ML18047A457

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Valve Inlet Fluid Conditions for Pressurizer Safety & Relief Valves in Westinghouse-Designed Plants, Interim Rept
ML18047A457
Person / Time
Site: Palisades Entergy icon.png
Issue date: 02/28/1982
From: Meliksetian A, Sklencar A
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
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ML18047A442 List:
References
V102-19, NUDOCS 8207160352
Download: ML18047A457 (53)
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Text

1-VM.VE INLET FLUID CONDITIONS FOR PRESSURIZER SAFETY AND RELIEF V/iLVES IN WESTINGHOUSE-DESIGNED PLANTS NP-RESEARCH PROJECT V102-19 (PHASE C)

INTERIM REPORT~ FEBRUARY 1982 Prepared by WESTINGHOUSE ELECTRIC CORPORATION Nucl.ear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 PRINCIP/iL INVESTIGATORS A. Meliksetian A. M. Sklencar Prepared for PARTICIPATING PWR UTILITIES and ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue Palo Alto, California 94304 EPRI Project Manager J. Hosler Nuclear Power Division

  • 8207160352 820401 PDR ADOCK 05000255 P PDR

NOTICE This report was prepared by the organization(s) named below as an account of work sponsored by the Electric Power Research Institute, Inc. (EPRI) and participating PWR Utilities. Neither EPRI, members of EPRI, participating PWR Utilities, the organization(s) named below, nor any person acting on behalf of any of them:

(a) makes any warranty, express or implied, with respect to the use of any informa- -

tion, apparatus, method, or process disclosed in this report or that such use may 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 report.

Prepared by Westinghouse Electric Corporation Pittsburgh, Pennsylvania

ABSTRACT

. rhe overpressure transients for Westinghouse-designed NSSSs are reviewed to

  • determine the fluid conditions at the inlet to the PORV and safety valves.

The transients considered are:

1. Licensing (FSAR) Transients
2. Extended Operation of High Pressure Safety lnjectio11 System
3. Cold Overpressurization The results of this review, presented in the fonn of tables and graphs, define the range of fluid conditions expected at the inlet to pressurizer safety and power-
  • operated *relief valves utilized in Westinghouse-designed PWR units. These.results will provide input to the PWR uti1ities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI/PWR Safety and Relief Valve Test Program indeed envelop those expected in their units
  • EPRI PERSPECTIVE PROJECT DESCRIPTION This report, developed* under RPV102-19 in support of the EPRI/PWR Safety and Relief Valve Test Program, presents the expected range of fluid inlet conditions for pressurizer safety and relief valves utilized in PWR units designed by Westinghouse. These conditions are determined based on consideration of FSAR, Extended High Pressure Liquid Injection, and Cold Overpressurization Events.

PROJECT OBJECTIVE The objective of this. report is to assist PWR.utilities with Westinghouse plants in demonstrating that the fluid conditions under which their valve designs are tested as part of the aforementioned program envelop those expected in their unit(s).

PROJECT RESULTS FSAR events are found to result in challenges to ~oth relief and safety valves under steam conditions with valve inlet pressures as high as 2682 psia. Liquid discharge through relief and safety valves is predicted for only one FSAR event, the faedline break accident. Liquid temperatures and surge rates for this event range fran 553 to 672 degrees Fahrenheit and 224 to 2989 gallons per minute, res pee tive ly.

Extended High Pressure Liquid Injection events are found to result in no relief and safety valve challenges in two loop plants, relief valve challenges in both three and four loop plants, and safety valve challenges only in four loop plants.

In cases when the valves are challenged, liquid discharge is also predicted for these events. Liquid temperatures and surge rates for these events range from 498 to 598 degrees Fahrenheit and O to 1104 gallons per minute, respectively.

Cold overpressurization events challenge only relief valves. Liquid discharge is predicted for these events at pressures ranging from 280 to 2350 psia with temper-atures ranging fran 100 to 650 degrees Fahrenheit.

John Hosler, Project Manager Mur.1i:i;ar Pnwi:ir nivhinn

ACKNOWLEDGEMENT The contributions of J. M. Thompson, s. L. Ellenberger and C. Allen in the preparation of this report are gratefully acknowledged

  • CONTENTS ection 1 INTRODUCTION 1-1
1. 1 Background 1~1 1.2 Objectf ve 1-1 1.3 Scope.of Work 1-2 1.4 Quality Assurance 1-4 2 GENERAL DESCRIPTION OF EVENTS THAT HAVE POTENTIAL FOR CHALLENGING SAFETY AND RELIEF VALVES 2-1 2.1 Licensing (FSAR) Transients 2-1 2.1.1 Class II Events 2-1 2.1.2 Class IV Events 2-3 2.2 Extended High Pressure Injection Events 2-5 2.2.1 Transients Expected to Result in Initiation of High Pressure Safety
  • Injection and Challenge the Relief or Safety Valves 2-5 2.2.2 Spurious Initiation of High Pressure Safety Injection at Power 2-6 2.3 Cold Overpressure Transients 2-7 2.3.1 Mass Input Event 2-7 2.3.2 Heat Input Events  ! 2-9 3 GROUPING OF WESTINGHOUSE-DESIGNED NSSSs 3~1 3.1 Definition of Critical Parameters 3-1 3.2 Basis for Selection of Reference Plants 3-3 4 METHODOLOGY USED IN DETERMINING RANGE OF EXPECTED PORV AND SAFETY VAL VE INLET CONDITIONS 4-1 4.1 Licensing-Type Transients 4-1 4.1.l Transients Resulting in Steam Discharge _4-1 4.1.2 Transients Resulting in Liquid Dfscharge 4-4 4.2 Extended High Pressure Injection Events i 4-5 I

4.3 Cold Overpressurization Events I 4-6

CONTENTS (cont)

-~

5 EXPECTED FLUID CONDITIONS AT SAFETY & RELIEF VAL VE INLETS 5-1 5.1 FSAR Transients. Resulting in Steam Discharge 5-1 5.1.l Reference Plant -- 2-Loop Group 5-1 5.1.2 Reference Plant -- 3-Loop Group 5-3 5.1.3 Reference Plant -- 4-Loop Group 5-3 5.2 Plant-Specific Valve Inlet Conditions for Main Feedline Break 5-4 5.3 Extended High Pressure Injection Events 5-4 5.4 Plant-Specific Valve Inlet Cond1tions Resulting from Cold Overpressurization Events 5-8 6 REFERENCES 6-1

SUftltfARY

.This ~eport provides documentation of the expected range of pressurizer safety and relief valve fluid inlet conditions for Westinghouse designed plants. It is intended for use by PWR utilities with Westinghouse units in their justification that the fluid inlet conditions under which their valve designs are tested, as part of the EPRI/PWR Safety and Relief Valve Test Program, envelop those expected in their unit(s). These conditions are determined based on consideration of FSAR, Extended High Pressure Liquid Injection, and Cold Overpressurization events.

The methodology used to determine these conditions includes the grouping of Westinghouse PWR units by design and layout philosophy *. For each group, a reference plant is then selected based on maximizing a nondimensional parameter which incorporates the critical plant parameters affecting the severity of FSAR overpressurization events resulting in steam discharge. Valve fluid inlet condi-tions resulting from limiting FSAR events, which result in steam discharge and an Extended High Pressure L)quid _Injection event and which may result in liquid dis-charge, are presented for each reference plant. These conditions envelop those expected by the plants represented by each reference plant *

  • Fran the FSAR events that may result in liquid discharge, the feedline break acci-dent is considered. For this event, plant specific valve inlet conditions are presented where applicable.

Fluid inlet conditions are presented for Cold Overpressurization events which envelop those expected fn all units for which Westinghouse has provided the plant specific Cold Overpressurization Protection System design and analysis. These analyses consider the limiting mass and heat input events for each unit evaluated *

- ~---

-* Section 1 INTRODUCTION

1.1 BACKGROUND

Following the Three Mile Island Unit 2 (TMI-2) incident, the Nuclear Regulatory Conmission (NRC) published NUREG-0578, "TMI-2 Lessons Learned - Task Force Status Report and Short-Term Recommendations, 11 which required utilities operating and in the process of constructing pressurizer water reactor (PWR) power plants to develop a program to demonstrate the operability of power operated relief valves (PORVs) and self-actuated safe~ valves (PSVs) used in the protection of reactor coolant systems. The requirements of NUREG-0578 were later ~larified in NUREG-0737. In response to NUREG-0578 and NUREG-0737 requirements, ~he PWR utilities assigned EPRI the responsibility of conducting a comprehensive test program to demonstrate the operability of the various types of PORVs and safety valves used by participating utilities. The primary objective of that program is to obtain performance data applicable to each of the various types of reactor

.coolant system safety and relief valves in PWR plant service for the range of conditions under which they may be required to operate.

As part of their response to the NUREG requirement, each PWR licensee or applicant is required to provide evidence that the conditions under which valves representative of those installed in their unit(s) are tested, are representative of those expected in their units. Such conditions include valve inlet piping configurations, backpressure and dynamic loading as well as fluid inlet state, pressures, and temperature.

To assist in the development of test conditions to be applied to the valves selected for testing, each PWR NSSS vendor was contracted to develop a "Pl ant Conditions Justification Report" describing the range of fluid conditions expected at the inlets of relief and safety valves installed in plants of their design.

l .2 OBJECTIVE The objective of this report is to document the justification for a set of limiting fluid inlet conditions to be used as input to the selection of fluid conditions for testing power-operated relief valve (PORV) and safety valve designs used in 1-1

Westinghouse plants. This report wi11 be referenced by ?WR utilities with Westinghouse plants fn their justification that the f1ui d conditions under which their valve designs are tested, as part of the EPRI program, envelop those expected in their untt( s).

1.3 SCOPE OF WORK The evaluation of expected fiu1d fn1et conditions is based an consideration of FSAR events. Extended Safety Injection events, and cold overpressur'fz:at1on eYEnts. - Standard licensing methodology is used to determine the ftuid conditions at the inlet of the PORV and safety valves for FSAA events. The events evaluated fn a plant's FSAR or in later licensing subm1ttals that have the potential of challenging such valves are considerede Because of the large number of 'Westinghouse plants w1th varying design and layout, reference Planu have been selected for evaluation.. ibe analysis performed on the

  • reference Plant represents the expeeted behavior of a11 plants in that gl"Cup .. The, reference plants are selected by pei'"form1ng d1mens1ona1 ana1yses on their critical parametars and considering the simi1ar1~ of characteristics, .design, and 1ayout ..

Reference plants are used for licensing-type accident analyses (with the exception of the feed1fne break accident) and extended high pressure Hquid fnject'fan events. Plant specific: f1uid conditions are presented for the feedHne break accident, 'A'here app1fcab1e.

  • Not all Westinghouse units are eovered by this report since several units no longer uti1fze Westinghouse fue1 or have their reload analyses performed by Westinghouse. Those that are* covered with respect to FSAA and Extended Safety Injection. ~v~_ts are shown below. Of these p1ants, those for which f1uf d c:~nditions for cold overpressurfzat'fon events are presented are identified with an asterisk. Conditions resulting frcm co1d oYel"fJressure transients are presented an1y for plants for which Westinghouse performed the speeiffc design and analysis of their cold overpressurization protectian systen.

Two-Loop Plants Name RGE R. E. Gfnna Rec::hester Gas &E1ec:trie Corp.

'.tlE? Point aeach fl ~fsconsin Electric: ?ewer Utilities n

'lllIS Point Beac:n Wisconsin Electric Power Uti1fties

Two-Loop Plants Name Owner NSP Prairie Island #1 Northern States Power NRP Prairie Island #2 Northern States Power WPS Kewaunee Wisconsin Public Service Three-Loop Plants Name Owner SCE San Onofre #1 Southern California Edison CPL H. B. Robinson #2 Carolina Power &Light Co.

FPL Turkey Point #3 . Florida Power &Light Co.

FLA Turi:ey Point #4 Florida Power &Light Co.

VPA Surry #1 Virginia Electric &Power.Co.

VIR Surry #2 Virginia Electric & Power. Co.

DLW . Beaver Valley #1 Duquesne Light Co. ,, ,* l~

VRA North Anna #1 Virginia Electric &Power Co.

ALA Joseph M. Farley #1 Alabama Power Co.

VGB North Anna #2 Virginia Electric &Power Co.

APR Joseph M. Farley #2 Alabama Power Co.

CGE Virgil C. SLD11111er #1 South Carolina Electric &Gas DMW Beaver Valley 12 Duquesne L.i ght Company

  • CQL Shearon Harris #1 Carolina Power & Light Co.

CRL Shearon Harris #2 Carolina Power &Light Co.

CSL Shearon Harris 13 Carolina Power &Light Co.

CTL Shearon Harris 14 Carolina Power &Light Co.

Four-Loop Plants Name Owner IPP Indian Point #2 Consolidated Edison Co. of New York INT Indian Point #3 Power Authority, State of New York CWE Zion 11 Commonwealth Edison CCM Zion #2 Commonwealth Edison AEP Donald C. Cook #1 American Electric Power Co.

AMP Donald C. Cook #2 American Electric Power Co.

PGE Diablo Canyon #1 Pacific Gas &Electric Power 1-3

Four-Loop Plants

-Name OMier PEG Diab1o Canyon #2 Pacific Gas & Electric Power POR Trojan Portland General Electric TVA

  • Sequoyah 12 Tennessee Valley Authority PSE Sal* #1 Public Service Electric I Gas PN.J Salem #2 Public SerY'fce E1ectr'fc I Gas OAP W. Bo McGuire #1 Duke Powel" Co.

DBP V. B. McGuire #2 Duke Power Co.

WAT

  • Watts Bar #1 Tennessee Valley Authority WBT '* Watts Bar #2 "Tennessee Valley Authority CAE .* Byron #1 Ccamonwealth Edison Co.

w Alvin We Yogtle fl aeorg*f'a Power Co.

&BE A1v1n Vo Yogtle 12 Georgia Power Co.

Nm M111stane #3 Northeast Ut11ftfas NAH Seabrook #1 Public Service Co. of New Hampshire NCH Seabrock #2 Public Service Co. of New Hampshire DC? Catawba #1 Duke PoM!r Co.

CDP TBX TCX CCE Catawba #2 Comanche Peak #1

  • eananche Peak
  • Braidwood #1
  1. 2 Duke Power Co.

Texas Uti1f ti es Texas Utilities Coamonwealth Edison Coo CDE

  • Braidwood #2 Caimonwealth Edison Coo TGX
  • South Texas #1 Houston Lfght &Power THX
  • Scuth Texas #2 Houston Lfght &Power PSJ Marble Hi11 #1 Pub11c Service of Indiana PCJ Marble Hf11 #2 Publfc Senrice' of Indiana CSE
  • ByYoon #2 Co11111onwealth Edison Co.

SAP "' Wo1 f Creek (SWPPS) Kansas Gas &E1ectrfe Coo SCP

  • Ca11away #1 (SNUPPS) Unf on Electric Co.

SF?

  • Call away 12 (SNUPPS) Union Electric Co.

1.4 QUALiif ASSURANCE The wcr!c perfonied in the deve1opment or this report is in accordance with 10Cr~SO

  • Appendix B, Quality Assurance requirements.

Section Z GENERAL DESCRIPTION OF EVENTS THAT HAVE POTENTIAL FOR CHALLENGING SAFETY AND RELIEF VALVES The events that cause*overpressurization of the reactor coolant system are grouped into licensing (FSAR} transients, transients that result from automatic initiation of the high pressure injection system, and cold overpressurization transients.

I .

2.1 LICENSING (FSAR) TRANSIENTS The transients that result in the actuation* of safety and relief valves and are normally analyzed for safety analysis repo.rts can be groupe~ under Cl ass II and Class IV events.

2.1.l Class II Events Class II events are incidents of moderate frequency that may occur during a calendar year for a particular plant. The transients in this class are described in the following paragraphs.

2. 1. l. l Loss of Load. In the ev~nt of the loss of external electrical load without .bYP.a~s, a. sudden reduction in steam flow will cause an
  • Both the pressurizer safety valves and main steam safety valves may open for the loss of load event. Only steam is discharged from the pressurizer safety and relief valves and no water discharge is

. observed.

2.1.l.2 Loss of.Normal Feedwater. A loss of normal feedwater results in a reduction of the secondary system's capability to remove the heat 2-1

generated in the ~actor core. If an alternative supply of feedwater were not supplied, residual heat following reactor trip would heat the primary system water to the point where water relief frcm the pressurizer occurs.

The analysis non11111y shews that the pressurizer steam space does not ccmp1ete1y f111 with W1ter 1 and therefore only steam diseharge through PatY and safety valves fs observedo 2.1.1.3 Accidental Depressurization of the Secondary System. The accidental depressurization of the main steam system may result fran the inadvertent opening of a single steam dunp, relief, or safety ,

valv*. This event results fn a sma11 increase f n naminal steam flow.

A much larger- increase in steam f1ow can be caused from steam 1fne rupturee Initial increase fn steam f1ow increases the energy removal rate frem the primary ~ystem and causes a reduction of reactcr coolant tempera-ture and pressure. In the presence of a negative moderator tempera..

ture coefficient, the cooldown results fn a positive reactivity inser-tion. If the most reactive rod cluster contro1 assembly fs assumed stuck in its fully withdrawn position after r~actcr trip, there fs an fne~eased possibilit\Y that the core wi11 became critical and return to power-. The eo.-. is ultimately shut down due to boric acid injection by the Safety Injec:tf on System o At hot shutdowns actuat1 on of the safety 1nJection system occurs early enough ta prevent erft1c:a11ty.

The above discussion 11so covers Minor Steam Line Rupture, wtlf-ch -1s a c1ass III event. 1 For both events the- extended *operation- of the

  • > ) ~ > ~

Safety Injection System can result 1n pressurization of the primary system 'Nhfch will result first fn steam discharge and 1ater in watar discharge, if safe"t;y valves and PORVs are actuate<io 1c1ass III events are events ~hich may occur very infrequently during the 1ife of the p1ant.

2-2

2.1.1.4 Loss of Off-Site Power. In the event of complete loss of offsite power and turbine trip, there will be a loss of power to the station auxiliaries, (reactor coolant pumps, condensate pump, and so forth). Reactor coolant flow coasts down to natural circulation flow rates. Main feedwater flow is lost and the auxiliary feed pumps auto-matically start. As the system pressure rises (due to decay heat input) following the trip, the system's PORYs are automatically opened. If the steam flow rate through the PORYs is not adequate, safety valves -may lift to dissipate the excess energy by passing steam.

2.1.1.5 Uncontrolled Rod Withdrawal at Power. A continuous uncontrolled rod cluster control assembly (RCCA) withdrawal at power due to faulty operator actions or malfunction of the reactor instru-ments will result in an increase in the core heat flux. Following the event, the steam generator heat removal rate will lag behind the core power generation rate until the steam generator pressure reaches the main steam safety/relief valve setpoint. This unbalanced heat removal rate will cause the reactor coolant temperature and pressure to rise and eventually may actuate the PORYs. If the steam flow through the PORVs is not adequate, safety valves may also be actuated to dissipate*

the excess energy by passing steam.

2.1.2 Class IV Events Class IV events are limiting faults that are not expected to take place, but are postulated because their consequences include the potential for the release of significant amounts of radioactive mate-rial.

2.1.2.1 Main Feedwater Pipe Rupture. A major feedwater line rupture is defined as a break in the feedwater line large enough to prevent the addition of sufficient feedwater to the steam generator to main-.

tain .shell side fluid inventory in the steam generators. If the break is postulated to occur in the feedline be_tween the check valve and the steam generators, fluid from steam generators will be discharged through the break. Feedwater flow to the steam generators may be reduced. This can cause the reactor coolant temperature to increase prior to reactor trip. For certain locations and sizes of the 2-3

postulated breaks the PORVs and safet'J valves are challenged and wi11 pass steam followed by slightly subc:oo1ed water.

2.1.2.2 s*team Line Rupture. The scenario for this accident is the same as discussed fo~ accidental depressur1zation of the secondary system due to fnactve~tent opening of a single steam dump, relfef, or safety valvea The effects of minor ser:ondar-y system pipe breaks are bounded by the analysis pl"esented for this event. The safeties and PORVs~ when actuated, w111 d1s~harge steam and fn the long term (beyond scope of FSAR analysis} they may also discharge water. As far as 11qu1d discharge through the safety and relief valves is concerneds depres-surizatfon of the secondal""J systsm due to steam 11ne rupture or inadvertent opening of main steam safety/relief valves is bounded by the analysis performed for feed1fne break and spurious actuation of the safety fnjeetion system at pewero 2.1.2.3 Lccked Rotor. Thfs accident fs postulated to result from a sudden locking of one rotor on one of the primary reactor cao1 ant pumps. Thfs causes a rapid reduction fn core flow rate, reducing the heat transfer rate fn one steam generator and. fncreasing the tempera-ture of the coolant, eausing. severe pressure increases. Departure frao nucleate bo111ng may oceur due ta flow reduction and the resul~

tant powr-coolant mismatch. Both the PORYs and safet"J valves ar'e cha11enged and are required to flow steam.

2~1.2.4 Rod Ejection. This accident 1s the result of the assumed mechanical failure of a control rod mechanism pressure hcusfng,* such

~at the reactor coolant system pressure ejects the control rod and drive shaft to the fully withdrawn position~ This mechan1ca1 failure at most 1eads to a Fap'fd rtact1vity insertion teg~ther with a higheF core p~er distribution peak and high reactor coolant pressure.

As the system pressure increases the PORVs may be actuated to discharge steam. HaweveF, if the Ste~ flow through the PORVs is not

-**adequate, safet'J valves may open on steam to prevent excessive pressuri:ati on.

2.2 EXTENDED HIGH PRESSURE INJECTION EVENTS The safety injection system (SIS) is designed to provide emergency core cooling in the case of a LOCA or steam break accident. The system is designed to maintain its protective capability in case of single failure. The system layout varies somewhat.for different groups of plants. However, they retain the same basic functional design criteria, the main difference befog the use of certain pumps for different purposes. The system operation is initiated by the Safety Injection Signal, which can be actuated by any of the following:

o Low pressurizer pressure o High containment pressure o High steam line differential pressure o Low steamline pressure o Manual actua.tion fran control board

.:.*In the following sections, incidents that actuate the SIS signal are considered.

  • 2.2.1 Transients Expected to Result in Initiation of High Pressure Safety Injection and Challenge *the Pressurizer Relief or Safety Valves For events described in this section steam is discharged through PQRV and safety valves when the valve is first lifted, but later, when the pressurizer is filled with water, water discharge is al so predicted.

2.2.1.1 Accidental Depressurization of the Secondary System/Steam Line Rupture. An accidental depressurization of the main ' ,.

steam system may result from the ruptu*re of steam 1i.ne ., * .

  • l ' > '

inadvertent opening of main steam safety/relief valves, or a single steam dump. The steam release results in an initial increase in steam flow that decreases during the accident as the steam pressure falls. The temporary increase in energy removal fran the reactor coolant system causes a reduction of cool ant temperature and pressure *. Due to the negative moderator coefficient, the cooldown results in a positive reactivity insertion. If the most reactive rod cluster control assembly is assumed stuck in its fully withdrawn 2-5

position after reactor trip, there fs the possibility of the core returning ta power. The primal""J cooldown and pressure drop actuates the SIS, which eventually results in a system repressuration as the pressurizer fills with lfquid *

. Extended operation of the SIS would result in cycling of the PalYs (or safety valves if PORVs wre assumed* unavailable) on steam followed by subcooled watero ..

2.2.1.2 Main Feed11ne Rupture Accident. _This aceident fs discussed 1n paragraph 2.. 1.2 .. 1 and later in paragraph 4o 1.2.1.

2.2.2 ~ul"ious Initiation of High_ Pressure Safety Injection at Power Inadvertent or spurious actuation of the safety injection system at powr can be caused by operator error or a fa1 se e1ectr1ca1 actuating signal.

Following the spurious actuation, the coolant charging pumps force highly concentl"lted boric acid solution through injection lines into the cold legs of each*1oop. Depending on the t'Jpe of the plant, the residual heat removal pmps, safety fnject1on pumps.s and the passive 1njeet1on system are also actuated but provide no flow when the RCS fs at normal pressureo An SIS signal results 1n reactor trip followed by a turbine trip; hoM!ver, for e~tra eonservativ1sm another case was reviewed where it was assumed that the trip was delayed.

Both PORY and safetf valves may be eha11anged depending o~ the J)ressure--head characteristics of the safety fnject'f en system. Va1 ves, if actuated, 11ft on steam, and for extended operation of the safet;y injection system, subceoled water dfsc:harge may be observed. This event is more limiting frcm the viewpoint of surge flow and range of liquid temperatures at the v~ive inlet, and is selected for subsequent analysis.

2.3 COLD OVERPRESSURE TRANSIENTS 2.3.1 Mass Input Events Based on probability of occurrence and in-plant operating experience, the most credible mass input events producing a net injection of mass into the reactor coolant system (RCS} involve failure in the air supply system, which causes the charging flow control valve to open,

    • and/or isolation of letdown. Mass injection based on single charging pump operation is the most likely mass input mechanism, producing.

typical charging rates up to 120.* gpm following isolation of letdown, and higher rates for air supply system failure.

Although precluded at low temperature by administrative procedure, two-charging-pump operation was considered in all plants to develop maximum input capability and thus provide additional flexibility in the operation of the cold overpressure mitigation system. Maximum input capability associated with this mechanism as applied to all plants analyzed to date is shown in Figure 2-1. The PORV inlet conditions presented in Figure 5-1 also include this mechanism.

Operation of the PORV at a predetermined setpoint pressure is employed by Westinghouse in the Cold Overpressure Mitigation System (OMS) to arrest the pressure transient caused by the above mechanisms.

Mitigation of the transient on valve opening results in the RCS .

pressure turning over. This produces a transient peak overpressure.

The PORV continues to open until valve capacity matches the net mass injection rate, after which the reset pressure is reached and the valve begins to close. PORV closure arrests the decreasing RCS pressure and reinitiates the pressure increase to complete the pres-sure transient cycle. This mimimum pressure is termed the transient pressure undershoot _and is determined by the blowdown setting of the PORVs (nominally 20 psi). Pressure cycling continues until action is taken to remove the mass input mechanism.

Selection of ?ORV setpoints for pressure contrcr1 of mass input-induced transients are based on a water-solid. reactor coolant system, which produces pressure excursions significantly higher than for a RCS with 2-7

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Valve opening and closure times of 2 seconds are assumed, and valve setpoi'nts are staggered such that operation of the first valve wi11 mitigate the event so that the other valve will not be challenged.

2.3.2 Heat Input Events

  • The heat input case which has the potential for the most severe pres-sure transient is that in which the steam generators exhibit a higher temperature than the remainder of the reactor cool ant system. The magnitude of the difference in temperature is dependent on the means by which the temperature asynnetry was achieved, but a typical difference is considered to be about S0°F.

For the heat input transient with the initial reactor coolant tempera-ture so°F less than the temperature in the steam generator.$ and with all reactor coolant pumps off, one of the ~o reactor coolant pumps is

  • started to circulate the reactor coolant through the wanner steam generators. As the* coolant flow begins, the wann water in the tubes of the steam generator in the active loop is forced out and into the reactor coolant PLDllP where it is pumped into and mixed with the.colder reactor coolant. In the inactive loops, the wanner water from the tubes of the steam generator is forced out in a reverse direction due to the backflow in the inactive loops, and also mixed with the cooler reactor coolant. This initial mixing of the wann water with the larger volume of cooler water causes an initial shrinkage effect which tends to decrease the initial coolant pressure.

Simultaneously, the cooler reactor coolant that enters the steam generator begins to be heated as it moves through the tube bundle. As heat is added to the coolant due to heat transfer from the secondary water in the steam generator, the cool ant attempts to expand and cause a resultant pressure increase. The net effect of the expansion due to the heat transferred to the coolant and the shrinkage effect due to the mixing of the warm water*with the cooler coolant is a relatively

  • 2-9

constant coolant pressure in the initial few seconds of the transient.

Then, as the flow rate increases and the heat transfel"' mechanism becomes predominant, the coolant p1"9ssure increases rapidly.

As fn the mass input scenario described fn Section 2G3.1, the reactor cool ant pressure 1nc1'"H.ses unt11 the pressure reaches the PORY set...

point. The valve opens and pressure f s mitigated. When the valve apens suff1cient1y to provide a capacity in excess of the expansion rate of the coolant and the coolant pressure decreases rapid1y after reaching an overshoot, PORY closure at the reset pressure setpoint is then requi ~ to arrest the pressure decay and start the pressure transient cycle over again.

The heat input transient due to temperature asymnetry in the reactor coolant system 1s unique in that it is se1f.. Hm1t1ng; foe .. 9 when the temperatures are brought to equilibrium by the reactor coolant f1ClWG the transient is ended. The use of a relief valve to mitigate the pr~sure t1'"111sient results fn a va1ve cycling of the coolant as ft fs heated, but the va1ve is only required to cycle a few times until the temperatures fn the system are brought tc equilibrium and coo1ant expansion ceases. The first cycle results fn the 1arge$t setpofnt overshooto Subsequent valve cycles result fn diminishing overshoots as the eoo1ant expansian rate diminishes unti1 eventually the valve closes and remains elosedo The heat input event is considered in the algorithm utilized for cold overpressure system !etpofnt selection for the re1fef valves as desc: ribed above G

Section 3 GROUPING OF WESTINGHOUSE-DESIGNED NSSSs A complete analysis of all overpressure transients in Wes*tinghouse-de~igned plants is prohibitive due to the large number of these plants. Therefore, a method was established to group the Westinghouse NSSSs un.der selected reference plants.

Various methods could have been utilized to group the plants and select reference plants. The method utilized in this report has two important objectives:

a. to maintain the similarity of plant characteristics and design philos-ophy within each group, and
  • b. to represent the performance of the plants in each group with the results of transient analyses perfonned on the respective reference plant.

The following method was used to generate the groups and to select reference plants for each group.

3.1 DEFINITION OF CRITICAL PARAMETERS

. The parameters that may have an effect on peak reactor cool ant pressure and the rate of pressurization for overpressure transients are listed in Table 3-1. The list was developed fran sensitivity analyses and engineering judgment.

The grouping of Westinghouse plants was done in two steps.

In the first step, the Westinghouse-designed NSSSs are collected into groups according to differences in design and physical layout. This initial grouping was made by number of loops. The number of loops is important, from the overpressure transient analysis view, since the limiting overpressure .transients are not the same for plants with different numbers of loops.

3-1

The number and capacities of safety and relief valves in the pri ma r:t system a~

also different for plants with different numbers of coolant loops.

Table 3-1

  • LISTING OF CRITICAL PARAME'ltRS Pa meters Units a**
1. Htaber of L.cops

..c ...

2. Steam Generator Type 3e NSSS Power MWt 4Q Vessel Average Temperature o,
5. Coolant Flow Rate lb/hr
6. Number of Pressurizer Safety Valves 7o Capae1ty of Pressuri:ef Safety Valves 1b/hr SQ Nuaber of PressurizeF R~1fef Valves O<=->
9. Capacity of Pressurizer Relief Valves 1b/hf'
10. Safet;y Valve Qpening Setpoint psig Relief Valve Opening Setpoint p~1g 11o 12., Safet;y InJ~t1on Charging Rate Versus Pressure gpm There are significant differen~es in the design af the SIS for the varioYs twa~p three-, and four*1ocp plants while they retain the same basic functional design crite~ia. The main difference f s the use of cer4'...ain pumps for single duty in one type and multiple dut'/ in others. Other differences inc1ude the number, shut-~ff

pressure, and size of passive injection system and the number, shut-off pressure, and run-out f1ow rate of 1ow and high head safety injection pumps.

There are other differences between plants with different numbers of loops that have less effect on the overpressure transients, such as the auxiliary and feedwater system, the containment pressure transient, and the protection logic.

Therefore, it is logical and necessary to group the Westinghouse designed NSSSs under two-, three-, and four-loop plants.

In the second step, it is necessary to select reference plants for each group considering the renaining critical parameters, which are:*

o ,Steam Generator Type o NSSS Power Generation o Coolant Flow Rate o Vessel Average Temperature o Safety and Relief Valve Opening Setpoint

  • 0 Safety and Relief Valve Number and Capacity Within Each Group The above critical parameters are utilized in the next section to select reference plants for*each group.

3.2 BASIS FOR SELECTION OF REFERENCE PLANTS The parameter that utilizes the effect of the above critical parameters for selecting the reference plants is the ratio of asymptotic surge rate to safety valve capacity.

The asymptotic surge rate is defined as follows:

W = {Reactor power) x {Total volUl'lletric expansion per unit temp. change)

Combined heat capacity of primary and secondary system Selectipn of the plant within a _group having the higher W should result in the 11 11 most severe valve inlet conditions for the plants in that group. The volumetric expansibn is considered in three parts: {l) cold volume, which includes the I

I 3-3

volume of the steam generator outlet plenum, the volume of cold legs, the volume of the do"1'1comer, and the volume of the reactor vessel 10.....!r plenum; (2} the medium temperature volume, which includes the volume of steam generator tubes, the active fuel region, and the volume of bypass f1ow region; and (3} hot volume, which is made up of the volume fran the active fuel region discharge ta steam generators.

A linear variation of volume,:tr'ic expansion versus temperature fn a small nef ghbof""'

hood about an average temperatun in each region is assumed, from which the tota1 volumetric expansion of the primary c:ool ant due to temperature c:hange is ea1 cu-1ated .. The heat capacity of the p.-inary system fs evaluated fran the basics of thermodynamics and that of the secondary system is obtained using empirical corre-lations:

o Steam Generator Type ..:- The effect of the steam generator type enters through the empfrical correlation used to ca1eu1ate the heat eapae1~ of the secondar.Y system .. The sensitivity analysis performed on the effect of the existence of preheaters fn the steam generators indicates that peak pressure reached during the course of overpressure tran$ients f s insensitive ta the existence of the preheaters. Therefore, steam generator type is not considered a lcey parameter fn this evaluation and was not used in plant groupingc a NSSS Power Generation .... This parameter appears fn the definition of the asymptotic surge rate ex;il1c'ft1y.

Vessel Average Temperatures and Ncmina1 Pressure -- The effects of these parameters are fnc:1 uded fn the ca1cu1atf on of vo1 umetric expansion rates calcu1ated at average temperature and pressure in each regiono o

  • Coolant Fiow Rate -- The effect of coolant flaw ra~ is fncorpo~ated thraugh pressure drop ea1cu1 at'fons that take into account the losses across the care fn1et, outlet, and upper and lower tie plates, and fric~

t1onal fuel spacers, hydrostatic and other accelerationJosses through~

1 out the primary locpe Except for the hydrostatie losses~ all other pressure drops are obtained frau ex?erimenta1 results and correlation$

are based on coo1ant f1ow rate. The eooiant f1cw rate and er.er;y transfer are intimately related to the heat transfer film coefficient, *nhic:h h

strongly dependent on the coolant velocity. Therefore, the use of tem-perature and pressures in calculating the volumetric expansion rate implicitly incorporates the effect of coolant flow rate.

o Safety and Relief Valve Capacity and Opening Setpoint -- The effects of these parameters are considered not in the asymptotic surge rate, but in

  • the ratio of asymptotic surge rate to the valve capacity. The volumetric discharge rate of the valve is calculated at the valve opening setpoint*

and depends on the capacity (area of the valve).

Therefore, by maximizing one nondimensional parameter (asymptotic surge rate to safety valve capacity) reference plants for each group ca.n be selected which would be expected to have fluid conditions enveloping those for the other plants in that group.

The results of this analysis are presented in Table 3-2, where the values of critical parameters and the ratio of asymptotic surge rate to PSV capacity for each plant are listed. The reference plants for two-, three-, and four-loop plants whose parameters are shown in Tables 3-3, 3-4, and 3-5 have the ratio of asymptotic surge rate to safety valve capacities of 1.563, 1.424, and 1.619 respectively. For 2 and 4-loop plants, reference plants were selected with the highest surge rate to valve capacity ratio for that group. For 3-loop plants the reference plant selected represents the majority of the plants in that group, rather than being the plant with the largest asymptotic surge rate to valve capacity for the following reasons:

1. More than 87 percent of the 3-loop plants are represented by a plant with an asymptotic surge rate of 10424.
2. The results of transient analysis perfonned for 4- and 2-loop plants (with surge rate to valve capacity of 1.619 and 1.563 respectively) envelop the results of transient analyses for 3-loop plants. (See table 5-1.

The reference plants from this point fon1ard will be considered generic plants representing two-, three-, and four-loop plants *

  • 3-5

TABLE 3-2 CRITICAL AND NONDIHENSIONAl PARAMETERS FOR WESTINGHOUSE-DESIGNIED NSSSs Ratto of Asymptottc Vessel Surge Rate Nunbe!I" Ste'a11 NSSS Average Coola"t PORVs Safety Valves to Safety of Generator- Power Tempo now Setpot111t tacac1tv setpol~t tacacf ty Valve Plant LOOI!_! Ty2e (Kit) fOf) (lb/hrdoB! N1111ber lesta>> n /hr) Nllllher* (psta 11 /hr) Ca(!llC ttl RGE 2 44 1520 573 .. 5 0600 2 2350 178900 2 2500 . 288000 1.563 WEP 2 44 151805 57309 0671 2 2350 179000 2 2500 288000 1.563 WIS 2 44 1518.5 573 .. 9 .671 2 2350 '879000 2 2500 288000 1.563 NSP 2 51 1650 567 .. l 0682 2 2350 119000 2 2500 345000 1.259 HRP 2 5ll 1650 56703 .682 2 2350 179000 2 2500 345000 1.259 WPS 2 51 . Ui50 56703 .682 2 2350 210000 2 2500 350000 1.244 SCE J 21 1l51l 575 1.38x 2 2350 108000 3 2500 240000 NA w CPL 3 44 2200 574.2 1.015><1 2 2350 210000 3 2500 288000 1.425 A\ FPL 3 44 2208 574.2 .. 965 2 2350 210000 3 2500 288000 1.425 FLA 3 44 220, 574.2 .965 2 2350 210000 3 2500 288000 1.425 VPA l 51 2441 574 .. 3 1.001 2 2350 210000 3 2500 293330 1.517 VIit 3 51 2441 574.3 1.007 2 2350 210000 3 2500 293000 1.517 DUI 3 51 2660 576.2 1.008 3 2350 210000 3 2500 . 345000 1.415 VllA 3 51 2785 580.3 1.05 2 2350 210000 3 2500 380000 1.425 At.A l 51 2660 577.2 1.007 2 2350 210000 3 2500 345000 1.415 VGD 3 51 2785 580.3 1.05 2 2350 210000 3 2500 380000 1.425 APll 3 51 2660 571.3 1.007 2 2350 210000 3 2500 345000 1.415 CGE 3 Ol-1 2785 587.4 1.096 3 2350 210000 3 2500 420000 1.425 om l 51 2660 576.2 1.008 3 2350 210000 l 2500 345000 1.415 Ct)L J 04-2 2785 587.5 1.092 3 2350 210000 3 2500 380000 1.425 CUL 3 SD4-1 2785 587.5 1.092 3 2350 210000 3 2500 380000 1.425 CSL l SD5 2785 587.5 1.092 3 2350 210000 3 2500 380000 1.425 CTL 3 05 2785 587.5 io092 3 2350 210000 3 2500 380000 1.425

  • TABLE 3-2 (cont)

CRITICAL AND NONOIMENSIONAL PARAMETERS FO~

WESTINGHOUSE-DESIGNED NSSSs Ratio of Asymptotic Vessel Surge Rate Number Steam NSSS Average Coolant PORVs Safet)'. Valves to Safety of Generator Power Temp. Flow Setpof nt C:apacftY Setpo1nt Capacity Valve Plant Loops Tlpe (MWt) iOf) (1b/hrx108l Number (psi a) Jlb/hr) Number (psia (lb/hr) Capacit)'.

IPP 4 44 2758 569.5 1. 361 2 2350 179000 3 2500 408000 1.222 INT 4 44 3025 571.5 1.363 2 2350 179000 3 2500 420000 1.334 CWE 4 51 3250 562.2 1.350 2 2350 2:10000 3 2500 420000 1. 352 COM 4 51 3250 562.2 1.350 2 2350 210000 3 2500 420000 1.352 AEP 4 51 3250 567.8 1.350 3 2350 179000 3 2500 420000 1. 352 AMP 4 51 3403 573.8 1. 346 3 2350 2:10000 3 2500 420000 1.483 PGE 4 51 3350 576.6 1.329 3 2350 2:10000 3 2500 420000 PEG 4 51 3423 577.6 1.339 3 2350 2:10000 3 2500 420000 1.619

>> PDR 4 51A -3423 584.7 1.326 2 2350 2:10000 3 2500 42000() 1.619

~TVA 4 51 3423 578.2 1.38 2 2350 179000 3 2500 420000 1.619 TEN 4 51 3423 578.2 1.38 2 2350 179000 3 2500 420000 1.619 PSE 4 51 3350 576.8 1.323 2 2350 210000 3 2500 420000 1.619 PNJ 4 51 3423 578 1.322 2 2350 2:10000 3 2500 420000 1.619 OAP .4 02 3425 588.2 1.448 2 2350 2:10000 3 2500 420000 1. 619 DBP 4 03 3425 588.2 1.448 2 2350 2:10000 3 2500 420000 1.619 WAT 4 03-2 3425 588.2 1.448 2 2350 210000 3 2500 420000 1.619 WBT 4 03-2 3425 588.2 1.448 2 2350 210000 3 2500 420000 1.619 CAE 4 04 3425 587.7 1.405 2 2350 2~10000 3 2500 420000 1.619 GAE 4 F 3425 588.5 1.421 2 2350 2:10000 3 2500 420000 1.619 GBE 4 F 3425 588.8 1.421 2 2350 2~10000 3 2500 420000 1.619 NEU 4 F 3425 587.1 1.408 2 2350 210000 3 2500 420000 1.619 NAU 4 F 3425 588 1.42lx 2 . 2350 2~10000 3 2500 420000 1.619 NCll 4---- -*-f 3425 588.5 1.421 2 2350 2:10000 3 2500 420000 1.619 DCP 4 03-2 3427 590.8 1.434 3 2350 2~10000 3 2500 420000 1.619

TABLE 3-2 (cont>>

CRIJICAL AND NONDIMENSIONAL PARAMETERS FOR MESTINGUOUSE-DESIGNED NSSSs Ratfo of Asymptotic Vessel Surge Rite N1111ber Steam NSSS Average toobnt PORVs Safety Va hes to Safety of Generator Powerr Tempo now Setpotnt tag1cilb setpof nt tacactty Valve I.oops (Kit) !Of) (lb/hrx108! Nllllber (psta) n /hr) Number (psfa Cl /hr) Capac tty, Plant Type 210000 2500 420000 1.619 DDP 4 06 3421 500 .. 8 lo434 3 2350 3 TDX 4 04-2 3425 *1588 1 .. 403 2 2350 210000 3 2500 420000 1.619 TCX 4 05 3425 588.,5 10421 2 2350 210000 3 2500 420000 1.619 CCE 4 1>5 3425 58707 10405 2 2350 210000 3 2500 420000 1.619 COE 4 05 3425 58107 1.405 2 2350 210000 3 2600 420000 1.619 OPS 4 f 3425 588.5 1.403 2 2350 210000 3 2500 420000 1.619 OQS 4 f 3425 508 .. 5 1 .. 403 2 2350 210000 3 2500 420000 1.619 lGX 4 £2 3811 593 1.396 2 2350 210000 3* 2600 420000 1.619 w TUX 4 E2 38H 59l 1 .. 396 2 2350 210000 3 2500 420000 1.619

& POJ 4 04 3425 581.,7 1.,405 2 2350 179000 3 2500 420000 1.619 Pf.J 4 05 3425 158707 1 .. 405 2 2350 210000 3 2500 420000 . 1.619 COE 4 05 3425 58707 1.405 2 2350 210000 3 2500 420000 1.619

. SAP 4 f 3425 588.5 lo421. 2 2350 210000. 3 2500 420000 1.619 SCP 4 f 3425 588.5 1.421 2 2350 210000 3 2500 420000 1.619 SfP 4 f 3425 !58805 1.421 2 2350 210000 3 2500 420000 1.619

Table 3-3

-* REFERENCE PLANT FOR WESTINGHOUSE TWO-LOOP PLANTS NSSS Power (MWt) 1520 Thermal Design Flow (gpm) 83700 Reactor Coolant Pressure (psia) 2250 Reactor Coolant Temperature C°F)

Core Outlet 612.2 Vessel Outlet 609.8 Core Average 583. 7 Vessel Average 581.2 Vessel/Core Inlet 552.5 Stearn Generator Outlet 552.5 Stearn Generator Type 44 Steam Temperature (°F) 521.2 Steam Pressure (psia) 821 Steam Flow (10 6 lb/hr total) 6.62

. Feed Temperature C°F) 435.7 Zero Load Temperature (°F) 547 Pressurizer Safety Valves Number 2 Set Points (opening/closing) 2500/2500 Capacity (lb/hr) 288000 Pressurizer Relief Valves Number 2 Set Points (opening/closing) 2350/2350 Capacity (lb/hr) 175000 Ratio of Asymptotic Surge.Rate to SV Capacity l. 563 3-9

Table 3.. 4 REFERENCE PlANT FOR WESTINGHOUSE nfREE*LOOP P1.ANTS MS.SS Powel'" (Milt) 2787 Thermal De.sign F1 cw, ( gpn) 95000 Reactor Coolant Pressure (psi a) 2250 Reactor Cool ant Temperature (°F)

Core Outlet 622.8 Vessel Outlet 620.1 Core Average 591.1 Vessel Average 587.8 Vessel/Core Inlet 55505 Steam Generator Outlet 55505 Steam Generator Type 51 Steam Tenperature c°F} 532.0 Steam Pressure (psia) 900 Steam Flow (106 lb/hr tata1) 12.2 Feed Tenperature c°F) 437 Zero Lead Temperature c°F) 541 Pressu~izer Safety Valves Nunber 3 Set Points (opening/closing) 2500/2500 Capacity Clb/hr) 380000 PressurfzeF Relief Valves Nunber 3 Set Points (opening/c1os1ng} 2350/2350 Capae1 ty ( 1b/hi") 210000 Ratio of Asymptotic Surge Rate to SV Capacity 1.425

.Table 3-5

-* REFERENCE PLANT FOR WESTINGHOUSE FOUR-LOOP PLANTS NSSS. Power (?4'1t) 3425 Thermal Design Flow (gpn) 94400 Reactor Coolant Pressure (psia) 2250 Reactor Coolant Temperature c°F)

.Core Outlet 621.1 Vessel Outlet 617.8 Core Average 5910 l Vessel Average 587.7 Vessel/Core Inlet 557.6 Steam Generator Outlet 557.3 Steam Generator Type 04 Steam Temperature c°F) 543.3 Steam Pressure (psia) 990 Steam Flow (106 lb/hr tota1) 15.13 Feed Temperature c°F) 440 Zero Load Temperature c°F) 557 Pressurizer Safety Valves Number 3 Set Points (opening/closing) 2500/2500 Capacity Ob/hr) 420000 Pressurizer Relief Valves Number 3 Set Points (opening/closing) 2350/2350 Capacity (lb/hr) 210000 Ratio of Asymptotic Surge Rate to SV Capacity 1. 619

  • 3-11

Table 4-1 TRANSIENT RESPONSE PARAMETERS OF OVERPRESSURE EVENTS COMPARED FOR A TYPICAL PLANT (2*-Loop)

Peak Pressure Maximum Valve Rate at Enthalpy

  • Discharge Inlet Safety Valve of Fluid Fluid Pressure Opening Discharge Transient Condition (psia) (psi/sec) (Btu/lbm)

Locked Rotor Steam 2675 236 l123 Loss of Load Steam 2553 73 1126. 7 Loss of Nonna1 Steam 2529 7 1126.1 .

Feedwater Station Steam 2529 7 1126.1 Blackout Rod Ejection

  • Steam 2341 N/A N/A Rod Withdrawal Steam 2504 9 1125. 6 at Power
  • Accident is not nonna11y analyzed for overpressurization.

4-2

Section 4 ME'THODOLOGY TO DETERMINE RANGE OF EXPECTED PORV AHO SAFm VA.YE IHLET COND ITlONS In thfs section the methodology used in detennining the range or inlet fluid con-ditf ons at the PORV and safety valves f s di sc:ussed. Two different methods are used; one for the extended operation of the high pressure injection system and transients that result in steam discharge through safety valves, and another for a main feedline. rupture &ecidento 4e1 LICENSING-TYPE ~SIEHTS The transients that are analy?ed far reload and licensing can be divided into two groups; those that result in steam d1schaf9e and those that result in liquid dis*

charge through PORV and safety val veso 4o1o1 rransients Resultin9 in Steam Discharge Th~ resu1 ts of analyses af overpressure transients that result in steam discharge thl"Ough PORV and safety valves are sumnarfzed fn Table 4m1 foF a typfca1 two~loap p1anto The standard Safety Analysis Report (SAR) type analysis was uti11Ie~ *. listed are the peak pressure and enthalpy of steam being discharged and the pressurization rate at valve opening~ . ,

  • Ca1par1son of the peak pressure for each event indicates that the 11mf t1ng transients for steam di~harge are loss af 1aad and locked ro"Wri.

4o1.1.1 Loss ot Load. ihi$ transient fs ana1yied to make certain that the reactt:ir eoolant and steam generatcrs are not over~

pressurized and the increase of reactor coolant system temperature does not resu1t in Departure from Nucleata Soiling

(DNB) in the core. In addition, fuel temperature and fuel clad strain limit should not be exceeded *

  • The initial power level is assumed to be at maximum allowable value plus 2 percent uncertainty. However, the sensitivity analysis performed shows that the RCS peak pressure for loss of load is relatively insensitive to.the initial conditions of temperature and pressure.

For standard DNB design procedures, the reactor cool ant average temperature corresponds to the initial power level and the pressure is at nominal, including allowance for calibra-tion and instrument errors.

Control *systems are assumed to function only if their opera-tion results in more severe accident results.

Two cases are analyzed, both with and without automatic pres-sure control, to assure that the reactor is protected for both modes of plant operation. Cases are also analyzed for maximum and minimwn reactivity feedback.

To maximize the peak pressure and pressurization rate for loss of load transients for SAR-type analyses, no steam dump is asswned, and minimum values for overal 1 heat transfe.r coeffic-ients are used. For the control system, it is assumed that the pressurizer spray system and heaters are off, and no credit is taken for the rod control system. The safety injec-tion system and auxiliary feed water system are not assumed to operate. The same conservative asswnpti ans are made for a locked rotor incident.

4.1.1.2 Locked Rotor/Loss of Flow. For this accident, two separate procedures are used. One procedure is used for calculating reactor coolant pressure and clad temperature and the other is used to calculate the number of rods in the DNB. For this accident, the reactor coolant pressure is expected to remain

  • below 110 percent of design pressure and the clad temperature 4-3.

must remain below 2700°F. The fnitial power level is assmned to be at maximum allowable value plus 2 percent uncertainty

  • The initial system pressure for RCS pressure calculation is assumed to be a nominal pressure plus allowance far c:al ibra-tian and instrument error. Control systam is assumed ta function only if the1 r operations increases the ~verity of the event..

4.lo2 TFansiefits Resulting fn Liauid Discharge Past analysis indicates that the mast limiting transient resulting fn Hquid discharge through the PORY and safeey valves fs the fee_dline break aceidento Water discharge through safety and relief valves fs predicted during standard SAR analysis of feed1f ne breako 4e1o2~1 Main Feed11ne Ruptureo The purpose of ana1yrfng a feedwater line rupture incident is to ensure that the plant fs main*

tafned fn a safe condition for a range of feed1fne breaks up to and including a break equivalent in area to double-ended J"Upture of the largest water line. The ea1culate4 radioactive llli.ter1a1 release of these events should not exceed the guide-line value of 1_0CFR100. Ta achieve this, the pressure fn the reactor coolant system fs conservatively maintained below 110 percent of the design pressure, and the fuel damage that may occur during the course af the transient must be limited so that the care wf11 remain geometrically intact with no 1oss of ~are cooling capabi11~. Based on sensitivitf analysis, two cases wtffch represent the worst cond'f ti ons are presented 1n the SAR& Bo~h cases assume a daub1e~ended rupture oceurT~

ing downstream of the main feedl'fne check valve with the co~

at 102 percent pot11er. The RCS f~ aw is assumed at thermal .

design flow with temperature and pressure at nominal condi~

tion with allowance for Ci.11bration and instrument error.

Following a main feedline ruptul'"9, steam line pressure and steam generator water level begin to drop. The iew ste~~ 1ine pressure signal initiates a steam 1ine iso1ation signal and a

safety injection signal, which initiates flow of borated water into the RCS.

No reactor control system is assumed to function during the accident unless its function results in a more severe tran-sient. Core nuclear parameters are chosen to maximize the energy input to the coolant.

4.1.2.2 Small Steam Break. This accident is similar to feedline break and considers the event of potential reactor cooldown resulting fran a secondary pipe rupture.

\

This accident is bounded, as far as the rate of liquid

  • discharge through PORVs and safety valves is concerned, by the feedline break analysis.

4.2 EXTENDED HIGH PRESSURE INJECTION EVENTS As discussed earlier, the limiting Extended High Pressure Injection event is spurious SIS actuation at power. This event is analyzed to make sure that the critical heat flux is not exceeded. The pressures calculated in the reactor coolant and main steam system ate below 110 percent of the design pressure, and fuel temperature and fuel clad strain limits are not exceeded. The peak linear

  • heat generation rates are below a value that could cause the fuel centerline to melt. The range of safety and relief valve inlet conditions fo~ this event depends on the maximum safety and relief valve opening setpoi nt, initial core boron concentration, boron worth, and maximum s~fet_x.injection flow rate versus reactor cool ant pressu~e. As* brie~lY.

0 described before, the current design of the safety injection system (SIS) for two-s three-, and four -loop Westinghouse~designed NSSSs vary in layout and philosophy. The initial power is assumed to be at maximum allowable NSSS power plus 2 percent uncertain~. The reactor coolant pressure and average temperature are assumed to be at nominal temperature and pressure including allowance for calibration and instrument errors. Pressurizer water volume corresponds to the prograrmned reactor coolant average temperature, and feedwater temperature corresponds to the initial power level.

The accident is simulated by initiating injection of borated water into each of

  • rea.ctor cool ant cold legs with no direct reactor trip resulting from the SI signal.

4-5

The only protection system assumed avai1ab1e is the low pressurizer pressure reactar trip. The control systems are assumed ta function only if their operation results in more severe accident results. For this accident, no cont1"1)1 system are assumed operable.

T~ cases wfth maxi111111 and lllfnimum boron worth for a given boron cQncentrat1on 9

'ldl1ch yields a high boron lllDrth per ppm fncre.~se fn borcn c:oncentrat"fonll wre analy%8d. Minimum values of overa11 fuel heat transfer ccefficients are used that delay the heat addition to reactor coolant system, whfc:h fn turn increases the primary coolant expansion rate and thus the discharge rate through safety and relief valves.

Maximum f1ow rate is assmed for the safety injection system versus reactor coolant pressure wfth a11 pumps running and 111 lines fnjectingc

. f1owrates presented for this event f nc1ude punp f1ow as we11 as any system Surge expansion which might occur during this eventG 4o3 COLD 0¥£RPRESSURIZATION EVEMTS The range of relief valve inlet conditions for cold overpressure operation fs dictated by several factors: Cold Overpressure Mit'fga-tian System (OMS) r-eHef va1ve.setpo1nt, the existence of a steam bubble in the pressurizer, minimum a11aw~

able systam pressure, and maximu1 relief va1ve setpointe (JllS setpoints are based on consideration of the mast severe~ credible mass fnpyt and heat input events. Analytical assunptions 1nc1ude *pressurization of a water-solid, cold reactor coolant system; staggered valve operation, which takes credit for only single valve operation; and reactor vessel and reactor coolant ..

punp no. 1 seal pressure limit eonstnintso Setting OMS setpoints' on the* above bases provides the system the capabfHty to accoamodate a11 expected tf'ansients and fluid conditions at the valve in1ete The setpoints also depend on acceptable performance of at least one of the PORVse The ma~imum pressure expected to oc:eYF during a cold overpressure incideHt ~t any RCS temperature fs that associate4j with actuation of PORV f2 only, the re1fef va1ve at the higher set~oint. This va1ve setpoint has an upper limit of 2335 psig, 4-6

which corresponds to the normal (high temperature, high pressure) setpoint. PORV 12 operation would nonnally occur during cold overpressurization in the event of

. . *failure of PORV #1, the relief valve at the lo..,r setpofnt.

The CMS setpoint is variable with RC_S temperature, from a low value at low tem-perature to- the upper 2335 psig limit. The highest possible setpoints were deter-mined to provide the plant operator with maximum pressure margin for plant operation during shutdown.

The minimum pressure expected during any cold overpressure incident at any appl i-cable RCS temperature is the reactor cool ant pump no. 1 seal pressure limit used in setpojnt detennination. Theoretically, operating pressure may be established as low as this limit and the minimum temperature in the pressurizer operating with a steam cushion is based on this limitG While setpoints are determined for the conservative situation whereby the RCS is assumed to be in a water-solid configuration, in reality, a steam cushion could exist in the pressurizer via compliance with administrative procedures to amelior-ate the effects of cold overpressurization events.

Depending on the progress*made by the operator in implementing these procedures at the time of a* cold overpressure event, and the duration of the event, the state of the pressurizer could be saturated steam, saturated water, or subcooled water.

The condition of the fluid at the PORV inlet during a cold overpressu~ event can therefore vary fran saturated steam to subcooled water, as discussed in Section 5. 4.

4-7

    • Section 5 EXPECTED FLUID CONDITIONS AT SAFETY & RELIEF VAL VE INLETS The expected fluid conditions at the inlet of PORvs* and safety valves for both steam and liquid discharge are discussed. Normally for SAR-type analyses of over-pressure incident, no credit is taken for the operation of PORVs. Therefore, SAR-type analysis is used to obtain fluid condition at the inlet to the safety valves. To obtain the inlet condition to the PORVs, the above analysis is repeated *with PORVs assumed operational.
  • In no case is credit taken for the pressurizer ~prays.

Two limiting transients, loss of load and locked rotor, are considered for steam discharge. Both accidents are analyzed for each reference plant. The results are presented in tables. For water discharge ~hrough the valves, the extended operation of the high pressure injection system and feedline break events are analyzed. The Extended Operation of High Pressure Injection Event is analyzed for each reference pl ant, whereas ..for feedl i ne break, only results of analyses for plants that have feedl ine break analyses are presented. This is further discussed in appropriate sections.

5.1 FSAR TR.AHSIENTS RESULTING IN STEAM DISCHARGE The FSAR transients that result in steam discharge, locked rotor, and loss of load

/ . '

are selected as the limiting events in* detennining the inlet fluid conditions for ,

PORV and safety valves. The reasons for this selection,* as discussed before, are the peak pressurizer pressures reached during the course of the transients and the rate of pressurization reached.

5. 1.1 Reference Plant -- Two-Loop Group The inlet fluid conditions for PORV and safety valves are presented for locked rotor and loss of load in Table 5-1. The inlet fluid conditions *are also summarized where credit is taken for PORV
  • operation. The limiting transient for steam discharge in two-loop plants is the locked rotor, during which peak reactor coolant pressure 5-1

Table 5-1 VALVE INL.ET CONDITIONS FOR FSAR EVENTS RESULTING IM STEAM DISCHARGE:

Mu1mm Maximum

.Valve .PT"'eSSUMZf!r' Pnssure-Reference Openiny Pnssure (psfa)/ Rate (psia/sec}/

Plant Pressures psi a) *Limiting Event Lfmi ting Event SAF£TY VALVES ONLY 2 .., Loop . 2500 2682/Locked Rotor 240/Loc:ked Rotor 3 "' Leap 2500 2592/loeked Rotor 216/l,oc:ked Rotor 4..,. Loop 2500 2555/loss of Load 144/Locked Rotor SAFm AND RELIEF VM..VES 2 ... Leep 2350 2573/Loc ked Roto,. 202/Lccked Rotor' 3 - Loop 2350 2555/Locked Rotor 200/loc:ked Rotor 4 - laop . 2350 2532/Loss of Load 130/1..ocked Rotar

and maximwn pressurizations rates are observed. A peak pressure of 2682 psia and a pressurization rate of 240 psia/sec is observed when no credit is taken for the operation of PORVs. Assuming the PORVs operable reduces the peak pressure by 4. 1 percent and the rate of pressurization *by 16.5 percent.

5.1.2. Reference Plant -- 3-Loop Group Table 5-1 also presents the results of loss of load and locked rotor analysis for _plants with three loops. As for two-loop plants, the fluid inlet conditions expected at the inlets of the safety valves and PORVs are also included in Table 5-1. Locked rotor can also be considered as the limiting overpressurization transient fo~ three~1oop plants.

The peak pressure reached during the locked rotor accident is 2592 psia and the maximum rate of pressurization is about 216 psia/sec.

When PORVs are assumed operational, a *1*.4 percent and 7.4 percent reduction is peak pressure and maximum rate of pressurization are observed, respectively. It should be noted that the result of analy-sis for the t~-loop reference plant envelops the predicted valve inlet conditions of three-loop reference plant for transients resulting in steam discharge.

5.1.3 Reference Plant -- 4-Loop Group In terms of peak pressurizer pressure, loss of load is the limiting transient for the four-loop plants. However, the rate of pressuriza~

tion is higher for the locked rotor*compared to that for loss of 1oad. The flufd conditions expected at tne inlets of the safety valves and PORVs are presented in Table 5-1.

The peak pressure reached during the loss of load accident for four-loop plants is 2555 psia. The maximum rate of pressurization of the reactor coolant system for loss of load was below that for the locked rotor transient. Hence, the maximum pressurization rate of 14'!

psia/sec from locked rotor analysis for four-loop plants is recorded in Table 5-1. If PORVs are assumed operational, a 0.9 percent

  • reduction in peak pressure and 9.7 percent reduction in the rate of reactor coolatn pressurization is observed.

5-3

5.2 PLANT-SPECIFIC VJJ.VE IMLET CONDITIONS FOR MAIN FEEDLlNE BREAK The fluid conditions at the inlet to the safety valves for feedline rupture acci-dents are sunmarized in Table 5-2 for plants that have feedline break accident analyses. The information presented fn Table 5-2 must be considered in light of the following discussion.

The feedl i ne break analyses fo!9 Beaver Va11 ey _Unit 1 and North Anria Units 1 arid 2 were done in a solllfthat mare conservat1 Ye fashion due to the Cade lfmitationso For these plants, it was ~sumed that the safety valves did nat open until 2575 psia instead of their actual setpoint of 2500 psia. These analyses also overspe:'f fied the 11qu1d enthalpy in the pressurizer. More recent analysis have used better mcdels ta predict valve opening characteristics and f1uid enthalpy.

For a11*cases presented 0 maximum pressurization rates are taken when valves open an water (the valves fn1t1a11y open on steam; ho.wevera the pressu'l"izatf on rate h.

enve1 oped by those presentad for the 1oc:ked rotor and loss of 1oad events). When the pressurizer is filled and begins diseh3~g1ng 11quid9 the pressurization rate is sma11.

For the feed1fne break 9 as mentioned above, the rssults of standard FSAR*analysis are reportedo The range of pressures for Hquid discharge is* frm '2500 ta 2575

-*psi a and entha1 pf es frcm 570 to 742 Btu/1b e Thi$ de ff nes the range of f1 ui d temperatures possible fraa 570°F, which eol"l"esponds to subcoo1ed water a~ 2500 psi a and 570 Btu/lb entha1 py, and OJ3°F 8 which corresponds to saturated water at 2580 psfa and 741.9 Btu/lb enthalpy. The range of the pressurization rates fs fran 1.6 ta 12 psia/sec. The range of surge rates through the pressurizer wheft valves are discharging liquid fs frcm 0.. 6129 to 6.66 ft3/sec (224 GPM ta 2989 GPM).

5.3 .EXTENDED HIGH PR.ESSURE IMJECTXON EVEMTS The limiting Extended High Pressure Injection Event was the spurious activation of the safety injection system at power. This transient f s a Condition II event which, at worst, wi11 resuit 1n ! reactor shutdoWi1 with the p1ant capable of returning to operation. Condition II event~ should net cause mere serious events~

that is, Ccndftiori III or IV events. Other criteria stated fn the USNRC Standarod Review ?1an are stated in Section 4. ihe results of the analyses are prasented in Table 5*3.

  • Table 5-2 SAFETY VALVE INLET CONDITIONS FOR FSAR EVENT RESULTING IN LIQUID DISCHARGE (MAIN FEEl~LINE BREAK)

Maximum Liqu1d Surge Safety Maximum Surg1e Rate Into Range of Liquid Valve Pressurizer. Maximum Pressurizer When Temperatures at Opening Pressure Pressur1 zatfoh Valve Is Passing Valve Inlet (1st or Plant Setpo1nt (ps1a) (ps1a) Rate Cpsia/sec) Liquid (GPM) Subsequent Openings) (Of)

CAF./CBE/ 2500 2507.7 3.5 569.1 615.0 635.1 CCE/CDE PBJ/PCJ DAP/DBP 2500 250L8 5.0 659~3 611.9 622.5 DCP/DDP 2500 2507~7 5e0 .543.1 613.7 631.3 NAH/NCH 2500 2504.9 3.0 275. 1 568.7 584.1 U1 Q

01 TBX/TCX 2500 2503.2 5.0 1109.5 608.2 614.9 TGX/THX 2500 2505.4 6.0 408.2 607.1 609.6 CQL/CRL/ 2500 2504.0 4.0 313.7 620.1 623.4 CSL/CTL SNUPPS 2500 2535.0 12.2 2512.5 613.4 632.7 DMW 2500 2503.7 8.0 224~4 553.8 572.0 VRA/VGB 257~ 2575.0 4.0 507.2 634.5 636.6 CGE 2500 2510.9 6.0 535.9 623.6 644.3 DLW 2575 2575.0 1.7 2010 *.8 644.6 672.0 ALA/APR 2575 2575.0 5.2 2989.2 646.0 672.0

Table 5-2 (Continued)

SAFETY VALVE INLEV ~ONDITIONS FOR fSAR £VENT RESULTING IN llQUID ORSCHARGE (MAIN FEEDLINE BRIEAk)

M1xtm111 Ltqutd Surge Safeey Maximum Surge Rate Into Range of Uqu1d Valve Pressurizer MaxflWll Pressu~~zer When Temperatures at 01>en1ng P&"essure Prea;surh1tfiibn Valve Is Passtng Valve Inlet (1st or Ltguicfl (GPM) Subseguent 02entn9s) (Of)

Plant Set~o1Blt <<2sta) <<2sta) Rate i(!sfa/Hc) 2575 2575 .. 0 3.. 6 1575 ..-4 H6.0 672.0 POR 2575 2575.0 Je4 646.,3 66402 658.0 PGE/PEG/

PSE/PNJ/

TVA 2675 2575e0 1 .. ~ 430 .. ~ 630~8 637.0 WAT AMP No Wateir Dhcil1rge Observed rn 8

Ob

  • Table 5-3 SAFETY &RELIEF VALVE INLET CONDITIONS RESULTING FROM SPURIOUS INITIATION OF HIGH PRESSURE INJECTION AT POWER WHEN VALVES ARE DISCHARGING LIQUID Range of Valve F1uid Maximum Range of Surge Rates
  • Range of Opening State Pressurizer Pressurization When Valve Liquid Temperature Reference Setpoint on Valve Pressure Rates Is Passing LJquid at Valve Plant (psi a) Opening (a) (psi a) (psi/sec} (GPM} Inlet (OF)

SAFETY VALVES 2-Loop No Discharge 3-Loop No Discharge U1 4-Loop 2500 Steam/L1qu1d 2507 0-4 0.0-628.3 567-572 I

RELIEF VALVES 2-Loop No Discharge 3-Loop . 2350 Steam/Liquid 2352 0-12 0.0-781 498-502 4-Loop 2350 Steam/L1qufd 2353 0-4 113.1-1104. l 565-569

a. First/subsequent openings

The fluid conditions at the inlet tc safety valves range from 567° tc 572°r at 2507 ps1a with a maximum discharge rate of 628.3 gpm. No liquid discharge fraa*

the safety valves of the 2* and 3-loop reference plants was observed during the analysis. The fluid conditi.ons at the inlet tc PORVs range from 498°F to 569°F at 2353 psi a with a maximum discharge rate of 1104a 1 gpm. In this cas*e no liquid discharge fT'CID the PORYs of the 2-laop reference plant is observed duY"ing the fntertal that the transient was analyzed.

In general valves open on steam and no liquid. discharge is observed unt11 the pressurizer becCllles water solid. .This is plant dependent and can va'f"'J anywhere frcm 20 minutes to mare than six hcurs.

So4 Pt.ANT-SPECIFIC VALVE IHI.ET CONDITIONS RESULTING FROM COLD OV£RPR£.SSURIZATION EVEHTS 0

Setpo1nts for the cold overpressurization mitigation system are cafiservat1ve1y determined ta accomodate the rapid PM!$Surization rates (up to 100 psi/see) pro..,

dueed by cold overpressure transients (Section 2e3) du~ing watel'"-so11d, 1ow temperature operation of the reactor coolant systemo In praetiee, ho~ver, f1uid candftfons at the relief valve inlet are not restricted to low temperature, subc cooled watero A variable fluid condft1on (steam or water) and temperature Csatur~

ated to subcooled) at the va1ve. 1n1et 1s possible due to administrative require=

ments for maintaining a pM!ssurfzer steam b'ubb1e during low temperature operation~

-lllhen pressure excursions due to cold overpreS$i&r1zatfon events are a possibility (Seeticn 4o3}o The maximum range of potential cold overpressure fluid conditions at the relief valve 1n1et 9 covering all Westinghouse plants analyzed ta date, may be inferred frcm Figure SG)l. These plants include: Cananche Peak Units 1 and 2, SNUPPS 9 Sequoyah Un1't$ 1 and 2, 'Watts Bar Units 1 and 2p South Texas Units 1 and 2, and Byran/Braidwood Units 1 and 2. A description of the indexed curves used ta define
  • the range Qf potent1a1 fluid conditions is presented be1ow.

Legend Aca1fcab1e To F1gul"'e 5°1 Index Oescriotion 1 Locus of maximum primary system preS$Ures deve1oped fol1cwirig ?C!W

~2 o~eration (limiting condition/ watarQsolid RCS)

2000 Cl 2

i 0_,

..J 0

u..

w 160().

a:

)

~ (,:,

W-a:en

""" Cl.

== ,..:

.wz 1-W 1200 en>

>w enw Ca:

w ::l I- en ww CJ "'

CL. a:

x Cl.

w a: 800

Ew
) >
Eo

-c x ..J

<o 1

lE CJ 400 3

o--~~ ........~~~--~~--~~~--~~~........~~_..~~~-

o 100 200 300 400 500 600 FLUID TEMPERATURE, OF

  • Figure 5-1. Potential Cold Overpressure Fluid Conditions at the Relief Valve Inlet 5-9

Index Description 2 Potential steam/saturated liquid conditions fn pressurizer per recomnended administrative procedure 3 M1n1111.1111 operating pressure 11m1t ta ensure reactor coolant pump Noo 1 Seal integrft\f 4 Maximum relief valve setpofnt based an high temperature operation It should be noted that although possible 9 11quid discharge at temperatures tower than 100°r aM! extreae1y un1fke1y because of the Hm'fted time the plant is c:ald and fJr a condition capable of being pressurized (f Qe., the RV head fs off or the*

RCS f s open for maintenance)o

SECTION 6 REFERENCES Tables 6-1 and 6-2 list the sources of data and the date that the analysis was perfonned. For FSAR events resulting in steam discharge (except feed1ine break),

and spurious actuation of high pressure safety injection system, reanalysis was perfonned for this report. The* feedline break analysis was performed as part of the FSAR analysis as required by Regulatory Guide 1.70 *

  • 6-1

TABLE 5-1 DATA SOURCE FOR FSAR EVENTS RESULTING IN STC:.4M OISCHARGE AND SPURIOUS ACTUATION OF TifE HIGil PRESSURE INJECTION SYSTEM Date Reanalysis Based Analysis Reference P1ant On Done 4-1.ocp FSAR 1981 3-Lcop Cycle 3 Reload 1981 z..Lcop FSAR 1981

TABLE 6-2 DATA SOURCES FOR FEEDLINE BREAK DATE ANALYSIS DONE ANALYSIS PLANT FOR DONE CA£/CBE/ FSAR 1980 CCE/CDE PBJ/PCJ DAP/DBP FSAR 1979 DCP/DOP FSAR 1979 NAH/NCH FSAR 1981 TBX/TCX FSAR 1976 TGX/11iX FSAR 1977 CQL/CRL/ FSAR 1979 CSL/CTL SNUPPS FSAR 1980 OMW FSAR 1978 VRA/VGB FSAR 1976 CGE FSAR 1978 DUI FSAR 1974 ALA/APR FSAR 1974 POR FSAR 1974 PGE/PEG FSAR 1974 PSE/PNJ TVA WAT FSAR 1977 AMP FSAR 1978

  • 6-3
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