ML19312E977

Mark II Design Assessment Rept, Revision 8

ML19312E977
Person / Time
Site:	LaSalle
Issue date:	06/01/1980
From:	COMMONWEALTH EDISON CO.
To:
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ML19312E976	List: ML19312E975 ML19312E977
References
NUDOCS 8006180281
Download: ML19312E977 (37)

Text

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J s LSCS-MARK II DAR Rev. 8 6/80 LA SALLE COUNTY STATION MARK II DESIGN ~ ASSESSMENT REPORT INSTRUCTIONS FOR UPDATING YOUR LSCS-MARK II DAR i

Changes to the LSCS-MARK II DAR are identified by a vertical line in the right margin of the page. To update your copy  ;

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of the LSCS-MARK II DAR, remove and destroy the following pages and figures and insert the pages and figures indicated.

REMOVE INSERT Page iv Page iv '

Pages vil and viii Pages.vii and viii Page x Page x ,

4 After page 3.4-4 Page 3.4-4a l l After page 3.4-8 Pages 3.4-9, 3.4-10, and 3.4-11

Page 4.3-1 Page 4.3-1 Page 4.4-7 Page 4.4-7 Page 5.3-2 Page 5.3-2 Pages 6.1-1 through 6.1-5 Pages 6.1-1 through 6.1-5 Pages 6.2-1 through 6.2-17 Pages 6.2-1 through 6.2-14

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Figures 6.2-1 through 6.2-8 Figures 6.2-1 through 6.2-6 f

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LSCS-MARK II DAR Rev. 8 6/80

() TABLE OF CONTENTS (Cont'd)

PAGE 5.4 BOP PIPING ANALYSIS 5.4-1 5.5 EQUTPMENT (BOP) 5.5-1 5.5.1 Reevaluation and Design Assessment Methods 5.5-1 5.5.1.1 Analysis 5.5-1 5.5.1.1.1 Static Analysis 5.5-1 5.5.1.1.2 Dynamic Analysis 5.5-1 5.5.1.1.2.1 Acceptance Criteria 5.5-2 5.5.1.2 Testing 5.5-2 5.5.1.2.1 Single Frequency Testing 5.5-2 5.5.1.2.2 Random Frequency Testing 5.5-2 5.5.1.2.3 Acceptance Criteria 5.5-2 6.0 SUPPRESSION POOL WATER TEMPERATURE MONITORING SYSTEM 6.1-1 6.1 SYSTEM DESIGN 6.1-1 6.1.1 Safety Design Basis 6.1-1 6.1.2 General System Description 6.1-1 6.1.3

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() 6.1.4 Normal Plant Operation Abnormal Plant Operation 6.1-3 6.1-3 6.2 SUPPRESSION POOL TEMPERATURE RESPONSE 6.2-1

, I 6.2.1 Introduction 6.2-1 I 6.2.2 Temperature Response Analysis 6.2-2 l

, 6.2.2.1 Model Description 6.2-2 6.2.2.2 General Assumptions and Initial Conditions 6.2-3 6.2.2.3 Description of Non-LOCA Events 6.2-5 6.2.2.3.1 SORV at Power 6.2 5 6.2.2.3.2 Isolation / Scram 6.2-5 6.T;.2.4 Small Break Accident , 6.2-6 6.2.3 Results/ Conclusions 6.2-6 7.0 PLANT MODIFICATIONS AND RESULTANT IMPROVEMENTS 7.1-1 7.1 STRUCTURAL MODIFICATIONS 7.1-1 7.2 BALANCE OF PLANT PIPING AND EQUIPMENT 7.2-1 7.3 NSSS PIPING AND EQUIPMENT 7.3-1 l 7.4 SRV DISCHARGE QUENCHER 7.4-1 8.0 PLANT SAFETY MARGINS 8.0-1 1 /~N 8.1 CONSERVATISM IN PLANT DESIGN 8.1-1 iv

LSCS-MARK II DAR Rev. 8 6/80 LIST OF TABLES

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NUMBER TITLE PAGE 1.0-1 Mark II Containment Supporting Program:

LOCA-Related Tasks 1.0-4 1.0-2 Mark II Containment Supporting Program:

SRV-Related Tasks 1.0-6 1.0-3 Mark II Containment Supporting Program:

Miscellaneous Tasks 1.0-7 1.1-1 Primary Containment Principal Design Para-meters and Characteristics 1.1-2 3.1-1 Functional Capability Acceptance Criteria (Equation 9 of NB-3652 and NC-3652) 3.1-6 3.2-1 SRV Discharge Line Clearing Transient Para-

meterization 3.2-18 3.2-2 SRV Bubble Dynamics Parameterization 3.2-19 3.2-3 Second Actuation Assumptions 3.2-20 3.2-4 Second Actuation Results 3.2-21 3.4-1 Conformance of the LSCS Design to NUREG-0478 Criteria 3.4-2 4.1-1 Design Load Combinations 4.1-4 4.3-1 LOCA and SRV Design Load Combinations -

Reinforced Concrete Structures Other Than Containment 4.3-3

(')

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4.3-2 LOCA and SRV Design Load Combinations -

Structural Steel Elastic Design 4.3-4 4.4-1 Load Combinations for EOP' Piping 4.4-7 l 4.4-2 Load Combinations and Allowable Stress Limits for BOP Equipment 4.4-8 4.5-1 Load Combinations and Acceptance Criteria for NSSS Piping and Equipment 4.5-3 5.1-1 Margin Table for Basemat for All Valves Discharge 5.-13 5.1-2 Margin Table for Basemat for 2 Valves DisJaarge 5.1-14 5.1-3 Margin Table for Basemat for ADS Valves Discharge 5.1-15 5.1-4 Margin Table for Basemat for LOCA Plus Single SRV 5.1-16 5.1-5 Margin Table for Containment for All valves Discharge 5.1-17 5.1-6 Margin Table for Containment for 2 Valves Discharge 5.1-18 5.1-7 Margin Table for Containment for ADS '

Valves Discharge 5.1-19 5.1-8 Margin Table for Containment for LOCA Plus Single SRV 5.1-20 5.1-9 Margin Table for Reactor Support for All Valves Discharge 5.1-21 (3

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J LSCS-MARK II DAR Rev. 8 6/80

() LIST OF TABLES (Cont'd) s NUMBER TITLE PAGE i 5.1-10 Margin Table for Reactor Support

[ for 2 Valves Discharge 5.1-22

! 5.1-11 Margin Table for Reactor Support

for ADS Valves Discharge 5.1-23 5.1 12 Margin Table for Reactor Support for LOCA Plus Single SRV 5.1-24 5.1-13 Margin Table for Drywell Floor and SRV and LOCA Loads 5.1-25 5.2-1 Summary of Containment Wall Liner

! Plate Stresses / Strains for All SRV Cases 5.2-5

5.2-2 Summary of Containment Wall Liner i Anchorage Load / Displacement for All SRV Cases 5.2-6 ,

) 5.3-1 LOCA and SRV Design Load Combinations Reinforced Concrete Structures Other Than Containment 5.3-10 4

5.3-2 LOCA and SRV Design Load Combinations i

Structural Steel Elastic Design 5.3-11 5.3-3 Capability of Concrete (Other Than Containment) and Steel Structures 5.3-12 O 6.2-1 Pool Temperature Conditions -

Case la 6.2-8 1 . 6.2-2 Pool Temperature Conditions -

! Case Ib 6.2-9

.l 6.2-3 Pool Temperature Conditions -

Case 2a 6.2-10 1 6.2-4 Pool Temperature Conditions -

i Case 2b 6.2-11 ~

1 6.2-5 Pool Temperature Conditions - ,

i Case 3a 6.2-12 l 6.2-6 Pool Temperature Conditons -

Case 3b 6.2-13 l 6.2-7 Pool Temperature Analysis Results 6.2-14 l 7.2-1 Retested HVAC Equipment 7.2-3 8.2-1 Margin. Factors for Containment j During Maximum Transient Condition 3.2-2 i l i

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LSCS-MARK II DAR Rev. 8 6/80

() LIST OF FIGURES (Cont'd)

NUMBER TITLE 5.1-4 Radial Variation of Moment in Drywell Floor Due to Concentrated Radial Moment Applied at Radius 23'-3" 5.1-5 Circumferential Variation of Moment in Drywell Floor Due to Concentrated Circumferential Moment Applied at Radius 23'-3" l' 5.1-6 Radial Variation of Moment in Drywell Floor Due to Concentrated Circumferential Moment Applied-at Radius 23'-3"

5.1-7 Base Mat Plan - Top Reinforcing Layout 5.1-8 Base Mat Plan - Bottom Reinforcing Layout 5.1-9 Containment Wall Post-Tensioning Tendon Layout 5.1-10 Containment Wall Reinforcing Layout 5.1-11 Reactor Support - Concrete Plug 5.1-12 Reactor Support - Reinforcing Layout Before Modification

5.1-13 Unit 1 Drywell Floor Reinforcing Layout

5.1-14 Unit 2 Drywell Floor Reinforcing Layout 5.1-15 Design Sections - Primary Containment and Reactor Support A 5.1-16 Design Sections - Drywell Floor V 5.1-17 Representative Base Mat Interaction Diagram 5.1-18 Representative Containment Interaction Diagram 5.2-1 Base Mat Liner Detail 5.2-2 Base Mat Liner Stiffener Detail 5.2-3 Containment Wall Liner Detail

5.3-1 Downcomer Vent Bracing at Elevation 721'-0" l

5.3-2 Downcomer Vent Bracing at Elevation 697'-0" 1

5.3-3 Partial Downcomer Bracing Model Inner Rings 1 and 2 5.3-4 Partial Downcomer Bracing Model Outer Rings 3 and 4 6.2-1 Pool Temperature Response - Case la SORV at Full Power, 1 RHR Available 6.2-2 Pool Temperature Response - Case lb SORV at Full Power, 2 RHRs Available 6.2-3 Pool Temperature Response - Case 2a Isolation / Scram, 1 RHR Available 6.2-4 Pool Temperature Response - Case 2b Isolation / Scram, 2 RHRs Available 6.2-5 Pool Temperature. Response - Case 3a SBA, 1 RHR Available 6.2-6 Pool Temperature Response - Case 3b SBA, 2 RHRs Availablc

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%s Q' TABLE 3.4-1 (Cont'd)

FAPX II CWF.K GIDUP 1890 PWTTW S*AS's IA SAW WIT!N OF AMAM fRM4_

IaAp op m m w ,,,v, LOC FitcIFTNTTM Acceptable Sinusoidal pressure fluctuotion b) Medium Steam Fluz added to local hydrostatic. An-Emade p11tude uniform below went exit-linear attenuation to poc1 surfmee.

7.5 psi peak-to-peas amr11tude.

5,6 Hs frequencies.

Acceptable pendirg e) Chugging Loads Representative pressure flue. resolution of TSI tuntion taken tree AT test concerns.

added to local hydrostatie,

- unifors loading Maximum saplitude unifors below condition vent exit-linear attenuation to pool sur.* ace. +4.8 psi maximum overpressure, 4 .0 psi maximum under pressure, 20-30 Hz frequency. h g

- asyuznetrie loading Maximum say11tude uniform below g condition vent emit-linear attenutation to ,

M w pool surfmee. 20 pst mezimum

's- overpressure. -1k psi maximum g h underpressure,20-30 Hs fre- w

! quency, peripheral variation of amplitude follows observed statistical distritution with maximum and miniums dia-metrica11y cTposed.

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@ @ 9 TATLE 3.61 (Cont'd)  !

MARK I! ChTEM OPOL ? IA SALLF FMT"!04 ON 40tTPTAM PRTTTPIA fila *1 ST*f*1 C ?.TTCTv NFC FJVIF4 STAS C LOA?) OR 19tv>WN I7 .teconiary Loais Negligible load - none speelfied Acceptable A. Sonic Wave Load Acceptable B. Compressive have Lead Kefligible Lai - ocne specified

" M t unique load #ddressed to respeces te (%estions C. Post Svell Wave Load No generic load providei specification an:1 C20.8 and c20.44 (Appers'iz B)

Nic review.

Plant unique load Addressed la response ts. Question C. Leissie Slosh Loa 3 No generic losi provided specification and OM.44 (Appendix B)

EPC review.

Acceptable E. Fallteck loai on Sutserged Ne611gible load - none rpecified Boundary Acceptable F. Thrust Loads Momentu:n talance Acceptable G. Friction Trag Lcsis Standard .rictier .4 rag calculatices

• E w on Vents 5

• s- Negligible Load - none specifies Acceptable

.'o H. Tent Clearing Loads U E

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1 TABt2 3.k-1 (Cent'd) .

MPE, II CWhT.!C GILEP IA ?ALIF FW7T!ON ON AMTMANr'? cp!TyptA

!#D SFWIFTrATTfM Nac En'IrW 7AUM LOAr OR P"FPME'f0M Interi:n technical Acceptable, Rodabaugh criteria may be used.

FUNCTI0ft\L ptsition (7/19/76)

CAPABILITY Verify using ELAPL / Acceg. tail.

MASS-ENERGY MOLS RZ:L'.EE POR ANEut3 PE3S.

15% peak broadening Acceptable

, Q%7 IONS te be used.

!. EB-2, MEB-5 l closely spaced modes Acceptable. MSSS scope uses modifed summation MEB-3 EB-5 combined Per 1.92 per approved GE3*AR.

tynamie analysis Acceptable MEB-1 methods acceptable l

OBE Damping - Level i EB-2 A or B SSE Lamping - Level Acceptable {

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!s C or D >

1 ?

e Seismie slosh-p' art Acceptable h ME3 6 unique review M

Load Combinations: Acceptable. See load ecmbination table for Case #2 I

MEP-7a and b AP+ SSE and 7 OBE+SRV 5

Functional capability S*e load combination tat,le, i yza-8 and piping ace *ptance criteria 1

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TABIE 3.k.1 (Cont'd) val % El CMRS GMYJP ttAD CITTTWATTON NW FWTTW STAT'C 1A FAI.fE WinCN ON AffEP'"UcE CPTTTPIA LOAC MTNNPCM 4

N+S Wx To B Acceptable 1.

N*SWx+0BE to B Acceptable Approved CESSAR approach used

2. for iCS3.

N+SWall+SSE to C Acceptable 3

N+S Wads +0BE+IBA to C Acceptable k.

N+SWads+0BE+IBA to C Acceptable H+SWads+SSE+IBA to C Acceptable 6.

N+SSE+tBA to C Acceptable 7.

N to A Acceptable 8.

N+0BE to B Acceptable

9. E N+SWa*SSE+pBA to C Applied to mntairunent stnicture only (See M 000.22 0

10. and IFFR 5.2.h, and letter to R. J. Mattson front h y L. J. Sobon dated Feb. 22, 1979). g r

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LSCS-MARK II DAR Rev. 8 6/80 4.3 OTHER STRUCTURAL COMPONENTS 4.3.1 Concrete Structures The load combinations, including pool dynamic loads, considered in the reassessment of concrete structures (other than contain-ment concrete structures) such as shear walls, slabs, and beams are shown in Table 4.3-1.

For concrete structures, the peak effects resulting from seismic and pool dynamic loads were combined by the conservative ABS method even though the SRSS method is more appropriate, since the probabil-ity of all peak effects occurring at the same time is very small.

Acceptance Criteria The acceptance criteria used in the reassessment of reinforced concrete structures other than containment and internal concrete structures are the same criteria defined in Subsection 3.8.4.5 of the LSCS-FSAR and are identified in Table 4.3-1 for each load combination. The stresses and strains are limited to those specified in ACI 318-1971. As indicated in Table 4.3-1, ultimace strength design method has been used for all load combinations. No overstress is allowed for seismic loads. As stated in the FSAR, when a LOCA occurs outside the containment as in load combinations 4, 4a, 5, Sa, 7, and 7a, yield line theory is used to design reinforced concrete walls and slabs.

4.3.2 Steel Structures The load combinations including pool dynamic loads considered in the reassessment of steel structures such as framing, and con- '

tainment galleries, are listed in Table 4.3-2. The conduits, cable trays, ducts and their hangers have been designed for load combina-tions 1, 2, 4, 4a, 6, 7 and 7a of Table 4.3-2.

4.3-1

O O O TABLE 4.4-1 LOAD COMBINATIONS FOR BOP PIPING OBE SSE SBA( IBA DBA Acceptance Criteria Load Case P W ALL ADS Upset 1 X X X X Upset 2 X X X X Emergency II 3 X X X X X Emergency (

4 X X X X X Emergency (2) g 5 X X X X X X (1) X I1I Emergency I}

, 6 X X X X Emergency I}

7 X X l

N O

P - Pressure W - Weight SRV - Safety Relief Valve ADS - Automatic Depressurization System OBE - Operating Basis Earthquake SSE - Safe Shutdown Earthquake g SBA - Small Break Accident e i IBA - Intermediate Break Accident 4 i DBA - Design Basis Accident m

1. IBA or SBA, Whichever Governs m Faulted for Non-Essential' Subsystems

2. D i 3. Chugging

! 4. Condensation Oscillation or Annulus Pressurization i

2

LSCS-MARK II DAR Rev. 8 6/80 The percentages given in the table represent the portion of

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the structural elements which have the reserve strength to sustain the SRSS/ ABS loading.

5.3.2 Steel Structures The load combinations including pool dynamic loads considered in the reassessment of steel structures such as framing, and con-tainment galleries, are listed in Table 5.3-2. The conduits, cable trays, ducts and their hangers have been designed for load combina-tions 1, 2, 4, 4a, 6, 7 and 7a of Table 5.3-2.

For cable tray, conduit and HVAC duct hangers, the peak effects resulting from seismic and pool dynamic loads were combined j l by the SRSS method. Structural steel framings in the drywell  !

and in the reactor building and their corresponding embedments

, had the peak effects of the seismic and pool dynamic loads l combined by the absolute sum method, and the stiffening of

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structural elements for such a combination is in progress.

5.3.2.1 Acceptance Criteria For steel structures, stresses, and strains in accordance with'the 1969 AISC specifications are used for load combina-tions 1 through 3 defined in Table 5.3-2. No overstress is allowed for seismic loads. For load combinations involving abnormal or extreme environmental loads as in load combinations 4 through 7a of. Table 5.3-2, 'he steel stresses were conserva-tively limited to 0.95 f . No plastic deformations were alluwed.

y 5.3.3 Downcomers and Downcomer Vent Bracing The downcomer vents are subjected to static and dynamic loads due to note.al, upset, emergency, and faulted plant conditions.

The downcomer vents are braced at elevation 721'-0", well O'- '. above the pool swell impact zone to reduce the forces and moments.being transmitted through the downcomers to the 5.3-2

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-LSCS-MARK II-DAR Rev. 8 6/80 6.0 SUPPRESSION POOL WATER TEMPERATURE MONITORING SYSTEM 6.1; SYSTEM DESIGN 6.1.1. Safety Design Basis 4

i The safety design basis for setting the temperature limits for the suppression pool temperature monitoring system are

. based on providing the operator with a,dequate time to take i the necessary action required to assure that the suppression pool temperature will always remain below the 200* F temperature limit established by the NRC in NUREG-0487. The system design also provides the operator with necessary information regarding

! localized heatup of the pool water while the reactor vessel

is being depressurized. If relief valves are selected for j actuation, they may be chosen to assure mixing and uniformity of heat energy Injection to the pool.

O 6.1.2 General Sys tem Description The suppression pool temperature monitoring system monitors

! the pool temperature in order to prevent the local pool water temperature from exceeding 200* F during SRV discharge and

provides the operator with the information necessary to pre-vent excessive pool temperatures during a transient or accident.

j Temperatures in the pool are recorded and alarmed in the main j

control room. The instrumentation arrangement in the suppres-sion pool consists of two bulk and 14 local temperature sensors i mounted on the pool and pedestal wall.

i

n. The two bulk temperature sensors are dual-element chromel 1

constantan thermocouples located at elevation 683 feet 0 inch, and at azimuths 17' and 197*, respectivcly. These sensors provide signals which are used to indicate to the operator the_, bulk temperature of.the suppression pool.

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'?CS-MARK II DAR Rev. 8 6/80 The local temperature sensors consist of 14 dual-element,

(]) 100 0, platinum RTD's located 1 foot below the low water level, at elevation 698 feet 10 inches. Ten of the sensors are located on the outer suppression pool wall at azimuths 0*, 30', 67', 113*, 150', 180', 210', 247*, 293*, and 330*.

The other-four are located on the pedestal at azimuths 0*,

I 90*, 180 , and 270'.

The sensors are powered from ESS-1 and ESS-2 divisions and local discharge areas are monitored by two sensors, one from each division. This represents a conservative measure-ment of local pool water heatup. All instrumentation will 1 be qualified Seismic Category I. The time constant of the l thermocouple installation will be no greater than 15 seconds.

The time from output of sensor to initiation of function will be no greater than 0.5 second. The difference between measure-ment reading _and actual temperature will be within + 2 F.

The samp1ing technique for monitoring the pool temperature

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is to continuously record the measurements made by each of the 14 RTDs. The discharge locations and spacing are such l

that the number of sensors and their arrangement provides

, conservative monitoring of localized suppression pool water heatup in addition to bulk pool temperature.

The quenching of the steam at the quencher discharge forms jets that heat the water and generate convection currents in the suppression pool. These currents eventually rise and displace cooler water near the pool surface.

During an extended blowdown, a large temperature gradient is expected initially near the quencher. After a short time the pool gradients will stabilize with a bulk to local temper-ature difference of about 10' F. (Bulk and local temperatures are defined in NUREG-0487.) The adequacy of the temperature

{) monitoring system will be confirmed by the in-plant SRV testing.

6.1-2

, LSCS-MARK II DAR Rev. 8 6/80 6.1.3 Normal Plant Operation

)

The temperature monitoring system is utilized during normal plant operation to ensure that the pool temperature will remain low enough to condense all quantities of steam that may be released in any anticipated transient or postulated accident.

When rams head devices were specified for design, there was an NRC concern that high pool temperature might result in high pool dynamic loads during SRV discharge because of unstable steam condensation. Installation of T-quenchers has eliminated this concern. The local pool temperature (temperature measured on the containment wall at the elevation of the T-quencher) limit for SRV discharge is specified to be 200* F in accordance with the NRC Lead Plant Acceptance Criteria (NUREG-04e23 During normal plant operation, the system is in continuous operation recording the suppression pool water temperature in the main control roem. An alarm is actuated in the control room to allow the operator tc take appropriate action and prevent the exceedance

\

of the pool temperature limit.

6.1.4 Abnormal Plant Operation BWR plants take advantage of the large thermal capacity of the suppression pool during plant transients which require relief valve. actuation. The discharge of each relief valve is piped to the suppression pool, where the steam is condensed.

This results in an increase in pool water temperature but with a negligible increase in containment pressure. However,

. certain events have the potential for substantial. energy addition to the suppression pool and could result in a high local pool temperature if timely corrective action is not taken. When rams head discharge devices are used, test tasults and operating experience indicate that high magnitude oseillatory loads may occur when a high steam mass flux is injected into a pool with local temperature above 170

• F. Although analysis

. demonstrates that the pool temperature will remain below 150* F when the steam mass flux is high enough to cause these loads, 6.1-3

LSCS-MARK II DAR Rev. 8 6/80 T-quenchers have been installed instead of the rams heads

.(m\

\_/ to provide additional margin to the pool temperature limits.

Most'of the transients that result in energy d'scharge to the suppression pool are of short duration and have little effect on the suppression pool temperature. However, three events have the potential for substantially high energy release to the pool that could result in undesirably high pool temperatures if timely corrective action is not taken.

These events are: (1) stuck-open relief valve (SORV) at power cases; (2) isolation / scram cases; and (3) small break accidenc (SBA) cases. A brief description of each of these events follows:

1. Stuck-Open Relief Valve (SORV) Cases The steam flow rate through a safety / relief valve (SRV),

is proportional to reactor pressure. One method of

() terminating energy input to the pool is to scram the reactor and depressurize the RPV in the event the relief valve cannot be closed. During the energy dump, the pool temperature will increase at a rate determined by the RPV pressure, the flow capacity of the SRV, the primary system heat-removal system capability, and the suppression pool water heat-removal capability.

l

2. Isolation / Scram Cases l h

Isolation / scram cases which include NRC events (c) and (e), '

are analyzed to demonstrate that the loss of the main condenser by the sudden closure of the MSIV's and subsequent SCRAM, SRV openings at set pressure, and manual depres-surization will not result in high pool temperature.

Two single failures are considered separately; one is '

the loss of a RHR HX and the other is the failure of

() a SRV to reclose (SORV) . While NRC. event (c) considered 6.1-4

LSCS-MARK II DAR Rev. 8 6/80

, a SORV at hot standby as the initiating event with a

(_')N e single failure of one RHR HX,-it is clear that the decay heat load once the reactor is already in hot standby would be small. As a result, peak pool temperature, even with one RHR HX available, in bounded by the isolation /

scram cases considered. For both cases analyzed, manual depressurization is not assumed to begin until the pool temperature reaches 120* F.

I Detailed assumptions for isolation / scram are discussed in Subsection 15.2.2.3.2.

3. Small Break Accident l SB' A) Cases SBA cases are analyzed to demonstrate that SRV discharge required to depressurize the reactor coolant system following a small break will not result in high pool temperatures. SRV discharge is terminated prior to

() reaching peak pool temperature.

There are six plant depressurization transients which were considered as limiting events for energy released to the suppression pool. These events are numbered for ease of reference and are described as follows:

Event la - SORV_at power and loss of one RHR heat exchanger.

Event lb - SORV at power and spurious closure of the main steam isolation valves.

Event 2a - Isolation / scram and loss of one RHR heat exchanger.

Event 2b - Isolation / scram and SORV.

Event'3a - Small break accident and loss of one RHR heat exchanger Event 3b - Small break accident and loss of shutdown cooling.

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6.1-5

LSCS-MARK II DAR Rev. 8 6/80 Q .6. 2 SUPPRESSION POOL TEMPERATURE RESPONSE 6.2.1 Introduction The La Salle County Station (LSCS) (Units 1 and 2) take '

advantage of the large thermal capacitance of the suppression pool'during plant transients requiring safety / relief valve

(SRV) actuation. The discharged steam is piped from the reactor pressure vessel (RPV) to the suppression pool where it condenses, resulting in a temperature increase of the pool water, but a negligible increase in the containment pressure. Most transients that result in relief valve actu-

~

ations are of very short duration and have a small effect on the suppression pool temperature. However, certain postulated events.with conservative assumptions present the potential for substantial energy additions to the suppression pool that

~

could result in high pool temperature.

( 1 T-quenchers have been installed to provide additional margin j and to conform with NUREG-0487, for the quencher the NRC has established a local pool temperature limit of 200* F. The i

results of the quencher calculations are presented in the fol-I

, lowing subsections to demonstrt-te the adequacy of the LSCS design.

Both LOCA and non-LOCA events were investigated. The events l

- consist of

1. Stuck-Open Relief Valve

a. From power operation.with loss of one RHR heat exchanger.

b. From power. operation with spurious closure of MSIV's.

6.2-1 ,

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

LSCS-MARK II DAR Rsv. 8 6/80 i

,_ 2. SRV Discharge Events During RPV Isolation k_/ a. Isolation and reactor depressurization with loss of one RHR heat exchanger.

b. Isolation and a stuck-open relief valve.

2

3. Small break accident (0.01 ft liquid break)

a. Loss of one RHR heat exchanger.

b. Loss of shutdown cooling.

6.2.2 Temperature Response Analysis This analysis was performed for the quencher SRV discharge device. Pool temperatures were calculated until a peak pool temperature was reached.

6.2.2.1 Model Description O Non-LOCA Events To solve the transient response of the reactor vessel and suppression pool temperature due to the postulated events, a coupled reactor vessel and suppression pool thermodynamic model was used. The model is based on the principles of conservation of mass and energy and accounts for any possible flow to and from the reactor vessel and the suppression pool.

The model incorporates a control volume approach for the reactor pressure vessel and suppression pool. It is capable of tracking a collapsed reactor vessel water level and having a rate of change of temperature or pressure imposed on it.

The various modes of operation of the residual heat removal (RHR) system can be simulated, as well as the relief valves, HPCS, RCIC, and feedwater fusctions. The model also simulates system setpoints (automatic and manual) and operator actions and accepts as input the specific plant geometry and equipment capability.

6.2-2

LSCS-MARK II DAR Rev. 8 6/80 Small Break Accident Model O

In the small break accident analysis, the' mass and~ energy conservation laws are applied to a control volume which includes all of the reactor vessel contents and its walls.

This control volume is subjected to the boundary conditions of decay heat input. The break and the safety / relief valve i

flow rates and the associated fluid enthalpies are derived from the state of fluid in the control volume undergoing the transient and the specified flow areas and locations.

The time-dependent break and safety / relief valve mass and energy flows are then input to another control volume con-taining the suppression pool. The pool temperature transient is obtained using the energy and mass balance equations on the suppression pool.

6.2.2.2 General Assumptions and Initial Conditions O

The following common assumptions were used throughout the analysis of the LSCS suppression pool temperature response:

a. Decay heat per ANS 5-20/10.

b. Full crudded RHR heat exchangers.

c. RCIC and HPCS water source is the condensate storage tanks.

d. Condensate storage tank temperature was 80' F.

e. Wetwell air temperature equal to the suppression pool water temperature.

f. Feedwater in excess of instantaneous pool temp-erature is assumed to maintain level rather than condensate storage tank inventory via RCIC and-HPCS. This assumption maximizes heat addition

[ }' to the pool.

6.2-3

- ~ _ - - , _ - - - . -- -- -. . _ . . . - . . - .- .___

i LSCS-MARY. II DAR Rey, 8 6/80

g. In calculating the overall heat transfer coefficient

[}

of the vessel wall and internal structures, it is assumed that the heat transfer is dominated by conduction. The heat transfer area of the reactor internals is obtained by assuming that they have the same metal thickness as that of the vessel, which is assumed to be 0.333 foot uniformly.

h. The control volume of the reactor includes the l reactor vessel, the recirculation lines, the feedwater lines from the vessel to the nearest feedwater heaters, and the steamlines from the vessel to the inboard main isolation valves (MSIV).

i. The initial water level in the reactor vessel is calculated based on the assumption that the voids in the two-phase region collapse. Therefore, the ECCS ON/OFF volumes are based on the total

() liquid volume of the reactor vessel, the feedwater lines and the recirculation lines combined.

j. The specific heat of the reactor vessel and the internal is assumed to be 0.123 Btu /lbm/*F.

The metal density is assumed to be 490 lbm/ft .

k. A stuck-open relief valve can be detected and the corresponding quencher within the suppression l chamber identified.

1. Additional safety / relief valves are manually opened as necessary to depressurize the reactor.

m. Minimum technical specification suppression pool water level.

n. Maximum suppression pool initial temperature which was 100* F during power operation and

- 120

• F at hot standby.

O.

(_ ! o. 122.5% rated ASME safety / relief valve flow rate.

6.2-4 l

- - . ., . - . . + , . . - - -- . . ,_

LSCS-MARK II DAR Rev. 8 6/80

() 6.2.2.3 Description of Non-LOCA Events Th!s subsection describes the safety / relief valve discharges for non-LOCA events, (Subsection 6.2.2.4 describes the LOCA event). A complete description of the sequence of events for all of the cases, i.e., Events la, lb, 2a, 2b, 3a, 3b, is given in Tables 6.2-1 through 6.2-6.

6.2.2.3.1 SORV at Power

1. The SORV is the initiating event and two single failures are considered separately:

a. Loss of one RHR HX.

b. MSIV isolation signal at t = 0.

2. In accordance with the Technical Specifications, manual scram occurs at 100* F. Manual scram is accomplished

[]}

in a single manipulation by transferring the mode switch from "run" to " shutdown."

3. Pool cooling initiated at t = 10 min.

4. For 1 (a) above, main condenser remains available.

5. for 1 (a) above, the operable RHR HX is placed in shut-down cooling mode. For 1 (b) above, two RHR HX are available and no shutdown cooling is used in the analysis.

6.2.2.3.2 Isolation / Scram

1. Isolation / scram is the initiating event and two single failures are considered separately:

a. Loss of one RHR HX.

O

(_) b. Spurious failure of a safety / relief valve in the open position (SORV) .

6.2-5

- - - - . - g- ,

LSCS-MARK II DAR Rev. 8 6/80 i;

2. Pool cooling initiated at t = 10 min.

j}

3. .Offsite power assumed to be unavailable.

4. A reactor depressurization is initiated at 120' F.

5. For 1 (b) above, the SORV is assumed to occur at t = 0.

6. For 1 (a) above, the operable RHR HX is placed in shutdown cooling mode. For 1 (b) above, two RHR HX are available and no shutdown cooling is used.

6.2.2.4 Small Break Accident

1. Two single failures considered separately:

a. Loss of one RHR HX.

b. Loss of shutdown cooling mode.

)

2. Offsite power assumed to be unavailable. l

3. SCRAM on high drywell pressure and MSIV closure signal assumed at t = 0.

<4 . At t = 10 min, pool cooling is initiated.

5. A reactor depressurization is initiated at 1203 F.

6. For 1 (a) above, the operable RHR HX is placed in shutdown cooling mode.

l 6.2.3 .Results/ Conclusions l The results obtained from the LSCS suppression pool temperature analyses are depicted in Figures 6.2-1 through 6.2-6. Summary i

{) ~results are presented in Table 6.2-7. Conservative assumptions

, 6.2-6

LSCS-MARK II DAR Rev. 8 6/80 rv were used for all transient events presented in this report.

(l For1 example, 122.5% of ASME rated steam flow h.r SRV discharge, maximu:i. initial pool temperature, minimum initial pool mass, continued addition of feedwater energy into the reactor vessel, and the initial reactor power corresponding to 105%

of rated steam flow (105% rated steam flow is equivalent to 103% of rated thermal power) are conservative parameters that will aft'ect the pool temperature.

For the case of a stuck-open relief valve from power (Figures

6. 2-1 and 6. 2-2) , calculated suppression pool temperatures are below the bulk pool temperature limit of 190' F for the quencher.

The cases of isolation / scram are given in Figures 6.2-3 and 6.2-4. The peak bulk pool temperature for thern cases are 184* F (loss of one RHR) and 175* F (SORV).

O ss The cases of small break accident without ADS (Figures 6.2-5 and 6.2-6) yield the lowest suppresson pool temperature for the events examined. To maximize pool temperature these SBA analyses were performed without actuation of ADS.

In conclusion, the calculations indicate that the bulk pool I temperature can be maintained below 190* F for all events j I

considered. With the in-plant testing to confirm the 10 F bulk l to local temperature difference, the results described herein j demonstrate conformance with the quencher pool temperature limits established by NUREG-0487.

6.2-7 j w e - - - y w y -

LSCS-MARK II DAR Rev. 8 6/80 TABLE 6.2-1 (O

_) POOL TEMPERATURE CONDITIONS - CASE la SORV at full power, 1 RHR available Manual Scram at Tpool = 110' F.

Mechanistic closure of-the turbine stop and bypass valves (Product Line Unique - BWR-4 or 5). .

One RHR in pool cooling 10 minutes after high temperature alarm.

Main condenser reestablished through bypass system 20 minutes after scram using plant specified bypass capacity.

. Main condenser available using full bypass capacity until reactor vessel permissive for RHR shutdown cooling.

RHR out-of-pool cooling when pressure permissive for RHR

(~>) shutdown cooling is reached. Sixteen minute delay for RHR transfer to shutdown cooling. (Additional SRVs opened as required during switchover to assure no repressurization during switchover.)

D)

\-

1 6.2-8

LSCS-MARK II DAR Rev. 8 6/80 TABLE 6.2-2

() POOL TEMPERATURE CONDITIONS - CASE lb SORV at full power - 2 RHRs available Manual Scram at Tpool = 110' F.

Non-mechanistic isolation at scram with 3.5-seconds main isolation valve closure.

Two RHRs in pool cooling 10 minutes after high pool temper-ature alarm.

When Tpool = 120' F, begin manual depressurization by opening additional SRVs so that SORV plus cycled SRVs  ;

result in depressurization rate of approximately 100 F/hr unless SORV alone causes faster depressurization.

RHR shutdown cooling not initiated.

1 s

l i

I

(}

l 6.2-9 i

LSCS-MARK II DAR Rev. 8 6/80 TABLE 6.2-3

' POOL TEMPERATURE CONDITIONS - CASE 2a Isolation-Scram (non-mechanistic) 1 RHR available Isolation Scram at t =-o non-mechanistic with 3.5 seconds

-main isolation valve closure.

One RHR in pool cooling 10 minutes after the event.

When Tpool ='120' F, begin manual depressurization by opening j

additional ^ valves as needed. Depressurize at 100*T/b .

RHR out-of-pool cooling when pressure permissive for RHR shutdown cooling is reached. Sixteen minute delay for RHR transfer to shutdown cooling. (Additional SRVs opened as required during switchover to assure no-repressurization during switchover.)

O i

l l

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A i

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l 6.2-10 1

9 1

LSCS-MARK II DAR Rev. 8 6/80 ,

TABLE 6.2-4 O

(_/ POOL TEMPERATURE CONDITIONS - CASE 2b Isolation Scram (non-mechanistic) 2 RHRs available Isolation Scram at t = o non-mechanistic with 3.5 seconds main isolation valve closure.

SORV at t = o.

Two RHRs in Pool Cooling at 10 minutes after the event.

When Tpool = 120* F begin manual depressurization by opening additional valves. Depressurize at 100'Fc RHR shutdown cooling not initiated.

O l

l I

L' 6.2-11 l

LSCS-MARK II DAR Rev. 8 6/80 TABLE 6.2-5

. k- - POOL TEMPERATURE CONDITIONS - CASE 3a GBA Event Mode 1 RHR Available Scram at t = o on high drywell pressure.

Isolation at t = o (non-mechanistic) with 3.5 seconds main isolation valve closure.

One RHR in pool cooling 10 minutes after high pool tempera-ture alarm.

When Tpool =~120* F, begin manual depressurization by opening additional SRVs as needed. Depressurize at 100 F.':.r.

RHR out-of-pool cooling when pressure permissive for RHR shutdown cooling is reached. Sixteen minute delay for RHR transfer to shutdown cooling. (SRVs opened as required during switchover to assure no repressurization during

() switchover.)

Automatic RHR switchover to LPCI mode when RPV pressure is less'than RHR pump flow head. If switchover occurs after ten minutes, assume 10 additional minutes to convert manually back to pool cooling. If switchover occurs during the first 10 minutes while operator is attempting to initiate pool cooling, assume no additional time.

I l

1 l

l l

l O) b 6.2-12 ,

LSCS-MARK II DAR Rev. 8 6/80 TABLE 6.2-6 POOL TEMPERATURE CO DITIONS - CASE 3b SBA Event 2 RHRs Available Scram at t = o on high'drywell pressure. 5 Isolation at t r- o (non-mechanistic) with 3.5 seconds main isolation valve closure.

Two RHRs in pool cooling 10 minutes after high pool tempera-ture alarm.

Automatic RHR switchover to LPCI mode when RPV pressure is less than RHR pump flow head. If switchover occurs after 10 minutes, assume 10 additional minutes to convert manually back to pool cooling. If switchover occurs during first 10 minutes while operator is attempting to initiate pool cooling, assume no additional time.

When Tpool = 120* F, begin manual depressurization by opening

(]) SRVs as needed. Depressurize at 100*F/hr. l

\

RHR shutdown cooling not initiated.

l 6.2-13

LSCS-MARK II DAR Rev. 6/80 q TABLE 6.2-7 D POOL TEMPERATURE ANALYSIS RESULTS LA SALLE BULK POOL PEAK CASE NUMBER TEMPERATURE

1. SORV at Power - Loss of 1 RHR HX la 159* F

2. SORV at Power - Spurious Isolation- lb 177* F

3. Isolation / Scram-Loss of 1 RHR HX 2a 184* F

4. Isolation / Scram - SORV 2b 175' F

5. SBA-Loss of 1 RHR HX 3a 186* F

6. SBA - Shutdown Cooling Not Available 3b 177* F O

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POOL TEt1PERATURE RESP 0flSE - CASE lA l SORV AT FULL POWER, 1 RHR AVAILABLE

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('; FIGURE 6.2-3 U

POOL TEf1PERATURE RESPONSE - CASE 2A ISOLATION / SCRAM, 1 RHR AVAILABLE

REV. 8 6/80 POOL TEMPERATURE 'F

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('; FIGURE 6.2-5 V

P00L TEMPERATURE RESPONSE - CASE 3A SBA, 1 RHR AVAILABLE

POOL TEMPERATURE *F 5  : O C  ;

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%v FIGURE 6.2-6 P00L TEMPERATURE RESPONSE - CASE 3B i SBA, 2 RHR'S AVAILABLE l l

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