margin to demonstrate 10 CFR 50.46(b)(1 -4) criteria (Reference 4). This manual RCP trip study was performed consistent with the Combustion Engineering Owners Group guidelines described in Generic Letter 86-06 (Reference 8) where compliance with 10 CFR 50.46 is demonstrated when operator action to trip the RCPs is taken within 2 minutes after the RCP trip criterion is reached using the 10 CFR 50 Appendix K (Reference 7) method.

Also, consistent with Generic Letter 86-06, additional delayed RCP trip sensitivity studies were performed to find out the maximum delay time for operator action under a more realistic scenario. Best-estimate assumptions were applied using the same model-as the Appendix K RCP analysis with relaxation in two areas: decay heat multiplier reduction from 1.2 to 1.0 and critical break flow model change from Moody to the Homogeneous Equiiibrium Model.

Relaxing these conservative parameters gives an analysis which is more representative of a realistic response. A range of RCP trip delay times was examined for both hot and cold leg break locations.

The results demonstrated that longer delay times could be accommodated and still meet the 10 CFR 50.46 criteria.

Section 4.3.1 discusses the cold leg breaks RCP trip sensitivity study. Section 4.3.2 discusses the hot leg breaks sensitivity study.

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensincq Report Paqe 4-5 4.3.1 Cold Leg Breaks RCP Trip Sensitivity Study For the cold leg break RCP trip sensitivity study, the spectrum of break sizes from 2.0 inches to 9.49 inches in diameter were re-analyzed with a 2-minute RCP trip delay following loss of subcooling margin in the cold leg pump suction. The limiting cold leg break size for this sensitivity study was a 4.02-inch diameter break, producing a PCT of 1644°F. Therefore, the cold leg breaks RCP trip sensitivity study has shown that there is at least 2 minutes to trip ali four RCPs after loss of subcooling margin in order to meet the 10 CFR 50.46(b)(1-4) criteria.

In addition, relaxation of Appendix K assumptions demonstrated that longer delay times of up to 15 minutes after losing subcooling at the RCP suction could be accommodated and still meet the 10 CFR 50.46 criteria for a spectrum of cold leg break cases.

4.3.2 Hot Leg Breaks RCP Trip Sensitivity Study For the hot leg break RCP trip sensitivity study, a spectrum of hot leg break sizes from 2.0 inches to 9.49 inches in diameter were analyzed with a 2-minute RCP trip delay following loss of subcooling margin in the cold leg pump suction. The limiting hot leg break size for this sensitivity study was the 5.0-inch diameter break, producing a POT of 1575°F. Therefore, the hot leg breaks RCP trip sensitivity study has shown that there is at least 2 minutes to trip all four RCPs after loss of subcooling margin in order to meet the 10 CFR 50.46(b)(1-4) criteria.

In addition, relaxation of Appendix K assumptions demonstrated that longer delay times of up to 10 minutes after losing subcooling at the RCP suction could be accommodated and still meet the 10 CFR 50.46 criteria for a spectrum of hot leg break cases.

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinq~ Report Pacqe 4-6 4.4 Attached Piping Break Sensitivity Study Although breaks in the attached piping are not typically PCT limiting, they do result in reduced EGOS flows available to mitigate the event.

Therefore, an analysis of the limiting break size and location in an attached piping was performed.

For Millstone Nuclear Plant Unit 2, the limiting break location and size for an attached piping is considered a double-ended guillotine break of an SIT line. The break was located in the SIT line connected to Loop 2B.

For the double-ended guillotine break in the SIT, the calculated POT is 1239°F, which is bounded by the limiting POT of the break spectrum. The minimal HPSI and LPSI flow rates modeled were sufficient to prevent a subsequent heatup after the initial quench from the SIT discharge.

4.5 Safety Injection Low Fluid Temperature Sensitivity Study

AREVA Inc.

Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis I ir.i.n.*inn Re~nnrt ANP-331 5NP Revision 0 Paae 4-7

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis LikeRnsina Renort Paae 4-8 17 -

r.-.....

7--

Table 4-1 Summary of SBLOCA Break Spectrum Results Break diameter (in) 2.00 3.00 3.60 3.70 3.75 3.76 Break Area (ft2) 0.02182 0.04909 0.07069 0.07467 0.07670 0.07711 Peak Clad Temperature (0 F) 1135 1384 1472 1501 1599 1651 Time of PCT (sec) 4330 2451 2008 1978 1892 1857 Time of Rupture (sec) 1772 1654 Transient Local Maximum Oxidation (%)

0.0864 0.4249 0.6647 0.7543 1.9966 2.6718 Total Local Maximum Oxidation (%)1 2.337 2.676 2.915 3.005 4.247 4.922 Core Wide Oxidation (%)

0.0042 0.0111 0.0124 0.0139 0.0254 0.0314 PCT Elevation (ft) 10.52 10.52 10.77 10.77 11.02 11.02 Break diameter (in) 3.78 3.785 3.79 3.90 4.02 4.40 Break Area (ft2) 0.07793 0.07814 0.07834 0.08296 0.08814 0.10559 Peak Clad Temperature (0F) 1707 1690 1476 1606 1596 1581 Time of PCT (sec) 1824 1852 2035 1954 1746 1284 Time of Ruptu re (sec) 1563 1591 1852 1684 1270 Transient Local Maximum Oxidation (%)

3.5273 3.2474 0.6689 1.3276 1.1046 0.6721 Total Local Maximum Oxidation (%)1 5.778 5.498 2.920 3.578 3.355 2.923 Core Wide Oxidation (%)

0.0396 0.0368 0.0119 0.0175 0.0162 0.0110 PCT Elevation (ft) 11.02 11.02 10.77 11.02 11.02 10.52 Break diameter (in) 4.60 4.80 5.00 5.30 5.50 6.00 Break Area (ft2) 0.11541 0.12566 0.13635 0.15321 0.16499 0.19635 Peak Clad Temperature (0F) 1621 1637 1615 1591 1398 1535 Time of PCT (sec) 1076 939 832 699 674 501 Time of Rupture (sec) 1047 907 814 691 Transient Local Maximum Oxidation (%)

0.7874 0.8477 0.6456 0.5072 0.1142 0.2316 Total Local Maximum Oxidation (%)l 3.038 3.098 2.896 2.758 2.365 2.482 Core Wide Oxidation (%)

0.0108 0.0111 0.0091 0.0075 0.0020 0.0051 PCT Elevation (ft) 10.77 10.77 10.52 10.52 10.27

10. 27 Break diameter (in) 7.00 8.00 9.00 9.49 Break Area (ft2) 0.26725 0.34907 0.44179 0.49120 Peak Clad Temperature (0F) 1510 1540 1399 1437 Time of PCT (sec) 353 258 202 180 Time of Rupture (sec)

Transient Local Maximum Oxidation (%)

0.2133 0.2247 0.1097 0.1367 Total Local Maximum Oxidation (%)1 2.464 2.475 2.360 2.387 Core Wide Oxidation (%)

0.0049 0.0046 0.0018 0.0024 PCT Elevation (ft)

10. 27

10. 27 10.02 10.02 1 Includes the pre-transient oxidation of 2.2507%

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensino Reoort Paoe 4-9 Table 4-2 Sequence of Events for Break Spectrum (seconds)

=

• =

=

I o

C IC C.0 18 00 29 30 38 5

63 83 9

4C8)..

512 552 02 25 20 3.6 142 0

2 2

2-6 5

74 6IC8 41 384 C39 570 In3 In0 20 3.78 169 IC 9

2 7

44 5

2 6

IC IC 0

408

• 5 36 43 1396 191 185 18 2.0 10

0. 181 5

4 50 70.5

0.

43 36 230 116 3004 58 8215 9

2.00 1159 0

6617 792 90 149 124 242 516 70 125 1088 1100 173 41421684 1433 3030 3.00 1384 0

29 30 38 540 63 83 98 418 600 23 9512 552 127 12308

--7 12845 280 3.60 14721 0 20 22 29 46 545 74 468 414 250 250 347 394 57008 1638 104 20086 208 3.70 1501 0

19 214 27 46 52 7264 46 412...

48 364 374 5102 1572 90 1978 200 3.75 1659 0

1920 271 46 52 724 64 84 428 30 71 3606 370 8466 1483 8177 1892 190 3.76 16591 0

19 20 27 3 4635 72 62

-1420 2

25 358 2368 4638 1428 165418579 188 3.78 1707 0

19 120 27 44 52 72 62 692 408 436 204 270 356 366 4580 1367 156 1824 186 3.785 1690 0

19 0 127 44 52 72 62 51 408 356 13661453481396 159-51 185218 3790 1476 0

9 18 20 2143 51 71 62 360 408 944 1950 1330 3746 35 1680 2353 186 3.90 16406 0

18 19 1 2514 508 7 58 40 434 369 24 11 7

6 35019424 1598 185 1954 192 4.02 1596 0

17 18 24 42 49 69 56 206 43 700

-6 125 1027 340 1738 1452 168 17246 184 8

4.40 1581 0

14 16 22 40 47 67 48

-1842-600 76 120 954 292 1278 1230 127 1284 142

AREVA Inc.

Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Reoort ANP-331 5NP Revision 0 Paae 4-10 Figure 4-1 Peak Cladding Temperature versus Break Size (SBLOCA Break Spectrum) 0~

2200.0 2000.0 1800.0 1600,0 1400.0 1200.0 1000.0 800.0 600.0 400.0 200,0

~~0' 00-°

°

- -°.....

0o-O So Breaks _ Transition Break

• Breaks <Transition Break~

- Transition Break-0.0 2.50 3.00 2.00 2.0 3,0350 4.00 4.50 5.00 5.50 6.00 6.50 7.00 7.50 8.00 8.50 9.00 9.50 Break Size, in

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Lice~nsina Reoort Paae 4-11 i i i

f.

Figure 4-2 Reactor Power - 3.78-inch Break Reactor Power 3000.0,-.-.----------- -

200.

2000.0

....... Reactor Power a

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n

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0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Report ANP-331 5NP Revision 0 P~ni. 4-12P (3_

Figure 4-3 Primary and Secondary System Pressures - 3.78-inch Break System Pressures 2 5 0050 0.0

`

`

2250.0 *-

1750.0i RV Upper Head

• SG-1 10..................................~...

SG-2 1500.0 **'

1500.0

• J*

7 5 0.0 0-

° 0

500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (a)

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Report Paae 4-13 2000.0 1500.0 r

1000.0 0.

Figure 4-4 Break Mass Flow Rate - 3.78-inch Break Break Flow

.....................Flo j....

.. I...

S 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 500.0 0.0 i0

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis I ien~n.ina Re.onrt Paae 4-14

• '='i * "*r=" v..

, i Figure 4-5 Break Vapor Void Fraction - 3.78-inch Break 1.0 Break Vapor Void Fraction

-U

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AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis I ir'i~n~inn Renort

= *

=

Figure 4-6 Loop Seal Void Fraction - 3.78-inch Break Loop Seal Void Fraction 1.0 r 0.8 !

• 064

--- a Loop 1A

-....... Loop 1B

• • Loop 2A

........4Loop 2B 0.2 i

1 0

500}

1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Report ANP-331 5NP Revision 0 Paae 4-16 40001 30001 20001 W

10001 Figure 4-7 Total Core Inlet Mass Flow Rate - 3.78-inch Break Total Core Inlet Mass Flow Rate 30, Flow 0.0 300 0.0

'S.~4*~U.

~ -a.. a.-

-- in u-a-u--U.u-. ~

-10000.0 0

500 1000 1500 2000 30 50 40 50 50 2500 Time (s) 3000 3500 4000 4500 5000

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis I inr.~n.ina Re~nnrt Paae 4-17 Figure 4-8 Downcomer Collapsed Liquid Level - 3.78-inch Break Downcomer Level 40.0............

35.0

....... DC Level - Average U,

"0.

U, 30.0

"=-'="='

25.0 20.0

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0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinq Report Paqe 4-18 Figure 4-9 Inner and Outer Core Collapsed Liquid Level - 3.78-inch Break Inner/Outer Core Liquid Level r

u nerCr 71i

....... Outer Core 8.0 H,.-

0)n o=

6.0 "

J!

i 0

500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

APaqe 5NP9 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Reoort Paq~A~-i9 Figure 4-10 Reactor Vessel Mass - 3.78-inch Break RV Mass 250000.0.................... ---

1 1

200000.0 Mass E

n w

150000.0

-s a

tv/i h

a

• • U=.....a l

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0 500 1000 1500 2000 2500 3000 Time (s) 3500 4000 4500 5000

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingq Report Paqie 4-20 Figure 4-11 RCS Loop Mass Flow Rates - 3.78-inch Break RCS Loop Fiow Rates 9000.0 7000.0 000.0....

Loop 1 A 5000' ii

......... -*Loop21B

[Z 4 Loop 2AB.

3000.0 i

-1000.0

--3000.0..

-0000 500 1000 150 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Report ANP-331 5NP Revision 0 Pane 4-21 I I I

m Figure 4-12 Steam Generator Main Feedwater Mass Flow Rates -

3.78-inch Break SG MFW Flow Rates 2 0 0 0.0........ -.----------................,.- :--*--,--,

1500.0 1000.0 Loop 1 4

Loop 2

.0 0*

(t-500.0 0

500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis I irnp~~inn R*.nnrt Paae 4-22 Figure 4-13 Steam Generator Auxiliary Feedwater Mass Flow Rates -

3.78-inch Break Auxiliary Feedwater Flow

...... Loop 1

.......* Loop 2 cr 0

LL 10.0 u-

• -*w*m**m*=.-

5.0 0

500 1000

-.*.D-*,I*..I&*

-* I*--*.ID-III* II @*

• 114q1.,*11--4*-,*i,.*1 --481,.,1*-.

3500 4000 4500 5000 1500 2000 2500 3000 Time (s)

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis I iC'.n.flinlO RP.flort PainA 4-23I

,-i " -vr--.....

"1.

Figure 4-14 Steam Generator Total Mass - 3.78-inch Break 150000.0......

140000.0 SG Total Mass Loop 1

.......*'Loop 2

.a..' 3* -U*

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.j

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i.

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./

/

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.4-100000.0 L........L..

0 500 1000 1500 2000 2500 3000 3500 4000 Time (s) 4500 5000

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Report P~rn* 4-9A ii mI Figure 4-15 Steam Generator Narrow Range Level % - 3.78-inch Break SG NR Level %

',""*~~~~~~~~~~~~~~~~~~.......................................r'"r""'"F""r'"r""'"*

60,0

..,,Intact Loop SG Level

...e Broken Loop SG Level a,

-J 40.0 20.0 0

500 1000 1500 2000 2500 3000 3500 Time (s) 4000 4500 5000

AREVA Inc.

APa3e 4-NP Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensina Report Paae 4-25 Figure 4-16 High Pressure Safety Injection Mass Flow Rates - 3.78-inch Break HPSI Mass Flow Rates 2 0.0 r--

,,,- r.......... -- - -, =- - -,........ --"............--r --

r..... ", "'"...... "........

15.0 Er 1,0, n_

5.0 -

0 500

/

..........- Loop 1A 4=

..........,*Loop 1B jl" Loop 2A

.......... 4Loop 2B 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

ANP-331 5NP Revision 0 Miiistone Unit 2 M5 Upgrade, Smaii Break LOCA Analysis Licensingq Report Paqe 4-26 Figure 4-17 Low Pressure Safety Injection Mass Flow Rates - 3.78-inch Break LPSI Mass Flow Rates 30,0i

• Loop 1A

• Loopl1B 20....-...

Loop 2A

• Loop 2B 1r 0.0"

_oa

0.

-4IJ_4

~'

f

~

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4,@$~

~'

f'Q4~

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensino Reoort ANP-331 5NP Revision 0 Paae 4-27 Figure 4-18 Safety Injection Tank Mass Flow Rates - 3.78-inch Break 600.0 *...... -----.............r SIT Injection T............""-.

"T r

' r '

00

....... Loop lA Loopi1B

• .Loop 2A K

Loop 2B 100.

0,0 *.*.**4B*.*. * *`.*

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• } *: * *:'*'*'

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0 500 1000 1500 2000 2500 Time (s)

.o.o--

5oo ooo-oo oo

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensino Reoort AJ)Oi I

Figure 4-19 Integrated Break Flow and ECCS Flow - 3.78-inch Break Integrated Break Flow and Total EGG Flow 800800000.0

......... *........., : r * : ""

600000.0 -

400000.0

.m Interated Beak Flo

@.81,..1....].....'1']

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E n

0

~.0 Cu Vw Cu us C

r

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4i*

0.0 04 *÷.I* **.i-(

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0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (e)

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinq Report Paqe 4-29 Figure 4-20 Hot Assembly Collapsed Liquid Level - 3.78-inch Break Hot Assembly Collapsed Level 12.0

10.

Liquid Level 4.0 200 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 2.0 (s

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensinqi Report Paqe 4-30 Figure 4-21 Hot Assembly Mixture Level - 3.78-inch Break Hot Assembly Mixture Level Mixture Level 10.0 t*£/"

S 9.0i 0

500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

ANP-331 5NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis

[ ice.nsina Renort Paoe 4-31 Figure 4-22 Peak Cladding Temperature at POT Location (11.02 ft) -

3.78-inch Break 2000.0 i,

Peak Cladding Temperature

• .-./.i.....,..-.-....-.--....- --.. :..U--- -.:.- -.:..¢.-:.-*-..*--:..,--:--:.-..,.........-- --.-..-.

1500.0 U-a) a~

0~Ea)

F-p U

1000.0

• 500.0 F u-na--m, a a,...--nm a a.

• r~n--m--s.

IIl*,.ill...m -.I 0.0 L..

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

AREVA Inc.

ANP-3315NP Revision 0 Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensingl Report Paqe 5-1

5.0 REFERENCES

1.

AREVA Inc. Topical Report EMF-2328(P)(A) Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001.

2.

AREVA Inc. Topical Report EMF-2328(P) Revision 0, Supplement 1, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2012.

3.

AREVA Inc. Topical Report BAW-10240(P)(A) Revision 0, Incorporation of M5 Properties in Framatome ANP Approved Methods, May 2004.

4.

Code of Federal Regulations, Title 10, Part 50, Section 46, Acceptance Criteria For Emergency Core Coollng Systems For Light-Water Nuclear Power Reactors, January 2010.

5.

AREVA Inc. Topical Report XN-NF-81-58(P)(A) Revision 2, Supplements 1 and 2, RODEX2 FUEL Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.

6.

AREVA Inc. Topical Report ANF-81-58(P)(A) Revision 2, Supplements 3 and 4, RODEX2 FUEL Thermal-Mechanical Response Evaluation Model, Advanced Nuclear Fuels Corporation, June 1990.

7.

Code of Federal Regulations, Title 10, Part 50, Appendix K, ECCS Evaluation Models, March 2000.

8.

Nuclear Regulatory Commission Generic Letter 86-06,

Subject:

Implementation of TMI Action Item I1.k.3.5, 'Automatic Trip of Reactor Coolant Pumps" (Generic Letter No. 86-06), Mary 29, 1986.

Serial No.15-411 Docket No. 50-336 AREVA APPLICATION FOR WITHHOLDING AND AFFIDAVIT DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

AFFIDAVIT COMMONWEALTH OF VIRGINIA

)) ss.

CITY OF LYNCHBURG)

1.

My name is Gayle Elliott. I am Manager, Product Licensing, for AREVA Inc.

(ARE VA) and as such I am authorized to execute this Affidavit.

2.

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

3.

I am familiar with the AREVA information contained in the Licensing Report ANP-3315P, Revision 0, entitled, "Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis,"

dated April 2015 and referred to herein as "Document." Information contained in this Document has been classified by ARE VA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4.

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by ARE VA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6.

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

(a)

The information reveals details of ARE VA's research and development plans and programs or their results.

(b)

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

(c)

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

(d)

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

(e)

The information is vital to a competitive advantage held by ARE VA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(c) through 6(e) above.

7.

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

8.

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

9.

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

SUBSCRIBED before me this____h:

day of 4pr v-]

2015.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg. # 7079129 IMy 70701a9nE~mlOct 31, 2018