ML20076G690

Forwards Draft FSAR Page Changes,In Response to Reactor Sys Branch Draft SER Item 1, Odyn Analysis of Pressurization Transients. Changes Will Be Incorporated Into FSAR Rev Scheduled for Jul 1983

ML20076G690
Person / Time
Site:	Limerick
Issue date:	06/13/1983
From:	Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8306160001
Download: ML20076G690 (58)

Text

7 c.

4 8

l PHILADELPHIA ELECTRIC COMPANY 23O1 M ARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.19101 /

EtDW ARD G. BAUER, J R.

(215)841-4000 noesas ALcounset EUGENE J. BR ADLFY assessava esussaL counsas RON ALD ELANMEN CLUDOLPH A. CHILLEMI

3. C. KI R K H A LL T. H. M AHER CO RN ELL PAUL AUERBACH assesvant osmanas counsso C DW ARD J. CULLEN, J R.

THOM AS H. MBLLER. J A. June 13, 1983 BIEN E A. McMEN N A assestaNT Counsst s Mr. A. Schwencer, Chief Licensing Branch No. 2 -

Division ot Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Limerick Generating Station, Units I&2 Response to Reactor Systems Branch Draft Sa fety Evaluation Report (DSER)

References:

1) A. Schwencer to E. G. Bauer, Jr.,

letter dated March 11, 1983

2) E. J. Bradley to A. Schwencer, letter dated April 27, 1983 File: GOVT l-1 (NRC)

Dear Mr. Schwencer:

The attached documents are draft FSAR page changes in response to Reactor Systems Branch (RSB), Draft Safety Evaluation Report (DSER) Item I, "0DYN Analysis o f Pressurization Transients."

l Due to logistical difficulties we were unable to incorporate l these FSAR page changes into the June revision. However, these page changes will be incorporated into the FSAR revision scheduled for July, 1983.

Sincerely, Eug ne . Bradley RJS/gra/II I Copy to: See Attached Service List I

r306160001 830613 PDR ADOCK 05000352 E PDR -- - . .. _

r cc: Judge Lawrence Brenner (w/o enclosure)

Judge Richard F. Cole (w/o enclosure)

Judge Peter A. Morris (w/o enclosure)

Troy B. Conner, Jr., Esq. (w/o enclosure)

Ann P. Hodgdon (w/o enclosure)

Mr. Frank R. Romano (w/o enclosure)

Mr. Robert L. Anthony (w/o enclosure)

Mr. Marvin I. Lewis (w/o enclosure)

Judith A. Dorsey, Esq. (w/o enclosure)'

Charles W. Elliott, Esq. (w/o enclosure)

Jacqueline I. Ruttenberg (w/o enclosure) -

Thomas Y. Au, Esq. (w/o enclosure)

Mr. Thomas Gerusky (w/o enclosure)

Director, Pennsylvania Emergency Ma'nagement Agency (w/o enclosure)

Mr. Steven P. Hershey (w/o enclosure)

Donald S. Bronstein, Esq. (w/o enclosure)

Mr. Joseph H. White, III (w/o enclosure)

David Wersan, Esq. (w/o enclosure)

Ro' v ert J. Sugarman, Esq. (w/o enclosure)

Martha W. Bush, Esq. (w/o enclosure)

Atomic Safety and Licensing Appeal Board (w/o enclosure)

Atomic Safety and Licensing Board Panel (w/o enclosure.)

Docket and Service Section (w/o enclosure)

O f

1 0

+

I i

LGS FSAR TABLE 4.4-1 (Page 1 of 2),

THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS OF THE REACTOR CORE LIMERICK SUSQUEHANNA BWR/4 BWR/4 BWR/4 GENERAL OPERATING CONDITIONS (251-764) (218-560) (251-764)

Reference design thermal 3293 2436 3293 output, MWt Power level for engineered 3435 2558 3435 safety features, MWt Steam flow rate, at 420.0F 14.159 10.5 13.48(1) final feedwater temperature, millions lb/hr Core coolant flow rate, 100 77.0 100 millions lb/hr 10.4 13.44 Feedwater flow rate, 14.127 millions Ib/hr System pressure, nominal in 1020 1020 1020 steam dome, psia System pressure, nominal 1035 1035 1035 core design, psia Coolant saturation temperature 548 549 549 at core design pressure, OF Average power density, kW/ liter 48.7 49.2 48.7 Maximum linear heat generation 13.4 13.4 13.4 rate, kW/ft Average linear heat generation 5.3 5.4 5.3 rate, kW/ft Core total heat transfer area, 74,871 54,879 74,871 ft3 Maximum heat flux, Btu /hr-sq ft 361,600 361,600 361,600 l Average heat flux, Btu /hr-sq ft 141,100 145,060 143,700 Design operating minimum critical Jedi" 1.22 1.25 power ratio (MCPR) l ,12.

i Core inlet enthalpy at 4200F 526.1 526.9 521.8(1)

i a 5.2.--l AMJ

• ED * #'2'~/b MJ. w - M e f MM*

LGS FSAR O A5 w.

facility.

DRAPT These models include the hydrodynamics of the flow loop, the reactor kinetics, the thermal characteristics of the fuel and its transfer of heat to the coolant, and all the principal controller features, such as feedwater flow, recirculation flow, reactor water level, pressure, and load demand. These are represented with all their principal nonlinear features in models that have evolved through extensive experience and favorable comparison of analysis with actual BWR test data.

A A.

tailed description 5 of bh+s modef ts documented in-licerrin; tepic:1 repcrt N500-10902, "'..:1ytical ":th;d Of P1:nt Tr:ncien.

--E"rlustic"e fr: th; ~Z-sJK," (Kcf 5.2 "). MSRVs are simulated in l a nonlinear representation, and the model thereby allows full j

investigation of the various valve response times, valve capacities, and actuation setpoints that are available in applicable hardware systems.

The typical valve characteristic as modeled is shown in Figure 5.2-2. The associated bypass, turbine control valve, and MSIV characteristics are also simulated in the model.

5.2,.2.2.2 System Design A parametric study was conducted to determine the required steam flow capacity of the MSRVs based on the following assumptions.

5.2.2.2.2.1 Operating Conditions

a. Operating power = 3435 MWt (104.3% of nuclear boiler rated power)

b. Vessel dome pressure P <1020 psig

c. Steamflow = 14.86 x 10* lb/hr (105% of nuclear boiler rated steamflow)

These conditions are the most severe because maximum stored energy exists at these conditions. At lower power conditions the

' transients would be le'ss severe, i 5.2.2.2.2.2 Transients The overpressure protection system must accommodate the most ,

severe pressurization transient. There are two major transients, the closure of all MSIVs and a turbine-generator trip with a coincident closure of the turbine steam bypass system valves, that represent the most severe abnormal operational transients resulting in a nuclear system pressure rise. The evaluation of transient behavior with final plant configuration has shown that the isolation valve closure is slightly more severe when credit is taken only for indirectly derived scrams; therefore, it is used as the overpressure protection basis event,:nf ;h;.:n in 5.2-4

LGS FSAR Figure 5.2-1. Table 5.2-8 lists the sequence of events, tor tne3 main steam line isolation closure event with flux scram and with -

the installed MSRV capacity.]

5.2.2.2.2.3 Scram

a. Scram reactivity curve - Figure 5.2-3 h RED [ M

b. Control rod drive scram motion Figure 5.2-3 I'1"N" ODY 5.2.2.2.2.4 MSRV Transient Analysis Specifications

a. Valve groups: 3 ..

b. Pressure setpoint (maximum safety limit):

1. 1142 psig - group 1

2. 1152 psig - group 2

3. 1162 psig - group 3 The setpoints are assumed at a conservatively high level above the nominal setpoints as shown by Table 5.2-2. This is to ,

account for initial setpoint errors and any instrument setpoint

n. Typically the assumed drift that might occur during oper setpoints in the analysis are 1% to above the actual nominal setpoints. Highly conservative MSRVf* response characteristics are also assumed.

l 5.2.2.2.2.5 MSRV Capacity Sizing of MSRV capacity is based on establishing an adequate margin from the peak vessel pressure to the vessel code limit (1375 psig) in response to the reference transients (Section 5.2.2.2.2.2).

5.2.2.2.3 Evaluation of Results 5.2.2.2.3.1 MSRV Capacity The required MSRV capacity is determined by analyzing the

! pressure rise from an MSIV closure with flux scram transient.

The plant is assumed to be operating at the turbine-generator design conditions at a maximum vessel dome pressure of 1020 psig.

The analysis hypothetically assumes the failure of the direct isolation valve position scram. The reactor is shut down by the backup, indirect, high' neutron flux scram. For the analysis, the safety set ints are assumed to be in the range of 1142 to 1162 psig. analyses indicateg that the design valve capacity is capable of maintaining an adequate margin below the peak ASME Code allo ble pressure in the nuclear system (1375 psig).

(W RED / 4 Obyp 5.2-5

LGS FSAR Figure 5.2- shows curves produced by nalysis. The..

sequence of events in Table 5.2-8Aassumed in this analysis, was investigated to meet code re the pressure reliefsystemexclusively.@quipementsand & ODyA/ valuatg Av  % r.1-/k m /

T.4.c.1-d. 6 Under the General Requirements for Protection Against Overpressure as given in Section III of the ASME B&PV Code, credit can be allowed for a scram from the reactor protection system. In addition, credit is also taken for the protective l

circuits that are indirectly derived when determining the required MSRV capacity. The backup reactor high neutron flux scram is conservatively applied as a design basis in determining the required capacity of the MSRVs. Application of the direct position scrams in the design basis could be used, since they' qualify as acceptable pressure protection devices when determining the required safety / relief valve caoacitv of nuclear, ,

vessels under the provisions of the ASME Cod TA4 '

The parametric relationship between peak vesselu(bettomi ptwessure y 4

>- and MSRV capacity for the MSIV transient with high and

, position trip scram is described in Figure 5.2-4. Also shown in Figure 5.2-4 is the parametric relationship betwee peak vessel

(bottom) pressure and MSRV capacity for the turbine trip with a l coincident closure of the turbine bypass valves and direct scram, i which is the most severe transient when direct scram is considered. Pressures shown for flux scram result only with multiple failure in the redundant direct scram system.

The time response of the vessel pressure to the MSIV transient with flux scram and the turbine trip with a coincident closure of the turbine bypass valves an direct scram for 14 valves is illustrated in Figure 5.2-5. - h e shows that the pressure at the vessel bottom exceeds 1250 p ig for less than 5.8 seconds, which is not long enough to transf r any appreciable amount of heat into the vessel metal and wh ch is at a temperature well below l 5.500F at the start of the tr ansient.

I l

5(.z.e.4.a.4IAl.SEfCT eressueA/vwp EXT M(2 n Inlet ME47-andl Discharge l Prebure drop on the piping f m the reactor vessel to the MSRV

u. taken into account in calcu ting the maximum vessel pressures. Pressure drop in th discharge piping to the suppression pool is limited by roper discharge line sizing to prevent the backpressure on eac MSRV from exceeding 40% of the valve inlet pressure, thus ensu ing choked flow in the valve orifice and no reduction of va ve capacity due to the discharge piping. Each MSRV has its ow separate discharge line.

't[ W AEDy nA opM m unktfhwA2la.e.L als.A.4 mcg. ,

cWkr

• v '

5.2-6 1

LGS FSAR INSERT FOR 5.2.2.2.3.1 DRAFT Safety requirements demand very high reliability in the reactor SCRAM functions. Recognition of this reliability to contribute to vessel pressure protection is reflected in the new Section III Code provisions.

The actual design conservatively applies the code provisions such that sufficient margins exist beyond the code limits, further assuring the reliability of vessel pressure protection.

The design basis for sizing safety / relief valves with indirect SCRAM credit is technically sound. It is allowed under Section III of the ASME Boiler and Pressure Vessel Code, which has been adopted in the design of this Boiling Water Reactor, The '-hig -kr-"-------"- - ' ' - -- -Predithd vessell -- *pr#1sva ' -- " y' ' ' REDY tha n by 0DY N for the same MSIV transient indicates that it is conservative to apply the REDY SRV sizing results to demonstrate compliance with overpressure protection code requirements.

l i

l l

l l

l 5.2-6a(INSERT)

PCY:im/5E-3 5/13/83

LGS FSAR DRAFT t

5.2.5.12 Conformance to Reculatory Guide 1.45 (May 1973)

- Reactor Coolant Pressure Boundary Leakace Detection System The leak detection system design is in conformance with the guidelines of Regulatory Guide 1.45 except that, with reference to Paragraph C.2 of the guide, the containment airborne radiation monitors may not always be capable of detecting a leak rate of 1 gpm in one hour, as explained in Section 5.2.5.5.

The procedures and technical specification limits recommended by the guide will be followed during operation.

5.2.5.13 Seismic Capability of Leak Detection System The RCPB leak detection system is designed to seismic Category I criteria to remain functional following a safe shutdown earthqualze (SSE). Instrumentation associated with the plant

. drainage system is not designed to seismic Category I criteria.

5.

2.6 REFERENCES

5. 2- 1 pu, R. Linford, " Analytical Methods of Plant Transient Evaluation for the General Electric Boiling Water Reaction," NEDO-10802 (April 1973).

{ 5.2I J.M. Skarpelos and J.W. Bagg, " Chloride Control in BWR Coolants," NEDO-10899, (June 1973).

5.2-3 W.L. Williams, Corrosion, Vol 13, p. 539t (1957).

5.2-4 M.B. Reynolds, Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flows, GEAP-5620 (April 1968).

5.2-5 " Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants,"

NUREG 76/067/ NRC/PCSG (October 1975).

g2_.ghk4f 6 dLe 2'-2-d bubo

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Wb, #, NOO-14/K4. [0W 197E) t 5.2-45

LGS FSAR h

TABLE S.1-kA-( h h.

SEQUENCE OF EVENTS FOR FIGURE 5.2-1 A TIME-SEC EVENT 0 Initiated closure of all MSIVs 0.3 MSIVs reached 90% open and initiated reactor scram.

However, hypothetical failure of this position scram was assumed in this analysis.

1.66 Neutron flux reached the high APRM flux scram setpoint and initiated reactor scram.

2.41 Steam line pressure reaheed the safety valve pressure setpoint, and safety valves started to open.

i 2.65 Recirculation pump drive motors were tripped due to

high vessel pressure.

3.0 The MSIVs completely closed.

3.08 All safety relief valves opened.

5.02 Vessel bottom pressure reached its peak value.

17.77 The safety relief. valves closed.

24.0 (est) Wide-range sensed water level reached the Level 2 setpoint.

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LGS FSAR TABLE 5.2-8b(ODYN)

SEQUENCE OF EVENTS FOR FIGURE 5.2.-lb TIME (SEC) EVENT 0 Initiated closure of all MSIVs 0.3 MSIVs reached 90% open and initiated reactor scram.

However, hypothetical failure of this position scram was assumed in this analysis.

1.66 Neutron flux reached the high APRM flux scram setpoint and initiated reactor scram.

2.45 Steam line pressure reached the safety valve pressure setpoint, and safety valves started to open.

l 2.51 All safety relief valves opened.

2.57 Recirculation pump drive motors were tripped due to high vessel pressure.

3.0 The MSIVs completely closed.

3.93 Vessel bottom pressure reached its peak value.

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I

I multinode, single channel thermal hydraulic model that requires simultaneous solution of the partial differential equations for the conservation of mass, energy, and momentum in the bundle, and O that accounts for axial variation in power generation. The primary inputs to the model include a physical description of the bundle, channel inlet flow and enthalpy, and pressure and power ,

generation as functions of time. l l

A detailed description of the analytical model may be found in ,

Appendix C of Ref 15.0-2. The initial condition assumed for all '

full power transient MCPR calculations is that the bundle is operating at or above the MCPR limit (ttryH . Maintaining MCPR7 /,g greater than the safety MCPR limit is a. sufficient, but not .

necessary, condition to assure that no fuel damage occurs. This is discussed in Section 4.4.

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

These correlations are substantiated by fuel rod failure tests, and are discussed in Section 4.4 and Section 6.3.

15.0.3.3.2 Input Parameters and Initial Conditions for Analyzed Events In general, the transients or accidents analyzed within this

(( section have values for input parameters and initial conditions as specified in Table 15.0-2. Analyses that assume data inputs different from these values are designated accordingly in the appropriate event discussion.

The transient analyses herein presuppose, for the end-of-equilibrium-cycle (EOEC) conditions being simulated, that the plant design includes a recirculation pump trip (RPT) actuated by either fast closure of the turbine control valves or closure of the turbine stop valve. An EOEC/RPT system is not in the current plant design. Adoption of this feature is under review by the applicant. The transient analyses presented herein will be revised to reflect the outcome of this review.

15.0.3.3.3 Initial Power / Flow Operating Constraints The analysis basis for most of the transient safety analyses is the thermal power at rated core flow (100%), corresponding to 105% nuclear boiler rated (NBR) steam flow. This operating point is the apex of a bounded operating power / flow map that, in response to any classified abnormal operational transients, will yield the minimum pressure and thermal margins of any operating point within the bounded map. Referring to Figure 15.0-1, the apex of the bounded power / flow map is point A, the upper bound is the design flow control line (105% rod line A-D'), the lower bound is the zero power line H'-J', the right bound is the rated

.k ,

15.0-7

gras. .u e ,

utilized to accommodate the transients, or acciden ,

effects, and the systems involved in the protective actions.

2 )

Interdependency of analysis, and cross-referral of protective actions, are an integral part of this chapter and the sect-ion.

A summary table that classifies events by frequency only (i.e.,

not just within a given category such as decrease in core coolant temperature) is in Section 15.9.

15.

0.6 REFERENCES

15.0-1 United States Nuclear Regulatory Commission Regulatory Guide 1.70, Revision 2 (Preliminary), " Standard Format and Content of Safety Analysis Report for Nuclear Power Plants, Light Water Reactor Edition," September 1975.

, *n 15.0-2 " General Electric BWR Thermal Analysis Basis.(GETAB):

Data, Correlation, and Design Application,'

November 1973 (NEDO-10959 and NEDE-10958).

' I f, e - 3 R.B.b Med, Ant (ybcal Mt4hed2 of Plant Tretn 3ient Evalv aS.u 4e r Me G e ne *=( Electric Ec'iling W shr Resche, N Epo-l o 8 o t, G en e r al E f e ch .'s , A p. i f i 9 73 ,

i s.o-4 c,enee sf Etect es c, Qo sM i c Af'*' n 'N % e 0 " c ' O '" '" 'ic" 4!

c. . w adem+ M d e f f e < 3 4.ft ,,% W de r h a tt w, 3 N eo c - 2415+ ,

v Oc4eber f978.

f5 ,e - fr F. O da< et al, " Safe f y h5 0 8o" b %

• Ciee'er4I U h tiri (

Top h gl R tp.. f ; Q u=G t 'c a t h n of tac on,-beken s.h F Co , e Tr=44.ent Ar. del 4.< 8estia3 Wde r Renef e-a "

N E t>o - 2 4 r 54. e et U fi O F. -If 15 + ~ 0, Wlu mes I, H , II2 , ' 7 6 0.

15.0-12

a e 4

J .

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l ll r

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t Fasts 15..-1 (page 1 of 33 sesotts sWeens? or Tennstant svsufs apettcasts To suas l

NARINWE Cons Avtases sush-

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) EDs 50- 283gs1FT1oM rtos reassoas passsues Passsess 5 or .

Fas00suCT stow. scene 15.,

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15.1.2 m n. o, ,.e..et.r 1 no. co. trol at.r.

,g g g g ,. ,,,. . ...

15.1-3 Peed.ater Controller rai ere, M*'1"*"

"'/ jl.*JN [ III.5 ll&I ll% lllC 0C.0 (g) M Md D * **'

15.1 3 15.1-4 Pressure segulator 0.04 re11ere - opes 14.3 1149.. 1165..

1 114.. . 1.. 3 >g 4 5 3.2 15.1.4 -

Ina49ertent opening of t 0A6 1

safety or selief Valve see Text b

15.1.6 -

Inadvertent BNs Shetdown ' , *.

Cooling operation see Test -*

a g

15. 2 -

INCs Ass IN esACTo. PasSSUse 15.2.1 -

Pressere negolator Fallare - closed see 15.2.2 and 15.2. 3 gerpass ont a 15.2.2 15.2-1 Generator toad sejection, ..

Tripscram$)efpass,and ar? - on ( $$ lib 9 ll93 li k4- l0l.~2. Mr 1

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,2,3)ct- 0 .L se? - on (5p 1225.. 11 .0 M ,gg d b 14 1.7 15.2.3 15.2 3 Turbine Trip, Trip scram, and Prf - on 163. 3 1174.0 1996.. 1969.0 1 2.0 .l. 4e"q, a 14 5.s 15.2.3 15.2-4 Turbine Trip, Trip screst, sypass - orr, are - on(U 1T8.4

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L35 PSAR TASta 15.0-1 l Cont'41 (page 3 of 33 914531NR0 Cops AVBERSE 303A-IERIIWit setFACE y 50. OF T105 Innsrasu maarmen mAntman sTsam asAT vnLyss or 7434- 18tU7305 BONE VBSSEL LIIIB FLUE 1ST SLOID-GRAPE FIS005 FLUX POBSSUBS FEB5SURE POSSSUSB 5 OF F8809tKY SLDED- total Wa E9a IIIG13211E 1 L8DE 1991sl JP9L9L Jaelst IBITIAL AGES GAE399EI M E9WE .iansk 15.4.2 . sofs - At fouer See Test b s <. 3 - Contro! 8o4 setemperation see 15.4.1 and 15.4.2 b

6. 4. 4 15.4-5 Abnormal startup of Idle socirculation Loop e54.9 981.0 996.8 977.0 150.8 . Mets"(G) a e 9.0 15.4.5 15.4-7 tocircolation Flow Control i Failure - Increasing Flow 302.3 982.0 1991.0 928.0 145.1 +trSE' a O 9.0 15.4.7 - letoplace4 Bendte Accident see Test b I'

J 15.4.9 - tod trop Acc14ent see Test c 15.5 - IIOCBBAsE IN BEACTOS

, COOLAllT IIIVIIITOSY 15.5.1 15.5-1 Inadvertent NPCI Peep (N l start 127.0 1023.e tese.e 1013.0 107.7 .6,97-<0.16 a e e.e j 15.5.3 = EnfB Transtante see appropriate events in 15.1 and 15.2 tes a = Incidente of moderate fregeency

! b = Infrequenc incidente c = Limiting feelte

( CPR.s ^^t .d a cPR 'M WU, '

i o3w- u A J A g%. p 2 L J n A L L W L,( A MCfRN

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V '

TABLE 15.0-2 (Pd'ge1of3)

+

INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS

, 1. Thermal power level, MWt Warranted value 3.293 x 10+3 Analysis value - radiological 3.458 x 10+:

consequence

- transient analysis 3.435 x 10+3

2. Steam flow, Ibs per hr Warranted value 1.416 x 10+7 Analysis value 1.486 x 10+7 ,

3., Core f15w, lbs per hr 1.0 x 10+8

4. Feedwater flow rate, Ib per sec

Warranted value 3.958 x 10+

! Analysis value 4.129 x 10+:

l j

n U

5. Feedwater temperature,OF 4.zS W x 10+:

t.oto

6. Vessel dome pressure, psig .1 046 x 10+3

, i,o 3 l

7. Vessel core pressure, psig L O?? x 10+3

8. Turbine bypass capacity, percent NBR 2.5 x 10+1

9. Core coolant inlet enthalpy, g, q 4, Btu per Ib JrH6 x 10+a

10. Turbine inlet pressure, psig 9.6 x 10+a

11. Fuel lattice p 8 x 8 F.

12. Core average gap conductance, Btu /sec-fta_oy # o,i744

13. Core leakage flow, percent 10.

14. Required MCPR operating limit 4-2t 8" I*3 ISi0 -3 And Ti b:e t fp - 5

15. MCPR safety limit 1.06

16. Doppler coefficient -4/0F Nominal EOC-1 2.29 x 10-1 Analysis dataG) 2.06 x 10-2

LGSFSAROL RAFT t

TABLE 15.0-2 (Cont'd) (Page 2 of 3)

17. Void coefficient -4/% Rated voids Nominal EOC-1 7.61 Analyses data for p wer increase eventsw 1.271 x 10+8 Analyses data for power decrease events 3.63 4 ,3 +5-

18. Core average rated void fraction, percent 4 rett x 10+8

19. Scram reactivity, $Ak Analysis dataf3) Figure 15.0-2

20. Control rod drive speed, position versus time -

Figure 15.0-2

21. Jet pump ratio, M 2.0

() 22. Safety / relief valve capacity, percent NBR At 1142.0 psig 8.74 x 10+1 Manufacturer Target Rock Quantity installed 14 l

23. Safety relief function delay, seconds 4.0 x'10-2 s.T

24. Safety relief function response, seconds 4,4P x 10-1

25. Setpoints for safety / relief valves (a)

Safety relief function, psig 1142.0, 1152.0, 1162.0

26. Number of valve groups simulated Safety relief function, Number 3

27. High flux trip, percent NBR 12.1.0 . l.2.62.

Analysis setpoint (+tef0x1.043) 1.23-2 x 10+2 l

l 28. High pressure scram setpoint, psig 1.071 x 10+3

29. Vessel level trips, feet above bottom of separator skirt (42.83 ft above vessel zero) i l \ Level 8 - (L8), feet 5.022 6,018 l Level 4 - (L4), feet Sv447- 3.615 Level 3 - (L3), feet -2.167 s.75e Level 2 - (L2), feet -GrO4 2- (-) 3 7 c 8

-,,----n- ,w-- .- - -p_,.--,,..ww-.-----,,

o me, DRAFT .

TABLE- 15.0-2 (Cont 'd) (P 'ge 3 of 3)

30. Recirculation pump trip delay, 4

. seconds .

1.75 x 10-a

31. Recirculation pump trip inertia time constant for analysis, seconds (2) 4.5 4.o s s'

32. Total steamline volume, fta 4,4M x 10 + 3 (2) The inertia time constant is defined by the expression:

. t = 2Jnn gTo where .t = inertia time constants (sec)

Jo = pump motor inertia (Ib-fta) n = pump speed (rps) g = gravitational constant (ft/sec a)

To = pump shaft torque (Ib-ft)

, ca) Safety analyses are co'nservatively based on these setpoints.

Actual setpoints are 1130, 1140 and 1150 psig.

_ & -K M ~

,_ _ Km n _ _

(3) Appttemble to evevsta n eg ol y z,ec( ti,d n ) m,d,[ 4, *y

'I n rc4creoce ILo-3 l

O .

l

pF; AFT LGS FSAR TABLE 15.0-3

SUMMARY

OF ACCIDENTS FAILED FUEL 4'2f GE NRC WORST PARAGRAPH CALCULATED CASE NO. TITLE VALUE ASSUMPTION 15.3.3 Seizure of One None -

Recirculation Pump -,

15.3.4 Recirculation Pump None -

Shaft Break 15.4.9 to^bslRod Drop Accident < 770 770 4 -t:: ;f re:)

Y-

,_ 15.6.2 Instrument Line Break None None 15.6.4 Steam System Pipe Break None None l Outside Containment i

15.6.5 LOCA Within RCPB None 100%

15.6.6 Feedwater Line Break None None N/A 15.7.1.1 Main Condenser Offgas N/A Treatment System Failure 15.7.3 Liquid Radwaste Tank N/A N/A Failure I 15.7.4 Fuel Handling Accident < 124 124 15.7.5 Cask Drop Accident N/A N/A 15.8 ATWS Special Event Eee_It_x_t -

l O

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

DRAET TABLE 15.0-5 REQUIRED OPERATING LIMIT CPR VALUES Pressurization Events:

CPR (Option A)* CPR (Option B)*

Load Rejection Without Bypass 1.19 1.11 Turbine Trip Without Bypass 1.17 1.10 Feedwater Controller Failure 1.17 1.14 (127% Flow)

Load Rejection 1.14 1.07 Non-Pressurization Events:

CPR Rod Withdrawal Error *** 1.21 Loss of Feedwater Heater 1.22**

• Includes adjustment factors as specified in Reference 15.0-5.

i ** Required OLCPR using either Option A or Option B adjustment factor l regardless of frequency category of the turbine-generator trip events with bypass failure.

[

• • • OLCPR value is obtained for the 107% Rod Block setpoint, Control Cell Core analysis.

1 I

(INSERT) l l PCY:im/5E-1 5/13/83 l

t

(

h h WB M C'u j

$N$" y 0 ,.

l 120 -

l A

. ,00 _

1 i

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l g TYPICAL POWERIF LOW DER ATE LINE r

& K g _

D ALLMED OPERATM D.

r REGION

( 1 40 -

u f CAVITATION REGION O e,

O 20 - N ATUR AL (PUMP) CIRCULATION $ $

MINIMUM PUMP SPEED H H' J' J l l l 40 60 80 100 120 0 20 CORE FLOW (% rated)

LIMERICK GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT TYPICAL POWER / FLOW MAP l

FIGURE 15.0-1

] ...pg LIMERICK MINIMUM OPERATING CPR LIMIT

--- EVENT HINIMUM OPER. LIMIT I 1.25 1.25 LFWH e _ _ _ _ st .

~

>o 1.2o ],pg i

/

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C

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1.05 1.05 z.,as c.,. r, r 0 c. l FIGURE 15.0-3 MINIMUM OPERATING CPR LIMIT VS. SCRAM SPEED INSEAT)

LGS FSAR 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE '

15.1.1 LOSS OF FEEDWATER HEATING 15.1.1.1 Identification of Causes and Frequency Classification

/

J5.1.1.1.1 Identification of Causes Feedwater heating can be lost in at least two ways:

a. Steam extraction line valves to one heater string are closed.

b. A feedwater train is isolated automatically by high water level in either the first or second heaters.

The first case p/oduces a gradual cooling of the feedwater. The second case produces a slight cooling of the feedwater. In both cases the reactor vessel receives cooler feedwater. The maximum l number of feedwater heaters that can be isolated by a single

! event represents the most severe transient for analysis considerations. This transient has been conservatively estimated to incur a loss of up to 1000F of the feedwater heating capability of the plant and causes an increass in core inlet subcooling. This increases core power due to the negative void reactivity coefficient. The transient can occur with the reactor in either the automatic or manual control mode. In automatic control, some compensation of core power is realized by modulation of core flow, so the transient it lae= covere than in manual control. ()Lc 41 p* d 4 75"'

4>* A

15. 1.1.2 Frequency Classificatio d The probability of this transient is considered 15w enougn to 2

, warrant it being categorized as an infrequent incident. However, i because of the lack of a sufficient frequency data base, this transie t i anal zed a an incident of m g naJ erate frequency; A /pp'f n .

[ Fa 1.1.2 Se5uence of ve es and vstem operation 15.1.1.2.1 Sequence of Events Tables 15.1-1 and 15.1-2 list the sequence of events for this transient and its effect on various parameters is shown in Figures 15.1-1 and 15.1-2.

In the automatic flow control mode, the reactor settles out at a lower recirculation flow with no change in steam output. An average power range monitor (APRM) neutron flux or thermal power x alarm alerts the operator to insert control rods to get back down to the rated flow control line, or to reduce flow.if in the manual mode. The operator should determine from existing tables 15.1-1

LGS FSAR f the maximum allowable turbine-generator (T-G) output with , sS---

feedwater heaters out of service. If reactor scram occurs 6 as itJ oes in manual tiew ennerni mode > the operator should monitor the reactor water level, pressure controls, and T-G auxiliaries during coastdown.

15.1.1.2.2 System Operation In establishing the expected sequence of events and simulating plant performance, it is assumed that normal functioning occurs in the plant instrumentation and controls, plant protection, and reactor protection systems.

'~~

Required operation of engineered safeguards f atures ( F) is hot i e pected in,eijher case f 's tr sient /b$tAf @ ct.

=~ m L. eg.p L. .

A.1.2.3 m y. k TheIE

=n'ffec h) f Sincie ailur s and Toedtor Errds ,'~

This transient generally leads to an increase in reactor power L level. Single failures are not expected to result in a more severe transient than analyzed.

See Section 15.9 for a' detailed v- discussion of this subject.

15.1.1.3 Core and System Performange 15.1.1.3.1 Mathematical Model ,

l The predicted dynamic behavior has been determined using a l computer simulated, analytical model of a generic direct-cycle l boiling water reactor (BWR). This model is described in detail l in Ref 15.1-1. This computer model has been verified through '

extensive comparison of its predicted results with actual BWR test data.

The nonlinear, computer simulated, analytical model is designed to predict associated transient behavice of this reactor. Some of the significant features of the model are:

a. A point kinetic model is assumed with reactivity feedback from control rods (absorption), voids (moderation), and Doppler (capture) effects,

b. The fuel is represented by three four-node cylindrical elements, each enclosed in a cladding node. One of the cylindrical elements is used to represent core average power and fuel temperature conditions, providing the source of Doppler feedback. The other two are used to represent " hot spots" in the core to simulate peak fuel center temperature and cladding temperature,

c. Four primary system pressure nodes are simulated. The nodes represent the core exit pressure, vessel dome 15.1-2

=

LGS FSAR pressure, steam line pressure (at a point repr'esentative of the main steam relief valve (MSRV) location), and turbine inlet pressure.

d. The active core void fraction is calculated from a relationship between core exit quality, inlet subcooling, and pressure. This relationship is generated from multinode core steady-state calculations.

A second-order void dynamic model with the void boiling sweep time calculated as a function of core flow and void conditions is also utilized.

e. Principal controller functions such as feedwater'TTow, recirculation flow, reactor water level, pressure and load demand are represented together with their dominant nonlinear characteristics.

f. The ability to simulate necessary reactor protection system functions is provided.

,,. 15.1.1.3.2 Input Parameters and Initial Conditions These analyses have been performed, unless otherwise noted, with

plant conditions as tabulated in Table 15.0-2.

~

l The plant is assumed to be operating at 105% nuclear boiler rated (NBR) steam flow and thermally limited conditions. Both automatic and manual modes of flow control are considered.

The same void reactivity coefficient conservatism used for pressurization transients is applied since a more negative value conservatively increases the severity of the power increase. The values for both the feedwater heater time constant and the feedwater time volume between the heaters and the spargers are adjusted to reduce the time delays since they are not critical to the calculation of this transient. The transient is simulated by programming a change in feedwater enthalpy corresponding to a 1000F loss in feedwater heating.

15.1.1.3.3 Results In the automatic flux / flow control mode, the recirculation flow control system responds to the power increase by reducing core flow so that steam flow from the reactor vessel to the turbine remains essentially constant. In order to maintain the initial steam flow with the reduced inlet temperature, reactor thermal

power increases above the initial value until the flow-referenced l

high flux scram setpoint is reached. The peak heat flux is 113%

l of its initial value and the peak fuel center temperature l increases 4560F. The MCPR reached in automatic control mode is l greater than for the more limiting manual flow control mode.

15.1-3 l

. . nO A gll=1 LJn HE E 5;c e%zt rf/s#m&dJ.C satn w. k !

LGS FSAR iOp g4gg .

O Y0;- PJJEM.

The increased core inlet subcooling aids t'hermal margi6s, -and the smaller power increase makes this transient less severe than the {

manual flow control case given below. Nuclear system pressure does not change and, consequently, the reactor coolant pressure I boundary (RCPE) is not threatened. Since scram occurs, the results are very similar to the manual flow control case. This transient is illustrated in Figure 15.1-1.

In manual mode, no compensation is provided by core flow and thus the power increase is greater than in the automatic mode. A scram on high APRM neutron flux occurs. Vessel steam flow increases and the initial system pressure increase is slightly larger than yj Peak heat flux is JJ4%'%g in the automatic flow control -'

mode.

fu;I rnter

_ '.__--, a;---f its initial The value,,r '

increased core ini t subcooling 7--

aids core thermal margins and minimum MCPR is J 44*./ herefore, is satisfied. The transient responses of the the key plant variab for this mode of operation are shown in d ci;r 5 :i;les Figure 15.1-2. ( g, w com aower inattal power levels for i Tnts stansiwni 1. iw== avve6e two main reasons:

'N a. Lower initial power levels will have initial MCPR values greater than the limiting case initial value assumed.

b. The magnitude of the power rise decreases with lower ,

initial power conditions. Therefore, transients from J lower newer levels *will be less severe,-

l i

15.1.1.3.4 Considerations o!! Uncertainties Important factors (such ,as reactivity coefficient, scram characteristics, magnitude of the feedwater temperature change) are assumed to be at the we.rst configuration so that any deviations seen in the actual plant operation reduce the severity of the transient.

15.1.1.4 Barrier Performance As noted above and shown in Figures 15.1-1 and 15.1-2, consequences of this transient do not result in any temperature or pressure transient in excess of the criteria for which the i fuel, pressure vessel, or containment are designed; therefore, these barriers maintain their integrity and function as designed.

15.1.1.5 Radioloqical Consequences l

Since this transient does not result in any additional fuel failures or any release of primary coolant to either the l

secondary containment or tne environment, there are no radiological consequences associated with this transient.

- 15.1-4 l

m, ,,,, DRAFT 15.1.2.2.3 The Effect of Single Failures and Operator Errors e

In Table 15.1-3 the first sensed event to initiate corrective action to the transient is the vessel high water level (LB) trip.

Multiple level sensors are used to sense and detect when the water level reaches the L8 setpoint. At this point in the logic, a single failure will not initiate or prevent a turbine trip signal. Turbine trip signal transmission, however, is not built to single failure criterion. The result of a failure at this point would have the effect of delaying the pressurization

" signature." High levels in the turbine's moisture separators will result in a trip of the unit before high moisture levels enter the low pressure turbine. However, if excessive moisture enters the turbine it will cause vibration to the point where it would also trip the unit.

Scram trip signals from the turbine are designed such that a-single failure neither initiates nor impedes a reactor scram trip initiation. See Section 15.9 for a discussion of this sutryect. _

15.1.2.3 Core and System Performance 15.1.1.3.1 Mathematical Model E<T7W3 PA45

~$

o _

[q -._

%_- c p y ect @ 5 M h Ao j 15.1.2.3.2 Input Parameters and Initial Conditions 1

These analyses have been performed, unless otherwise noted, with the plant conditions as tabulated in Table 15.0-2.

(

t The same void reactivity coefficient used for pressurization transients is applied since a more negative value conservatively increases the apparent severity of the power increase. End of in A~tl l cycle ' 11 red: cuthscram characteristics are assumed. The MSRV opening is conservatively assumed to occur with higher than nominal setpoints. The transient is simulated by programming an upper limit failure in the feedwater system such that M i U '4 feedwater flow occurs at a system design pressure of 1060 psig.

15.1.2.3.3 Results The simulated feedwater controller transient is shown in Figure 15.1-3. The high water level turb trip and feedwater pump trip are initiated at approximate 1 4 .. seconds. Scram occurs l simultaneously from stop valve closure and limits the neutron i flux peak and fuel thermal transient so that no fuel damage

[ +9xxee=

occurs. MCPR remains} bev: ' Da :nd paak fUe1 < enter-temper 4t

'^ s than M5ay The turbine bypass system opens to limit peak pressure in the steam line near the MSRVs to 4+N M g "

and the pressure at the bottom of the vessel to abouty e psig.

g c n3idecdly deve fM 5.M y h*-s R, l 15.1-6 i

l

INSERT M Ir. l t.1 l- p.iyt.

s i

DRAFT The predicted dynamic behavior has been determined using a computer simulated, analytical model of a generic direct-cycle BWR. This model is described in i

f detail in Reference 15.1-1. This computer model has been improved and veri-fled through extensive comparison of its predicted results with BWR ,

l test dasa. .

i i The nonlinear computer simulated analytical model is designed to predict associated transient behavior of this reactor. Some of the signift, cant

(

l features of the model are: ~

I

( a. An integrated one-dimensional core model is assumed which includes a detailed description of hydraulic feedback effects, axial power shape changes, and reactivity feedbacks.

I-

b. The fuel is represented by an average cylindrical fuel and cladding m/.c.

, for each axial location in the core. ,

c. The steam lines are modeled by eight, pressure nodes incorporating mass and momentum balances which will predict any wave phen <eena present in 8 the steam line during pressurization transient.

d. The core average axial water density and pressure distribution is calculated using a single channel to represent the heated active j - flow and a single channel to represent the bypass flow. A model, representing liquid and vapor mass and energy conservation and six-

- ture momentum conservation. is used to describe the thermal-hydraulic behavior. Changes in the flow split between the bypass and active

- channel flow are accounted for during transient events.

c l

s

[

i IA{ SERT FbK /r.l.2.3.)

f.1 of 2-l

.~~. ***

Di4 AFT

e. Principal controller functions such as feedwater flow, recirculation flow, reactor water level, pressure and load demand, are represented together with their dominant nonlinear characteristics.

f. The ability to simulate necessary reactor protection system functions is provided.

l l

g. The control systems and reactor protection system models are, for the most part, identical to those esployed in the point reactor model, which is described in detail in Reference 15.0.4-2 and used in analysis for other transients.

l l

0 g--.,---,,---,n,-w,.,,--

6

} LGS FSAR 15.1.6.4 Barrier Performance DRAFT As noted above, the consequences of this transient do not result in any temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel, or containment are designed, therefore, these barriers maintain their integrity and function as designed.

15.1.6.5 Radiolocical Consequences Since this transient does not result in any fuel failures, no analysis of radiol.ogical consequences is required for this transient.'

15.

1.7 REFERENCES

15.1-1 R.B. Linford, " Analytical Methods of Plant Transient )

Evaluations for the General Electric Boiling Water i Reactor," April 1973 (NEDO-10802).

I f, l- 2 G e.ne a l E( eefdc. , Q v di Wc cNe n J 4h. One Nmen Jonst Ceee Tv g A sie J M.d e I fo r 13eiling W ct% r Reacf. 2,

) N EDO - 2,4 I 94 3 Oc4 *er I?78 l

l l

l l

I l

l - j l

..)

15.1-15

I LGS FSAR TABLE 15.1-1 D~RrFT .,

SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATING (FIG. 15.1-1)

TIME-SEC EVENT l 0 Initiate a 1000F tetaperature reduction in the feedwater system.

l I 5 Initial effect of unheated feedwater starts to raise core power level but AFC system automatically reduces core flow to main-tain initial steam flow.

I 166.3 APRM initiates reactor scram on flow referenr'ed high neutron flux.

b 4.,%& $.

l l

l l

l

LGS FSAR TABLE 15.1-2 DRAET i

SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATING (FIG. 15.1-2) l TIME-SEC EVENT 0 Initiate a 1000F temperature reduction into the feedwater system.

5.0 Initial effect of unheated feedwater to I

NO Qw raise c re power level a d steam kn?- flo &w os.,

73 9 Q e.

JtPRM initiate reactor scram on higM '

neutron flux. -

m.

-+ih4 984 "';rbin: centrel 1 1 :: etsrt *^ Niere-to

, N:- &. ulate pr;;; :;.

• V m r!= & m&

90-t-/0/,g- ..";rr;  ;;r.;; -etes le.el e-c;;d; L8 high 1 level setpoint ::rltin; ir tripret the l main turbine and feedwater turbines.

l - Turbine bypass operation initiatedJ 80.2 (RPTinitiatedwhenturbinestopvalves reach the 90% open position.

40 4/0/,4 'TLebine st valves clospd O O .

N- r -- t u,M "0. " /o/,7 Rec ationpu[pmotorcirquit reakers open causing decrease in core flow to natural circulation.

Turbine inlet pressure exceeds pressue regulator setpoints and bypass valvae been to reamin nrm ==nre control)

. (Sb.

>;20.0 Reactor variables settle into new steady state.

l

DRAFT LGS FSAR TABLE 15.1-3 SEQUENCE OF EVENTS FOR FEEDWATER CONTROLLER FAILURE

_ (FIG. 15.1-3)

TIME-SEC EVENT 12"I O Initiate simulated failure of M6%

upper limit on feedwater flow.

s.7 - t 49rt- L8 vessel level setpoint trips main turbine and feedwater pumps. Turbine byptre ; N ;.. .. initict-A.

2.7,'Lt ._

18,-a- Reactor scram trip actuated frem main

. turbine stop valve position switches.-

2. 7. 2.1

49 rib RPT actuated by stop valve position switches.

27,3

+9 Main turbine stop valves closed and l

f turbine bypass valves start to open.

! \ 2a . 4-

, 44v&- Recirculation pump motor circuit breakers open causing ferr rr: in-ecre-y 'lew 2 eweflow te n:turalb sketcir:deti^ne t cen t4 dewn y e c.i'e c u bM *1 21.4 )l5RVs actuated due to high a Qdeg /p' sur ' T '

'/

I n R /\ '

1 27.4(est)

All' MSRV lo\ ,syd )

7,50.0 , Turbine bypass valves closed'.

', L '

v i j.

750.0 -

d Mainsteamlineisolationand~RCI) wide ranje Cap /

HPCIAystems low level *(L2).in}iReli4f iation on groSp 1,' l' nn.n and closed on pressure. cyc /es.d) 21.0 Firsr s+ safet dat +e3rsoo kis h eveu vre. yj'etineV v' M s,J of' l

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LIMERICK GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT LOSS OF 1000 FEEDWATER HEATING AFC FIGURE 15.1 1

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LIMERICK GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT LOSS OF FEEDWATER HEATER, U MANUAL FLOW CONTROL FIGURE 15.12

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. d g4~ LIMERICK GENER ATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH LEVEL TURBINE TRIP FIGURE 15.1-3 l

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ORAFT LGS FSAR O designed to satisfy single failure criteria. An evaluation of the most limiting single failure (i.e., failure of the' bypass system) was considered in this transient. Single failure analysis can be found in Section 15.9.

15.2.2.3 Core and System Performance 15.2.2.3.1 Mathematical Model.

A The computer model described in Section 15.l y2'.3.1 was used to simulate this transient.

15.2.2.3.2 Input Parameters and Ini'tial Conditions

! These analyses have been performed, unless otherwise noted, with the plant conditions tabulated in Table 15.0-2.

The turbine electrohydraulic control (EHC) system detects load l

rejection before a measurable speed change takes plac q Theclosurecharacteristics'oftheTCVsareassumedtohaN'a

full stroke closure time, from fully open to fully closed, of i 0.15 seccnds.

Auxiliary power would normally be independent of any turbine-generator overspeed effects,.and be continuously supplied at rated frequency as automatic fast transfer to auxiliary power supplies occurs. For the purposes of worse case analysis, the recirculation pumps are assumed to remain tied to the main generator, and thus increase in speed with the turbine-generator overspeed until tripped by the RPT system.

The reactor is operating in the manual flow control mode when load rejection occurs. Results do not significantly differ if the plant is operating in the automatic flow control mode.

The bypass valve opening characteristics are simulated, using the specified delay together with the specified opening characteristic required for bypass system operation.

Actual closure of a main steam isolation valves (MSIVs) as caused by low water level trip (L2), and actual flow from initiation of reactor core isolation cooling (RCIC) and high pressure coolant injection (HPCI) core cooling system functions do not occur during the duration of the simulation. If these events occur, they will follow sometime after 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.

O 15.2-5

LGS FSAR 15.2.2.3.3 Results D R- . A P Y 3 15.2.2.3.3.1 Generator Load Rejection with Bypass Figure 15.2-1 shows the results of the generator trip from 104.3%

rated power. Peak neutron flux rises to ?"0.7% of NBR conditions. . 17 8,5 */,

y a o s.

The average ' surface heat flux peaks at vet 4% of its initial.

value, d J1CPR does not icantly decrease below its initial value. y L fo & p % M_ o,e3) 15.2.2.3.3.2 Generator Load Rejection'with Failure of Bypass 106 1 tts.F l

f IFigure 15.2-2 shows that, in the case of bypass failure, peak l 1 neutron flux reaches abou 194=P%, and average surface heat flux 1

reachesAt0^.7%, of the initial value. Because this transient is classified as an infrequent incident, it is not limited by the GETAB criteria, and the minimum critical power ratio (MCPR) limit is permitted to fall below the safety limit for incidents of moderate frequency. "CI* reaches--*--64-fte thir trans44hW. NMOg d CDL f Consideration

% uM k M% oM k Q fM a r,e 6 5.2.2.3.4 of Uncertainties /

l The full stroke closure time of the TCVs of 0.15 seconds is conservative. Typically, actual closure time is more like 0.2

)#

seconds. Thus, the shorter time chosen for closure results in a more severe pressurization effect.

l All systems utilized for protection in this transient were l

assumed to have the poorest allowable response (e.g., relief setpoints, scram stroke time, and work characteristics).

Anticipated plant behavior is, therefore, expected to reduce the actual severity of the transient.

15.2.2.4 Barrier Performance 15.2.2.4.1 Generator Load Rejection Peak pressure remains within normal operating range and no threat to the barrier exists. ,

l 15.2.2.4.2 Generator Load Rejection with Failure of Bypass l 119 0-Peak pressure at the safety relief valves reaches 4497- psig. The peak nuclear system' pressure reaches 1225 psig at the bottom of the vessel, well below the nuclear barrier transient pressure limit of 1375 psig.

8 15.2-6

LGS FSAR 15.2.3.3 Core and System Performance k

15.2.3.3.1 Mathematical Model .

The computer model described in Section 15.1.1.3.1 was used to

. f-v*. n by r=w*sth simulate these transients, >. Mia.cm.(

we eby..,.e wiu si t.id.J

+= css, hia per.4;. r , The sn.t i.. ,, , im Gg, e p.

, , ., ,e e.. ,

15.2.3.3.2 Input Parameters and Initial Conditions ,,,3,y, These analyses have been performed, unless otherwise noted, with

. plant conditions as tabulated in Table 15.0-2.

2 Turbine stop valves full stroke clossre time is 0.1 second.

l l

A reactor scram is initiated by position switches on the stop l valves when the valves are less than 90% open. This stop valve I trip signal it automatically bypassed when the reactor is below 30% N3R power level. -

Reduction in core recirculation flow is initiated by position switches on the main stop valves that trip the recirculation pumps.

15.2.3.3.3 Results ,

15.2.3.3 3.1 Turbine Trip A turbine trip with the bypass system operating ncrually is ,

simQlated at 104.3% NBR stear flow conditions in Figure 15.2-3.

Neutron flux increases rapidly. because of the void reduction caused by the pressure increase. However, the flux increase is limited to 163.3% of rated value by the stop valve scram and the RPT system. Peak fuel surface heat flux does not exceed 102% of its initial value. v."CPR Th e afec th.i icenei;nt i: %w 15.bi.e frip sui % gss ,Jbyyur f aile,e c FR. s l = 85 4 A** be 15.2.3.3.3.2 Turbine Trip with Failure of Bypass deu

..d n .. ni w ea ebget,,,,,* ~ lsech s.t..U.9 l

A turbine trip with failure of the bypass system is Nm*u"lat)$ Yb,;,",,"'"-

104.3% NBR steam flow conditions in Figure 15.2-4.[ gg 19 5.+ l_r,wri - v Peak neutron fluz reaches 368 2% of its rated value, anu peax sudace lie (l ft:1 :::ter t ;rrett:: in::::::: ;;;;;;;;;inct:1y tSt^r., Because this transient is classified as an infrequent incident, it is not l limited by the GETAB criteria and the MCPR limit is permitted to l fall below the safety limit for incidents of moderate frequency.

However,1th; ""PC fec ^ :.i; t::::icat i; ' .00, tich i: j : t O t-th; ::f;ty limit frr i :id: t ;f ::f:::t: f. ; ..n.y. andy th;;;i : , th: ft:i : brei: ' =tia!i :.

9 I% e. ca . e in c PR 4., +kis and is o.o 6, en el MCPR for 41.'s i 4ra n see h .)

s <c - .q- ims a.- usa.s na + .. v 15.2-10

m LGS FSAR vessel bottom. Therefore, the overpressure transient is., clearly below the reactor coolant pressure boundary transient preisure e

limit of 1375 psig. 2,. l..... es... . 2. . . . . . . - - - - J 'J ; ^

W 15.2.3.4.2.1 Turbine Trip with Failure of Bypass at Low. Power Qualitative discussion is provided in Section 15.2.3.3.3.3.

15.2.3.5 Radiolooical Consequences i

Although the consequence of this trancient does not result in

fuel failure, it does result in the discharge of normal coclant l nctivity to the suppression pool via MSRV operation. Because l this activity is contained in the primary containment, there is l no exposure to operating personnel. Tais transient does not l result in an uncentrolled release to the environment, so the l plant operator can choose to leave the activity bottled up in the containment or discharge it to the environment under controlled release conditions. If purging of the containment is chosen, the j release will be in accordance with established technical 3 specifications and, at the worst, would on y result in a sr.all increase in the yearly integrated exposure level.

15.2.4 MSIV CLC3URES 15.2.4.1 Identification of Causes and Frecuency Claksification 15.2.4.1.1 Identification of Causes Various steam line and nuclear system malfunctions, or operator actions, can initiate MSIV closure. Examples are: low steam line pressure, high steam line flow, high steam line radiation, low water level, or manual action.

15.2.4.1.2 Frequency Classification 15.2.4.1.2.1 Closure of All Main Steam Isolation Valves This transient is categorized as an incident of moderate frequency. To define the frequency of this transient, as an initiating transient and not the byproduct of another transient, only the following contribute to the frequency: manual action (purposely or inadvertent), spurious signals such as low pressure, low reactor water level, low condenser vacuum, ete; and, finally, equipment malfunctions such as faulty valves or operating mechanisms. A closure of one MSIV may cause an immediate closure of all the other MSIVs, depending upon reactor conditions. If this occurs, it is also included in this category. During the main steam isolation valve closure, position switches on the valves provide a reactor scram if the valves in three or more main steam lines are less than 90% open, 15.2-12

LGS FSAR I -

TABLE 15.2-1  %

SEQUENCE OF EVENTS FOR T-G LOAD REJECTION WITH BYPASS (FIGURE 15.2-1)

TDfE-SEC EVEMY

, (-)O.015 Turbine-generator detection of loss of electrical (approx.) load gg g 0.0 Turbine-generator [LU) devices trip to ir.itiate, TCV fast closure 0.0 Turbine-generator PLU trip initiates main turbine bypass system operation O.0 N [ control valve ( CV) closure initiates scram I

trip CV closure initiates a recirculation pump trip (RPT) i 00 0.07 TCV closed WD II to i

4. 87 F DE,bfr;eh,yp,tsgvalygst,ajg;v,o,gg,,,p,,n,,g,,$

cut.c M d.,, $ d E ws+ e 4

4 gt M %.o Gro;ip 1 MSRVs actua ted e.e.t.d.w4 4 ,4 2.,15 Group 2 MSRVs actuated 4re 2,40 Group 3 MSRVs actuated I Gv.-r 1 M SRVs cI*se.

4,4- 3 90 AMI :: EVs-etamed O

, LGS FSAR - A Tg l M TABLE 15.2-2 '-

SEQUENCE OF EVENTS FOR TG LOAD REJECTION WITHOUT BYPASS (FIGURE 15.2-2)

TIME-SEC EVENT ,

l

! (-)0.015 -lurbine-genatator detection of loss of electrical l (approx.) load '

l pwer Ion / .m u,sfrove

'.0 O J.irbine Ner.erator 4 (PLU) devices trip to initiate b4'i e v (' +

co ROVt;utvc fast closure -

l 0.0 Tet.bine bypers valves f eil to operate F g.t ... nte.t v 4 e ( f= c v ) cl u o re.

0.0 ATpew t.l .initiates s v r t. ceram trip'.~ p.my -h ,'p reci-cvt.dn 0.0 -

FCV 4 initiates M. PI). -

0.07 t

t

o. s v s- 'h,.cv

<c eotoclose..d 4 e..m eei oto(..

d..ti, . f Erc~m.:

p m r..

4i. ,+.c...+4. ,*

3 ,e e., co In A. e l +r?- 3.+ 1 Group i l'9RVa actuated M ; .6 o Group 2 55RVs aletuated 4,4 J . (= f Group 3 MSRVs actuated G-e u r 1

%$61 AAH MSRVs closed -

40.O T "" r;:1ie ::t : tin ;r. g ;;:::: d::;r.d 1

Q

t

"" "^"

DRAFT TABLE 15.2-4 SEQUENCE OF EVENTS FOR TURBINE TRIP WITHOUT BYPASS (FIGURE 15.2-4)

E TIME _-SEC EVEN'P 0.0 Turbine tefp inittates closure of main stop valves ,

l 0.0- Turbine bype;s ve.1ves fail to cptwrate

! 0.01 Main turbine stop valves reach 9 3 open position l

And initiate reactor occar trip l 0.01 -

valves resch 9 % open position t

Main tur hine a4 and initiate sto(N;FT).

I rees're* M <e.>>> re 'r 4.">f 0.j~5 c.

Turbine stop valves closed . -

c c r, . g4 < 6 as P'mr gm *t'" rf**'i* 6"**"*'s ******'"'3 W-

! 4 r+- f,47 dEo~u['1 N5RVN"*achu~a' tid l

• 4:4 1.43 Group 2 MSRVs actuated '

me- s .l o Grcup I MSEVs actuated G <e o ys i h 6.1- A444- MSRVs closed t0.7 MRV ciclic rrtuatir  ;;;;; . J; .;;;'

l U

l l

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LIMERICK GENERATING STATION ph UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT

. GENERATOR LOAD REJECTION, TRIP SCRAM, l BYPASS - ON, RPT - ON FIGURE 15.21

,,--- ~ . ,

l l

~

i I EUTID4 FLUX 1 VeYSEL IWS HISE IPSil 2 PEAK FUEL EENTER Tele 2SintM S RISE (P513 150. - 3 RVE SURfl C11(RT FLUX 3 MIEI' '; I'3 300* ts (qDCv- V i 4 FEEDMAIEI FLOW g.:

i

' S VESSEL $1EfM FLOW 5 erre:S V VE FLOW (el n I'fdC SIE Fli>! IPCil 100. 4 O 200.< ~ - - -- --

y---

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TIE 15Ecl Titt 15ECl f e ,g o r e l g k o. G CI tIE J C4 'd r b # 8 lell (M16 LB C0i 15'4 DIY 672C16 -

one seer

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o a DRAFT

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V y LIMERICK GENER ATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT s

GENERATOR LOAD REJECTION, d WITHOUT BYPASS FIGURE 15.2 2

I

• l l EUTRON FLUX ( l  : VESSFL PSIE9 RISE (PSil 2 PEAK FUEL EENTER Tele  !

l

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( g g4 kr,C bsttd $ 4 *Lt5.*'a Neb 01 W IW1 IMOS LH COI 151 DRF 672CiG l 4;*.

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s DRAFT' i :

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LIMERICK GENER ATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT TURBINE TRIP WITHOUT BYPASS, O' TRIP SCRAM FIGURE 15.24

I lj ,

i NEUTRON FLUX 2 PEIM FUEL CENTER TEMP 1 VES$ft PIE'i RISE IPSin 2 T1H LINE IHES RISE (PSil ISO * - 3 HVE SURFI CE,_lEAT Fj.UX 300* --  ? SU' EII 'I I3l --f

- 3 0W tel 4 f ELLINRill FLON ti'llLL il Vi l.VE ILOW fel 5 VESSEL SIEFN FLOW 9 I:HWSS Vi'L VE FLOW tel 511:II) Siff at FLOW IPC11 5 100. 200. - - --

i Y f4 I

+R/\f+5-.h-M

= 50. { x, .

100. -

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fi f 5 8 35 t-.SE _ . .. _

-t O. 2. 4. 6. 8. D. 4. u. 6. a.

TIE (SEC) , T!K t$FCI I LEVELilWH-REF-SEP-SMIRT l 2 W R SENSEO LEVELIINCHESI ll VO!D

.2 DIVPLERF1EArilVITT FEntTiy Tt 200* 3 N R SENSED LEVELIINCIES) y* aN 3 TIA HUM.W 4 CORE 1HLE I FL(Di IPCT) , ~ T l @ AL DFT G v! F 5 DRIVE FLCW I (PCTI N J -

L 1 D 100 0W'x -

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O. 6. 8. O. 1. 2.

TIE ISEC) ,

TIME IStri turhane b.Q u .bM bYOi' 53 1911 OM06 TT C01 1 9 DN 672CIS

, TURBINE TRIP HITHOUT BYPRSS. N ansesnis Figure. 15.2 -4

LGS FSAR e

4 flow in the active jet pumps increases considerably and produces approximately 154% of normal diffuser flow and 58% of rat core flow. Minimum critical power ratio (MCPR) remains -et- oltre< ,

2;; n ..e =ly " ;. :the operating limit; W 4 thus, the fuel thermal limits are not violated. During this transient, level swell is not sufficient to cause turbine trip and scram.

15.3.1.3.3.2 Trip of Two Recirculation Pumps [

Ff.gure 15,3-2 shows this transient, with minimum specified rotatire; iner tia, in graph form. MCPR remains unchanged at + 24.

No scram is initiated directly by punp trip. 'ine vessel autea-level spell due to rapid flow coastdown is expceted ta reac.n tr.e high level trip, thereby shutting down the main turbine and . feed pump turbines and scramming the reactor. Subsequent events, such ,

as main steam line isolation and initiation of RCIC and HPCI systems occurring later in this transter.t, have no sigt:ificant <

effect on the results.

15.3.1.3.4 Consideration of Untertainties i

Initial ronditions chosen for these anal .aes are cor.servative nad 3 '<

tend to torce analytical resuj ts to be not e s+ vere than those ext:ected under actual plant conditiens.

Actual pump and pump motor driveline rotatlag antertias are expected to be somewhat greater than the minimum design values assumed in this simulation. Actual plant daviations regarding inertia are expected to lessen the severity of the results indicated by this analysis. Minimum design inertias were used as well as the least negative void coefficient, because the primary concern here is flow reduction.

15.3.1.4 Barrier Performance 15.3.1.4.1 Trip of One Recirculation Pump Figure 15.3-1 results indicate a basic reduction in system pressures from the initial conditions. Therefore, the RCPB barrier is not threatened.

15.3.1.4.2 Trip of Two Recirculation Purps The results shown in Figure 15.3-2 indicate that peak pressures stay well below the 1375 psig limit allowed by the applicable code. Therefore, the barrier pressure boundary is not threatened.

15.3.1.5 Radiological Consequences While the consequence of this transient does not result in fuel failure, it does result in the discharge of normal coolant 15.3-4

o, e LGS FSAR f k

of the failure. Pressure regulator failures are discussed in Sections 15.1.3 and 15.2.1.

The effect of a single failure in the level control system has rather straightforward consequences including level rise or fall by improper control of the feedwater system. Increasing level. '

wil' trip the turbine end automatically trip the HPCI systea of f.

This trip cignature is v.Jready described in the failure of feedvater controller with increasing flow. . Decreasing level vill automatically initiate scram at the L3 level trip and will have a signature similar to loss of feedwater control - decreating flow, l

15.5.1.3 , Core and Svr_tgm Perfngpance ,

15.5.1.3.1 Mathematical V.odel Tto detailed nonlinear dynamic model described briefly ir Secticn 15 2.2.3.1 in used tc simulete this tranutent.  ;

23.5.1.3.,1 Input Parameter and 1citial Conditions

Tcis analysis has been perforned, unless ntherwise noted, with plant conditions tabulated in Table 15.0-2 The water temperature of the HPCI systen was assumed to be 4CoF 4 with an enthalpy of 11.0 Etu/lb. ,

Inadvertent startup of the HPCI system wks chosen for analysis since it provides the greatest auxiliary source of cold water for the vessel.

15.5.1.3.3 Results Figure 15.5-1 shows the simulated transient for the manual flow

control mode. It begins with the introduction of cold water into i the upper core plenum. Within one second the full HPCI flow is established at approximately 19.7% of the rated feedwater flow l rate. No delays were considered because they are not relevant to the analysis.

Addition of cooler water to the upper plenum causes a reduction in steam flow which results in some depressurization as the pressure regulator responds to the event. In the automatic flow control mode, following a momentary decrease, neutron power settles out at a level slightly above operating level. In the manual mode, the flux settles out slightly below operating level.

In either case, pressure and thermal variations are relatively small and no significant consequences are experienced. MCPR)

?'-- st=> 2e e.0 therefore fuel thermal margins are maintained.

ggg

(

_}

15.5.1.3.4 Consideration of Uncertainties /b$d[j j 15.5-2 L