ML20198G716

Rev 0 to EMF-2145, Millstone Unit 2 Large Break Loca/Eccs Analysis with Replacement SGs & Plant Modifications

ML20198G716
Person / Time
Site:	Millstone
Issue date:	11/30/1998
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-2145, EMF-2145-R, EMF-2145-R00, NUDOCS 9812290150
Download: ML20198G716 (44)

Text

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SIEMENS EMF-2145 Revision 0 D

Millstone Unit 2 Large Break LOCA/ECCS Analysis With Replacement Steam Generators and Plant Modifications November 1998

,; ^ r 111 R A P hK hohh336 ' )\

PDR i:

Siemens Power Corporat. ion

%5h Nuclear Division

. .: . . er t; Siemens Power Corporation - Nuclear Division il/p W R__

EMF-2145 Revision O Millstone Unit 2 Large-Break LOCA/ECCS Analysis With Replacement Steam Generators and Plant Modifications Prepared: M ii [w l 96 R. A. Shaw, Engineer Date PWR Safety Analysis Contributors: T. E. Guidatti, Consultant PWR Safety Analysis R. D. Hentzen, Consultant PWR Safety Analysis T. H. Chen, Engineer PWR Safety Analysis Approved: N. C, ll/24[$

R. C. Gottula, Manager Date # '

' PWR Safety Analysis Approved:

J. S i.

olm/ Manager Il U / l[

Dat'e' Pro et Licensing Approved: (

L. E. Hansen, Mana@er

[ .2 Y f8 Date Customer Projects

dar 1

l. ,

Customer Disclaimer important Notice Regarding Contents and Use of This Document Please Read Carefully Siemens Power Corporation's warranties and representations concerning the subject matter of this document are those set forth in the agreement between Siemens Power Corporation and the Customer pursuont 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, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or

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

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

In order to avoid impairment of rights of Siemens Power Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writing by Siernens Power Corporation or until after six (6) months following termination or expiration of 4

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.

l

Millstona Unit 2 L:rg3 Br :k LOCA/ECCS An Iysis With R:pl ctm:nt Sts:m EMF-2145 Revision O G:nzrators and PI:nt Modifications Pagei Nature of Changes Paragraph item or Page(s) Description and Justification

1. All New document 1

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1 Millst:n] Unit 2 Lcrom Brack LOCA/ECCS EMF-2145 An: lysis With R;pl:c:m:nt Str m Ravision 0 Generators and Plant Modifications Page -si 1

j 1

l Contents i i

1.

I n t r o d u cti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1

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

3. Analysis........................................................................................................3-1 3.1 D e sc riptio n o f th e LB LO C A Eve nt ... .. . . ... . . . . . . . . .. . . . . . . .. .. .. ............... ... . ... . . . . 3- 1 3.2 D e s c riptio n o f An alytic 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 Expo su re Stud y R e sults . . .. ... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 5 3.6 EO C Co a std own O peratio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 5 3.7 Fu el D e sig n A ss e ssm e nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-6

4. C o n c l u si o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

5.

References....................................................................................................5-1 Tables 3.1 K e y An a ly sis Pa ra met e r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -7 l 3.2 Fu el D e sign Pa ramete rs . . . . . . . . . . . . .. . . . . .... . . ... . .. .. . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 i

3.3 SIS Delivery Curves for Loss-of-Diesel Single-Failure......................................... 3-8 3.4 SIS Delivery Curves for Loss of-LPSI Pump Single-Failure .................................. 3-9 3.5 Changes to the Containment Model Used in This Analysis............................... 3-10 3.6 Break Spectrum and Axial Shape Analysis PCT Results................................... 3-11 l

3.7 Summary of Results for 1.0 DECLG Break with EOC Axial and Lo s s-o f-Die s el Sin gle-Fa ilu re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 2 3.8 Calculated Event Times for 1.0 DECLG Break with EOC Axial and Lo ss-o f-Die sel Sin gle-Fa ilure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 3 Siemens Power Corporation - Nuclear Division

Millstone Unit 2 Lerge Break LOCA/ECCS EMF-2145 Analysis With Replacernant Stearn Revision 0 Gensrators and Plant Modifications Page lii Figures -

2.1 Allowable LHGR as a Function of Axial Location............................................... 2-1 3.1 No rm a iiz ed Po w e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3.2 S IT D is c h a rg e R a t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.3 H P S I Fl o w R a t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

i 3.4 L P S I Fl o w R a t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3.5 Upper Plenum Pressure During Blowdown ...................................................... 3 18 3.6 Total Break Flow Rate During Blowdown ....................................................... 3-19 3.7 Avera ge Core inlet Flow Rate During Blowdown............................................. 3-20 3.8 Hot Channel inlet Flow Rate During Blowdown............................................... 3-21 .

3.9 PCT Node Fluid Quality During Blowdown........................................

............. 3-22 3.10 PCT Node Fuel (Average), Cladding and Fluid Temperatures During Biowdawn..................................................................................................3-23 3.11 PCT Node Heat Transfer Coefficient During Blowdown.................................... 3-24 3.12 PCT Node Heat Flux Du ring Blowd own . .......... . . ... . . .. . . . . . .. . . . . .. . . .. . . ................... 3-2 5 3.1 3 C o n t a i n m e n t Pre s s u re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3,14 Upper Plenum Pressure Af te r Blowdown ... .............. ............ ........ . .................. 3-2 7 3.15 D own com e r C oila p s ed Liq uid Level . . . . ...... .. .. . . . . . . . . . .. . . . .. .. . . . .. . .. . . .. . . . . . . .. .. . .... .... 3 2 8 3.16 C o re E f f e etive Flo odin g Rate . . . .. . . .. . .. . . . . .. . . . . . . . . .. .. . . . .. . . . . . .. . . . . . . . . .. . .. . . .... . .......... 3 2 9 3.1 7 C o r e M i xt u re Le v e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

' 3 .1 8 C o r e Q u e n c h L e v e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3.19 PCT N od e He a t Tra n si e r Coe f ficie nt . .. ... . . ... . . . . ... . . . .. ... . . . . . .. . . . ... . . . . . .. . . ... ..... .. ... . 3 3.20 PCT Node and Ruptured Node Cladding Temperatures..................................... 3-33 Sirmens Power Corporation - Nuclear Division

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Millstons Unit 2 Large Break LOCA/ECCS Analysis With Replzcarnant Steam EMF-2145 Revision O Gensrators and Plant Modifications Page 1 1

1. Introduction This report describes an analysis of a postulated large-break loss-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 [ Reference 11, as modified by the changes described in Reference 2, was used to perform the analyses.

The analysis supports operation during Cycle 13 and future cycles, unless changes in the l 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 Cycle 13 analysis [ Reference 3].

For example, o

The SEM/PWR-98 LBLOCA methodology was used; e

New steam generators have been included in the models; e

implementation of new limiting axial power shapes for middle-of-cycle (MOC) and end-of-cycle (EOC) exposures; The resinter densification has been increased; e

Changes in Safety injection System (SIS) delivery curves; Decreased the SIS delay times: HPSI from 40 see to 25 sec, LPSI from 60 see to 45 sec; e

Revision of the initial containment pressure from 14.7 to 14.27 psia; Increased the CAR fan performance (including single-failure criterion);

e Changes in the containment spray rates from 1650 to 1900 gpm/ pump (including single-failure criterion);

e Removal of paint from the containment surfaces; e

Miscellaneous corrections in various models.

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

Si mens Power Corporation - Nuclear Division

Millstone Unit 2 Large Break LOCA/ECCS EMF-2145 Anslysis With Replacement Steam Revision O Gsnerators and Plant Modfications Page 21

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 1.0 in conjunction with an EOC axial shape and failure of a single diesel generator was found to be most limiting. The peak cladding temperature (PCT) for that limiting case was calculated to be 1814*F.

The analysis supports full-power operation at 2700 MW (plus 2% uncertainty) with a total symmetric steam generator tube plugging level of up to 500 tubes per steam generator.

The analysis supports average peak assembly exposures of up to 56,000 mwd /MTU. The analysis domonstrates that the 10 CFR 50.46(b) criteria are satisfied with an axial and exposure-independent linear heat generation rate (LHGR) limit of 15.1 kW/ft (as shown in Figure 2.1).

20 Unacceptable Operation y 15 -

g 15.1 kW/ft 5 10 a:

o 35 Acceptable Operation 0

0 0.2 0.4 0.6 0.8 1 o Fraction of Active Fuel Helgist Figure 2.1 Allowable LHGR as a Function of Axial Location Sitmens Power Corporation - Nuclear Division

Millst::na Unit 2 Largs Brisk LOCA/ECCS EMF-2145 An: lysis With RIpirc: mint Stum Rsvision O Generators and Plant Modifications Page 31

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 amo~unt of fuel cladding which reacts chemically with water or steam does not exceed 1% of the total amount of zircalloy in the core. i

3. The clariding temperature transient is terminated at a time when the core geometry is still arnenable 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, as required 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 l results of the break spectrum and axial shape analysis. Section 3.5 summarizes the results of the exposure analysis. So tica 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 O.a to 1.0, is typically analyzed. Single-ended split (SECLS) breaks are also considered in LBLOCA analyses. The limiting break location is on the pump discharge side l of a cold leg pipe.

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

The break initiates a rapid depressurization of the RCS. A reactor trip signalis issued i

when the Low Pressurizer Pressure trip setpoint is reached; however, reactor trip and scram are conservatively neglected in the analysis.

Early in the blowdown, the core experiences flow stagnation and reversal, which causes the fuel rods to pass beyond the critical heat flux. Subsequently, the fuel rods dissipate Siemens Power Corporation - Nuclear Division

Millstons Unit 2 Largi Brak LOCA/ECCS EMF-2145 An" lysis With R;;plicam:nt Stam Rsvision 0 Generators and Plant Modifications Page 3-2 heat via transition or film boiling modes of heat transfer. Any rewet that may occur during the blowdown is conservatively neglected in the analysis, as required by Appendix K of l 10 CFR 50. l l

l A Safety injection Actuation Signal (SIAS) is issued when the High Containment Pressure i 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 bvpass 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 arid, 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 transfa The reflood period begins at the BOCREC time. ECCS fluid filling the dcwns omer provides the driving head for reflooding the core. As the liquid level moves up the . e, 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 ref!ood rate. The fuel rods are cooled by radiation and, eventually, by convection as the quench front moves up the core.

3.2 Description of AnalyticalModels The EXEM/PWR Evaluation Model [ Reference 1), as modified by Reference 2, was used to perform the analysis. This model consists of the following computer codes:

1. RODEX2 for computation of the initial fuel stored energy, fission gas release, and fuel-cladding gap conductance;

2. RELAP4-EM for the system blowdown and SITISIS flow calculations:

Siemens Power Corporation - Nuclear Division

Millstone Unit 2 Large Break LOCA/ECCS Analysis With Raplacernsnt Steam EMF-2145 Revision O Generators end Plant Modifications Page 3 3

3. CONTEMPT /LT-22, as modified in accordance with NRC Branch Technical Position CSB 6-1, for computation of the containment back-pressure:

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 appropria'e 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 required by Appendix K. Appropriate conservatism specified by Appendix K was incorporated in all of the 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-tuba utsam gene.ators, 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 regions, interconnected by flow paths or " junctions." The t'uo cold legs connected to the intact loop steam generator were assumed to be symmetrical and were modeled as one intact cold leg with appropriately scaled input. The SITS, pressurizer, and steam generators (both primary and secondary sides) were modeled. The HPSI and LPSI pumps were modeled as fill junctions in the SlT lines, with conservative flows given as a function of RCS backpressure. The RCP performance curves are representative of CE-plant RCPs.

The core was modeled with an average-core assembly and a hot assembly as parallel flow channels.

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 expected to also bound the plugging levels of future cycles.)

Sismens Power Corporation - Nuclear Division

Millsto.m Unit 2 Lsrgs Braik LOCA/ECCS EMF-2145 i Anilysis With R;plictmInt Stum Rsvisior. O !

Generators and Plant Modifications Page 3-4 l Values for system parameters and fuel design parameters used in the analysis are given in Tables 3.1 and 3.2, respectively.

I Two single-failures were examined in this analysis:

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I

1. Loss of one diese! generator, and j

2. Loss of one LPSI pump in previous 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 I 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 above; and a separate calculation was performed for each of the single-failures. Tables 3.3 and 3.4 present the SIS delivery curves for both single-failure cases. Table 3.5 presents the revised containment parameters used in this analysis.

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 l

loss-of-diesel scenario. These have minor competing offects 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 fully analyzed in the current analysis.

3.4 Break Spectrum and Axlal Shape Study Results Calculations were performed for O.4,0.6,0.8, and 1.0 DECLG breaks and 0.8 and 1.0 SECLS breaks at a peak LHGR of 15.1 kW/ft with the following bounding combinations of stored energy and axial shape:

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

2. Bounding BOC stored energy (where maximum densification 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 u relative core height of 0.85.

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Siemens Power Corporation - Nuclear Division

Millstan) Unit 2 Large Brt:k LOCA/ECCS An:ly:Is With Riplicsmant Stssm EMF-2145 Rsvision 0 Ginirstors and Plant Modifications Page 3-5 The bounding MOC and EOC axial shapes were determined from projected axial shapes for  ;

1 equilibrium and xenon oscillation transient conditions. '

l The PCT results of this break spectrum and axial shape study are shown in Table 3.6. i i

Note that complete break spectruen and axial exposure calculations were performed for i both single failures. Those results indicate that the 1.0 DECLG break with the EOC axial shape (and MOC stored energy) and loss-of-diesel generator single-failure was tha limiting  !

case.

l. i As can be seen in Table 3.6, the loss-of-diesel generator single failure was the more limiting. The PCT for the limiting case was calculated to be 1814*F. The hot rod results, event times, and transient plots for the limiting case are shown in Tables 3.7 and 3.8 and Figures 3.1 through 3.20.

I 3.5 Exposum Study Results The calculations described in Section 3.4 support exposures out to EOC. This section addresses exposures out to end-of-life (EOL) with a maximum average peak assembly exposure of 56,000 mwd /MTU.

l A separate EOL calculation was performed for the limiting break spectrum case l (1.0 DECLG with the loss of one diesel generator). The EOL rod was conservatively Essumed to be at 66% of the Technical Specification hot rod limits.

That EOL calculation resulted ;n a PCT of 1362*F, or 452'F less than the overall limiting case. Therefore, a peak LHGR of 15.1 kW/ft is supported for average peak rod exposures of up to 56,000 mwd /MTU. {

r*. l 3.6 EOC Constdown Operation -

l l

A separate calculation was performed, using the limiting case determined in the break spectrum and axial shape calculations, to assess EOC avera0e temperature coastdown cperation. in that calculation, full-power operation with 12*F reduction in primary coolant temperature was analyzed. That analysis resulted in a PCT of 1758'F, or 56'F less than the overall limiting case.

i This analysis therefore supports full-power EOC coastdown operation with a up to 12'F primary coolant temperature reduction.

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l Siemens Power Corporation - Nuclear Division

l Millston) Unit 2 L rga Brzk LOCA/ECCS EMF-2145 AnIlysis With R;placsmint Steam R1 vision 0

! Generators and Plant Modifications Page 3-6 3.7 FuelDesign Assessment ~

l Cycle 13 is the first reload with the new fuel design; thus, approximately two-thirds of the fuel in 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 F, 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.

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. Siemens Power Corporation - Nuclear Division

I Mills'ons Unit 2 Lirgs Br:sk LOCA/ECCS Anilysis With Ripl:camInt Stssm EMF-2145 Rsvision O

, Generators and Plant Modifications Page 3-7 Table 3.1 Key Analysis Parameters Reactor Power 2700 MWi '

RCS Flow Rate 1.36 x los Ibm /hr (360,000 gpm)

RCS Volume 11,000 ft3 '

RCS Pressure 2250 psia Core inlet Coolant Temperature 549*F

! Reactor Ves.el Volume 4538 ft' l Pressurizer Total Volume 1500 ft' l Pressurizer Liquid Volume 800 ft' l- SIT Total Volume 2019 ft3 (one of four) l SIT Liquid Volume 1150.5 ft3 (one of four) l SlT Pressure 238.5 psia' SlT 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 ft2 l Steam Flow Rate Broken Loop (50% of total reactor power) 6.04 x 10* lbm/hr intact Loop (50% of total reactor powel)_,,,

,,, 6.04 x 10* lbm/hr Steam Generator Pressure Broken Loop 878 psia Intact Loop 878 psia Feedwater Temperature 435 F RCP Rated Head 271.8 ft RCP Head (DIL)

• 230.4 ft

• RCP Head (SIL,BL)

• 233.0 ft

• RCP Rated Torque 31,560 ft-Ibi RCP Rated Speed 892.0 rpm l initial RCP Speed 866.8 rpm '

RCP Moment of inertia 100,000 lbm-f t*

~

Containment Volume 1.938x10' f t' Containment Temperature 101.6 *F SIS Fluid Temperature 72.B

• F HPSI Delay Time 25 s LPSI Delay T me 45 s

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

• Transcription error correction (239.5 psia previously). No impact on final results.

DIL e double intact loop; SIL = single intact loop; BL e broken loop.

Value(s) used in RELAP4 for initialization.

I I-1 Siemens Power Corporation Nuclear Division

i Millstona Unit 2 Largs Brask LOCA/ECCS _ EMF-2145 Antlysis With RiplicsmInt Stsam Rsvision 0 Generators and Plant Modifications Page 3-8 i

Table 3.2 Fuel Design Parameters i

Cladding Outside Diameter O.440 in.

Cladding Inside Diameter O.334 in.

t- Cladding Thickness 0.028 in. '

Pellet Outside Diameter 0.377 in.

Pellet-Cladding Diametrical Gap O.007 in. -

Pellet Density 95.0 % of Theoretical Active Fuel Length 136.7 in. -

i Table 3.3 SIS Delivery Curves for Loss-of-Diesel Single-Failure' l-  ;

RCS Dll Sll BL DIL SIL 15L 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 i 1100.00 61.38 30.69 30.69 0.00 0.00- 0.00 1050.00 87.75 43.88 43.88 0.00 0.00 0.00 1000.00 108.65 54.33 54.33 0.00 0.00 0.00 900.00 137.89 68.95 68.95 0.00 0.00 0.00 700.00 178.35 89.18 89.18 0.00 0.00 0.00 L 500.00 213.13 106.57 106.57 0.00 0.00 0.00 i 300.00 243.43 121.22- 121.22 0.00 0.00 0.00 .,

l

-200.00 256.67- 128.34 128.34 0.00 0.00 0.00 1 4

150.00 263.57 131.31 131.59 0.00 498.35 605.38

~

100.00 270.04 -133.80 134.54 0.00 881.92 103003_

! 50.00 276.41 136.26' 137.44 0.00 1156.35 1349.tl4 14.70 280.83 138.01 139.47 0.00 1310.58 1524.6L L O.00 280.83 138.01- 139.47 0.00 1310.58 1524.6L SIS delivery to specific loops was chosen to ensure conservative results and thus does not reflect the actual plant cold leg / SIS train arrangement. For example, the larger of the two LPSI

flows under loss-of-diesel conditions was directed to the broken loop. The model is insensitive l . to intact loop / SIS train assignments.

l Siemens Power Corporation Nuclear Division

-. ~ -- , - _ . , -

Millst:ro Unit 2 Lsrgs Brask LOCA/ECCS Anilysis With R;plictmrnt Sts:m EMF-2145 Rsvision 0 GIntrstcrs and PIInt Modifications Page 3-9 Table 3.4 SIS Delivery Curves for Loss-of-LPSI Single-Failure

• RCS Dll SIL BL DIL 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 0.00 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 SIS delivery to specific loops was chosen to ensure conservative results and thus does not reflect the actual plant cold leg / SIS train arrangement. For example, the largest of the LPSI flows under loss-of-LPSI conditions was directed to the broken loop. The model is insensitivt, to /ntact loop / SIS train assignments.

Siemens Power Corporation - Nuclear Division

Millstona Unit 2 Largs Brsak LOCA/ECCS EMF-2145 Anzlysis With Ripiscamsnt Stsam Rsvision O Generators and Plant Modifications Page 310 1

Table 3.5 Changes to the Containment Model Used in This Analysis General:

Initial Containment Pressure (psia): l 14.27  !

Additional metal mass associated with the replacement steam generators Paint was removed from the containment heat conductors Fan Coolers:

Activation: 0.0 s after break initiation

~

Containment CAR Cooling Unit Energy Removal Rates (Btu /s)

Air-Steam Mixture 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 ES 180 3.311 E4 6.622E4 150 1.528E4 3.056E4 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) 1900 gpm (264 lbm/s) 3800 gpm (528 lbm/s)

I

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Siemens Power Corporation - Nuclear Division

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Millstona Unit 2 Lcrgs Brask LOCA/ECCS An: lysis With Riplacimint Staim EMF-2145 Ravision O GIntrstors and Plant Modifications Page 3-11 i

Table 3.6 Break Spectrum and Axial Shape Analysis PCT Results ' l

, Break Break Axial Single-Failure PCT

! Configuration Co or Size Shape (*F)

BOC 1662

,. 0.4 MOC 1711 EOC 1722 BOC- 1664 0.6 MOC 1759 l DECLG EOC 1770 BOC 1694 Loss of 0.8 MOC 1784

one LPSl EOC 1806 pump BOC 1688 1.0 MOC 1786 EOC 1811 BOC 1606 0.8 MOC 1669 SECLS EOC 1703 BOC 1630 1.0 MOC 1731 EOC 1755 BOC 1661 0.4 MOC 1703 i EOC 1713 BOC 1662 0.6 MOC 1753 DECLG EOC 1766 l

BOC 1692

. Loss of 0.8 MOC 1782 ons diesel EOC 1804 g:nerator BOC 1686 1.0 MOC 1785 EOC 1814 BOC 1605 0.8 MOC 1639 SECLS EOC 1713 BOC 1629 1.0 MOC 1732 EOC 1759 l

Siemens Power Corporation - Nuclear Division i

Millstona Unit 2 Lsrga Brsak LOCA/ECCS EMF 2145 An: lysis With R pitcImsnt St2am Rsvision O Generators and Plant Modifications Page 3-12 i

-Table 3.7 Summary of Results for 1.0 DECLG Break with EOC Axial-i and Loss of Diesel Single-Failure l

i  !

l Hot Rod Burst Time (sec) 43.0 l

Elevation (ft) 9.77 Channel Blockage Fraction O.46 PCT -

Temperature (*F) 1814 Time (sec) 135.8 Elevation (ft) 10.52

! Metal Water Reaction' Local Maximum (%) 2.36 Elevation of Local Maximum (ft) 10.52 Core Maximum (%) < 1.0 l l

At 450 seconds.

Siemens Pos r Corporation - Nuclear Division l

Millstona Unit 2 Largs Brzek LOCA/ECCS EMF-2145 Antlysis With R:plactmInt Stacm- l GIntr:; tors End PIInt Modifications Ravision O 4 Page 313 Table 3.8 Calculated Event Times for 1.0 DECLG Break with EOC Axial and Loss-of-Diesel Single-Failure Ever.t Time (sec)

Analysis began 0.00 Break opened 0.05 SIAS issued 0.7 Broken loop SIT injection began 9.5 Double intact loop SlT injection began 14.9 Single intact loop SlT injection began 14.9 Refill began (EOBY) 18.1 Reflood began (BOCREC) 31.3 Fuel rupture occurred 43.0 Broken loop SIT emptied 48.1 Double intact loop SIT emptied 51.3 Single intact loop SIT emptied 51.9 PCT occurred

, 135.9 ,

Siemens Power Corporation Nuclear Division

Millstrna Unit 2 L:rgs Brisk LOCA/ECCS EMF-2145 ,

An:lyris With RIpt:cimInt Stz:m Rsvision O  !

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Millst:na Unit 2 L:rgi Brrk LOCA/ECCS EMF-2145 Analysis With RiplicimInt Stum Rsvisi n O Generators and Plant Modifications Pege 315 i

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Millst:n3 Unit 2 Lcrgs Brick LOCA/ECCS EMF-2145 AnIlysis With Riplictm:nt Stram Rsvislan 0 Generators and Plant Modifications Page 3-18 1

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Figure 3.7 Average Core inlet Flow Rate During Blowdown 1

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Millst:ne Unit 2 Larga Bre:k LOCA/ECCS l An: lysis With Riplic:m:nt St:Im EMF-2145 Rwision O l Gin:rators and Plant Modifications Page 3-21 l

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Figure 3.8 Hot Channelinlet Flow Rate During Blowdown l

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Millstons Unit 2 Ltrge Brick LOCA/ECCS EMF-2145 Analysis With R;plic:m:nt Stscm RIvision O Generators and Plant Modifications Page 3-22 l

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Figure 3.9 PCT Node Fluid Quality During Blowdown l

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Figure 3.11 PCT Node Heat Transfer Coefficient During Blowdown l

Siemens Power Corporation Nuclear Division

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! Millst:n3 Unit 2 L:rga Brnk LOCA/ECCS EMF-2145

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Figure 3.12 PCT Node Heat Flux During Blowdown 1

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l Millstuna Unit 2 Larg) Brnk LOCA/ECCS EMF-2145 Analysis With R pl c:m:nt Stiam Revisi:n 0 Generators and Plant Modifications Page 3 26 t

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1 Millstrne Unit 2 krgs Brnk LOCA/ECCS Anilysis With R:plictmInt St:Im EMF 2145 ,

G:nirct:rs cnd Plant Modifications Rwisi:n 0 '

Page 3-27 1

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Siemens Power Corporation - Nuclear Division

Millst:n3 Unit 2 L rga Brz k LOCA/ECCS EMF-2145 An: lysis With R:pi:ctmInt StI m Rsvision 0 gin:r: tors cnd Plant Modifications Page 3 28 e

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Millstona Unit 2 Larg: Br:ck LOCA/ECCS EMF-2145 An" lysis With R plicIm:nt Stiam RIvision O GInIrrt:rs cnd Plant Modifications Page 3-30 8.0 . . . ,

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~ Siemens Power Corporation - Nuclear Division

Millst:n2 Unit 2 L:rgs Bruk LOCA/ECCS EMF-2145 An:ly:Is With RiplictmInt StItm RIvision O Gin r:t:rs and Pl:nt Modifications Page 3-31 12.0 . . . , , . . . . . .

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Siemens Power Corporation - Nuclear Division

. Millstona Unit 2 Lsrgs Braak LOCA/ECCS l Anslysis With R:plicsmInt Stasm EMF-2145 GIntrztors cnd Plint Modifications Rsvision 0 Page 41

4. Conclusions The results of the LBLOCA analysis for Millstone Unit 2 show that a 1.0 DECLG break in conjunction with an EOC axial shape, MOC stored energy, and loss of one diesel generator is the limiting case.

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

l, The analysis supports an axial and exposure independent LHGR limit of 15.1 kW/ft. 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. The analysis also supports average temperature coastdown of up to 12'F at full-power conditions.

Operation of Millstone Unit 2 with SPC 14x14 fuel at, or below, the 15.1 kW/ft LHGR limit assures that the LOCA acceptance criteria (10 CFR 50.46(b)) will be met for LOCAs initiated by a spectrum of large pipe breaks uo to, and including, the double-ended i

guillotine break of a reactor coolant pipe.

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Millstona Unit 2 Largs Brsek LOCA/ECCS EMF-2145

! Anllysis With RiplIcsmint Stsam Rsvision O Ginirstors and Plant Modifications Page 5-1

5. References i

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

i a.

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

t.

l b. XN-NF-82-20(P)(A), Revision 1 and Supplements 1, 3, and 4, Exxon Nuclear l Company Evaluation Model EXEM/PWR ECCS Mode! Updates, Advanced  ;

Nuclear Fuels Corporation, January 1990. 1

c. XN-NF-82-07(P)(A), Revision 1, Exxon Nuclear Company ECCS Cladding i

Swelling and Rugture Model, Exxon Nuclear Company, November 1982. '

I

d. XN-NF-81-58(P)(A), Revision 2 and Supplements 1 and 2, RODEX2 Fue/ Rod Thermal-MechanicalResponse Evaluation Model, Exxon Nuclear Company, i 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, Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs, Advanced Nuclear Fuels Corporation, January 1990.

2. EMF-2087(P), Revision 0, SEM/PWR-98: ECCS Evaluation Modelfor PWR LBLOCA Applications, Siemens Power Corporation, August 1998.

3. EMF-94-023, Revision 3, Millstone Unit 2 Large Break LOCA/ECCS Analysis for a Revised fuelDesign, Siemens Power Ccrporation, May 15,1997.

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Siemens Power Corporation Nuclear Division

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Richland R. A. Shaw, 36 l- R. C. Gottula, 36 B. D. Stitt, 36 R. l. Wescott, 33/NUSCO (10)

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l' Siemens Power Corporation - Nuclear Division