ML18026A520

Forwards Revision to FSAR Chapter 15 Due to Reanalysis of Transients Using Odyn Code.Revision Will Be Incorporated in Future FSAR Amend.Review Completes Action on SER Outstanding Issues 16 & 66

ML18026A520
Person / Time
Site:	Susquehanna
Issue date:	06/10/1981
From:	Curtis N PENNSYLVANIA POWER & LIGHT CO.
To:	Schwencer A Office of Nuclear Reactor Regulation
References
PLA-801, NUDOCS 8106110408
Download: ML18026A520 (55)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

I ACCESSION NBR;8106110408 DOC.DATE: 81/06/10 NOTARIZED:

NO FACIL:50 387 Susquehanna Steam Electric Stationi Unit 1< Pennsylva 50 388 Susquehanna Steam Electric Station< Unit 2P Pennsylva AUTH'AME AUTHOR AFFII IATION CURTI SE N ~ W ~

Pennsylvania Power 8, Light Co, RECIP ~ NAME RECIPIKNT AFFILIATION SCHWENCER,AD Licensing Branch 2

DOCKET' 050.0.0387 M3).038 8>>

SUBJECT:

Forwards revision to FSAR Chapter 15 due to reanalysis of transients using ODYN codex'evision will be incorporated in future. FSAR amend. Review completes action on SER Outstanding Issues ib 8,

66

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Send ICE 3 copies FSAR 8 all amends' cy'.BWR LRG P~(CRIB) 05000387 05000388 RECIPIENT ID COOK/NAME.

ACTION:

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TWO NORTH NINTH STREET, ALLENTOWN, PA. I 8101 NORMAN W. CURTIS Vice presirtent-Engineering 3 Constnrctlon-Nuclear 770.5381 June 10, 1981 PHON ~

(2 I 5) 775.5 I 5 I OL

('=.,',@/Pp '

e Q)/I Mr. A. Schwencer, Chief Licensing Sranch No.

2 Division of Licensing U.S. Nuclear Regualtory Commission Washington, D.C.

20555 Docket Nos.

50-387 50-388 SUS(UEHANNA STEAM ELECTRIC STATION SER OUTSTANDING ISSUES 816 and 66 ER100450 FILE 841-2 PLA-801

Dear Mr. Schwencer:

Attached is a revision to Chapter 15 due to the reanalysis of transients using the ODYN code.

This revision will be incorporated in a future FSAR amendment.

This completes our action SER Outstanding Issues 16 and 66.

Very truly yours, Vice President-Engineering and.,Construction-Nuclear cc:

R.M. Stark gOOl S 1 A

8 PRNNSYLVANIA POWSR TLIGHT C,OMPANY QIANA l 0

The following sections in th Susquehanna FSAR the ODYN implementation:

are affected by o

4.4.1.3:

Requirements for Steady State Conditions Table 4.4-1 o

5.2.2:

Overpressure Protection "Table 5.2-9A

• Fig.

5.2-1A Fig.

5.2-4 ',

Fig.

5.2-5 o

15.0:

Accident Analyses Table 15.0-1 Table 15.0-2 "Table.15.0-5

• Fig.

15.0-3 o

15.1,.2:

Feedwater Controller Failure Haxinam Demand Table 15.1-3 Fig.

15.1-3 o

15.1.7:

References o

15.2.2:

Generator Load Rejection Tab1e 15.2-1 Table 15.2-2 Fi g.

15.2-1 Fig.

15.2-2 o

15.2.3:

Turbine Trip (Bypass - Off}

Table 15.2-4 Fig.

15.2-4 Besides the above sections, any section or table in the FSAR that explicitly mentions the value of operating, CPR limit or NCPR has been revised or deleted from the text.

• New tables or figures that are added.

SSES-FSA R

'YAQ LE Trl. R."~AL AND HYDRAULIC D-SLG.'J CHARACTi:RiSTZCS OF THE R~EAC7OR CORE Reference design thermal output, Poser Leve1. for Engineered Safety F atures, tlat Steam flow rate, at 3d30F final feedvater temperature, miLlion" lh/hc 3293 '0 3040 0

13<<%cd Coco coo1ant floe rate, mxllions. lb/hc Feedva te" flo~ race, millions 3.b/hc Syst: em pressu"e, nomina1. in steam

Come, psia System

pressure, nominal co e design, psia

%00.0 13 4v

$ 020 0

303i 0

Coolant sacu"ation temperature ac core es xqn pc~ssu cP o F Average po~ec density, x.H/Liter Maximum. Linea" iieat Generation:

Ra-e, kA/ft.

Average Linear Heat Generation Rate, kN/ft 5~9 0

48 7

l3 Core totaL heat transfer area, ft~

Average heat flux, Dtu/hr-sq ft.

Maximum Heat Flux, Btu/hc-sg ft Design Op@rating

."minimum Critical Po~er Ratio Core inlet enthalpy at 393~F'FV ~,

Btu/1!~

Core inlet tempectuce, at 383~F

FF47,

~F

'?0 3, 700 36l 600 s~C TqSK ~~0~

pug F1&. tR 0-3 521.

8 Core maximum exit voids within assem1>lies, Coze average void fraction, active coolant 70~0 0 410 Rev.

0 7/78

SS ES-ES AB The rated capacity of the pcessuce relieving devices shalL be sufficient to prevent a cise in pcessuce within. the protected vessel of zoc than 110% of the desiqn pcessuce (1

LO x L250 psjg.

1375 psiq) foc events defined in Subsection 0

3 l Pull account is taken af the pcessuce drop on noth the inlet and, dischacqe sides of the valves.

All combination safety'relief vai.ves dischacqe into the suppcession pool through,

a. discharge pi:pe fco:. each valve which is desiqned to achieve sonic f,lov, conditions throuqh the valve; thus pcovidinq'la~ independence zo d ischarqe' ipi nq losses Table 5 2-6 lists the systems which could ini"ate during. the desiqn basis overpressure event 5

2 2

2 Des~i. n Evaluation, 5

2 2. l nethon of K~n~lsis To design the pressure protection for the nuclea-" boiler system extensive analytical models cepcesentinq ali. essential dynamic characteristics of the system are simulated on a, large computing facility.

These models include the hydrodynamics of the floe loop, the ceactox kinetics the thermal chaxa.ctecistics.

oC, the fuel and its "ransfer of heat to the coolant, and. all the principal controller features such as feedvatec fLos recirculation flov, reactor Mater level,. pcessu-e and l.oad demand These are cepresented with all their principal nonlinear features in models that have evolved through extensive experien e

and favorable compacison of apalysis vith actuaL BVR test data.

A detailed description of ~. model is documented zn Reference

5. 2-1 Sa fetygrelief valves are simula"ed in a nonlinear representation, and the model thereby allow fuLL investigation of the various valve response

tines, valve capacities and actuation setpoints that are available in applicabLe hard~a=e systems.

The typical valve characteristic as modeled is sho~n in Pigure 5 2-2 foc the spring mode of operation The associated bypass turbine control valve, and main steam isolation valve characteristics ace ai.so simulated in the model 5

2 2 2~2S -tee see~i.

n A pacametcic study was conducted to determine the required steam flow capacity of the afety/relief valves based on the foLLo'~ing assumptions 5.2-0

SS ES-FS AB 5

2 2

2 2

1 Operating Conditions (1) onecatinq po~er

= 3439 N4t (104 4% of nucl,ear Boiler ra ted power),

(2) v..ssel dome pcessuce

( 1020 psig, a,nd

())

steamf lov

=

14 153 x

10~ lb/hc (105~ of nuclear boiler catei steamfloM)

These conditions are the most severe because maximum stcred.

enerqy exists at these conditions At love po~er conditions the transients would be less sevece.,

5 2

2 2 2.2 Transients The overpressuce pcotec'tion system must accor odate th most severe pressurization transient There a

e tvo maj"r transients, the closure of all main steamline isolation valves and a

turbine/qeneratoc trip vith a, coincident or the turbine steam bypass system valves tha t represent the most severe abnormal operational transient resulting in a nuclear syste pressure rise.

The evaluation of transient behavio-with final plant configuration has sho~n that the isolation valve closure is sliqhtlv mere severe chen ccedit is taken only for 'indirect derived

sccams, therefore, it is used as the overpressure protection basis event and shovn in Figure 5 2-l Table 5 2-9 lists the sequence of events of the vacious systems assumed to opecate ducinq the main steam line. isolation closure vith. flux serac event

~ g~g raSults

$.r W Zauve.

'tv +q.8'.>-tA; ca~5~ q ~ks

'~

+J,lc S'.w-RA 5

2 2

2 2

3 Scram f

(1) c"ae reactivity curve Fiqure 5 2-3

~~ M~feau+i S;>-lA +~~

QbYPJ (2) central rod drive scram motion Figure 5 2-3 5.2.2 2

2 0

Safety/Relief Valve Transient Analysis S oeci ficat ions (l) valve qrou ps:

spcinq-action safety 'mode 5 groups (2) pressuce setpoint (maximum safety limit):

5 2-5

SSES-PSAR sprinq-action safety mode 1177-1217 psig The set pcints, ace assumed at a conservatively high leveL above the nominal set points

. This is to account for initial set point errors and any instrument set point dcift that might occur dlJring operation Typically the assumed setpoints in the analysis are 1

to 2% above the actual nominal set points Highly conservative safety/relief valve response characteristics ace a1so assumed 5-2-2

2. 2 5

Safety Valve Ca acity Sizing of the safety valve capacity is based on establishing an adequate margin, from the peak vessel pcessuce to the vessel coDe limit (L375 psiq) in response to the reference transients Subsection 5.2 2

2 2.2

5. 2.2 2

3 Evaluation of Results 5

2 2. 2 3

3.

Sa~fet valve Caoacatv The required safety valve capacity is determined by anaLyzing the pressure rise from a a1SIV closure with flux scram transient The plant is assumed, to be operatinq at the tu bine-generator design conditions at a

maximum vessel dome pressure of 1020 psig The analysis hypothetically assumes the Zailuce of the direct isolation valve position scram The reactoc is shut down by the backup. indirect, high neu con flux scram For the analysis, the spring-action safety set points are assumed to be in the range of 1177 to 1217 psig The analysis indicates that the design Valve capacity is capable of maintaining adequate margin below the peak AStfE code allowable.pressure in the nuclear system (1375 psig)

Figure 5 2-1 showy curves produced by this analysis The sequence of events assumed in thCscanalys&.

Mas investiqated to meet code requirements and to evaluate the pressure relief system exclusively Under the General Requirements for Protection Against Ovecpressuce as given in Section XXZ of the ASME Boiler anct Pressure Vessel Code, ccedit can be allowed for a scram from the reactor protection system Xn addition, credit is also taken for the protective circuits which are indirectly derived when determining the required safety valve capacity The backup reactor high neutcon flux scram is conservatively applied.

as a

desiqn basis in determininq the required capacity oZ the pressu e

relievinq safety valves.

Application of the direct: position sccams in the design basis could be used since they qua.lify as acceptable pcessuce pcotection devices when determininq the 5 2-6

SS ES-FS AB required safety valve capacity of nuclear vessels under the provisions of the ASIDE code.

The parametric relationship between peak vessel (bottom)

Pressure and safety valve capacity for the llSIV transient with high flux and position trip scram is described in Pigure 5 2-4.

Also shown in Figure 5. 2-4 is the parametric relationship between peak vessel (bottom) pressure and safety valve capacity for the turbine trip with a coincident closure of the turbine bypass valves and direct scram, which is the most severe transient wh..n direct scram is considered.

Pressures sho~n for flux scram will result only with multiple failure in the redundant direct scram system.

The time response of the vessel pressure to the NSXV transient with flux scram and the turbine trip with a coincident closure of the turbine bypass valves and di ect 1.s illustrated in Piqure 5.2-5..

sho~s. that the pressure at the vessel bottom exceeds 1250 psig for less than 6 seconds which is not lonq enouqh to transfer any appreciable amount of heat into the vessel metal which was at a temperature well below 550~P at the start of the transient.

CSYQ resu.ltS a~ <ho shag> ~ ~eric

'ft j+<45-T4.

+ol c cA'fc445jw

)Lad gl ccLIct5 ~ Nor c s~~

luau/ i t N

+4.

A~teoat le pM~u,~

lsd',0 op ~ ~AY. Prese,t syst~

5. 2.2

2. 3. 2 Pressure Drop in Inlet and Discharge Pressure drop on the pipinq from the reactor vessel to the valves is taken into account in calculating the maximum vessel pressures.

Pressure drop in the discharge piping to the suppression pool is limited by proper discharqe line sizing to prevent backpressure on each safety/relieE valve from exceeding 40% of the valve inlet pressure, thus assuring choked fLow in the valve orifice and no reduction of valve capacity due to the discharqe pipinq.

Each safety/relief valve has its own separate discharge line.

Figure 5.1-3 is the PGID for the.Nuclear Boiler System including pressure-relieving devices S. 2 2. 4 ggu igment'nd Com~onen~tescr~it ion 5~2. -4 i De cliotion The nuclear pressure -elief system consists of safety/relief valves located on the main steam. lines between the reactor

5. 2-7

SSES ZSAR 5 2.5.7 Safest Interfaces The Balance of Plant-GE Nuclear Steam Supply. System safety interfaces for the Leak Detection system are the signals from the monitored balance of plant equipment and systems which are part of the nuclear system. process

barrier, and associated wiring and cable lying outside the Nuclear Steam Supply System Equipment These balance of plant systems and equipment include the main steam line tunnel.,

the safety'/relief valves, and the turbine build in@ s um ps.

5 2.5.8 Testin and Calibration Provisions for Testing and Calibration of the leak detection system. is,. covered in Chapter 14.

5 2.6 References

5. 2-1 5.2-2

5. 2-3

5. 2-0 5 2-5 R'. Linford, >>Anal.ytical ilethods of Pl.an t Transient Evaluation for the General Electric Boiling Mater Reactor,"

HED0-10802, April 1973..

J.H. Skarpelos and J.M.

Bagg,

>>Chloride Control in BMR Coo3ants,,>>

June

1973, NZD0-10899.

M.I.'il'liams, Corrosion, Vol 13, 1957, p-539t GEAP-5620, Failure Behavior in ASTN A106B Pipes Containing Axial Through-Mall. Plows, by N

B Reynolds April, 1968.

>>Investigation and Evaluation of Cracking in.

Austenitic Stainless Steel Piping of Boiling Mater Reactor Plants,"

NiJREG-76/067, HRC/PCSG, hated October 1975.

KNX.>

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SS ES-FSAH core.would not be expected to experience boiling transition (see Reference 15.0-2)

This criterion is met by demonstrating that

,incidents cf moderate frequency do not result in a minimal critical power ratio (HCPR) less than 1.06.

'"'P zJ /

tdB<<&Ms@

The ncpB Qusinq siqnificant ahnnnnal events is calculated using a transient core heat transfer analysis compute<

program The computer program is based on a multinode single channel thermal-hydraulic model which zequires simu1taneous solut'oa of the partial differential equations for the

, conservation of mass,

enezgy, and, momentum in the bundle

and, which accouats for axial variation in'ower generation The pzimary inputs to the model: include a physical description of the bundle and channel inlet flow and enthalpy, pzessuze and power generation, as functions of time A

et iled des ot on ntI the analytical

model, may 15p 0-2 be found in

,hp V

T V

'I/ I pt I

q yj For situations. in which fuel damage is sustained the event of damage is determined by correlatinq fuel energy content cladding temperature, fuel rod internal pressure, and cladding mechanical characteristics.

These correlations are substantiated by fuel rod failure tests aad are discussed in Section 4.4 and Section 6

3 15 0

3 3

2 In ut Parameters and Initial Conditions for Anal zed Events Xn general the events analyzed within this section have values, foz input parameters and initial conditions as specified in Table 15.0-2 Analyses which assume data inputs different than these values are designated accordingly in the appropriate event dz.scusszca The analysis basis for most of the transient safety analyses is the thermal power at rated core flow (1005) corresponding to 105~

Nuclear Boi3.er Rated steam flow.

This operating point is the 1.0-8

IHSERT 8 Determination of the steady-state operating limit is accomalished as follows:

1)

The change in the critical po~er ratio (ACPR) which would result.in the safety limit CPR (1.06) being reached, is calculated for each event.

These values are sho~n in Table 15 0-l 2)

The hCPR value is then added to the safety limit CPR value (1 06) to result in the event based RCPR except for events whose ACPR is calculated using ODYH.

3)

For events whose bCPR is determined by OQYB (all rapid pr s-surization events) the event based HCPR is determin d in conjunction sith correction factors, the ACPR and the safety limit CPR.

These correction factors are explained in detail in Section 3/4.2.3 of the Technical Specifications.

These results are given in Table f+Q-9~4 Figure 15 0-3 for the limiting transients.

The operating limit HCPR is the maximum value of tlie.

event RCPRs calculated from the transient analysis The maximum calculat d transient, MCPR is depicted by the solid line in Figure 15.0-3.

Haintaining the CPR operating limit at or above this operating limit assures that the safety limit CPR of 1.06 is never violated rf/626 2/4/81

TABLE 15.0"I RESULTS SUHHARY OF TRANSIENT EVENTS Page I of 3 ll Subsection I.D.

15.1 Figurc I.D.

D tutti tith DECREASE EN CORE COOLANT TEHPENATUNE Haximum Neutron F)ux HM Hax imum Dome Pressure

~tt Haximum Vessel Pr<<ssurc

~s~i Haximum Steam Line Prcssure Haximum Core Av<<rage Surface IIc at Flux X of Initial jkCPR Frequency

~tte r

Duration-oi'lowdovn sec Duration of Bioudown Nu. of Valves 1st Blou-doun 15.1

~ I 15.1.2 15.1-.2 15.1-3 Loss of Fecduat<<r !looter, Hanual Flov Control pc<<duster Cntl Failure, Hax Demand 119.1 1023 IOGI 1004

~lS,R EISA EISA mf'12.8 IIS;4 15.1.3 15.1.4 15.1.G 15.2 15.1-4 Pressure Regulator Fail Open Inadvertent Opening of Safety or Rclicf Valve RIIR Shutdoun Cooling Hal-function Decreasing Temp INCREASE IN REACTOR PRESSURE 103.5 1092 1123 Sce Text Scc Text 1092 100.2

~>

0

~r>, 0 15.2

~ I Pressure Regulator Fail - Closed Sco Subsoctions 15,2,2 aod 15.2.3 with By Pass on 15,2

~ 2 15.2.2 l5.2.3 15.2

~ 3 15.2.4 15'-1 15 ~ 2-2 Generator Load R<<)ection, Bypass. On, Generator Load Rc]ection, Bypass.Off, 2BII 7 A%.6 gg.S I

15,2-3 Turbine Trip, Bypass-00

, 15.2-4 Turbine Trip, llypass-Off 167,2 446.8 15,2-.5 Inadvertent HSIV Closure l63,6 1151 IIB7 114G I I l'1

~ l IS3 I LB&

I MS llS'I 4199 AGO'1/I3 1167 1132

~

I l85

+IS IISl

&7t 3-I90 4I 59 lo l.7 JO~

101,4 IISIS

$065 100.2 (ai C>,19 O 6'l 0.17 ~'>

kr09-

<6,o f 16 l6 f1 7 Kcv, 17, 9/BO

TABLE l5.0-1 (cont'd)

Pape 2 of 3 Subsection I.D.

15.2 '

15.2.6 15.2.6 15.2.7 Figure i.D.

~0tscr'io 15.2-6 Loss of Condenser Vacuum Haximma Xeutron Flux XBR 167.5

15. 2-7 Loss of Auxiliary Pover Transformer 104.5 15.2-8 Loss of All Grid Connections 107.2 15.2-9 Loss of All Fecdvater Flov 103.8 Haximum Dome Pressure

~i 1140 1145 Haxianm Vessel Prcssure 1165 1160 1094 1105 114io

~

1161 Ha ximum" Steam Line Prcssure

~~sl 1131 1140 Haximtua Core Averapc Suf face Hest Flux

$ of Initial 101.3 100. I.

1094 100.1 1130 100.1 HCPR 00) o,Oo) o 0)

Frequency C~at e!~or Duration-of Blovdovn sec 20 16 l5 13 17 Duration of Blovdovn Xo. of Valves 1st Blov-dovn l7 15.2.8 15.2.9 15.3 15.3.1 "15.3.1 15.3.2 15.3-1 15.3-2 Fecdvater Piping Break Failure of IUIR Shutdovn Cooling DECREASE IX RJACTOR COOLAXT SYSTEH FLOM RATE Trip of One Recirculation Pump Hotor Trip of Doth Recirculation Pump Hoturs Recirculation Flov Control Fa ilure Dc cress ing F1ov Sce 15.6 '

Sec Text 103.5 1113 Seo 15.3.1 103.6

. 1015 1053 998 1127 (I)

~ o.t) 100.0 n) 1109 loo.l

~0 10 28 17 15.3.3 15,3.4 15.4 15i3"3 Seizure of One Recirculation Pump 103eZ Recirc,pump Shaft Brcak See 15.3,3 RFACTIVITY AND POVER niSTKraDTiOX AXOHALIES 1126 0) 1137 1120 100.4 Pl.lo 13

[ii

~

15.4.1.1 RME Refueling See Text Rcvi 17 y 9/80

TABIE 15.0-1 (cont'd)

Pago 3 of 3

Subscctioa I.D.

Figure i.0.

i~iscr'ion Maximum Haut.ron Flux M BR Haximum Dome Pressure

~SL Haximum Vessel Prcssure

~si tlaximum Steam Linc Pressure

~s~i Haximum Core Average Surface Iieet Flur.

'X of Initial acpa" ~

Frequency C~Rt I

Durst'ion of Blovdovn sec Duration of Dlovdovn Ho. of Valves 1st Dlov-dovn 15.4.1.2 15.4i.2 15.4.3 RVE - Startup RVE - At Pover Control Rod ttisoperation Sec Text Sec Text See Subscct.iona 15.4.1 and )5.4.2 15.4.4 15.4.5 15.4-6 15,4-7 Recirculatioa Flov Control Failure - Iacrcasiag Flov 264,6 982 1008 Startup of Idle Recirculation 323.4 973 988 Loop 967 973 34.

130.3 0

I IS 0

) lS 15.4.7 15.5 ttisplaced Bundle Accident INCREASED IH REACTOR COOIAHT INVENTORY See Text 15.5.1

. 15.5.3 11.4 DMR Transients See appropriate Events in Sections 15.1 and 15.2 15.5-1 Inadvertent HPCI Pump Start 118.2 1023 1061 1004 fa - incidents of moderate frequency b - infrequent incidents c - limiting faults

~cpR.

basd on ~

inlk\\cg el'S

~hi<~ /ie~" >

4 0) equi iiAA

(>)

OVW res~lts

~lo4~t e>j~s'b~~t

(~)

~hese P.Ms

<W H.4<

Ruv, 17, 9(00

SSES-PSAR TABLE 15.0-2 INPUT PARAMETERS AND INITIALCONDITIONS FO>

T NA NSX E NTS 2

Thermal Po~er Level MQt Marran ted Value (HBR)

Analysis Value Steam Flo~, lbs per hr Analysis Value 14.153 x 10~

6. Vessel Dome Pressure, psig Vessel Core Pressure, psig 8

Turbine Bypass Capacity, NBH

9. Core Coola nt Inlet En thai py, Btu per lb

10. Turbine Inlet Pressure, psig=

ll Fuel Lattice-.

3. Core Plow, lbs per hr Feedvater Flo~ Pate, lb per sec Analysis Value

5. Feedvater Temperature, oP 3293 3439 (104 4>

(105 ~

NBR) 100 0 x l06 3921 386 9

1020 0

1030 0

25 521 1

960-0 Gx8 15 16 MCPB Safety Limit Doppler Coefficient f-,) 4/~P Hominal EOC-1 Analysis Data

12. Core Average Gap Conductance, Btu/sec-f t~-~F 13 Core Leakage

Flov,

14. Bequired MCPB Operating Limit 9 85

~~ ~+e g.Q-5 F';gw~a ~.o-2 1 06 0 2255 0 2142 17 Void Coef ficient (-) C/5 Rated Voids Nominal EOC-1 Analysis Data for Poser

'Increase Events Analysis Data for Poser Decrease Fvents 7 48 12 0

6 61 18.

Core Average Rated Void

Praction, 40 74

SSBS-PSAR TKBLP 15 O-2 ~Continu~ag Page 2

1 9. Sera n R ea cti vity,

S kk Analysis Da ta 20 Control Bod. Drive Speedr Position versus tine

21. Jet Pump Ratio, 22 Safety/Relief Valve Capacity',

5 HBR 1091 psig manufacturer Quantity Installed Figure l5 0-2 Figure 15 0-2 1 84 99 0

CRQS BX 16 23 Relief. Function Delay, seconds 20 Relief Function Response, seconds 25 Set.

P oints for Saf et } ReJ.ier Va ives, psig 26 Number of Valve Groupings Simulated 27 High Flux Trip,

~

NBR Analysis set. point (120 x 1.000),

NBR'8 High Pressure Scram Set Point, psig 0

0 15 tt tG r

1 f two 5

125 3

1,071.

tt~

~r r

~ t5'o 29 Vessel Level Trips, Inches Above

(+)

Separator Skirt Bottom Level 8 (L8), inches Level 0 (LQ) inches Level 3 (L3), inches LeveL 2 (L2), inches 4

30 h PRH Heutron Plux Scram Set Point

à "88 Below (;-)

~ +54

+30

+3.2 5

-38 125 0

3l

. Recirculation Pump Trip Delay, Seconds 0

3.75 32 Recirculation Pump Trip Inertia for Analysis seconds+

S The inertia tine constant is defined by the expression:

2mJ n t

o gT Mhere t Jo n

inertia tine ccnstaat (Sec) pump motor inertia (1 b-ft 2) ra ted pump speed (rps) gra vitat ion+l co nstan t (ft/sec 2)

REV. 11, 7(79

SSFS-FSAR TABLE 1S 0-2 /Con~tinned Page 3

Tp pump shaf t torque (1b-ft)

PnicAAmfers ~, ~

~ego'f en(y QDYAl ~t zen a~

.~l ~,'I a~m( i ~r iAA Cy~l~

C~J;e~

REV. ll, 7/79

NUCLEAR EN ER GY BUSINESS GROUP GiBF.RAL ';f'=

F.LFCTRIC REV SH NQ.

TABLE 15.0- 5 RE UIPED OPERATING LIMIT CPP, YALUES Pressurization Events:

CPP. 30Dtion Aj~

CPR (0 tion B

• Load Rejection Without Bypass Turbine Trip Without Bypass Feedwater Controller Failure Load Rejection 30*%

1.2S 1.22

1. 21

l. 20 24*~

1.14 Hon-Pressurization Events:

CPR Rod Withdrawal Error Loss of Feedwater Heater 1.18 Includes adjustment factors as speci ied in the NRC safety evaluation report on ODYH, HEDO-24154. and HEDE-24154-P

• " Required OLCPR using Option A and neglecting infrequent category of the turbine-generator trip events with bypass failure

• • • Required OLCPR using Option A with reclassification of the turbine-generator trip events with bypass failure.

• • • • Required OLCPR Using Option B.

15. 0-20

/55 RW4%0f PP&NTAVS eRq Ci~lr BPF4P A@V. OPm, infra

/35

/.SO

/30 l,a5

'o

/2S

~ ~

I

/2o

/go

~ <

/i~

l/~

Z~= cg sea&

M<D

~/g. /<.0 3

~~wINy~ cpm4r~

epg. uwr y~, ~c~~

sumo

SSZS-FSAB The increased core inlet subcooling aids thermal margins~

and, a

sna3.'ler pover increase makes this event less severe han the manual flov control case given belov Suclear syste~

pressure Goes not change and consequently the reactor coolant pressure boundary is not threatened.

This transient is i'lnstrated in Figure 15 1-3.

Xn nanua3.

node, no compensation is provided by core glov and. thus paver inc ea s is g ea r

han in t e automatic node

'jessal steam flov increases and the initial system pressure increase is slight3.y larger.

Peak heat flux is ll '5 of its initio.3. value and.

peak fu 1 center tenperatuce increases 108~P'he increased core inlet subcooling aids core thermal margins.

Therefore,, the design basis is satisfied the transient responses od the key plant va iabies Eor this node of operation are shovn in Pigure, 1.5 1-2.

This transient is less severe from lovec initial poser levels for tvo reasons=

(1) lover initial pover Levels vill have'nitial r4 values greater than h

Linitiag, initial value assumed and (2) the magnitude of the pover rise decreases vith lover initial pover conditions Theref ore, transients from Lover pc ver levels vi11 -be 3.ess se ve re.

1 i 3

e Cons'de" atior s op pncerta't.'es Xmportant factors

{s~!ch as reactivity coefficient, scraa characteristics, nagnitude of the.feedvater ten,ecature change) are assuned to be at the vo st configuration so that any deviations seen in the actuaL plant operation reduce the severity of the event 15 L.l 0 Bar iec Performance As noted above. and shovn in Figures

15. L-L and 15 1-2 the ccnse-guences of this event do not result in any temperature or pressure transient in excess of the criteria foc vhich the fuel pressure vessel or containment are designed; therefore these barriers maintain their integrity and function as designed Rev. Ll, 7/79

15. 1-4 a P

~ r

~

SSKS-PSAB (2)

Svitch the feedvater controller from auto to manual'ontrol in ocder to try to regain a cocrect output s iqna 1.

(3)

Identify causes of the failure and report all key plant parameters durinq the event Svsteas Operation In order to pcopecly simulate the expected sequence of events the analysis of this event assumes normal functioning of plan<

instrumentation and controls,, plant protection and reactor pcotection systems Important system operational actions for this event are hiqh water level trippinq of the main tucbine turbine stop valve scram trip initiation, recirculation.

punp trip (BPT). and lov water level initiation of the Reactor Core Isolation Cooling System (RCICS) and the High Pressure Coolant Xnjection System (HPCXS) to maintain long-term wa"e=- level control following tcipping of feedvatec Dumps 15.1.2-2-3 The Effeet of Single Failures an~dO era+or E rors In Table 15 1-3 the. first sensed event to initiate corrective action to the transient is the vessel high water level (L8) trip Nul iple level. sensors are used to sense and. detect vhen the vater level reaches the LS set point.

At this. point in ehe logic a single failure vill not initiate oc pcevent a turbine trip siqna1 Turbine trip signal transmission, however is not built to sinqle failure criterion The cesul.t of a Eailuxe at this point vould have the effect of delaying the pressurization "signature."

Hovevec, high moisture levels entecing the turbine vill be detected by high levels in the turbine's moistu e

separatocs which resul.ts, in a trip of the unit Scram tr"ip siqnals fccm the tucbine ace designed such that a single failure vill neither initiate noc impede a reac=or scram trip initiation.

See Appendix 15A for further discussion 15 1

2 3

Core and S~st ea Per foraance 15-3. 2.3.1 Iath matical Model IQS~+

lgsQRT The predicted dyna~ic behavfor has been deterafned using a computer simulated, analytical mdel o

a generic direct-cycle SN.

This riedel fs described in detail in Reference ig. )- p, This covyuter rodel has 5een i~rove0 nd Yel fied through extensive coryarfson of its predicted results with actual BLR tes data.

The nonlinear cci-."uter sinulated analytical ao"el is desi",reC to predict asso"fated transient behavior of this reactor.

Same of the significant fea tures of the nede f are:

An integrated one-dimensional core niochl fs ass ~ed wfifch fncludes a detailed descrfptfon of hydraulic feedback effects, axial power shape

changes, and reactivity feedbacks.

The fuel is repres nted by an average cylindrical fuel and cladding.

aedel for each axfal location in the cor.

C I

The stean 1fnes are-modeled by eight pressure nodes fncorporatfng nass.

and momentum balances which will predict any wave phenori na present fn the stew> line dur fng pressure fzatfon t.ansient d.

The core aver'age acfa1 water density and pressure distribution fs calculated using a single channel to represent the heated active flow and a single channel to represent the bypass flow.

A r:od 1, representing liqufd and vapor nass and en rgy conservatfoe and mix-ture aux. n-urn conservation, fs used to describe the themal-hydraulic

~ behavior.

Changes fn the flow split betw n the bypass and ective channel flow are accounted for during transient events

e.

Principal controlle+ functions such as feedkater flow, recfrcula"ion flow, reactor ~ater level. pressure and load de@and, are represent d

tog ther with their doefnant nonlinear character fstfcs.

The ability to simulate necessary reactor protection syst~ functions fs pl ovided.

9'-

The control systeas and reactor protection system nodels are, for the

\\

most. part, identical to those e~loyed in the point reactor ~odel which fs described in detail in Reference f5. f-f and used in analysis for other transients.

SSES-PSAR 15 1.2 3

2 Input Paraaeters and InitiaL Conditions These analyses ha ve been performed, unless other@

se noted, with the plant conditions tabulated in Table 15 0 The same void reactivity coefficient used foc pcessucization transients is applied since a nore negative value consecvativeLy incceases the a.pparent seve ity of the po~

c increa e

-nd of cycle (a3.1 rods out) scram characteristics are assum d

The saf ety-rel ief val ve action is consecva tively assumed to occuc vith h'ghee than nominal set points The transient is s mulcted by programming an upper limit failuce in the fee%>>'a.ter system such that 1355 feedvatec floe occurs at a system design pre=-auce of 3.060 psig 15 1

2 3

3 Results The siuu3.ated feedvater controller transient at J.05 '

2, rated steam, flo~ is. shown in Figure 15.1-3.

The high ~atec 3.evel tucbine trip and feedwatec pump trip are initiated at a,ppcoximately 10 sec Scram occurs simu'aneous3.y from stop valve closure, and limits the neutron flux peaR and fuel th=r al transient so that no fuel damage occurs The turbine bypass system opens to limit peak pressure in the steam line near the safety valves to ~~+ psig ?ild the pcessure at the bottom of the vessel to about

()8Q psig The nuclear system process barrier pressure limit is <<o" enJangered The. bypass valves subsequently close to re-usta blisn p"essure control in the vessel ducing shutdown The level viLl. gradually drop to the Los Level. isolation reference poin", acti va'"ing the BCXC/HPCI systems foc long term level contra>>

15 2 3

n Consideration-of Once tainzies All systems utilized for protection in this event vere assumed.

to have the most conservative al.lovable response

$e g.

- re3.ief setpoints, scram stroke time and vock chacacterist ics)

Plant behavior is, therefore, expected to lead to a les - sever transient.

Rev.

17, 9/80

15. 1-7

SSES-PSAR 15 1

6 3

Core and ~Sstem per%or..ance The increased subcoolinq caused by misoperation of the BAR shutdown coolinq mode could result in a slow power increase du<

to the reactivity insertion This power rise would he terui.nates by a flux scram before fuel thermal limits are approached.

Therefore only qualitative description is provided he e

15 1

6 4 Barrier Performance As noted, above the consequences of. this event do not result ia.

any tenperature or. pressure transient in excess of the criteria.

for which the fuel,, pressure vessel or containment are designs therefore, these barriers maintain their integrity and function as desiqned 15 l. 6 5

RadiolocCicag

~Cease ceo ces Since this, event does not result in any barrier failures no analysis of radiological. consequences is required for this eve~t 15-1-7 REFERENCES

15. 1-1 t9, I-~

R.B Linford. "Analytical tethods of P1.an" Transient Evaluations for the General Electric Boiling later Reactor," April 1973

(.'f EDO-10802)

~>~<, " Sht=ETY EAtL.QA'TtOR W~

6emegRL dmin-Rl <

UAURCA7taQ oF ~ mp-cgu~zQaL'OR Bd iud@

~AT@:P.

Z~~ZogZ ~

~E~< <<ted-p~-~n,g

)$ 8Ct 15 1-18

EEN ERAL ELECTRIC NVCLKARKHKRGY ENVISION 385HA807 RKV Table 15.1-3 SEgUEHCE OF EYEHTS FOR FIGURE 15.1-8 Time-sec Event

/o.V g (a.

gJ'D-.N~'(est)

/o.'7g 3Q;

//./s 5

]r.gz,f Initiate simulated failure of 135" upper limit on feedwater flow.

LS vessel level set point trips main turbine and fe dwater pumps.

Reactor scram trip actuated from main turbine stop valve position switches.

Recirculation pump trip (RPT) actuat d by stop valve position switches.

Hain turbine bypass valves opened.

acgac/e First group of safety/relief valves~

due to high pressure.

//. gs J'23/

2.

~ ~spy/ /,r ~/qs vvc/.Pc c/ne g.

ected yv'~j y ~"0-P 'f

/s K'ss vcvle

~P~//~fy'

~u z p~ssevtv

/gnPcFp ZAI/vE e/glltcs Ac/en'/Mc LD vs A

/v&cslvQ

.Pt'/ /y pvcsS e vt'..

15.1-15

150.

I NEUTRON 2 FCAK I'UEI 3~AY Sl!AFI l! FLI.LIINI)j) 5 VESSEL S

LUX CENTEA TEHP g HFAf FI.UX F LI)N En)I FLON 125.

I VES>'.ZL P SIH LINC

)II!lr>IIII I Liil>L IIIII

. col!c nvF.

0> TIJAIIINE CS AISE IPSI)

I'ACS A)SC IPSII ALS Al'>f. IPS>I)

I ~NI 'iitTIJILO)

VOIO rAAC g;)

fEflH FLU!I Sl o

100.

W h

Pu 50, LJ 75.

25.

0.0.

5.

10.

15.

TINE )SEC) 20.

-25,0.

10.

15.

TINE ISEC) 20.

150.

I LEYELII 2 )I A SENS 3 N A SCNSl MGIIE'HI:I 5 OAIYE FLI II-AEF-SEP-SIIIAT 0 LEVELIINCIICS) 0 LEVCLIINCNESI TVGNM I

)I I Qi)

I VOIO AEA 2 OI)PI'LFA 3 SCAr>H ACI II IIIli)LIK!

TIVITT rncflv)TT CT IVIIT L1TV111 100.

0.

50.

0. '~->.

0.

O.

)5.

IHE ISECI '0.

-2.

0.

5.

10.

15.

TINE )SEC)

I t tl'>I

, ga!t..n..

SUM)) NnhNA KAITTAECFX900 F<LIIII))LACONIAOLLEA FAII.UAE HflXIHUH OEHANO. Nl fll IIIGII NRIEfl LEVEL TIIIPS

+A i&I- >

ss~s-ps Aa 15 2

2 Generator Load R~e ection 15 2

2 L

Xdentification of Causes and Prequency Classif ication 15 2 2.1.1 Xdentification of Causes Past closure of the turbine control valves (TCV) is 'initiated vhenever electrical qrid disturbances occur which result in siqnificant loss of electrical load on the generator The turbine control valves are required to close as rapidly as possible to prevent excessive overspeed of he turbine-genera-or fT-G) rotor Closure of'he main turbine control valves>>i].L cause a sudden reducticn in steam flov vhich results in av.

increase in system pressure and reactor shutdovn 15 2

2 1-2 1

Generator Load n~e'ection wikL ar eit~o+~

This event is cateqorized as an incident of moderate frequency 1~.2 2

1 2

2 Generator Load Rejection vith ~Bpass Pailure This event is ca i rized as an infrequent inci 1 t vith the.

follovinq characteris~

P &LE=,TC Prequency:

0 0036/plant HTBE:

278 years P equency Basis:

Th " uqh searches of do-esti ant onerating records have rev ed three instances of bypass fai e durinq 628 bypass s

em operations.

This qives a probability bypa:.s failure 0.0048 Combininq the actual frequency of a qene "or load qection vith the failure rate of the bypass yields a

frequency of a qenerator load rejection vith bypass failure of 0

0036 eventgplant year.

15. 2-0

~

~

0 SSES-PSAR satisfy single failure criterion and credit is taken for these protection features.

The pressure relief system which operates the relief valves independently shen system pressure exceeds rel'ef valve instrumentation set points is assuned to functicn normally during the time period analyzed AL3. plant control systems maintain normal opera,tion unless specifically designated to the contrary.

Gene~a%or Load Rej~c~ion with Pai uze of Bvo'ass Same=- as Subsection 15 2.2 2 2.1 except that failure of the main turbine bypass valves is assumed for the entire transient 15 2

2 2

3 The Effect of Sin~le Failures and Operator Errors litigation of pressure increase is accomplished by the reactor protection systen functions Turbine control valve trip scram and RPT are designed to sa isfy the single failure criterion.

An evaluation of the most limiting s'ngle failure (i e failure of

'the bypass system}

was considered in this event DetaU.s of.

single failure analysis can be found in A.ppendix 15'5 2 2.3 Core and~S stem Performance 15 2

2 3 L Hathenatical

.ttodel 1

simulate this event 15-2 2

3 2

Eront Pananetees ancl Initial Conditions These analyses have been performed, unless otherwise noted vith the plant conditions tabulated in Table l5 0-2 The turbine electrohydraulic control system (EHC) po~er/Load imbalance device detects load rejection before a measurable speed change takes place.

The closure characteristics of the turbine control valves are assumed such that t'e valves operate in the full arc (PA) node Rev.

L7, 9/80 15 2-6

SS ES-PS AR and have a full stroke closure tine, fron fully open to full.y closed of 0 15 seconds Auxiliary qenerator frequency ur. plies n

analysis the nain q

overspeed (RPT).

power ~ould normally be independent of any turbine-overspeed effects and continuously supplied. at rated since automatic fast transfer to auxiliary'o~er ormally occurs "or the pu pose-o

~

~orst case the zecirculation pumps are assu=ed to remain tied. to enerator and thus increase in speed, with the

'Z-G until tripped by the Recirculation Punp Trip system The reactor is operatinq in the manual flo~-contrc3. acde shen load reject'ion occurs Results do not significancy differ if the plant had been operatinq in the autoaa.tic f3 ov-control mode The bypass valve opening characteristics are simulated usinq the specified delay together vith the specified opening characteristic required for bypass systea operation.

A3.thouqh the closure of main steam isolation valves as caused by los Mater leveL trip (L2) is included in the sinulaticn, the floes frcm initiation of RCXC and HPCX core cooling system functions are not included 'f these events

occur, they vill fn1Lnv sometime af ter the primary concerns of fuel margin and overpressure effects have passed and are expected to result in effects less severe than those already experienced.

by the reactor system 15 2. 2-3 3

Results 15 2

2 '3 3.1 Generato Load Re "ection vith 8

ass e

Ziquze 15.2-1 shoes the results of the generator tzip from rated Paver Peak eeu tree flux rises xka.'/

$ eke ed'se I(O The averaqe surface heat flux peaks at ~~ of its initial value and NCPR dces not siqnificantly decrease belo~ its initial value 15-2-2.3-3 2

Generator Load Rggectioo with Fai1uze of Egoass Piqure 15 2-2 shovs that, for the case of Bypass failure, peak neutron flur reaches about 4Q, X of rated, averaqe surface heat flux reaches

(~g 5 of its initial value o

BS L

c 7

1 L

~

age 15 2-7

SSZS-FSAB 15 2.2 3

4 Coasid ration of Uncertainties The full stroke "losuca rata of the turbine control valve o

0 15 s "oads is consarvativ Typic lly the actual closure ca,t is nore like 0 2 seconds "laarly the lass tine it ta."es to close the nore severe the pressurization effect All systens utilizai for protactioa in this event vere

assumed, to have the poocest'llovable response (e.g, relief set points scram stroke time and sock characteristics).

Plant behavior is tha cef oreexpe= ted to reduce the actuaL severi tr of the transient l5 2. 2. 4 Barr ier Parf ~rmaace 15.2.2 0 l Generator Load, Beje"tion

~ Peat psessate tesaias sittia aot ai safe y taape aaK ao threat to tha barrier exists l5.2 2. 0 2

"anarator Load Rajactioa vith Failure of B~ass i(89 Peak pressure at the valves ceaches ~

system pressure" rea=has Q.)Q p ig at tha veL1 balov tha auclaac barrier transient D Sl.ge psiy Lhe peak nuclear bottom of th vass l, pres ure l nit of 1375 15 2-5 Radial.'optical Coasegueaces awhile the consequence of this event does not result n fuel failures, it does result ia the discharge of nocna3. coolant activity to the suppcassioa pool via SPV'peration Since this a "tivity is contained ia th prinacy containment, there vi11 he no exposuce to operating parsonnal.

Siace this event does not result in an un" on trolled. release to the environment tha plant opecator can choose to laava the activity bottled up ia the contaianent oc discharge it to the environment under controlled release conditions.

ZE purging of tha containment is chosen the celaase vill have to ba in accordance vith established technical specificatioas; therefore, this avant, at the vocst

. vould only Rev. 16, 7/80 15 8

Tir. -sec Table

}S'. ~-}

S gUEhCE OF EVETS FO< FlGUPE }g,g-( NMzc7ronl, Z}-p~sg Event .(-)0.015 (aporox.) Turbine-generator detection of lo s of electrical load 0 Turbine-generator poser load unbalance PLU) devices trip to initiate turhin.control valve fast closure o.o/( Turbin~generator PLU trip initiates nein turbine bypass system operation. Fast control valve closure (FCY) initiates. scram trip o.oi( Fast control valve closure (FCY) initiates a recircu1a-tion=pump trip (RPT). 440 P y./55'.'/So /ms'-seo Turbine bypass valves start-to. open. Turbine control valves closed. Grouo 1 relief valves actuated. Group 2 relief valves actuated. Group 3 relief vaives actuated. Group. 4 relief valves actuated. Group 5 relief valves actuated. ie va ve se 15.2-10

~ ~ 385 HABG7 REV P M wo. 66 Tire-ei Table Jg. ~-m SEgUEHCE OF EVEtiTS FOR F?GORE t5,~-2. C043 gQWrorv, 2P yPhss Event (-)0..015 (approx.) Turbine-generator detection of loss of electrica.'l Toad. 0 Turbine-generator power load unbalance (PLU) devices trip to initiate turbine control valve fist closure 0 o, p(Q

o. or(,

-Q-.Y5 .0 7 Turbine bypass valves fail to operate. Fast control valve closure (FCV) initiates serac trip. Fast control valve closure (FCV) initiates a recircula-tion pump trip (RPT). Turbin control valves closed. o8oS Group 1 relief valves actuated Group 2 relief valves actuated. Group 3 relief valves actuated. Group 4 relief valves actuated Group 5 r lief valves actuated. r ~ v ~ 0 l I e ~ 15.2-11

)50. I NEUTflON LUX 2 PEAK fUE CENTEA TEHP 3 AVYE sUAF cE I)EAT FLUx 4 FtEUIII)18) I'LUH 5 YESSLL 5 EAH FLOH 300. I VESSEL P FS AISE tPS I) 2 5TH LINE PAES A)SE IPSI) 3 SAfg)T V I.vt FLOI) 0) It Iit.LI ( v)ivL tLfiH I)t' 5 Or'tnSS v)LYC rLAN F'.) 6 TUAAINE ! IEAH I'LOH OC) a IN. h I 50. 5 200. 100. 0, 20 4 ~ 6. TIHE ISEC) 8. 0,0. 24 4, 6. TIHE ISEC) 0. 200. I LEVELII 2 H A SENS 3 N tl SENSI 1E IIILl 5 0AIVE FL H-AEF-SEP-SIIIAT 0 LEVELtt))CHES) 0 LEVELIINCHESI 17CRTP l I VOID AEA 2 OOPPLCA f 3 SCIIAH AEt ijTilTITLHEI TIYITT EAC)tVITT CTIVIIT LTIvITT 100. 0. 0. -100. 0 2. 6, Il)IE tSECI 200 2. 3. TIHE ISECI 4, IIh bZlQ K4)efisib., SUSQUE)mNNI) HA)TTAECLAIOl ocNcfv)II)f) LoAO AEJEcT IGN IIITH 0TPAss Q

150. I HEUmON 2 PEAK FU 3 AVE SUAF( l'EEE)IIIIIEl 5 VESSEL 5 LUX CENTEA TEHP C '~AT FLUX OH EAH FLOH 300. I VESSEL P 2 5TH LINf: +5fif'EIT Vl 4 AtL(~EVI 5 OrrnSS Vf G TUfIOINE ! ES RISE (PSll PAES AISE (PSI) (VJ. FLON LVE FL()ll f/) LVE FLOH IEAH FLON (gJ o IM. W 5 50> 200. 100. 0. 2> 4 ~ 6. TINE (SEC) 0..(l5 0. 2> 4, 6. TINE (SEC) I LEVEL(l ~ ? N A SEHS 3 N A SENQ fllJHElHL' 081VE FL( -REF-SEP-SKI AT 0 LEVEL(IHCHESI 0 LEVEL(IHCIIES) TVL~%) Hl [Qo) I V010 BEA( I'IVITT 2 DOF'PLEA I CACTIVITY 3 SC(NH AEI CTIVITY lrTmx-Ifra rvrrv 0. 0. -1. I -1 00.0. 2> 4. 6. TINE ISEC) 8. 0 2.. 3 TIHE {5EC) 82Xt ~'plhlss>o> SOSQVEINM'Ifl 1U) IT TAECLBI 00 K"LMIO(lLOAD AEJECTIOH HlTNUT BTPASS , f./w

SS ES-ZSA R I result in a sma11 increase in the yearly integrated exposure level. 15 2. 3 TORBIN E TRIP 15.2.3.1 Identification of Causes and Frequency Classification 15.2.3 1 1 Identification of Causes vaciety of turbine or nucleac system malfunctions will initiate a turbine trip. Some eramples are moisture sepacato-and heater drain tank high levels, large vibrations, operator Lock out,'oss of control fluid pressure, low condenser vacuum and reactor high wa t er le ve 1. l5. 2 3.1.2 Preque~nc Classification 15.2.3el 2.1 'Turbine Tri ~ ov vien~~ +< This transient is cateqorized as an incident of moderate frequency. In defininq the frequency of this event, tucbine trips which occur as a byproduct of othec transients such's loss of condenser vacuum or reactor high level trip events are not included. However, spucious low vacuum or high level trip siqnals which cause an unnecessary turbine trip are included in defininq the frequency. In order to get an accurate eventby-event frequency breakdown, this type of division of initiating causes is required. 16-2. 3~.2.2 yuub~ine 'Zt ~vith pa~inta ut the ~agass nt disturbance is categorized as a nfrequent Pr ency is expected to be as lows= This tran incident. Frequency: MO. 064/plan at DE gg I p aTBB: 156 ye Frequency Basis: s discussed in wbsection

15. 2.2. 1. 2 2, the failuce rat the bypass is 0 0048.

Cezbining this with the turbine ap frequency of 1. 33 events/plane. y'ear ields the fre. ency of 0.0064/plant year.

15. 2-9

SSZS-PSA.R ],3.2.3 2 3'he Eft'ect of single pailures a~ad 0 eratog En~re I6 ) IS 3. 2 3 I Turbine Tr~is at Paver Levels Greater Than 30S gad litigation of pressure increase is acconplished by the reactor protection system, functions. lain stop valve closure scran trip and RPT are designed to satisfy single failure c"iterion 15

2. 3 2. 3 2

Tu bine Tri~sat Payee Levels Less Than 30 }IBB Same as Subsection 15 2.3.2.3.l except HPT and stop valve closure scran trip is normally inoperative. Since protection is still provided by high flux, high pressure,

etc, these will also continue to function and scran the reactor should a singLe failure occur.

15 2 3 3 core and~s stee Perfornance l5. 2 3 3.1 Mathematical Nodal '?he. conputer uodel simulate OP 8 '4< SMaLSef ~' $ 5.2.3 3 2 Engut described in Subsection 15.l.l 3 l was used to M3'5<Vi< +Y'iP mls .IPP'P~S a~4 ~.~ Q, (.2. 5 ( V~ ~J-Qr ~ ~34~ + 'P I ed~ These analyses have been perfornedunless otherwise

noted, with plant conditions tabulated in Table l5.0-2 Turbine stop valves full stroke closure tine is 0.1 second A reactor scran is initiated by position switches on valves when the valves are less than 90% open This scraa. trip signal is automaticaLly bypassed when the below 30%

NB rated'ower leveL the stop stop val ve reactor is Reduction in core recirculation f1ow is initiated by position switches on the main stop valves, which actuate trip circuitry which trips the recirculation pumps. Rev. 17, 9/80 15 2-12

SSBS-FSAR

15. 2. 3.3 3

Results 15.2.3 3.3.1 Turbine Trig A turbine trip vith the bypass system operating normally is simulated at 105% HB rated steam flov conditions in Figure 15.2-3. 'eutron flux increases rapidly because of the void reduction caused by the pressure increase

Hovever, the flux increase is limit d to 16.7Ã of rated.

by the stop valve scram and the RPT system Peak fuel surface heat flux does not exceed 101~ of its initial value. 15.2 3 3. 3 2 Turbine Tr~ivit~hailure of Bypass A turbine trip vith failure of the bypass system is simulated at 105% !tB rated steam flov conditions in Figure 15.2-0 peat neutroa flux reaches ~% of its rated value,. and vhs peA'urface k<v M 6/ H e 1 cl v ~ l 15.2.3 3e3..3 Turbine Trip with Bypass Valve Pailure, Lov Power This transient is less severe than a similar one at high pover Belov 30$ of rated

pover, the turbine stop valve closure and turbine control valve closure scrams are automatically bypassed.

At these lover power levels, turbine first stage pressure is used to initiate the scram logic bypass. The scram which terminates the transient is initiated by high vessel pressure. The bypass valves are assumed to fail; therefore, system pressure vill increase until the pressure relief set points are reached At this'ime, because of the relatively lov pover of this transient event, relatively fev relief valves vill open to Limit reactor pressure. Peak pressures are not expected to greatly exceed the pressure relief valve set points and will. be signif icantly belov the RCPB transient linit of 1375 psig. Peak surface heat flux and peak fuel center temperature remain at rel.atively 1cv values and MCPR is expected to remain veil above the GETAB safety Limit. 15 2-13

SSES;FSKR l5. 2.3 3.4 Considerations of Uncertainties Uncertainties in these analyses involve protection system

settinqs, system capacities, and system response characteristic-In all cases, the most conservative values are used in the analyses.

For example: (1) S).ovest allowable control rod scram motion is assumed. (2) Scram worth shape foc all-rod-out conditions. is assumed. v (3) Minimum specified valve capacities are utilized foc over-pressure protection (4) Set points of the safety/relief valves include errocs-(high) for all valves. l5 2 3 4 Barrier Performance 15.2.3.4el Turbine Trig Peak pcessure in the bottom of the vessel reaches 1167 psig, which is beloM the,ASIDE code limit of 1375 psig for the reactor coolinq pcessuce boundary. Vessel dome pressure. does not exceed ll43 psiq The sevecity of turbine trips from lover initial power levels decreases to the point where a scram can be avoided if auxiliary power is available from an external source and the power level is within the bypass capability 15.2 3 u.2 Tuthiue ~2ui v1th Pailute fath~ca pass The safety/relief valves are open and close sequentially as the stored energy is dissipated and the pcessure falls below the set points of the valves.. Peak nuclear system pressure reaches M9+ l2l3 psiq at the vessel bottom, therefore, the overpcessure transient is cleacly below the reactor coolant pcessuce boundary transient pcessure limit of 1375 psig Peak dome pressure does not exceed psig )l85 15 2.3-4.2.1 Turbine Trip vith Failure of Bypass at Low Power Qualitative discussion is provided in Subsection 15.2.3.3 3 3.

15. 2-14

'ELfh'-GAiQ ELECTR!s N~~~ R ENERGY D'.Vls'NOH 385HA807 Rav s ~.y8 Tir,e-sec 'able 15.4.-4 SEQUENCE OF EVEHTS FOR FIGURE 15-2.-4 y OR8zAC jRfp +~'~<o ~T D'AH<< ( Q 3 pH) Evert Turbine trip initiates closure of a@in stop valves; 0 Turbine bypass valves fail to operat O.C1 Hain turbine stop valves reach 90" open position and initiate reactor sc~am trip. 0.01 Hain turbine stop valves reach 90" open position and. initiate a recirculation pump (RPT) trip. 0.1 o,Ã7 /.z go /r43 Turbine stop valv s closed. Group 1 relief valves actuated. Group 2 relief valve actuated. Group 3 relief valves actuated. Group 4 relief valves actuated. Group 5 relief valves actuated. L8 vessel lovel set point trips feedv(ater pumps. 15.2-23

150. I NEUTAOII ? Pffi)f fUEl 3 AVf QNlFI 4 ttl 1)IIIIICI 5 VC.!!AL S Lur CCNTCA TEHP cg IIF,A~ILux t LIIW E'AH I UJII 3or). I VfSSEL P 2 51ll LINC 3 Slif f11 VI 4 I'0 ).Ill'i 5 Oilrr,S Vi fi Ill)if:INE ! FS A)% IPS!I) FAE5 illSI II'5 I ) l,yC FLei Wl I.vl. I I i)II l4) I.VC r 1 lilf gal IEfiil I I<III g4) a 100. LJ P!j 50. LJ 200. 100. 0.0. 20 4. 6. TIHC ISECI 0,: II 0. 2. 4. G. TIHE ISEC) I 'LEYELIIN 2 W A SENS 3 )I A SENS VYIIE 1IILl 5 OfllVE FL II-AEF-SEP-SIIIAT 0 LEYELI INCHCS I 0 LEVELl INCHES) ~L~I'l f/') I YOIO REA 2 DOPPLEA f 3 SCBAH ACI CTIITII.Til I 1 IVITT EACI I V ITT CT)VI'I'f CTTVI 1Y 100. 0. 0. .100. 2. II~ G. 'rl)E ISECI ' ~0. 2. 3. TlklE ISECI gw niMH!.>l. Siisr)UEHr)IINr) )IAITTBECTT?00 TUI+INE lAIP IIITHGUT f)YPASS ~. TRIP SC~ lq 1 -8 ~

SS ES-FSA R enterinq the turbine can cause vibration and trip the tucoine

via, turbine supervisory instrumentation Scram trip siqnals from the turbine are designed such that a

single failure vill neither initiate noc impede a reactoc scram trip initiation. See Appendix lSA foc further discussion. 15

3. I.

3 Cote and 5 ates pe fovaance

15. 3 l.. 3. 1 la thematical lodel The nonlinear, dynamic model described briefly in Subsection 15.1.l 3 l is used to si"ulate this event.

15 3 1 3 2 ~fn ut Pave eetevs and Initial Conditions These analyses have been performed, unless otherwise noted, vith plant conditions tabulated in Table 15.0-2. Pump motors and pump rotors are simulated. with minimum specified rotatinq inertias 15.3.1 3.3 Results Piquce

15. 3-1 shows the results of losinq one recirculation pump The tripped loop diffuser flow reverses in approximately 5.7 seconds Ho@ever, the ratio of diffuser mass floe to pump mass flow in the active jet pumps increases considerably and produces approximately 103~ of normal diffuser flow and 72% of cated core floe.

NCPB remains approximately + the Operatinq Limit. thus the fuel therma.l limits are not violated During this tcansient, level sve11 is not sufficient to cause turbine trip and scca m.

SS ES-PSAR Q ~f IL) L Hhi Fiqure 15 3-2 shows graphically this transient with minimu specified rotating inertia. liCPR cemains unchanged at No scram is initiated directly by pump trip The vessel water Level swell due.to rapid flow coastdown is expected. to reach the high level tcip thereby shutting down the main turbine and feed pump

turbines, and indirectly initiating scrams as a result of the main turbine trip.

Subsequent events,- such as main steam line isolation and initiation of RCXC and HPCZ systems occurring late in this event, have no significant effect on the results 15 3 1.3 4 Consideration of Oncectainties initial conditions chosen foc these anaLyses are conservative and tend tc force analytical results to be more severe than, expected under actuaL plant conditions, Actual pump and pump-motor drive line rotatinq inertias are expected to be somewhat greater than the minimum design values assumed in this simulation. Actual plant deviations regarding inertia are expected to 1essen the sevecity as analyzed. Minimum design inertias were used as well as the Least negative void coefficient since the primary interest is in the flow reduction 15.3.1 4 Barrier Performance 15 3.1 4 1 Tci of One Recirculation Pump Piqure 15 3-1 results indicate a basic reduction in system pressures from tne initial conditions. Therefoce, the RCPB barrier is not threatened 15.3.1 4 2 Tri of Two Recirculation Pu~m s The results shown in Pigure 15 3-2 indicate peak pres uces stay well below t'e 1375 psiq limit allowed by the applicable code. Therefore. the barrier pressure boundary is not threatened. i5 3-5

SSES-FSXB 15.5 g 2.2 System Operation f".'o pcopecly simulate the expected sequence of events the analysis of this event assumes normal functioning of plant instrumentation and controls, specif ically, the pcessuce cegulatoc and, the vessel Level control vhich respond directly to this event-Bequiced operation of engineered sa feguacds other than vhat is described is not expected for this tcansient event. The system is assumed to be in the manual flov control mode of operation.

15. 5; 1.2 3

The "-ffeet of~siu le Failures aa8 Operator E-rors Inadvertent operation of the HPCI results in a mild pcessur ization. Correcti ve action by the pressure regulator and/oc level. contcol is expected to establish a nev stable, operating state The effect of a single failure in the pressure regulator vill aggravate the transient depending upon the nature of the failure. Pressure regu'ator failures are discussed, in Subsections 15.1.3 and 15.2.1. A single failure in the level control system causes leve3. rise or fall. by improper control of the feedvater system Xncceasing level vill trip the turbine and automatically trip the HPCZ system off. This trip signature is already described in the failure of feedvatec controller vith increasing flov Decreasing level vill automatically initiate scram at the L3 level trip and vill have a signature similar to loss of feedvater contcol-decreasing, flov 15.5 1.3 Core and System Pe"focmance

15. 5. 1. 3. 1 Na the sa tical Model The detailed ao'nlineac dynamic model desccibed briefly in tS. ~. i. S. t

15. 5. L. 3 2

Zn~ut Parameter and Znitial Conditions This analysis has been pecfocmed, unless othecvise

noted, vith plant conditions tabulated in Table 15.0-2.

Rev. 3.7, 9/80

15. 5-2

SS ES-FSAR The, vater temperature of the HPCI system vas assumed to be 40~F vith an enthalpy of ll BTU/Lb. Inadvertent startup of the HPCI system vas chosen to be analyzed since it provides the qreatest auxiliary-source of cold vater into the vessel. 15.'5. 1 3 3 Results Piqure 15.5-1 shovs the simulated transient event for the manual flov control mode. It begins vith the introduction of cold vater into the feedvater sparger. Within 1 second the full HPCI flov is established at approximately 19% of the rated feedvater fLov rate. Ho delays vere consid'ered because they are not relevant to the analysis. Addition of cooler vater to the core causes the neutron flux to increase to a peak of 118$ NBR. a 15

5. 1 3.4 Consid~eatgon of Uncertainties Important analytical factors including reactivity coefficient and feedvater temperature change have been assumed to be at the vorst conditions so that any deviations in the actual plant parameters, vill produce a

Less severe transient 15 5-1.a Ba~iec pe~fccaaace Picture 15.5-1 indicates a slight pressure reduction fron initial conditions, therefore, no further evaluation is required as RCPB pressure margins are maintained. 15 5 1 5 R~dologic+ Consequences Since no activity is released during this event, a detailed. evaluation is not required. REV. 11, 7/79 15 5-3

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