ML20198G704

Rev 0 to EMF-2079, Millstone Unit 2 Large Break Loca/Eccs Analysis with Replacement Sgs

ML20198G704
Person / Time
Site:	Millstone
Issue date:	09/30/1998
From:	Gottula R, Holm J, Shaw R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20198G702	List: B17574, Forwards Rev 0 to EMF-2079 & Rev 0 to EMF-2145,both Providing Results of Evaluation of Large Break LOCA Event for Mnps,Unit 2.No New Regulatory Commitments Contained within Ltr ML20198G704 ML20198G716
References
EMF-2079, EMF-2079-R, EMF-2079-R00, NUDOCS 9812290148
Download: ML20198G704 (42)

Text

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f SIEMENS EMF-2079 Revision 0 ilA 4

Millstone Unit 2 Large Break LOCA/ECCS Analysis With Replacement Steam Generators 1

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EMF-2079 Revision O Issue Date:

, Millstone Unit 2 Large-Break LOCA/ECCS Analysis With Replacement Steam Generators Prepared: kW 9lM 98 R. A. ShMw, Engineer Date PWR Safety Analysis Approved: N C, d T/3/98 R. C. Gottula, Manager Date PWR Safety Analysis

[ I Approved: 6 /) $!7 J.S.H ,'an$ger M Date '

Produ t icensing Approved: [ 4/ k 9/8/9f L. E. Hansen, M'a' nager D6te Customer Projects i

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Customer Disclaimer important Notice Regarding Contents and Use of This Document Please Read Carefully Siemens Power Corporation's warranties and representations conceming the subject matter of this document are those set forth in the agreement between Siemens Power Corporation and the Customer pursuant to which this document is issued.

Accordingly, except as otherwise expressly provided in such agreement, neither Siemens Power Corporation nor any person acting on its behalf:

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

privately owned rights; or

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! apparatus, method, or process disclosed in this

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The information contained herein is for the sole use of the

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( until so authorized in writing by Siemens Power Corporation or

[ until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document.

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EMF-2079 Millst::na Unit 2 Largs Bruk LOCA/ECCS Ans. lysis RIvision O With Replacement Steam Generators Pagei l

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EMF-2079 Millstona Unit 2 Largs Brask LOCA/ECCS Anzlysis Ravision 0 With Replacement Steam Generators ;

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t. - Contents 1.

I n t ro d u c t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. .

Summary......................................................................................................2-1

3. Anatysis........................................................................................................3-1 3.1 De sc ription of the LBLO CA Event .. . .. . .... . .. . . .. . . . .. . . . . . . ... . . . . . . . . . . . . . . . . . . .... . . . . . 3- 1 3.2 De sc riptio n of Analytic al Mod els .......... .... . .. . . . .. . . . .. . . . . . .. . . . . .. .. . . .. . . .. . .... . . . . 3 2 3.3 Plant Description and Summary of Analysis Parameters ........................... 3-3

, 3.4 Break Spectrum and Axial Shape Study Results ...................................... 3-4 .

3.5 Ex po su re St ud y Re su lt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 3.6 EO C Coa std a wn Operatio n . . .. .. . . . . . . ... . . . . . .. .. . . .... . . .. . . . . . . . ... ..... . . . ... .. . . . . . . .. . . 3 6 4.

C o n cl u si o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

5.

References....................................................................................................5-1 Tables

.3.1 Sys tem Analysis Pa ra meters . . . . . . . . .. .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-7 3.2 Fuel D e s ig n Pa ra met e rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 8 3.3 SIS Delivery Curves for Loss-of-LPSI Pump Single-Failure .................................. 3-8 3.4 Changes to the Containment Model Used in This Analysis................................. 3 9 3.5 Break Spectrum and Axial Shape Analysis PCT Results................................... 3-10 3.6 Sun, mary of Results for 0.6 DECLG Break with MOC Axial and Loss-of-LPSI Pump......................................................................................................3-10 3.7 - Calculated Event Times for O.6 DECLG Break with MOC Axial and Loss-of-LPSI Pump.........................................................................................................3-11 Figures 2.1 Allowable LHGR as a Function of Axial Location............................................... 2-1 3.1 Normalized Power, 0.6 DECLG Break, MOC Axial ........................................... 3-12 3.2 SIT Discharge Rates,0.6 DECLG Break, MOC Axial, Loss-of-LPSI Pump........... 3-13 3.3 HPSI Flow Rates, 0.6 DECLG Break, MOC Axial, Loss-of LPSI Pump ................ 3-14 Siemens Power Corporation Nuclear Division

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, EMF-2079 Millstona Unit 2 Ltrgs Brssk LOCA/ECCS Analysis Ravision O With Replacement Steam Generators Page lii 3.4 LPSI Flow Rates, 0.6 DECLG Bresk, MOC Axial, Loss-of-LPSI Pump................. 3-15 3.5 Upper Plenum Pressure During Blowdown, 0.6 DECLG Break ........................... 3-16 3.6 Total Break Flow Rate During Blowdown, 0.6 DECLG Break ............................ 3-17 3.7 Average Core inlet Flow Rate During Blowdown O.6 DECLG Break, MOCAxiaI................................................................................................3-18 .

3.8 Hot Channelinlet Flow Rate During Blowdown,0.6 DECLG Break, MOCAxia!.................................................................................................3-19 .

l 3.9 PCT Node Fluid Quality During Blowdown,0.6 DECLG Break, i

MOCAxiaI................................................................................................3-20 3.10 PCT Node Fuel (Average), Cladding and Fluid Temperatures During Blowdown, 0.6 DECLG Break, MOC Axial, Loss-of-LPSI Pump ......................... 3-21 3.11 PCT Node Heat Transfer Coefficient During Blowdown,0.6 DECLG Break, M O C A xial, Lo s s-of- LPSI Pu m p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 2

3.12 PCT Node Heat Flux During Blowdown,0.6 DECLG Break, MOC Axial, Lo u s - o f - L PS I Pu m p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 2 3 3.13 Containment Pressure,0.6 DECLG Break, MOC Axial, Loss-of-LPSI Pump......... 3-24 3.14 Upper Plenum Pres 5Jre After Blowdown,0.6 DECLG Break, MOC Axial,  !

Lo s s-o f- L PS I Pu m p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 2 5 3.15 Downcomer Collapsed Liquid Level,0.6 DECLG Break, MOC Axial, Lo s s- o f - LPS I Pu m p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 2 6 3.16 Core Effective Flooding Rate,0.6 DECLG Break, MOC Axial,' Loss-of -

t LPSIPump..................................................................................................3-27 3.17 Core Mixture Level,0.6 DECLG Break, MOC Axial, Loss-of-LPSI Pump............. 3-28 -

3.18 Core Quench Level, 0.6 DECLG Break, MOC Axial, Loss-of-LPSI Pump............. 3-29 3.19 PCT Node Heat Transfer Coefficient,0.6 DECLG Break, MOC Axial, l Lo s s- o f- LPS I Pu m p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 3 0 i

3.20 PCT Node and Ruptured Node Cladding Temperatures,0.6 DECLG Break, M O C Axia1, Los s-o f-LPSI Pu m p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-31 l

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EMF-2079 Millstona Unit 2 Large Brask LOCA/ECCS Antlysis Ravision 0 With Rapliczmtnt Steam Generators Page 1-1 l

- 1. Introduction l

.This report describes an analysis of a postulated large-break losr-of-coolant accident (LBLOCA) in the Millstone Unit 2 nuclear power plant operating with Siemens Power Corporation (SPC) fuel. The SPC EXEM/PWR evaluation model[Ref. 21 was used to perform the analyses. The analysis supports operation during Cycle.13 and future cycles, unless changes in the Technical Specifications, Core Operating Limits Report, fuel design, plant hardware, or plant operation should cause the results presented herein to be invalid.

This analysis includes many changes from the previous [Ref.1] Cycle 13 analysis. For example, e New steam generators have been included in the models; e The maximum linear heat generation rate (LHGR) has been reduced from 15.1 kW/ft to 14.6 kW/ft; e

Implementation of new limiting axial power shapes for middle-of-cycle (MOC) and end-of-cycle (EOC) exposures; l e Changes in Safety injection System (SIS) delivery curves;.

• Changes in SIS delay times

.- high pressure safety injection (HPSI) 25 sec; e low pressure safety injection (LPSI) 45 sec; e Revision of the initial operating pressure in the containment; e Changes in the CAR fan performance (including single-failure criterion);

e Changes in the containment spray rates (including single-failure criterion); 1 1

• . Removal of paint from the containment surfaces;

• Miscellaneous corrections in various models.

A total steam generator tube plugging level of 500 tubes per steam generator (symmetric) was assumed in this analysis.

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i EMF-2079 Millstona Unit 2 Largs Bruk LOCA/ECCS Analysis Ravision 0 With Replacement Steam Generators i Page 2-1 I

2. Summary Based on calculations for a spectrum of break sizes and axial power shapes, a double-ended cold leg guillotine break (DECLG) with a discharge coefficient of 0.6 in conjunction with a MOC axial shape and failure of a single LPSI pump was found to be most limiting.

The peak cladding temperature (PCT) for that limiting case was calculated to be 1877 F.

The analysis supports full-power operation at 2700 MW (plus 2% r;ncertainty) with a total

, symmetric steam generator tube plugging level of up to 500 tubes per steam generator.

The analysis supports fuel assembly exposures of up to 56,000 mwd /MTU. The analysis demonstrates that the 10 CFR 50.46(b) criteria are satisfied with an axial and exposure independent LHGR limit of 14.6 kW/ft (as shown in Figure 2.1).

1 20 Unacceptable Operation j 14.6 kW/ft 5 10 - I ec 0

35 Acceptable Operation 1

0 . . , l 0 0.2 0.4 0.6 0.8 1 Fraction of Active Fuel Height Figure 2.1 Allowable LHGR as a Function of Axial Location Siemens Power Corporation - Nuclear Division

EMF-2079 Millstona Unit 2 Largs Bruk LOCA/ECCS Antlysis Revision O With Replacement Steam Generators Page 3-1

3. Analysis The purpose of the analysis is to demonstrate that the criteria stated in 10 CFR 50.46(b) are met. Those criteria are:

1. The calculated peak fuel cladding temperature does not exceed 2200*F.

- 2. The amount of fuel cladding which reacts chemically with water or steam does not  !

exceed 1% of the total amount of zircalloy in the core.

3. The cladding temperature transient is terminated at a time when the core geometry is i still amenable to cooling. The hot fuel rod cladding oxidation does not exceed 17%

during or after quenching.

~4. The core temperature is reduced and decay heat is removed for an extended period of time, ac ree.uired by the long-lived radioactivity remaining in the core.

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 Millstone Unit 2 plant and summarizes the system parameters used in the analysis. Section 3.4 summarizes the rcsults of the break spectrum and axial shape analysis. Section 3.5 summarizes the rcsults of the exposure analysis. Section 3.6 addresses EOC coastdown operation.

3.1 Description of the LBLOCA Event A LBLOCA is initiated by a postulated large rupture of the Reactor Coolant System (RCS) primary piping. A spectrum of breaks, DECLG with discharge coefficients (CD) ranging from 0.4 to 1.0, is typically analyzed. (Split breaks are also considered in LBLOCA analyses). The limiting break location is on the pump discharge side of a cold leg pipe.

A loss-of-cffsite power is assumed to occur coincident with the LBLOCA. Reactor Coolant Pump (RCP) coastdown occurs coincident with the loss-of-offsite powu.

The break initiates a rapid depressurization of the RCS. A reactor trip signalis issued when the Low Pressurizer Pressure trip setpoint is reached; however, reactor trip and scram are conservatively neglected in the analysis.

l Early in the blowdown, the core experiences flow stagnation and reversal, which causes l the fuel rods to pass beyond the critical heat flux. Subsequently, the fuel rods dissipate h:st via transition or film boiling modes of heat transfer. Any rewet that may occur during Siemens Power Corporation - Nuclear Division

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EMF-2079 l - Millstona Unit 2 Largs Brask LOCA/ECCS Analysis Rsvision O l- With Replacement Steam Generators Page 3-2 l.

. the blowdown is conservatively neglected in the analysis, as required by Appendix' K of 10 CFR 50.

A Safety injection Actuation Signal (SlAS) is issued when the High Containmera Pressure setpoint is reached. Due to the loss-of-offsite power, there is a time delay for diesel generator startup in addition to the time delays for HPSI and LPSI pump startup. The single-failure criterion is met by assuming that either one diesel generator fails or one LPSI .

pump fails.

When the RCS pressure falls below the Safety injection Tank (SIT) pressure, fluid from the

• SITS is injected into the cold legs. This fluid is assumed to bypass the core and flow to the break until sustained downflow in the downcomer is predicted to occur (which is used to identify the end-of-bypass (EOBY) time).

Following the EOBY time, Emergency Core Cooling System (ECCS) fluid (from the SITS and, later, SIS) fills the downcomer and lower plenum until the liquid level reaches the bottom of the core (which is defined as the beginning-of-core-recovery (BOCREC) time). >

During the refill period, the fuel rods are cooled by radiation heat transfer.

The reflood period begir.a at the BOCREC time. ECCS fluid fillir.g the downcomer provides the driving head for reflooding the core. As the liquid level moves up the core, steam is generated. Steam binding occurs as the steam flows through the intact- and broken-loop  ;

steam generators and RCPs. It is conservatively assumed that the rotor of each RCP has seized (per Appendix K of 10 CFR 50), which tends to reduce the reflood rate. The fuel '

rods are cooled by radiation and, eventually, by convection as the quench front moves up

[ the core. The reflood heat transfer rate is evaluated from experimentally-determined heat .

l' transfer and carryover rate fraction correlations.

l 3.2 Description of Analytical Afodels p

The SPC EXEM/PWR Evaluation Model (Ref. 2] was used to perform the analysis. This ,

model consists of the following computer codes:

1. HODEX2 for computation of the initial fuel stored energy,' fission gas release, and '

fuel-cladding gap conductance;

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2. RELAP4-EM for the system blowdown, hot channel blowdown, and SIT / SIS flow '

calculations; k

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Millstona Unit 2 Lcrgs Brcsk LOCA/ECCS Antlysis Rsvizion O With Replacement Steam Generators l Page 3-3 l

! 3. CONTEMPT /LT-22, as modified in accordance with NRC Branch Technical Position  !

CSB 6-1, for computation of the containment back-pressure; t

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4. REFLEX for the system reflood calculation;

5. TOODEE2 for the fuel rod heatup calculation during the refill and reflood portions of the LBLOCA event.

. The quench time, quench velocity, and carryover rate fraction correlations in REFLEX and the heat transfer correlations in TOODEE2 were based on SPC's FCTF data.

The governing conservation equations for mass, energy, and momentum transfer were used, along with appropriate correlations consistent with Appendix K of 10 CFR 50. The <

reactor core was modeled in RELAP4 with heat generation rates determined from reactor kinetics equations with reactivity feedback, and with actinide and decay heating as r: quired by Appendix K. Appropriate conservatism specified by Appendix K was I incorporated in all of the EXEM/PWR models.

3.3 Plant Description and Summary of Analysis Parameters Millstone Unit 2 is a Combustion Engineering (CE) designed pressurized water reactor (PWR), which has two hot legs, two U-tube steam generators, and four cold legs with one RCP in each cold leg. The plant uses a large dry containment.

The RCS model was nodalized into control volumes representing reasonably homogeneous rrgions, interconnected by flow paths or " junctions." The two cold legs connected to the intact loop steam generator were assumed to be symmetrical ano were modeled as one intact cold leg with appropriately scaled input. The SITS, pressurizer, and steam g:nerators (both primary and secondary sides) were modeled. The HPSI and LPSI pumps were modeled as fill junctions in the SIT lines, with conservative flows given as a function of RCS backpressure. The RCP performance curves are representetive of CE-plant RCPs.

7 The core was modeled radially with an average-core assembly and a Sot assembly as parallel flow channels, each with three axial nodes.

A symmetric total steam generator tube plugging level of 500 tubes per generator was

! assumed. (This conservatively bounds operation with the new steam generators and is l . cxpected to also bound the plugging levels of future cycles.)

i Values for system parameters and fuel design parameters used in the analysis are given in

Tables 3.1 and 3.2, respectively.

Siemens Power Corporation - Nuclear Division

EMF-2079 .

Millstona Unit 2 Largs Break LOCA/ECCS Analysis Revision 0 With Replacement Steam Generators Page 3-4 Two single-failures were examined in this analysis:

1. Loss of one diesel generator, and .

2. Loss of one LPSI pump l in pr'evious Millstone-2 analyses, single-failure considerations were combined into a single, bounding calculation using loss-of-diesel SIS delivery in conjunction with full CAR fan performance and containment spray flow rate conditions, in the current analysis, SIS -

delivery, containment spray rates, and CAR fan performance were explicitly modeled for each of the single-failures listed abnve; and a separate calculation was performed for each ,

- of the single-failures. Table 3.3 presents the SIS delivery curves for the limiting single-

' failure case in the current analysis. Table 3.4 presents the revised containment parameters used in this analysis.

i The loss-of-diesel single-failure results in (a) greater loss of SIS delivery and (b) less containment cooling than occurs in the loss-of-LPSI scenario. Conversely, the loss-of-LPSI

single-failure results in more SIS delivery and more containment cooling than occurs in the loss-of-diesel scenario. These have minor competing effects on reflood rate that make it difficult to know a prioriwhich of these single-failures will result in the higher PCT. For that reason, both single-failures were analyzed in the current analysis.

3.4 Break Spectrum and Axial Shape Study Results Calculations were performed for O.4,0.6,0.8, and 1.0 DECLG breaks at a peak LHGR of 14.6 kW/ft with bounding combinations of stored energy and axial shape:

1. Bounding BOC stored energy (where maximum densification occurs) with chopped cosine axial shape.

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2. Bounding BOC stored energy (where maximum densificction occurs) combined with a bounding MOC shape peaked at a relative core height of 0.77, and

3. Bounding MOC stored energy combined with a bounding EOC shape peaked at a relative core height of 0.85.

The bounding MOC and EOC axial shapes were determined from projected axial shapes for

- equilibrium and xenon oscillation transient conditions.

Only DECLG breaks were analyzed in this analysis because split breaks have been shown to be less limiting than DECLG breaks with the SPC EXEM/PWR LBLOCA methodology (Ref. 3]. Therefore, the reported analysis bounds split breaks.

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EMF-2079 4 Millstons Unit 2 Largs Break LOCA/ECCS Analysis Rsvision O  !

With Replacement Steam Generators '

Page 3-5 The PCT results'of this break spectrum and axial shape study are shown in Table 3.5.

Note that full break spectrum and axial exposure calculations were performed for the loss- '

of-diesel single-failure only. A single loss-of-LPSI calculation (corresponding to the limiting loss-of-diesel result) was then performed. This methodology can be justified by

. considering t'he PCT results shown in Table 3.5.

Those results indicate that the 0.6 DECLG break with the MOC axial shape (and BOC stored energy) was the limiting case it can be seen in Table 3.5 that under loss-of-diesel conditions, the 0.6 MOC shape calculation produced a PCT at least 50*F higher than the 0.4 MOC shape or O.8 MOC shape breaks. Similarly, the 0.6 MOC shape break resulted in a PCT at least 90'F higher than the 0.6 BOC shape or O.6 EOC shape calculations. As typically occurs, these differences are much larger than the difference in PCT observed between the two 0.6 MOC shape single-failure calculations (10*F). If the break spectrum differences (50*F ) or the axial shape differences (90'F) had been much closer to the single failure difference (10'F) additionalloss-of-LPSI calculations would have been performed. These PCT result e very distinct, however, which make complete loss-of-l L

LPSI break spectrurn and axit ,hape calculations unnecessary. l L .

As can be seen in Table 3.5, the loss-of-LPSI single-failure was the more limiting. The PCT l for the' limiting case was calculated to be 1877 F. The hot rod results, event times, and l transient plots for the limiting case are shown in Tables 3.6 and 3.7 and Figures 3.1 through 3.20.

3.5 Exposure Study Resulhe The calculations described in Section 3.4 support exposures out to EOC. This section

addresses exposures out to end-of-Efe (EOL) with a maximum fuel assembly exposure of 56,000 mwd /MTU. The SPC methodology predicts the maximum fuel stored energy to occur near BOC where the maximum densification occurs. Closure of the fuel-cladding l* gap at higher exposures significantly reduces the fuel stored energy. Beyond exposures of about 30,000 mwd /MTU, the stored energy begins to increase due to fission gas release

. to the gap, but the stored energy at such exposures is still significantly less than the stored energy at MOC. In addition, the power level of a fuel rod at EOL is significantly lower than that of the peak-power fuel rod, resulting in a significantly lower stored energy.

i The reduced stored energy at high exposures outweighs any adverse effects of higher rod l internal pressure. Thus, the PCTs at high exposures are lower than the PCT of the limiting

. case reported for the break spectrum and axial shape study in Section 3.4 (which assumes l Siemens Power Corporation - Nuclear Divmon l

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i an MOC fuel stored energy). Therefore, a peak LHGR of 14.6 kW/ft is supported for assembly exposures of up to 56,000 mwd /MTU.

Cycle 13 is the first reload with the new fuel design; thus, approximately two-thirds of the fuelin the reactor was manufactured with the old design. A review of the maximum F, and F, for both Cycle 13 and previously installed assemblies shows that the new Cycle 13 assemblies have at least 12% higher F, and Fr values than the previously installed fuel. .

This difference in power is judged to be much more significant than differences in fuel design; therefore, it is concluded that the current analysis bounds operation with the older ,

fuel.

3.6 EOC Coastdown Operation The previous licensing calculation [Ref.1] supported EOC coastdown operation at full power with a 12"F reduction in primary coolant temperature. In the current analysis, the average fuel temperature at MOC conditions (limiting case) exceeds the average fuel temperature at EOC by more than 200*F. That difference in stored energy more than offsets any adverse effects of 12'F EOC coastdown operation, thus obviating a detailed analysis of this condition. This analysis therefore supports full power EOC coastdown operation with a 12 F primary coolant temperature reduction, i'

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l EMF-2079 Millstona Unit 2 Largs Bror.k LOCA/ECCS Analysis Revision 0 With Replacement Steam Generators Page 3-7 Table 3.1 System Analysis Parameters l Reactor Power 2700 MWi' RCS Flow Rate 1.36 x 10' lbm/hr (360,000 gpm)

RCS Volume 11,000 ft

• l RCS Pressure 2250 psia Core inlet Coolant Temperature 549 F

. Reactor Vessel Volume 4538 ft Pressurizer Total Volume 1500 ft*

Pressurizer Liquid Volume 800 ft*

SIT Total Volume 2019 ft (one of four)

SlT Liquid Volume 1150.5 ft* (one of four)

SIT Pressure . 239.5 psia SIT Fluid Temperature 106.8 F Total Number of Tubes per Steam Generator 8523 Number of Steam Generator Tubes Plugged Broken Loop 500 Intact Loop 500 Steam Generator Secondary-Side Heat Transfer Area Broken Loop 87130 ft2 Intact Loop 87130 ft:

Steam Flow Rate Broken Loop (50% of total reactor power) 6.04 x 10' Ibm /hr ~

j Intact Loop (50% of total reactor power) 6.04 x 10' Ibm /hr Steam Generator Pressure Broken Loop 878 psia intact Loop 878 psia Feedwater Temperature 435'F RCP Rated Head 271.8 ft l RCP Head (DIL) 230.4 ft

• RCP Head (SIL.BL) 233.0 ft

• RCP Rated Torque 31,560 ft-Ibv RCP Rated Speed 892.0 rpm Initial RCP Speed 866.8 rpm '

RCP Moment of inertia 100,000 lbm-ft' Containment Volume 1.938x10' f t'

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Containment Temperature 108.1 F I SIS Fluid Temperature 72.8'F HPSI Delay Time 25 s LPSI Delay Time 45 s

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- The reactor power used in the RELAP4-EM modelis 1.02 x 2700 = 2754 MWi, This volume is based on 500 plugged tubes per steam generator. Accumulators and lines are i not included.

Value(s) used in RELAP4 for initialization.

Siemens Power Corporation - Nuclear Division

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l' EMF-2079 Millstons Unit 2 Largs Brs:k LOCA/ECCS Analysis Ravision O With Replacement Steam Generators Page 3-8 Table 3.2 Fuel Design Parameters Cladding Outside Diameter 0.440 in.

Cladding inside Diameter 0.384 in.

Cladding Thickness 0.028 in.

Pellet Outside Diameter 0.377 in.

U Pellet-Cladding Diametrical Gap 0.007 in.

Pellet Density 95.0 % of Theoretical -

Active Fuel Length 136.7 in.

Table 3.3 SIS Delivery Curves for Loss-of LPSI Pump Single-Failure RCS DIL SIL BL Dll Sll BL Pressure HPSI HPSI HPSI LPSI LPSI LPSI (psia) (gpm) (gpm) (gpm) - (gpm) (gpm) (gpm) 1144.34 0.00 0.00 0.00 0.00 0.00 0.00 1100.00 122.57 61.13 61.32 0.00 0.00 0.00 1050.00 175.22 87.38 87.65 0.00 0.00 0.00 1000.00 216.94 108.18 108.52 0.00 0.00 0.00 900.00 275.49 137.38 137.81 0.00 0.00 0.00 700.00 356.28 177.66 178.22 0.00 0.00 0.00 500.00 425.78 212.31 212.98 'O.00 0.00 0.00 300.00 484.25 241.46 242.23 0.00 0.00 0.00 .

200.00 512.67 255.64 256.45 0.00 0.00 0.00 150.00 525.61 262.00 263.11 653.94 274.77 398.76 .

100.00 537.51 267.76 269.33 1155.32 512.61 664.06 50.00 549.31 273.50 275.44 1475.25 663.50 834.75 14.70 557.54 277.50 279.70 1669.90 755.14 938.85

! 0.00 557.54 277.50 279.70 1669.90 755.14 938.85 L

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Siemens Power Corporation Nuclear Division '

EMF-2079 Millstona Unit 2 Largs Brask LOCA/ECCS Analysis Rsvision O With Replacement Steam Generators Page 3-9 I

Table 3.4 Changes to the Containment Model Used in This Analysis General:

Initial Containment Pressure (psia): 14.27 Additional metal mass associated with the replacement steam generators Paint was removed from the containment heat conductors l

Fan Coolers:

Activation: 0.0 s after break initiation l RBCCW CAR Cooling Unit Energy Removal Rates (Btu /s) inlet Loss of One Diesel Generator Loss of One LPSI Pump Temperature (*F) (Two CAR Fans) (Four CAR Fans) 290 9.683E4 1.937ES 250 7.650E4 1.530E5 210 5.156E4 1.031 E5 l 180 3.311 E4 6.622E4 i 150 1.528E4 3.056E4 l 120 1.028E4 2.056E4 35 0.0 0.0 Containment Spray:

Activation: 0.0 s after break initiation Total Containment Spray Rates Loss of One Diesel Generator Loss of One LPSI Pump (One Pump) (Two Pumps) 1700 gpm (236 lbm/s) 3400 gpm (473 lbm/s) l l

l 1

1 4

Siemens Power Corporation Nuclear Division

EMF-2079 Millstona Unit 2 Largs Break LOCA/ECCS Analysis R3 vision O With Replacement Steam Generators Page 3-10 l

Table 3.5 Break Spectrum and Axial Shape Analysis PCT Results i I

l Failure Axial 0.4 DECLG 0.6 DECLG 0.8 DECLG 1.0 DECLG Loss BOC 1610 F 1696 F 1660 F 1660 F of one MOC 1628oF 1867 F 1818 F 1768 F Diesel EOC 1609*F 1757*F 1809 F 1770 F Loss BOC -

of one MOC 1877*F LPSI Pump EOC Table 3.6 Summary of Results for 0.6 DECLG Break with MOC Axial and Loss-of-LPSI Pump Hot Rod Burst Time (sec) 37.5 Elevation (ft) 8.72 Channel Blockage Fraction O.403 PCT 7

Temperature ('F) 1877 Time (sec) 55.3 Elevation (ft) 8.72 Metal-Water Reaction' Local Maximum (%) 2.80 ,

Elevation of Local Maximum (ft) 8.72 Core Maximum (%) < 1.0 l

At 450 seconds.

Siemens Power Corporation Nuclear Division

EMF-2079 Millstona Unit 2 Lergs Brsak LOCA/ECCS Antlytis Rsvision 0 With Riplicamint Ststm Generators Page 3-11 Table 3.7 Calculated Event Times for 0.6 DECLG Break with MOC Axial and Loss-of-LPSI Pump Event Time (sec)

Analysis began 0.00 Break opened 0.05 SIAS issued

. 0.8 Pressurizer emptied 9

Broken loop SIT injection began 15.1 Double intact loop SIT injection began 17.6 Single intact loop SIT injection began 17.6 Refill began (EOBY) 21.7 Reflood began (BOCREC) 35.2 Broken loop SIT discharge valve closed (SIS calculation) 52.7 Double intact loop SIT discharge valve closed (SIS calculation) 54.3 Single intact loop SIT discharge valve closed (SIS calculation) 54.4 PCT occurred 55.3 SITS emptied (RFPAC calculation) 60.1 e

l Siemens Power Corporation - Nuclear Division

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I EMF-2079 Millstona Unit 2 Largs Brask LOCA/ECCS Analysis Rsvision O With Replacement Steam Generators Page 3-12

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EMF-2079 Millstone Unit 2 Large Break LOCA/ECCS Analysis Revision O '

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EMF-2079 Millstona Unit 2 Lcrga Brs k LOCA/ECCS Antlysis Rsvision O With Replacement Steam Generators Page 3-17

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EMF.2079 Millstona Unit 2 Ltrg2 Br:ak LOCA/ECCS Analysis Rsvision 0 l With Replacement Steam Generators Page 3-20  ;

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EMF-2079 Millstone Unit 2 Large Break LOCA/ECCS Analysis Revision O With Replacement Steam Generators Page 3-21 8

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Siemens Power Corporation Nuclear Division

l EMF 2079 I Millstons Unit 2 Lcrgs Braak LOCAiECCS An: lysis Rsvision O )

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EMF-2079 Millstone Unit 2 Large Break LOCA/ECCS Analysis Revision O With R9 placement Steam Generators Page 3-23 h

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EMF-2079 Millstona Unit 2 Lirgs Bre k LOCA/ECCS Anilysis Rsvision O With Replacement Steam Generators Page 3-24 g . . . . . . , , ,

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' Siernens Power Corporation - Nuclear Divis'

EMF-2079 t Millstona Unit 2 Largs Brzak LOCA/ECCS Antlysis Rsvis;on 0 With Riplicim:nt Strim Gintrators Page 3-25 i

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EMF-2079 Millstona Unit 2 Ltrga Braak LOCA/SCCS Anilysis Rsvision O

_With Replacement Steam Generators Page 3-26

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' Siemens Power Corporation - Nuclear Division

EMF-2079 Millst:na Unit 2 Larg) Break LOCA/ECCS Antlysis Rsvision O With Replacement Steam Generators Page 3 27

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EMF-2079 ,

Millstana Unit 2 Ltrgs Braak LOCA/ECCS Analysis Rsvision 0 l With Replacement Steam Generators Page 3 28

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EMF-2079 Millstone Unit 2 Large Break LOCA/ECCS Analysis Revision 0 With Replacement Steam Generators Page 3-29

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EMF-2079 Millstona Unit 2 Ltrgs Brstk LOCA/ECCS Analysis Rsvision 0 With Replacement Steam Generators Page 3-30 m .

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EMF-2079 Millstona Unit 2 Lirg) Brick LOCA/ECCS An: lysis Ravision O With Replacement Steam Generators Page 3-31, 2000s .., , , , , , , . , -

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Figure 3.20 PCT Node and Ruptured Node Cladding Temperatures,0.6 DECLG Break, MOC Axial, Loss-of-LPSI Pump s'

G Siemens Power Corporation - Nuclear Division I

1

EMF-2079 Millstone Unit 2 Large Break LOCA/ECCS Analysis Revision O With Replacement Steam Generators Page 4-1

4. Conclusions The results of the LBLOCA analysis for Millstone Unit 2 show that a 0.6 DECLG break in conjunction with an MOC axial is the limiting case with current EXEM/PWR models.

The anatysis supports operation at a power level of 2700 MWt (plus 2% uncertainty) and a total symmetry steam generator tube plugging level of up to 500 tubes per generator.

The analysis supports an axial and exposure independent LHGR limit of 14.6 kW/ft (as

, shown in Figure 2.1). The analysis supports fuel assembly exposures of up to 56,000 mwd /MTU. The analysis supports operation during Cycle 13 and is expected to also support operation during future cycles.

Operation of Millstone Unit 2 with SPC 14x14 fuel at, or below, the LHGR limit shown in Figure 2.1 assures that the LOCA acceptance criteria (10 CFR 50.46(b)) will be met for LOCAs initiated by a spectrum of large pipe breaks up to, and including, the double-ended guillotine break of a reactor coolant pipe.

Sismens Power Corporation - Nuclear Division

EMF-2079 Millstone Unit 2 Large Break LOCA/ECCS Analysis Revision 0 With Replacement Steam Generators Page 5-1

5. Roferences

1. EMF-94-023 Rev. 3, Millstone Unit 2 Large Break LOCA /ECCS Analysis for a Revised fuelDesign, Siemens Power Corporation - Nuclear Division, May 15,1997.

2. EXEMIPWR LBLOCA Evaluation Model, as defined by the following references:

, a. XN-NF-82-20(P)(A), Revision 1, Supplement 2, Exxon Nuclear Company Evaluation ModelEXEMIPWR ECCS Model Updates, Exxon Nuclear Company, February 1985.

b. XN-NF-82-20(P)(A), Revision 1 and Supplements 1, 3, and 4, Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates, Aovanced Nuclear Fuels Corporation, January 1990.

c. XN-NF-82-07(P)(A), Revision 1, Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, Exxon Nuclear Company, November 1982.

d. XN-NF-81-58(P)(A), Revision 2 and Supplements 1 and 2, RODEX2 FuelRod Thermal-MechanicalResponse Evaluation Model, Exxon Nuclear Company, March 1984.

e. ANF-81-58(P)(A), Revision 2, Supplements 3 and 4, RODEX2 fue/ Rod ThermalMechanicalResponse Evaluation Model, Advanced Nuclear Fuels Corporation, June 1990.

f. XN-NF-85-16(P)(A), Volume 1 and Supplements 1 through 3; Volume 2, Revision 1 and Supplement 1, PWR 17x 17 Fuel Cooling Test Program, Advanced Nuclear Fuels Corporation, February 1990.

g. XN-NF-85-105(P)(A) and Supplement 1, Scal /np of FCTF Based Ref/ood Heat Transfer Correlation for Other Bundle Designs, Advanced Nuclear Fuels Corporation, January 1990.

3. ANF-88-118, Millstone Unit 2 Large Break LOCA/ECCS Analysis, Advanced Nuclear

, Fuels Corporation, August 23,1988.

I Siemens Power Corporation Nuclear Division

Millstone Unit 2 Large Break LOCA/ECCS Analysis EMF-2079 With Replacement Steam Generators Revision O Distribution Richland R. A. Shaw, 36 R. C. Gottula, 36 T. M. Howe, 38 (5)

B. D. Stitt, 36 R. l. Wescott, 33/NUSCO (5)

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