ML20055G680

Yankee Plant Small Break LOCA Analysis

ML20055G680
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Site:	Yankee Rowe
Issue date:	07/31/1990
From:	YANKEE ATOMIC ELECTRIC CO.
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NUDOCS 9007240065
Download: ML20055G680 (39)

Text

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1lI Yankee Plant Small Break LOCA Analysis i

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'I July 1990 Major Contributors:

S. Mihalu-Westerlind R. C. Harvey R. K. Sundaram ll I

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I lb 4 O M 'WedeVNWd 9 \\ \\%k SD Prepared By:

S. Mihaiu-Westerlind Nuclear Engineer (Date)

LOCA Analysis Group M Ib I

Prepared By:

R. C. Harvey, ' Senior [/ Engineer

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/(Date)

LOCA Analysis Group

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Prepared By:

R. K. Sundaram, Manager (bate)

LOCA Analysis Group Approved By:

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7//E/IO B. C. Slifer irectorf' f

(Date)

I Nuclear Engi ering Department I

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I Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 I

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J DISCLAIMER OF RES10NSIBILITY This document was prepared by Yaikee Atomic Electric Company

(" Yankee"). The use of information cortained in this document by anyone other r

then Yankee, or the Organization for waich this document was prepared under L

contract, is not authorized and, nith_utsyg.ct to any unauthorized ung, neither Yankee nor its officers, directors, ar,ents, or employees assume any obligation, responsibility, or liability or make any warranty or

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representation as to the accuracy or completeness of the material contained in this document.

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l I ABSTRACT I

This report summarizes Small Break Loss-of-Coolant Accident (SBLOCA) evaluations for the Yankee Nuclear Power Station, carried out to address TM1 Action Item II.K.3.31.

The analysis shows that the limiting small break LOCA for the Yankee plant is an 8-1/2" break in the cold leg at the Emergency Core Cooling (ECC) piping junction, with direct ECC spillage to the containment.

The results for this case are well within the 100FR50.46 acceptance criteria.

These evaluations also show that for the Yankee Nuclear Power Station, the I

licensing basis for ECCS performance continues to be determined by the large break LOCA analysis.

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J ACKNOWLEDGEMENTS The completion of this analysis was made possible by additional support J

and cooperation from several individuals.

Liliane Schor provided valuable L

technical insights in the RELAP5YA model development.

Pedro Perez developed the core power shape input data and reviewed the ECC performance calculations. The word processing center at Yankee Atomic Electric Company provided timely assistance in completion of the report.

We gratefully acknowledge the contribution of all of these individuals.

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IABLE'0F CONTENTS EA&R i

1 i

i D I S CINMR O F RES PON S I B I L ITY......................................

iii ABSTRACT..........................................................

iv ACKN OWLEDG F.M ENTS..................................................

V i

LIST OF TABLES....................................................

vi LIST OF FIGURES...................................................

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1.0 INTRODUCTION

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2.0 SBLOCA EVALUATION M0 DEL...........................................

4 2.1 Model teatures..............................................

2.2 SBLOCA Model Compliance With USNRC Guidelines...............

3.0 SBLOCA ANALYSIS RESULTS...........................................

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3.1 Analysis scope............................................ee 3.2 Phenomena Description.......................................

3.3 Analysis Resu1tc............................................

4.0 CONCLUSION

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i-1 LIST OF TABLES I

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Ninnhe r T$ tie lagt a

W 2.1 Initial Steady-State Operating Conditions 8

2.2 Code Input Options 9

2.3 Sunnary of Metal Heat Slab Representation 10 3.1 Analysis Assumptions for Yankee Plant SBLOCA Cases 20 3.2 Sequence of Events for Yankee Plant SBLOCA Cases 21 I

3.3 Yankee Plant SBLOCA Analysis Results 22 I

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1 LIST OF FIGURES Humber Title Eagm Yankee Plant SBLOCA Nodalization Diagram 11 2-1 i

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2-2 Yankee Plant SBLOCA Core Representation 12 2-3 Core Axial Power Shape 13 3-1 Primary and Secondary Side Pressure for Limiting SBLOCA Case 23 3-2 Break Liquid Fraction for Limiting SBLOCA Case 24 I

3-3 ECCS Mass Flow Rate for Limiting SBLOCA Case 25

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3-4 Break Mass Flow Rate for Limiting SBLOCA Case 26

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3-5 Reactor Vessel Mass Inventory for Limiting SBLOCA Case 27 3-6 Peak Clad Temperatures for Limiting SBLOCA Case 28 I

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I 1.0 INIROMCIl0B The analysis of postulated Loss-of-Coolant Accidents (LOCA) is part of

!I the safety analyses that determine the licensing basis for commercial U.S.

nuclear power plants.

Title 10 of the Code of Federal Regulations, Part 50, Article 46 (hereby referred to as 100FR50.46), stipulates acceptance criteria for LOCA analyses and also describes requirements for acceptable analysis methods.

Yankee Atomic Electric Company (YAEC) has used dCA analysis methods that conform to 10CFR50.46 requirements and has performed LOCA analyses for the Yankee Nuclear Power Station (YNPS) since 1977. These analyses I

investigated a spectrum of break sizes and locations and showed that the limiting LOCA event for the Yankee plant would be a large break in the cold les piping (large break LOCA or LBLOCA). Thus, the LBLOCA analyses have formed the licensing basis for evaluating performance of the Emergency Core Cooling Systems (ECCS) at the Yankee plant.

Subsequent to the Three Mile Island (MI) accident in 1979, the United I

l States Nuclear Regulatory Commission (NRC) developed additional requirements for the analysis of Small Break LOCA (SBLOCA) scenarios.

Reference 1 describes these requirements and contains a set of action items to which licensees are required to conform.

Specifically, MI Action Item II.K.3.30 requires licensees to improve analysis methods to more accurately simulate SBLOCA phenomena, and W I Action Item II.K.3.31 requires application of those methods to perform plant-specific SBLOCA analyses.

YAEC developed the RELAP5YA computer program (Reference 2) as an I

evaluation model conforming to the requirements of 10CFR50.46 and W I Action Item II.K.3.30.

The NRC approved the RELAP5YA Program for Pressurized Water Reactor (PWR) SBLOCA analysis (Reference 3) with some guidelines for application of the method (Enclosure (1) of Reference 3).

The RELAPSYA method has been applied to perform SBLOCA analysis for the Yankee plant in

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conformance with these guidelines.

The analysis addresses the regulatory requirements of TMI Action Item II.K.3.31.

This report presents a summary of the analysis.

L The scope of the analysis described in this report has been transmitted to the NRC (Reference 4) and consists of a break spectrum analysis at the limiting break location.

The limiting SBLOCA break location for the Yankee plant is in the cold leg near the connection of the ECC pipe, resulting in direct spillage of ECC fluid to the containment (References 5, 6, and 7).

Section 2.0 of this report describes details of the RELAP5YA model of the Yankee plant and contains a summary of conformance to regulatory requirements.

Section 3.0 presents the analysis results with a detailed description of the limiting break size case.

Conclusions from the analysis are presented in Section 4.0. WPP41/17

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2.0 SBLQCA EVALUATION MODEL This section describes the main features of the RELAP5YA model and its j

compliance with the specific guidelines outlined in Enclosure (1) of L

Reference 3.

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2.1 tiodel_fsatutes The Yankee plant system nodalization was selected to be consistent with that used to assess RELAP5YA against experimental data. A plant nodalization diagram is shown in Figure 2-1.

The major features of the model are 1.

Reactor core region:

Hydraulically, the core is represented as a single channel divided into six axial nodes.

The fuel rods in the core are represented as two RELAP5YA heat structures.

One heat structure represents a fuel rod at average core power and is divided into six axial nodes. The other heat structure represents 1

the hot rod and is divided into 20 axial nodes.

The nodes in the hot rod are at least 3" in length as required by regulatory requirements for calculation of metal-water reaction rates.

Figure 2-2 shows the core representation.

2.

Broken Loops one broken primary coolant loop and one secondary side associated with the broken primary loop consisting of the hot leg, steam generator tubes, cold leg and reactor coolant pump, steam generator, normal feedwater system, main steam line, nonreturn valve, and safety valves. The steam generator tubes contain eight axial nodes, at required by guidelines contained in Enclosure (1) of Reference 3.

3.

Intact Loopt three intact primary coolant loops and the associated three secondary side systems are combined into one equivalent primary / secondary system consisting of the same components and

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systems as in the broken loop.

In addition, the pressuriser and the surge line are attached to the intact primary coolant loop.

I 4.

ECCS: The ECCS is modeled as a time-dependent junction with the injection flow as a function of the primary system pressure.

The ECC flow rate was determined from a separate RZ1AP5YA model of the Safety Injection System. The ECCS water temperature is assumed to be 200'F in compliance with Reference 3.

I The worst active single failure is the failure of a diesel generator, resulting in the loss of one entire ECCS train. This results in the minimum inventory replacement during the SBLOCA I

event.

Loss of off-site power is also assumed in conformance with regulatory requirements, and this results in a delay in the start time of the ECCS pumps.

5.

Core power:

It is assumed that the average core and the hot rod fuel regions have fuel rod thermal properties characteristic of Core 20 fresh fuel.

This assumption assures maximum stored energy conditions for the break spectrum analyzed. A conservative axial power shape was derived to bound peak axial locations determined I

- the large break LOCA analysis.

Figure 2-3 presents the axial pe, 2 factor versus the axial location cf the average core and hot rod heat structures. The Peak Linear Heat Generation Rate (PLHCR) for the hot rod was determined to be 13.36 kW/ft.

It is assumed that the reactor has been operating continuously at a power level of 1.03 times the licensed power level-to allow for instrumentation error uncertainties, per regulatory requirements.

6.

Initial conditions: The initial conditions selected for the SBLOCA analysis were intended to bound the expected Yankee plant operating conditions.

In general, values were chosen to be conservative with respect to detemination of the PCT.

(The principal initial conditions are summarized in Table 2.1.)

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s 2.2 SEQCA Model Compliance With USNRC Cuidelinga J

The following section discusses the SBLOCA model compliance with the 12 3

specific conditions stipulated in Enclosure (1) of Reference 3.

L 1.

The REIAP5YA PWR SBLOCA computer program applicability is

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restricted to analyzing break sizes up to 0.7 ft2 The largest break size analyzed for the Yankee plant is a 10" equivalent 2

diameter break corresponding to a break area of 0.545 f t,

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2.

REIAP5YA is restricted to analyzing SBLOCAs where the accumulators do not empty to limit the noncondensible gas in the primary system.

The Yankee plant design isolates the accumulator before the entire volume of water is injected. Therefore, the noncondensible gas from the accumulator is not injected into the primary system.

3.

The condensation rate during refill of the system high points g

(i.e., pressurizer) was found to be overpredicted by the RELAP5YA l

code. Therefore, code calculations are restricted to the time periods before this refill phase.

For all cases analyzed, the peak clad temperature is reached well before any of the system high points are refilled.

I 4.

A requirement of the plant-specific model was to have enough detail to account for vapor superheat. The Yankee plant core model is represented by six axial fluid volumes.

For the Yankee plant core height of 91", this nodalization provides adequate detail. Vapor I

superheat was calculated to occur for two-phase mixture levels I

below the top of the core.

5.

To avoid overprediction of the condensation rate, when subcooled ECC is injected, an acceptable guideline is to model the ECC water temperature t 200'F.

For all cases analyzed, the ECCS water is modeled with a temperature of 200*F. WPP41/17 1

i I 6.

Certain RELAP5YA code modifications were determined to be I

acceptable with some restrictions applied. The Yankee plant analyses adhere to these restrictions. More specifically, the Moody critical flow option uses the donor cell static pressure and 1

enthalpy as input to the critical flow table; the multiple surface radiation model is not used, and the fuel behavior model uses a temperature offset which is no less than 2'F.

I 7.

The input option selected in the RELAP5YA calculations are

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presented in Table 2.2.

The options chosen are consistent with those used in the code assessment analyses, selected to produce conservative calculations, or determined to be acceptable in I

Reference 3.

8.

A requirement of the plant-specific submittal is to address the following items:

time stip selection; break size sensitivitiest metal heat slab representation; multiple surface radiation modelingt justification for any nonzero direct moderator heating fraction; and pressurizer nodalization. The time ti.eps selected in I

the Yankee plant calculation are generally smaller than those used in the code assessment analyses, therefore, sensitivitiet studies I

to evaluate the impact of small time steps are not warranted.

Dreak sensitivities have been performed and the results are presented in Section 3.3.

A list of the metal heat slabs r

represented and neglected is presented in Table 2.3.

All structures of significant mass are represented in the model consistent with previous anslyses (References 5, 6, and 7).

Surface-to-surface radiation heat transfer is neglected in the I.

Yankee plant analysh. The direct moderator heating fraction is set equal to zero. The pressuriser nodalization used (Figure 2-1) follows YAEC modeling guidelines, as is similar to nodalization used in the code assessment calculations.

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9.

The Yankee plant steam generator U-tubes are represented by eight fluid nodes. This nodalisation is adequate to predict s

potential low flow and reflux cooling conditions. A more detailed nodalization would not alter the PCT results significantly.

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10. The nodalization used in the Yankee plant analyses (Figure 2-1) is consistent with current modeling guidelines and those used in assessment calculations. The core region is represented by i

six fluid volumes (see Condition 4) and is detailed enough to predict vapor superheat.

11. The following models include known errors and were not used in the 1

Yankee plant analysis control variables as power into a heat structuret reactor kinetics with no power from gamma heating; valve components with iorm loss coefficientsI and control variables with reactor kinetics feedback.

12. A requirement, based on integral code assessment calculations, is to perform a plant-specific break spectrum study. A Yankee plant t

break size sensitivity study was performed and the results of this study are presented in Section 3.3.

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l TARLE 2.1 Initial Sleady-State Ooeratine Conditions t

Reactor Power (MWt) 618 l

Pressuriser Pressure (psia) 2.090 Pressuriser Level (inches) 120 l

Cold Leg Temperature ('F) 524.6

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Hot Leg Temperature ('F) 568.6 Total Loop Flow (4 Loops) (1bm/sec) 10.687.6 Secondary Pressure (psia) 533 3

Steam Generator Liquid Volume per SG (f t )

343 Steam Generator Plugged Tubes (%)

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Accumulator Pressure (psia) 477.7 3

Accumulator Water Volume. Usable (ft )

700 I

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TABLE 2.2

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Cgje Input Options 1.

All fluid volumes utilize the nonequilibrium option.

2.

Wall friction is computed for all fluid volumes.

3.

Smooth area change is specified at all junctions except valves which require abrupt area change.

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All junctions use the two-velocity option with full inertia treatment.

4.

5.

Passive heat structures exposed to the containment environment use the insulated baundary condition on one side and the convective boundary condition on the other (ambient heat losses neglected).

6.

Thermal properties for all structures are input in table format.

7.

All heat structures use the modified Biasi and the Griffith-Zuber Critical llent Flux (CHF) correlations.

8.

Core heat structures use the fuel behavior option with use of the code internal tables for fuel rod rupture and blockage.

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9.

Rewet-quench model is used for the core heat structures.

10.

Direct moderator heating fraction is set equal to 0.0 for all core structures.

11.

Radiation heat transfer (surface-to-surface) is neglected.

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TABLE 2.3 I'

EuKaary_ of Metal llant Slab Reptagantation A.

titial_ Ital Slabs. Represented in the Yankee P1&Rt SBLOCA Model Reactor Vessel

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1.

Reactor Vessel Wall 2.

Thermal Shield 3.

Core Barrel

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4.

Core Baffle L

5.

Core Support Barrels 6.

Core Support Plates

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Primary coolant Loops 1.

Loop Piping 3

2.

Loop Isolation Valves 3.

Loop Check Valve 4.

Reactor Coolant Pump 5.

Steam Generator Tubes 6.

Steam Generator inlet and Outlet Plena 7.

Pressuriser Surge Line 8.

Pressuriser Vessel Wall Steam Centrators (Secondary side) 1.

Tube Sheet 2.

Vessel Wall 3.

Separator Components B.

titial llent Slabs Not Represented in the SBLOCA Model 1.

Reactor Vessel Upper Head interncis 2.

Reactor Vessel Lower Plenum Internals 3.

Reactor Vessel Guide Tubes and Support 4.

Control Blades 5.

Inutive Fuel Structure Components 6.

i Steam Lines 7.

Steam Generator Baffle I !

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3.0 SBLOCA ANALYSIS RESULTS

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Tnis section. describes the results of the ?BLOCA analysis performed for the Yankee plant.

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3.1 Analysis Scope The Yankee plant SBLOCA analysis approach was transmitted to the NRC in Reference 4.

Following this approach. YAEC has utilized the RELAPSYA method to reinvestigate the limiting SBLOCA scenarios identified by prior licensing analyses of the Yankee plant (References 5. 6. and 7).

The limiting small break location for the Yankee plant-is in the safety injection line near or at the connection with the MCS cold leg. This break location results in direct spillage of ECC fluid to the containment as follows:

1.

Breaks in a small length of ECCS piping near the cold leg injection point result in primary coolant blowdown through the 2.25" Inside Diameter (ID) thermal sleeve and ECC spillage through ths ECCS piping. This results in the maximum spillage area, but the break size is limited to 2.25".

2.

Breaks at the connection of the safety injection lino and the cold leg can result in break sizes larger than 2.25". but the ECC spillage area is limited by the flow area of the thereal sleeve (2.25" ID).

Breaks in these categories are the most limiting because of 'the_high ECC spill rate and the lower ECCS header pressure which delays ECC injection into the primary coolant system.

l The scope of the analysis reported here is to reinvestigate the above cases with the RELAP5YA evaluation model.

This consists of (a) break size of 2.25" combined with ECC spillage through the ECCS piping and (b) a break spectrum of break sizes larger than 2.25" combined with ECC spillage through

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I the 2.25" ID thermal sleeve. These analyses assume, as in previous analyses, g

W that the worst single failure is the failure of a dieael generator with attendant failure of one-train of ECCS.

3.2 Ehnnomens Description

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Before presenting the results of the RELAPSYA analysis, the sequence of events in SBLOCA scenarios is sununarized, highlighting features of the Yankee plant.

For a typical modern Westinghouse PWR, small break LOCA phenomena have E

m-Based on these investigations, the sequence of events following a small cold leg break can bo considered in the following four phases:

(1) after a break occurs in the cold leg, the primary system depressurizes to near the secondary pressure. During this phase, thc break discharge flow changes from subcooled f

liquid to two-phase conditions, the reactor power decays rapidly, the main

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coolant _ pumps coast down (assuming they are tripped), the secondary pressure increases due to its isolation (loss of off-site power assumed) and the upper I

l vessel regions begin voiding.

(2) The primary and secondary pressures remain l

relatively constant while core decay energy is-primarily removed by the I

two-phase break discharge flow. During this phase, for smaller breaks (less than 10% of cold leg pipe area), liquid can be held up in the steam generator tubes preventing vapor venting. This causes a manometric core level depression and heatup of the upper core regions.

Pressure imbalances within the loops caused by continuing vapor production in the core eventually clear-the steam generator tubes and the loop seals free of'11guid. This leads to core level recovery up to the loop' seal elevation and arrests core heatup.

For larger break sizes, these phenomena are not encountered because the energy removal rate at the break exceeds the core decay energy. Tais higher break I

flow causes a continuous and steady depressurization of toe primary with continuous flashing of liquid into vapor uniformly within the system. Thin, j

phr.se ends when the brer.k flow changes f rom two-phase to mostly vapor.

(3) Due to predominantly vapor flow at the break, the primary pressure begins to decrease more rapidly. During this phase, the loops are clear and all the liquid in the system.is in the vessel, below the cold. leg elevation. As the I I WPP41/17 i

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primary pressure decreases, liquid in the core boils off causing core heatup.

This phase ends when the primary pressure drops to the accumulator setpoint causing injection of accumulator water.

(4) The accumulator water recovers I1 core level and arrests core heatup. During this phase, the primary pressure continues to drop and the low pressure ECC pumps begin injecting.

Eventually, the accumulators are emptied, but low pressure pumped injection provides for long-term cooling.

I The above sequence of events is generally observed in the RELAP5YA small break analysis of the Yankee plant, but with some deviations due to unique plant-specific features. One such feature is that the main coolant pumps at the Yankee plant are canned rotor pumps with very little inertia.

In I

the RELAP5YA analyses, the pumps are assumed to trip immediately after the break occurs. Due to the low inertia, the pumps coast down rapidly. This can cause an early core heatup due to Departure from Nucleate Boiling (DNB). The heatup is terminated as liquid is redistributed within the primary system and two-phase flow is established in the core.

Another unique feature of the Yankee plant is its ECCS configuration.

The ECC pumps are piggy-backed with the Low Pressure Safety Injection (LPSI) pumps feeding the High Pressure Safety Injection (HPSI) pumps, and there is I

one accumulator which is connected to all four loops.

The accumulator is fed by nitrogen bottles designed to maintain a constant accumulator pressure.

It is isolated on a low-level signal to prevent nitrogen injection into the system. The order of ECC injection, for nonspillage cases, is HPSI boosted by LPSI at 1,550 psia, LPSI through the low pressure header at 690 psia, followed by the accumulator (477.7 psia). For the case of direct ECC spillage, the I

actual pressure at which these ECC systems inject into the primary is reduced. The break location chosen for the RELAP5YA analysis is in the ECCS piping. Thus, immediately af ter the break occurs, the accumulator begins I

injecting into the ECCS piping; and this water spills directly to the containment. No accumulator water can be injected into the primary system until the primary pressure drops below the accumulator pressure. For smaller breaks, by the time this occurs, the accumulator is completely depleted and I

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s core recovery is due only to HPSI and LPSI pumps.

For larger breaks, more accumulator water is available for injection into the Primary System. The l

availability of accumulator water as a function of break size has an impact on the timing of core recovery.

It is also the main reason why this category of breaks (with direct ECC spillage) provide the limiting SBLOCA scenarios for the Yankee plant.

3.3 Analysis Results The RELAP5YA evaluation model features for the Yankee plant.

l as described in Section 2.0, was used a_

SBLOCA scenarios for a spectrum of break sizes. All these cases wr

's the category of breaks leading to direct ECC spillage, as discusseu in Sec 3.1.

The major analysis assumptions are summarized in Table 3.1.

Ft. che case of the largest ECC spillage area (break in ECCS piping), the break size analyzed corresponds to the inner diameter of the thermal sleeve (2-1/4").

This case generally followed the sequence of events described in Section 3.2 and resulted in a PCT of 1,101*F during the boiloff phase.

It is not further described here.

For the case of ECC spillage through the 2-1/4" thermal sleeve, the break sizes analyzed correspond to equivalent diameters of 5", 7",

8-1/2", and 10".

Of these, the 8-1/2" break size resulted in the highest PCT of 1,601*F.

This limiting SBLOCA case for the Yankee plant is further described below.

For the 8-1/2" break size case, Figures 3-1 through 3-6 show the-calculated history of major paramoters.

Figure 3-1 shows the primary and secondary pressure response.

The primary pressure decreases rapidly to near the secondary pressure within about 25 seconds.

During this time, the fluid at the break changes from single-phase liquid to two-phase conditions.

This can be seen in Figure 3-2 which shows the liquid volume fraction at the break location. This break size is large enough for the mass and energy outflow at the break to overcome the core decay energy and volumetric expansion in the primary due to flashing. Hence, the primary pressure continues to decrease WPP41/17

I steadily (Figure 3-1).

The fluid at the break location becomes mostly vapor I

at about 40 to 50 seconds (Figure 3-2) accelerating the primary pressure decrease (Figure 3-1).

I The Safety Injection Actuation Signal (SIAS) is received when the primary c essure drops ta 1,665 psia (1,015 psia accoitnting for uncertainties).

However, due to the assumption of loss of off-site power, there is a maximum of a 30-second delay in the delivery of ECC injection.

This includes the delay time for diesel startup, loading the HPSI pumps on the diesel, and time for the pumps to achieve full speed. -Hence, ECC injection I

into the primary does not begin until about 30 seconds.

Figure 3-3 shows the combined ECC injection flow rate including available accumulator injection.

i For this break size, the accumulator begins injecting at about 50 seconds and is depleted (and isolated) at about 110 seconds.

In thi RELAP5YA calculation, some of the injected ECC water is

~~

entrained and carried out of the break. This can be seen in Figure 3-2 and also in Figure 3-4 which shows the break = discharge mass flow rate. The ECC-injection rate (Figure 3-3) overcomes the break flow rate (Figure 3-4)

I starting at about 160 seconds, causing a net increase of inventory in the 2

vessel (Figure 3-5) and beginning the core recovery process.

The PCTc in the average core and hot rod are shown in Figure 3-6.

The early heatup is caused by bulk voiding in the core due to rapid depressurization for this larger break size. Accumulator injection arrests this heatup ond causes core quench; but as the accumulator becomes depleted, the core dries out and begins a second heat 3 This resu]*.s in the peak clad temperature of 1,601*F at about 330 seconds.

Beyond this time, the core cools I

slowly (Figure 3-6) because the net inventory gain in the vessel occurs at-a very slow rate (Figure 3-5).

The ECC flow rate does compensate for the break discharge flow rate beyond this time, as seen in Figures 3-3 and-3-4.

Since the core power continues to decay, this will eventually lead to complete core f

quench and core refill. Hence, the calculation was terminated at 500 seconds into the transient.

I f

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2 WPP41/17 B

I The calculations for the other break sizes (S",

7", and 10") produced similar results with the major differences being in the timing of key events.

Table 3.2 summarizes the sequence of significant events for all the break I

sizes analyzid.

The results of the analyses, including the PCT and cladding oxidation, are presented in Table 3.3.

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-19" WPP41/17 I

IAELE 3.1 l

Analysis Assumntions for Yankee Plant SBLOCA Cases l

1.

A small break located at the cold les ECCS injection location occurs at l

1.0 seconds. The Moody two-phase critical flow model and the RELAP5YA subcooled model are used to calculate the break flow rates.

2.

A coincident loss of off-site power occurs at 1.0 seconds.

3.

.A power level of 1.03 times the licensed power was used.

The heat generation rates from radioactive decay of fission products were assumed to be 1.2 times the values for infinite operating time in the 1971 ANS standard.

4.

The Baker-Just equation was used to determine the energy release and cladding oxidation from the metal-water reaction.

5.

All safety injection flow to the broken loop from the ECCS pumps ant accumulator is lost directly to the containment.

'I s

6.

The axial power shape utilized bounds the peak locations determined from the large break LOCA analysis, i

7.

PLHGR limit of 13.36 kW/f t bounds the large break LOCA analysis limits.

8.

Maximum fuel-stored energy is used, representative of Beginning-of-Life

{

(BOL) conditions.

9.

ECC water temperature is assumed at 200'F, consistent with guidelines of Reference 3.

10.

Single failure is failure of a diesel, causing loss of one train of l

pumped ECC injection.

11. No credit for operator action to start steam-driven auxiliary feedwater pumps.

12.

Containment pressure is assumed to be constant at 14.7 psia.

i

$ WPP41/17 l

TABLE 3.2 Seonence of Events for Yankee Plant SBLOCA Cases Time (Seconds)

Break Size (Inch ID)

Event 2.25 5

7 8.5 10 i

Break Opens / Loss of-Off-Site Power 1.0 1.0 1.0 1.0 1.0 l

Reactor. Scram / Steam Generator Flow 1.6 1.6 1.6 1.6 1.6 Isolation /RCPs Trip-Feedwater Flow: Isolation 2

2 2

2 2

ECCS Injection Begins 483 32 32 32 32 Loop Seal Clearing:

Broken /

160/

.'110/

60/

42/

37/

Intact 450

,110 60 42 37-Accumulator Injection Begins 124 70 49 39 Accumulator Empty (Isolated) 30.3 185 125 107 93 Time of Peak Clad lemperature 910-315 315 320 290 l

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I TABLE 3.3 Yankee Plant SBLOCA Analvais Resul_ts I

Break Size (Inches)

Event 2.25 5

7 8.5 10 Peak Cladding Temperature ('F) 1,101 1,223

- 1,425 1,601 1,558 Peak Cladding Temperature Location (Ft) 6.4 5.15 5.45 5.45 5.15 Local Maximum 2r/H O Reaction (%)

0.003 0.049 0.27 1.01 0.45 2

Total Zr/H O Reaction (%)

(1

<1

<1

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2 e.t e e.. t.re Time (sec>

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L Figure 3.2 Break Liquid Fraction for Limiting SBLOCA Case i

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Figure 3.5 i

~

Reactor Vessel Mass Inventory for Limiting SBLOCA Case a

N

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o i

I 4.0 CONCM110N 1

A spectrum of small break LOCA cases has been evaluated for the YNPS.

I The analysis was performed with the RELAP5YA computer program in conformance with regulatory requirements. The results of the analysis show that:

1.

The limiting small break LOCA case for YNPS is a break size of 8.5" equivalent diameter located at the connection of the ECC injection line with the cold leg.

2.

The SBLOCA analysis results are well within the 100FR50.46 acceptance criteria:

j I

a.

The peak cladding temperature is below 2,200*F.

b.

The maximum cladding oxidation is below 17% of the initial cladding thickness.

c.

Hydrogen generation due to cladding oxidation is below 1% of the maximum possible amount.

d.

The core retains coolable geometry because no significant clad swelling and rupture were calculated.

e.

Long-term cooling is assured because the ECCS is able to compensate for the break discharge flow, and long-term cooling systems are adequate to remove decay heat.

I 3.

The peak cladding temperature for the limiting SBLOCA analysis is lower than that for the LBLOCA analysis. The SBLOCA analysis was I

performed at a PLHGR of 13.36 kW/ft with an axial power shape that bounds the LBLOCA analysis.. Thus, for YNPS, the licensing basis for ECCS performance continues to be determined by the LBLOCA analysis.

I WPP41/17 E

I

5.0 REFERENCES

I 1.

" Clarification of MI Action Plan Requirements," USNRC Report NUREG-0737 November 1980.

2.

R. 7. Fernandez, et al.. "RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis," YAEC-1300P, Volumes 1, 2, and 3. October 1982.

3.

Letter, USNRC to YAEC on acceptance of Topical Report YAEC-1300P, dated October 14, 1988, with Enclosure.

1 4.

Letter, YAEC to USNRC on M I Action Item II.K.3.31. BYR 89-012, dated i

January 23, 1989.

5.

Proposed Change No. 145, Supplement No. 7. WYR 77-90, " Additional Yankee

[

Rowe Core XIII Small Break Analysis," dated September 21, 1977.

6.

Letter, YAEC to USNRC, " Core XVIII LOCA Analysis - Additional Information," FYR 85-131, dated November 19, 1985.

7.

YAEC-1437, " Reactor Coolant Pump Operation During Small Break LOCA Transients at' the Yankee Nuclear Power Station," July 1984.

8.

NUREG/CR-443b, EGG-2424,."Results of Semiscale Mod-20 Small-Break (5%)

Loss-of-Coolant Accident Experiments, S-LH-1 and S-LH-2," November 1985.

9.

NUREG/CR-4384. EGG-2416 (Draft), " Break Spectrum Analysis for Small Break Loss-of-Coolant Accidents in a RESAR-3S Plant," September 1985.

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WPP41/17 I

.