LD-96-024, Transmits Finalized Package of Changes to Sys 80+ Std Plant Design Control Document

Transmits Finalized Package of Changes to Sys 80+ Std Plant Design Control Document

ML20113C181
Person / Time
Site:	05200002
Issue date:	06/27/1996
From:	Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
FACA, LD-96-024, LD-96-24, NUDOCS 9607010016
Download: ML20113C181 (103)

Text

ABD Mpp June 27,1996 LD-96-024 Docket 52-002 Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Subject:

System 80+ Standard Plant Design Changes

Reference:

Letter, C. B. Brinkman to NRC," Changes to the System 80+ Standard Plant Design," LD-96-021, June i 1,1996

Dear Sirs:

The Reference letter forwarded draft changes to the System 80+ Standard Plant design for Staff review and approval.

This letter transmits a finalized package of changes to the System 80+ Standard Plant Design Control Document. These changes are listed on the attached table with details shown on the revised DCD pages. It is our understanding that Staff review and approval of these changes will not impact the design certification schedule, nor are any of these changes required to meet Commission regulations or to maintain the safety conclusions of the FSER.

All changes will be formally incorporated into the System 80+ DCD when it is reprinted at the conclusion of design certification rulemaking.

Please call me, Stan Ritterbusch (860-285-5206) or Virgil Paggen (860-285-4700) if you have any questions.

Very truly yours, COMBUSTION ENGINEERING,INC.

nb *

. n C. B. Brinkman Director, Nuclear Licensing cc: J. N. Wilson (NRC)

F. A. Ross (DOE; w/o enclosure)

\

l Attachments: As stated %j ) b *) Y l ABB Combustion Engineering Nuclear Systems i \

Contustion Enginnenng Inc. 2000 Day Hit Hoad Telephone (Bf,0) EM1911

^ " ' * ^ * * ' ' '

Po. Box 500 Fax (860) 285-5203 9607010016 960627 w,nosor. ci ortras-osco PDR ADOCK 05200002 A PDR a

Synt;m 80+ DCD Ching;s item Reference Design Change Description Tbl i .8-7; incorporate ASME B&PV Code Case N-498," Alternate Rules for 10 Year Hydrostatic 1 Tbl 3.9-1; Pressure Testing for Class 1 and 2 SystemsSection XI, Division 1," to permit a system 5.2.4.6 leakage test to be conducted at or near the end of each inspection period. This code case was approved in Reg Guide 1.147, Rev 09 2 Sect 6.5.3.4; Fig Revise SCS piping & valve sizes; reduce allowable containment spray runout flow to 6.3.2-1 A; Fi9 provide reqd NPSH when used for the containment spray function.

6.3.2-18.

3 Tbl 3.2-1, 2 Correct the safety and seismic classification of SDS components.

4 Tbl 6.2.1-22 Revise Tech Spec lower limit on IRWST temperature. Update Fig 3.5.4-1 (in Ch 16) to Fig 16/3.4.5-1; indicate IRWST minimum temperature of 60F. Correct illustrative data point given on App 16A page B3.5-25 in Appendix 16A .

5 CDM 2.2.2 Identify twelve (12) additional locations in core center region that could contain a 4-finger CDM Fig 2.2.1.3 CEA and four (4) existing CEA locations that could contain either a 4-finger or a 12-finger ADM-4.2.2.4 CEA Revise CEA locations shown in CDM Section 2.2.2 and Fig 2.2.1-3; Revise ADM Figs in 4.2,4.3 Section 4.2.2.4 and Figs 4.2-11, Fig 4.3-46 and Fig 4.3-47.

6 5.1; Fig 5.1.2-1; Add RCS mid-loop level monitoring via external hot leg tanks and instrumentation.

7.7.1.1.15; Revise RCS P&lD (Fig 5.1.2-1) to show hot leg connections. Revise level measurement App 19.8A discussion in Section 19.8A/2.8.3.2.1, Tbl 19.8A/2.8-1; Fig 19.8A/2.8-4.

7 Fig 5.1.2-3; Revise SDS line/ valve size (shown in P&lD Fig 5.1.2-3) from 6" to 4" per new Tbl 5.4.10-1; qualification data. Safety analysis flow rate is unchanged.

Tbl 6.7-1 8 3.6.3.7; 3.6.3.8; Add altemative method for leak-before-break evaluation.

App 3.9A; Fig 3.9A-12

]

9 CDM 2.7.16; Delete charging pump interlock. Implement ITAAC verification of maximum charging ADM 9.3.4.2.1; pump flowrate to the RCS.

15.4.6.3; Tbl 15.4.6-1 ,

3.7.1.3; Fig 3.7-10 32, App 3.9A Incorporate damping per ASME Code Case N-411-1 for uniform envelope response spectrum piping analyses.

)

l 11 Tbl 3.6-3 Delete PSV high-energy lines, items 40.. 43 since PSVs are mounted directly on pzr; correct items 58,59. No impact on Fig 5.1.2-3 or to ITAAC.

12 Tbl 3.9-2 Revise title of Table to include component supports in loading combinations.

13 Tbl 4.2-3 Add ANO inspection program scope for year 1989/ cycle 7.

14 Tbl 5.4.7-2 Correct failure mode entry to " fail closed" for SCS pump discharge isolation valve.

15 Pg 9.3-30 Correct statement regarding fluid retum to RCS when in shutdown cooling mode.

16 Pg 9.3-37 Revise CVCS system " redundancy" statements to be consistent.

17 Tbl 9.3.4-4 Change VCT norrnal operating pressure range to 20-50 psig.

18 10.3.2.3.2.1; Revise MSIV bypass valve closing time to be 5 seconds or less (rather than 10 seconds) 10.3.4 to be consistent with safety analysis.

19 Ch 19.7.5 Delete probability values from Tables 19.7.5.1-1 through 19.7.5.4-7.

20 Pg 19.11-145 Clarify statement regarding core uncovery following intersystem LOCA.

File cMedWed-dchg doc Date: June 27,1996 Page 1

System 80+ DCD - Potential Design Changes Item Number: 1 Summary

Description:

Delete Inservice RCS Hydrostatic Testing Affected DCD Section(s): CDM: None ADM: Table 3.9-1; Section 5.2.4.6 l

l Description of Chance:  !

l A recent ASME B&PV Code Case has been published which provides an alternative to performing the 10-year hydrostatic pressure test. Code Case N-498," Alternate Rules for 10 Year Hydrostatic Pressure Testing for Class 1 and 2 SystemsSection XI, Division 1," indicates that a system leakage test can be conducted at or near the end of each inspection period, prior to l reactor startup instead of the Ilydrostatic Pressure Test. This code case was approved by the i

NRC as indicated in Reg Guide 1.147, Rev 09. Currently the DCD (Table 3.9-1) specifies that '

15 RCS and secondary hydrostatic tests are included in the stress analysis of Code Class I and CS components. This number will be reduced from 15 to 10 occurrences during the plant life time.

The RCS and secondary leak tests have adequate margin in the number of occurrences listed in Table 3.9-1 to account for the additional leak tests which will be performea instead of the hydrostatic pressure tests. Table 1.8-7 "ASME Section III Code Cases applicable to System 80+

will be revised to included Code Case N-498. In addition, Section 5.2.4.6 of the DCD will be revised to include reference to the Code Case.

DCD Markups Attached? Yes l

l l

l dedchng4. doc 6/27/96

i tra a

^

Svstem 80+ Deslan Control Decament

Table 1.8-7 ASME Section III Code Cases Applicable to System 80+

e Case Title N-4-11 1133711] Specis! Type 403 Modified Forgings or Bars, Class 1 and CS; 7/13/87.

N-60-4 Material for Cois Support Structures 7/27/88.

N-71-15 Additional Materials for Subsection NF, Classen 1,2,3 and MC Cnama-* Suppons i l Fabricated by Welding; 12/16/89.

N-122 1 Evaluation of the Design of Rectangular Cross-Section Attachments on Class-1 Piping; 7/27/92.

N 192-2 Use of Braided Flexible Connectors, Class 2 and 3; 9/17/87.

! N 247 Certified Design Report Summarv for Component Standard Support, Class 1,2,3, and MC; i 1/21/88.

a _

l N-249-10 Additional Materials for Subsection NF, Classes 1,2,3 and MC Contponent Supports l Fabricated without Welding: 5/06/89.

! N-262 Resistance Spot Weldtag for Structural Use in Component Supports; 7/28/88.

1 l N-284 Metal Cnataia aent Shell Buckling Design Methods;Section III, Division 1, Class MC; l 8/25/80.

! N-309-1 Identification of Material for Component Supports; 7/28/88.

} N-313 Alternate Rules for half-Coupling Branch canaa+ ions, Class 2; 11/28/86.

I N 318-4 Evaluation of the Design of Rectangular Cross Section Attaebmaats on Class 2 or 3 Piping; 1 12/11/89.

p

- N-319-1 Evaluation of Stresses in Butt Welded Elbows for Class 1 Piping; 7/24/89.

! N 391-1 Evaluation of the Design of Hollow Circular Cross Section Welded Attachmaata on Class I

! Piping; 7/24/89.

i N-392-1 Evaluatson of the Design of Hollow Circular Cross Section Welded Attachamus on Class 2 j and 3 Pipings; 12/11/89.

N 393 Repair Welding Structural Steel Rolled Shapes and Plates for Components Supports; 7/30/89.

N-411-1 Alternative Damping Values for Response Spectra Analysis for Class 1,2, and 3 Piptag; j 2 G /89.

l N-420 Linser Energy Absorbmg Supports for Subsection NF, Class 1,2, and 3 Construction; 2/14/88

N-430 Alternative Requirmaants for Welding Workmanship and Visual Acceptance Criteria for

, . Class 1,2,3, and MC Linear-type and Standard Supports: 2/28/89.

N-433 Non-throeded Fasteners for Class 1,2, and 3 Components Paping Supports: 12/16/89.

N-474-1 Design Stress Intensities and Yield Strength Values for UNS NO6690 with a mmimum

} spectric Yield Strength of 35 kai, Class 1 Components; 3/05/90. '

N-476 Class 1,2,3, and MC Lineer en apaa-' Supports - Design Criteria for Single Angle i Members, Subsection NF; 5/06/89.

4

{ N-Ljqq Allem HJe L I.t s (Or to year Hydros %Hc Passoa j Te stm q Or class t as d 7.- Sq 6 tee s Secxws c ,

O l D[v is 'io ra 1 W % ==w- -  %. 7.seo

J i rre m i 1

j j System 80+ Deslan ControlDocument Table 3.9-1 Transients Used in Stress Analysis of Code Class 1 and CS Components O (Cont'd.)

, Test Conditionsm OccurrencesA i 1. RCS hydrostatic test 1

~

[-[ Primary pressure cycles from atmospheric to 3125 psia ath, / D,

.w. between 120 and 400*FR_ .f

2. RCS leak test r

M 200.

pf -

[Prunary pressure cycles from a-h-ic to 2250 psia at b.

~f==L=..between 120 and 400*8 _

3. Secondary hydrostatic test m g ._ [

[ Secondary pressure cycles from atmospheric to 1500 psia at a -

/O, Imum tamparature of 190*F) _ __

4. Secondary leak test n p .- (

[ 200.

Wa y pressure cycles from atmospheric to 1200 psia at a 3 maw t=- -

Eaure of 200*F1

5. SIS /SCS check valve operabihty test 500.

6. SIS /SCS preoperational and maintenance test 240.

Upset C-r ""- _-"I OccurnacesA

1. Decrease in feedwater temperature 20.

2. Inenese in feedwater flow rate 20.

3. Increase in steam flow rate 20.

4. Inadvertent opesung of a steem generator relief or safety valve 10.

5. Loss of load (turbine speed control system operates normally) 19.

[ Loss of electrical load and normal turbine / generator runbeck to house load]

6. Turbine trip 20.

7. Loss of condannar vacuum 20.

~

B. Ims of none AC power to the station annhanes 10.

9. Loss of normal feedwater flow 20.

2 [%%=* actuation and cycling of cold emergency feedwater to the steam generators]

10. Loss of forced reactor coolant flow 20.

11. Uncontrolled CEA withdrawal froun suberitical or low power condition 10.

Amarovesf Desen AderenW. Demyn et ASC Asps J.9-70

rre a 1

)

System 80+

Deslan controlDocument m 1

i.E 5.2.4.4 Inspection Intervals e

i t gg The IWA-2000 examination program for the 120 month inspection interval will be defined in the ISI plan.

g .7 ~

The ISI Plan for all Code Class I systems and components will be in accordance with the ASME Code g) 'Section XI edition in effect per 10 CFR 50.55a 12 months prior to the issuance of an operating license (initial interval).

W ?.

j 'Odf 5.2.4.5 Evaluation of Examination Results V t i Evaluation of ex=inadon results for Class I components will be conducted in accordance with Articles i iy d IWA-3000 and IWB-3000 of ASME Section XI.

1

$s

. v) S! Unacceptable indications will be repaired in accordance with the requiremems of Articles IWA 4000 and j (3 IWB-4000 of ASME Section XI. Criteria for establishing need for repair or replacement shall be per IWB-3000.

% .s .e -.

5.2.4.6 System Imakage and Hydrostatic Tests f#! og {[ v i Oj^ ' The hydrostatic and system leak tests for the reactor pressure vessel and reactor coolant pressure

} f boundary will be conducted in accordance with the requirements of Articles IWA-5000 and IWB-5000

{pO g, of ASME Section Examinations performed during these tests will be corAucted without the removal of msulation. ical Specifications requirements on operating limits during heatup, cooldown, and 1 E g- 3 3g system hydrostatic pressure testing shall be employed for these tests.

e

h5 5.2.4.7 Code Exemptions As provided in ASME Section XI, IWB-1220, certain portions of Class I systems are exempt from the volumetric and surface examination requirements of IWB-2500. De following components (or parts of components) are exempt from the volumetric and surface examination requirements ofIWB-2500:

1. Components that are connected to the reactor coolam system and part of the reactor coolant pressure boundary, and that are of such a size and shape so that upon postulated rupture the resulting flow of coolant from the reactor coolant system under nornul plant operating conditions is within the capacity of makeup systems which are operable from on-site emergency power; e piping of 1 inch nominal pipe size and smaller, except for steam generator tubing; e components and their connections in piping of 1 inch nominal pipe size and smaller;

2. reactor vessel head connections and associated piping,2 inch nominal pipe size and smaller, made inaccessible by control rod drive penetrations.

5.2.5 Reactor Coolant Pressure Boundary r enkare Detection Systems Means for the detection of leakage from the Reactor Coolant Pressure Boundary are provided to alert operators to the existence of leakage above acceptable limits, which may indi: ate an unsafe condition for the facility. The leakage detection systems are sufficiently diverse and sensitive to meet the criteria of O

Regulatory Guide 1.45 for leaks from identified and unidentified sources. De leakage detection systems are capable of performing their functions following seismic events that do not require plam shutdown.

Anmed Da&n hinerW = RCS and Connened Byneme Page 6.M

G/30/%

System 80+ DCD - Potential Design Changes Item Number: 2 Summary

Description:

NPSH for SCS and CSS Pumps i Affected DCD Sections: CDM: None ADM: Section 6.5.3.4 Figures 6.3.2-1 A and 6.3.2-1B Description of Change:

Reason for Change: There are two root causes for this design change:

1. While bidding System 80+ for the Lungmen Prc-ject in Taiwan, subsequent to the issuance of the Final Design Approval, ABB-CE received data from prospective pump vendors KSB and Ingersoll-Dresser. The data showed that the net positive suction head required (NPSHR) by the proposed shutdown cooling (SC) and containment spray (CS) pumps exceeded the NPSHR assumed in preparation of DCD Section 6.5.3.4. We had assumed a maximum NPSHR of 20 feet at pump runout flow of 6500 gpm. The vendors require approximately 25 feet at 6500 gpm.

2. During the engineering work to support CESSAR-DC, ABB-CE did not evaluate the net I positive suction head (NPSH) available to the SC pumps when aligned for containment spray.

As a result, two design concerns were identified:

1. There may be insufficient NPSH to the containment spray pumps if they operate at the maximum flow rate of 6500 gpm stated in the DCD.

2. There may be insufficient NPSH to the SC pumps when aligned for containment spray, due to large frictional losses caused by the length and diameter of the suction crossover piping. '

[The NPSH available to the SC and CS pumps during all other operating modes is adequate.]

Description:

The NPSH during containment spray operation was evaluated and some SCS/ CSS

~

suction lines were re-sized so that the minimum available NPSH exceeds the NPSHR specified by the pump vendors. Based on the revised NPSH calculations, and as shown on the attached markup of DCD Figures 6.3.2-1 A and IB, the changes are:

Increased the nominal diameter of CS suction piping, including valves SI-104,105,157, and 158, from 18 to 20 inches Increased the nominal diameter of CS/SC pump suction crossover piping, including valves SI-340 and 342, from 18 to 20 inches Increased the nominal diameter of SC suction piping, including valves SI-107 and 106, from 14 (or 18) to 20 inches Despite increasing these line sizes, it was not possible to obtain enough NPSHA for the CS or SC pumps during containment spray operation above 6000 gpm. 'iherefore, the maximum allowable TKSdedch. doc 2 6/11/96

I 2re ~ 2.

Rct v' o i

~

System 80+ DCD - Potential Design Changes 1

1 containment sprey flow rate was reduced from 6500 gpm to 5500 gpm. Safety and containment

' analyses use CS flow rates of $000 gpm minimum,6500 gpm maxi'num, so this change provides l

margin to the maximum assumed v61ue.  !

During the performance of these analyses, it was also determined that the IRWST water volume presented in DCD Section 6.5.3.4 was the volume above the pump suction piping. The volume will be represented as to the total volume remaining in the IRWST (198,000 gallons), to be consistent with the measurement in Technical Specification 3.5.4. The minimum IRWST level (75.5 feet) is unchanged.

The following table compares original and revised NPSH values for containment spray l operation- 1 i

l

NPSH Required (Feet) Minimum NPSH Available(Feet)

Flow Rate Assumed in Vendor CS Pump CS Pump SC Pump SC Pump (gpm) DCD Data Original Revised Original Revised

5000 none 16.5 24 21.9 n/c 20.7 l 5500 none 18 n/c 21.1 n/c 19.6 6000 none 20 n/c 20.3 n/c 18.5 2

6500 5 20 25 21.2 19.4 n/c 17.4 l There are no design changes to the pumps themselves. The SC and CS pumps are still l l identical and interchangeable. The minimum pump design head and design flow will be -

l confirmed by ITAAC, as presented in Certified Design Material Tables 2.3.2-1 and 2.4.6-1.

This change has no impact on safety. In fact, it refines the System 80+ design to ensure that the CS and SC pumps are more likely to meet their safety functions.

1 DCD Markups Attached? Yes 4

TKSdedch. doc 3 6/11/96

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i Tre m 2, eu of y Syst:m 80+ orsion ca.ntrolDocument 8

Figure 6.2.1-16), the long-term airborne elemental concentration is calculated to be less *.han 20 percent of the organic. Thus, depression of the sump pH due to the long-term production of hcl by the c[ irradiation of the electrical insulation and the radiation-induced nitric acid formation would not impact Q o the dose assessment.  ;

Nb g The transient spray removal lambdas for the 10 CFR 100 LOCA analysis are shown on Figure 6.5-6.

-d 6.5.3.4 Available Net Positive Suction Head (NPSH) v (gn The cooling

, d term IRWST is the modes of suction source post-accident for theAsSI operation. pumpsinand described CS 6.8, Section pumps duringVolume the Holdup short Tankterm injection an d (HVT) performs water collection services after an accident. Spillways allow accumulated water in the l .e HVT to spill back into the IRWST, thereby replenishing IRWST water volume during accident l . fan operations. He minimum available NPSH for the SI and CS pumps was determined based on the l 7 f minimum water level in the IRWST during accident conditions. In addition, the following conservative

! n3 assumptions are made:

l c.

'N *

'Y Fluid conditions in the IRWST are saturated: no credit is taken for an increase in contamment pressure.

l l

n

$8 The contribution of the volume of water spillage from the RCS and safety injection tanks is g

conservatively neglected.

i l

c.) 2

• With the CS system actuated, the reactor cavity is assumed flooded and the HVT full to a level that is just below the level at which water begins to return to the IRWST through the spillways.

j t Spray water is being held up on surfaces throughout the containment. Locations for the 6 accumulation of water inside the conenmment include water held up on horizontal surfaces, Q%

M clogged floor drains, water held up in containment spray piping, water in the containmant l 4giE' atmosphere, water film on vertical surfaces, puddles trapped on equipment, water soaked into gg l

insulation, and the containment free volume filled with steam.

5 The Si and CS pumps are located in the reactor building subsphere and are placed 1 ow enou below the

[LE' minimum IRWST fluid level to assure adequate available NPSH. De minimum 4RWST fluid volume

after an accident has been determined to be , allons. This corresponds to a water level elevation of 75.5 feet. Iy og

-d - 9(.9 pl. / e calculatedjavailabre NPSH for the CS pumps ranges from the design flow rate of 5000 gpm to feet at a pump runout flow o Thi gpm.t s exceeds the CSy_ , required NPSH of feet at runout flow. oo /r mmwn g he calcula avuame NPSH for the SI pumps is 26.9 feet at a pump runout flow of 1235 gpm. This y exceeds the S' y_ y required NPSH of 20 feetVat ut h 0 a adypd E P"P

{ During a LOCA, the reactor cavity will not be as w'as e 4 6 uetermining the mmimum l ]. ._ 3 water level of 75.5 feet. Because of this, an additional volume of water will be available to raise the i

I

[d0 minimum water level in the IRWST by approximately 2 feet, thereby increasing the available NPSH.

Prevention of the entrance of debris into the IRWST and SI and CS pump suction lines is described in dw Section 6.8.

4preved Design MManel- hymeered Seiny Feeauen Page 6.5-23

System 80+ DCD - Potential Design Changes Item Number: 3 Summary

Description:

Revise Safety and Seismic Classification of SDS Components from PSVs to IRWST Affected DCD Sections: CDM: None ADM: Tables 3.2-1 and 3.2-2 Description of Chanee:

This is a consistency change to correct errors, not a design change.

Reason for Change: The safety and seismic classification of the SDS spargers, vacuum breakers and piping from the SDS valves and PSVs to the IRWST were specified inconsistently. ABB-CE desires to change classifications so that all components in this portion of the SDS are classified consistently.

1

Description:

This change revises the safety class and seismic category of the spargers and <

vacuum breakers, and upgrades the seismic category of the piping, so that they have the same safety classification and seismic category. For the spargers and vacuum breakers, the safety class will change from 2 to Non-Nuclear Safety (NNS) and the seismic category will change from I to ,

II. For the piping, the seismic category will change from non-seismic (NS) to II. The quality  ;

class will change from 1 to 2 for all NNS components.

This change has no impact on safety: The spargers are not pressure retaining components, and do not perform and safety-related function. Changing the safety classification from 2 to NNS changes the design code from ASME B&PV Section III to ASME/ ANSI B31.1, which has similar design rules for structural integrity. Changing the seismic category from I to II requires that the spargers still maintain sufficient integrity during seismic events, such that they would not damage safety-related equipment during the Safe Shutdown earthquake.

The vacuum breakers do not perform a safety-related function. Changing the seismic category from I to II requires that the vacuum breakers still maintain sufficient integrity during seismic l events, such that they do not damage safety-related equipment.

This change provides consistent safety, seismic, and quality classification for these components.

As currently specified, there would be several interfaces between safety and non-safety equipment, which would be complex to design and construct. The proposed revisions would provide consistent interfaces and higher assurance of correct construction.

DCD Markups Attached? Yes TKSdedch. doc 4 6/11/96

' ~

l System 804 B# Cb &g W A4J Deglan_Cynypt D_QQum_qc_t Table 3.2-1 Classificadon of Structures, Systems and Components O i

Safety Seismic Qusuty l rwnpanant Identincation Class Category i Mn "* D Class" Reactor Coolant Systen Ranctor Vessel 1 I RC 1 Stamm Generators (prunary/azoodary) 1/2 [1]* I RC 1 Pressuruer 1 I RC 1 l Rasctor Coolant Pumps (2,3,9]* 1 I RC 1

Penng within Ranctor Coolart Pressure 1/2 [4] I RC 1 j Boundary [5]

Control Fla-a-t Drive Mechanisms [6] [6] RC 1 Core Support structures and 3 i RC 1 Intanais Stras==s (71

, Fuel Aassablies (8] 2 I RC 1 Control Klamant Assemblies (8] 3 I RC 1 Closure Head IlR Rig NNS II(10] RC 2 l Hasted Junction Thennocouple Probe 1/3 (12] I RC 1 Assably HJTC Pressure Houang 1 I RC 1 ICI Cable Trsy Support Fracas 3 I RC 1 ICI Holding Prume NNS NS RC 3

!CI Guide Tubes 1 I RC 1 IQ Guide Tube Supports 1 I RC 1 m~

ICI Seal Housmg 1 1 RC 1 i ICI Seal Table 1 I RC 1 Piping (27] 1/2 I RC 1 Valves (27] 1/2 I RC 1 fa.cantal=mant Wats Storage Systeen

.IRWST 3 I RC 1 Holdup Volume Tank 3 I RC 1 Pressure Rabsf Despers 3 I RC 1 Cavity stoodtes Systen Piping 2 I RC 1 Vdm 2 I E 1

" ^'

Safety E _ --- Systen l Valves ikg2 I/JE RC t1 2.,

l[15 Piping 1/2/NNS IE RC Sparsers INNS /.ZE RC ya 8 story Infection system -

Safety Iqleation Pumps 2 I RB 1 Safety Iq)ection Tanks 2 I RC 1 l Prping [24,27] 1/2 I RB/RC 1 Valves [27] 1/2 I RB/RC 1 0'

Refer to Notes at enct of table.

Appnsed DeaQn nieennd Dea 6m d SCC (2ARE) hoe 2.2 4

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

+

TTe M ^b i

Svstotn 80+ Desian CanoelDocument Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.)

Safety Seismic Quality

Component Identification Class Category IAcation" Class

• W

! Station Air Systesu j Air Compressors NNS NS SB 3

{ Air Dryers / Filters NNS NS SB 3 Air Receivers l

1 Piping [27]

NNS 2/NNS NS I/NS SB All 3

1/3 k

i Valves [27] 2/NNS UNS All 1/3 1

1 Breathing Air System

! Air Compressors NNS NS SB 3 I

Piping [27] 2/NNS UNS All 1/3

Valves [27] 2/NNS UNS All 1/3 j Air Receivers NNS NS SB 3 j Air Dryer / Filters NNS NS SB 3 i Compressed Gas Systems

. High Pressure Gas Cylinders NNS NS YA 3

, Pressure Regulators NNS NS YA 3

Leak Detection Systems NNS NS All 3 i Liquid Nitrogen Evaporators NNS NS YA 3 j

j Piping [26, 27]

Valves [27]

2/NNS 2/NNS UNS 1/NS All

' All 1/3 1/3 Fire Pmtection Systen Jockey Pump NNS NS FP 2 Backup Storage Tank NNS I NA 1 Fire Pumps NNS NS FP 2 Backup Fire Pump NNS I NA 1 Storage Tanks NNS NS FB 2 Water Spray Systems (Deluge and 2/NNS I/II/NS TB/NA/RC/RB/ 1/2 Spnnkler) Piping, Valves [16,27] DG/SB p Hose Systems / Standpipes [16,27] 2/NNS I/NS All 1/2 Portable Fire Extinguishers [16] NNS NS All 2 Exterior Distribution System Piping NNS NS YA 2 Vdvm NNS NS YA 2 Stramers NNS NS YA 2 Alternate AC Sourte/ Combustion NNS NS YA 2 Turbine-Generator g

.c

'o

,j DG Engine 1%sel Oil Systan [17]

l Fuel Oil Storage Tanks 3 1 DF 1 1 g Recirculation Pumps NNS NS DF 3 Booster Pumps 3 I DG 1

.s Fuel Oil Day Tanks 3 I DG 1

\

g emmyg We\ Gewenb 5% f g y M3e\ c,enenhs ^

i Anproved Deenger hineernet . Desiger of SSC Pope 3.2-13 l

Trew 3 System 80+ oesign Control Document Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.)

Safety Seismic Quality p-l Component Identification Class Category Location" Class" Main Steam Supply Systan Piping [21]

Steam Generator to MSIV's 2 I RC/MS 1 Other NNS NS MS/NA/TB 3 5d Main Steam Supply Systan Valves [21]

Safety Valves 2 I MS I MSIV's, MSIV Bypass Valves 2 I MS 1 Atmospheric Dump Valves 2 I MS 1 Valves 2/NNS 1/NS NA/MS/TB 1/3 Containment Hydrogen Recombiner System Hydrogen Recombiners 2 I NA 1 Hydrogen Analyzers 2 I NA 1 Hydrogen Recombiner Control Panel 3 I NA 1 gl Piping [27] 2 I NA/RC 1 Valves [27] 2 I NA/RC 1 Steam Generator Blowdown System

[22]

Flash Tank NNS NS TB 2 Heat Exchanger NNS NS TB 2 Filter NNS NS TB 2 Demineralizers NNS NS TB 2 e-l Piping [271 2/NNS 1/NS RC/TB/MS 1/2 Valves [27] 2/NSS 1/NS RC/TB/MS 1/2 Steam Generator Wet Layup Recirculation System [22]

Piping [27] 2/NNS 1/NS RC/TB/MS 1/3 Valves [27] 2/NSS I/NS RC/TB/MS 1/3 Hydrogen Mitigation System Hydrogen Igniters NNS I RC 2 beib> Potable and Sanitary Water Systems NNS NS YA 3 f Instmmentation and Control Systems Plant Protection System (PPS) s

'Be PPS includes the electrical and mechanical devices and circuitry (from sensors to actuation device input termmals) involved in generating the signals associated with the two protective functions defined below:

Approved Design Material- Design of SSC Page 3.2-18

i

'ITc m 3 System 80 + Design Control Document i

,O i V Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.)

Safety Seisraie Quality Component Identification Class Categor>

LocationM Class" l" Nuclear Annex Structure Control Area 3 I NA 1 l"

EFW Tank / Main Steam Valve House 3 I NA I Area Emergency Diesel Generator Areas 3 I NA 1 l

CVCS/ Maintenance Area 3 I NA 1 i

Fuel Handling Area 3 ,H' I NA 1 (

Other Structures Unit Vent NNS II NA/RB 2 l"  ;

Turbine Building NNS II TB 2 Radwaste Building [28] NNS # f- RW 2 l Station Service Water Pump / Intake 3 I SP 1 Structure Component Cooling Water Heat 3 I CX/YD I Exchanger Structures and Pipe Tunnels Diesel Fuel Storage Structure 3 1 DF 1 i i a nsn Services Building / Auxiliary NNS NS SB 3 l

! (O

( ,/

. oner Struc.,se Administration Building ADB l

, NNS NS 3

, Warehouse NNS NS WH 3 l j Fire Pump House NNS NS FP 3

• j 1

Alternate AC Source / Combustion NNS NS YA 2 Turbine-Generator Structure and Fuel Tank  !

Dikes Dike (Holdup, Boric Acid Storage and NNS II YA 2 Reactor Makeup Water Tanks) [28]

4 Dike (Condensate Storage Tank) [28] NNS II YA 2 Cranes

, Polar Crane NNS 11 RC 2 Cask Handling Hoist NNS 11 NA 2 New Fuel Handling Hoist NNS II NA 2 Component Supports [23] 1/2/3/NNS I/NS All 1/2/3  %

i  !

i

. \

4 Apneroved Design Matenal Design of SSC Page 3.2 21 4

System 80+

k OWa?& hw b b Desian controlDocument

.R

, Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.)

Component Location / Safety Selanic Quality Identification Description Class Category Class Pool Cooling and Purification System (PCPS) [1] (Cont'd.)

PC-208, 209 Cooling pump discharge isolation 3 I 1 PC-211, 212 Cooling HX inlet isolation 3 1 1 PC-213, 214 Cooling HX outlet isolation 3 I 1 PC-249 IRWST return line isolation 3 I 1 9 PC-257, 258 Refueling pool discharge isolation 2 1 1 PC-291, 292 Refueling pool inlet isolation 2 1 1 PC-300, 301, 302, 303 Cooling flow indication isolation 3 I 1 PC-320, 321 Cooling pump suction pressure 3 1 1 Safety Depressurization Systesa (SDS)

RC-406, 407, 408, 409 Rapid depressuruation 1 1 1 RC-410, 411, 412, 413 Pressurizar vent 1 1 1 RC-414, 415, 416, 417 Reactor vessel vent 1 1 1 RC-418 RCGVS vent to RDT 2 1 1 A RC-419 RCGVS vent to IRWST 2 I 1 RC-263, 264 RD pressure indication 2 I 1 RC-267 RCGVS pressure indication 2 1 1 g.f 2 'RC-XXX s

\

\

\

'3DS/ Safety ' valve spargerht vacham breaker \ x \

N .2' ON d,'N

• q s k RC-XXX SDSISdsey volve sparger linas2 I2N y' \

QfA N \ vacurm bmiker- N #NS NNS Safety hdection Systaan (SIS) [1]

SI-100,101 IRWST return check valve 2 I 1 SI-102,103 IRWST isolation valve test 2 1 1 SI-104,105 CS pump suction isolation 2 1 1 SI-106,107 SCS pump suction isolation 2 I 1 SI-108,109 SCS pump suction pressure indication 2 1 1 isolation SI-113, 123, 133, 143 Safety injection contamment check 2 I 1 l

SI 115, 125, 135, 145 SI flow indication isolation 2 1 1 l

SI 116, 126, 136, 146 SI flow indication isolation 2 1 1 SI 117, 127, 137, 147 SIT pressure indication isolation 2 I 1 ,.m l SI-119, 129, 139, 149 SIT pressure indication isolation 2 1 1 AmsmeetDeske AfstenW G,niyn of SSC Age J.242 ,

i l

System 80+ DCD - Potential Desien Changu Item Number: 4 Summary

Description:

Revision to Tech Spec Figure 3.5.4-1 Affected DCD Sections: CDM: None ADM: Chapter 16, Fig 3.5.4-1 ADM: Appendix 16A, Pg B 3.5-25 ADM: Chapter 6, Tbl 6.2.1-22 l

Descrintion of Change:

Figure 3.5.4-1 provides a curve that indicates the maximum allowed containment atmosphere  ;

temperature vs. the IRWST water temperature. The current figure has a allowable IRWST l' temperature range from 40 F to 110 F. The proposed change to the figure is to revise the scale for the IRWST temperature range to indicate an allowable region from 60 F to 110 F. The basis for the change is the minimum allowable IRWST temperature assumed in the containment safety analysis. The safety analysis was performed based on input found in document ALWR-FS-DD140," Plant Safety, Performance, Containment and Severe Accident Analyses for the System 80+ Standard Design". Although this document is not contained in the DCD, many of the assumption have been included in the applicable chapters,6,15,19, etc. The minimum allowable IRWST temperature is stated as an assumption in Section 6.2.1.5.3.4 " Active Heat Sinks"in the DCD as 60 F. As such, Tech Spec Fig 3.5.4-1 should be revised to be consistent with the stated assumptions. Table 6.2.1-22 should also corrected to be consistent with the

assumptions used in the Containment Pressure Analysis.

. The slope of the curve in Figure 3.5.4-1 was developed based on a parametric study that varied l j

the containment and IRWST temperatures. This slope will not change as a result of revising the IRWST temperature scale to indicate a minimum of 60 F. As an example, the revised figure ,

indicates a minimum containment air temperature of approximately 94 F is needed for operation

, when the IRWST temperature i.s 60 F.

In addition to the Figure, the tech spec bases B.3.5.4 provides an example in the " Applicable Safety Analyses section for use of Figure 3.5.4-1. The current example provides an IRWST temperature of 53 F with a containment temperature of 90 F; this should be revised because of the revision to the Figure. A new example indicating an IRWST temperature of 81 F with a containment temperature of 110 F is recommended for use in the Tech Spec bases section.

DCD Markups Attached? Yes i

1Tc M d

. IRWST 3.5.4 l 120 , , , , , , ,

MAXIMUM ALLOWED IRWSTTEMPERATURE 110 - ... .-

t l _ 100 - -

t

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OPERATION -

i Q- ALLOWED

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! ALLOWED j 70 - -

i 60 60 70 80 90 100 110 120 130 140 CONTAINMENT ATMOSPHERE TEMPERATURE (oF) i Figure 3.5.4-1 Allowed IRWST Temperature vs. Containment Atmosphere Temperature i

1 i

l SYSTEM 80+ 3.5-11 Rev. 00 16.3 Tech Spec

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

l Irex 4 l

l IRWST l B 3.5.4 l BASES l .

l l l APPLICABLE The final containment pressure after an inadvertent l SAFETY ANALYSES containment spray actuation is sensitive to the initial (continued) containment atmosphere temperature and the IRWST water J

temperature. Figure 3.5.4-1 shows the minimum allcwed IRWST 4 water temperature for a given containment atmosphr.re temperature. For example, if the containment atmosphere .

l temperature is ]F, the minimum allowed IRWST water I temperature i [ ]F. The maximum temperature of the IRWST l is [110*]F. gl- 1 JIo* ,

The IRWST satisfies Criterion 3 of the NRC Policy Statement.

l LC0 The IRWST ensures that an adequate supply of borated water ,

is available to cool and depressurize the containment in the  !

event of a Design Basis Accident (DBA) and to cool and cover the core in the event of a LOCA. The IRWST ensures the l reactor remains suberitical following a DBA.

To be considered OPERABLE, the IRWST must meet the limits established in the SR for water volume, boron concentration and temperature.

APPLICABILITY In MODES 1, 2, 3 and 4 the IRWST OPERABILITY requirements are dictated by the SIS and Containment Spray System OPERABILITY requirements. Since both the SIS and  :

Containment Spray System must be OPERABLE in MODES 1, 2, 3 and 4, the IRWST must be OPERABLE to support their operation. j In MODES 5 and 6 with RCS level < [120'-0"] the IRWST .

OPERABILITY requirements are dictated by the SIS. Two l trains of SIS are required in these MODES, therefore the ,

IRWST must be OPERABLE to support the SIS. j l

l l

i I

(continued)

SYSTEM 80+ B 3.5-25 Rev. 00 l

16A Tech Spec Bases (2/95)

Tre u 4 System 80+

p t :: Contmln w= = = t Table 6.2.1-21 Typical Passive Heat Sink Data (Cont'd.)

Notes:

111 This column of infonnation is intended to be used for analyses where a small free volu conservative.

[21 This column of information is intended to be used for analyses where a large free volum conservative.

151 Minunum surface area does not reflect IRWST surface while surface area does.

I'l Total surface area exposed to contmiment atmosphere. All walls assumed to be insulate Isl Typical material properties used in contammant pressure and temperature .malyses in S 181 4

Typical matenal properties used in matamment pressure and temperature analyses in S t

i J

i a

Table 6.2.1-22 Initial Conditions for Containment Minimunn Pressute Analysis 5O l Parameters Amann=1 Value l Initial temperature, 'F (max) 110 Initial pressure, psia (min) 14.3 Relative humidity, % (max)

Refueling water temperature, 'F (min) 100_ -

l No heat input to conta=maat from structures or prunary and

(. W 81 )

j secondary system components V

[

1 Ideal gas behavior of air in contanament 1

O A

MM W M

• W S0% feeheet & 6.2 176

l i

System 80+ DCD - Potential Design Changes Item Number: 5 Summary

Description:

Reserved CEA locations in Reactor Core i

Affected DCD Sections: CDM: Yes, Fig 2.2.1-3; CDM: Delete maximum number of CEDMs in Section 2.2.2 ,

ADM: Yes, Fig 4.2-11, Fig 4.3-46, Fig 4.3-47 l ADM: Revise Text in Section 4.2.2.4 l Descrintion of Change: '

The objective of the change is twofold: (1) to allow for the possibility of having 4 element CEAs at twenty (20) specific core locations; and (2) to allow for the possibility of replacing 4-element CEAs with 12-element CEAs at four (4) specific core locations. Not all locations identified as possibly containing a CEA would necessarily have a CEA installed. The minimum number of CEAs would remain at 93.

The proposed change would allow flexibility in choosing locations for additional CEAs. The addition of CEAs in the central core region would permit enhanced maneuvering capability. The addition of CEAs could also improve shutdown margin. All locations identified for additional CEAs satisfy the minimum reactor vessel head ligament requirement.

The ADM (Section 4.1.1) currently allows changes to certain features and evaluated parameters for the fuel system design, nuclear design, and thermal and hydraulic design of the initial core without prior NRC review and approval provided these changes are within certain acceptance criteria. With the propose change, Section 4.1.1 would remain unchanged.

The CDM (Figure 2.3.1-3) and the ADM (Figure 5.3-7) show a minimum of 103 CEDM and instrumentation nozzles, and a minimum of 2 HJTC probes. The proposed change will maintain these minimum values.

DCD Markups Attacha3? Yes dedchng4. doc 6/27/96

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} Nuclear Fuel System Arrangement Figure 2.2.1-3 a

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Certined Design Meterial Page 2.2-5 i

, l Ire 8 s l l 4./Me /% l System 80+ Design controlDocument l

2.2.2 Control Element Drive Mechanism Design Description The control element drive mechanism is a magnetic jack device that positions and holds the control element assemblies relative to the fuel assemblies.

l The primary safety-related function of the Control Element Drive Mechanism (CEDM) is to release the l Control Element Assembly (CEA) upon termination of electrical power to the CEDM. A minimum of l

93 CEDMs is required, hc==, ; .ms-"- af n= h'_'+:d = CEOM mou bu um ikJ. r-.

l l

The CEDM also acts as a primary pressure boundary as part of the Reactor Coolant System. Refer to Section 2.3.1 for CEDM primary pressure boundary aspects.

Inspections, Tests, Analyses, and Acceptance Criteria None The initial test program addressed in Section 2.11 will test the ability of the CEDM to release the CEA upon terminatiori of electrical power to the CEDM.

The Basic Configuration of the CEDM primary pressure boundary components will be verified as part of Section 2.3.1.

The CEDM pattern will be verified as part of Section 2.2.1.

i j

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I CertrMed Design Material Pope 2.2-7

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a spider structure which couples to the control element drive mechanism (CEDM) drive shaft extension. 1 The neutron absorber elements of a four-element CEA engage the four corner guide tubes in a single fuel assembly. The four-element CEAs are used for control of power distribution and core reactivity in the l power operating range. The twelve-element CEAs engage the four corner guide tubes in one fuel assembly and the two nearest corner guide tubes in adjacent fuel assemblies. The twelve-element CEAs make up the balance of the control groups and provide the core with strong shutdown rods. The control element assemblies are shown in Figures 4.2-3 through 4.2-5 and Figure 4.2-14. The pattern of CEAs (total of 93f,irshown in Figure 4.2-11. ^" *"* "- *^ W ,

d"^-^'

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Twenty-five of the 93 CEAs are part-strength CEAs (PSCEAs);

l / art-strength CEAs are differentiated from full-strength CEAs by using alphanumeric serialization instea of the numerical system used on the full-strength CEAs.

All control elements are sealed by welds which join the CEA cladding to an inconel 625 nose cap at the bottom, and an Inconel 625 connector at the top which makes up part of the end fitting. The end fittings, in turn, are threaded and crimped in place by a locking nut to the spider structure which provides rigid I

lateral and axial support for the control elements. The spider hub bore is specially machined to provide a point of attachment for the CEA extension shaft.

The control elements of a twelve-element full-strength CEA consist of an Inconel 625 tube loaded with a stack of cylindrical absorber pellets. The absorber material consists of 73% TD boron carbide (B4C) pellets, with the exception of the lower portion of the elements, which contain reduced diameter B4 C pellets wrapped in a sleeve of Type 347 stainless steel (felt metal).

The design objective realized by the use of felt metal and reduced diameter B4C pellets in the element tip zones is that as the B 4C pellets swell due to irradiation, the felt metal sleeve compresses as a result l of the applied loading. His compression limits the amount ofinduced strain'in the cladding. "herefore, buffering of the CEA following scram, which occurs when the element tips enter the reduced diameter portion of the fuel assembly guide tubes, is not affected with long term exposure of the CEA to reactor ,

operating conditions.

)

During normal power operation, all of the twelve-element CEAs are expected to be in the fully withdrawn position. Thus, the local B-10 burnup progresses at a lower rate, and CEA life is prolonged. Above the absorber column is a plenum which provides expansion volume for helium released from the B4 C. The plenum volume contains a Type 302 stainless steel holddown spring, which restrains the absorber material against longitudinal shifting with respect to the clad while allowing for differential expansion between the absorber and the clad. He spring develops a load sufficient to maintain the position of the absorber material during shipping and handling.

He control elements of a four-element full-strength CEA consist of an Inconel 625 tube loaded with a stack of cylindrical Ag-In-Cd absorber bars. This CEA design is used for the regulating banks. Two design objectives are realized by use of Ag-In-Cd absorber over the full active length:

o CEA Cladding Dimensional Stability l

Because of its high ductility and low strength, the Ag-In-Cd will not deform the CEA cladding.

Buffering of the CEA following scram, which occurs when the corner element tips enter the reduced diameter portion of the fuel assembly guide tubes, is not degraded with long-term exposure of the CEA to reactor operating conditions.

Anwoved Design Meterial Reactor Pere 4.2-34

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P2 - PART STRENGTH GROUP 2 (LEAD)

Pg - PART STRENGTH GROUP 1 .

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Anwoved Desier, Material- Reactor Page 4.3-97

1 System 80+ DCD - Potential Design Chann_es 48 te ,

Item Number: [h Summary

Description:

Add RCS Mid-loop Level Measurement Tank and Instrumentation Across the Hot Leg Affected DCD Sections: CDM: None ADM: Chapter 5  ;

Chapter 7 Appendix 19.8A Description of Change:

]

A detailed engineering evaluation was performed to confirm the feasibility of adding tanks connected directly across each of the RCS hot leg pipes to measure the hot leg water level during Mid-Loop operations using HJTC instruments. The current DCD Chapter 5 includes the RCS P&lD and the cross section of the Reactor Vessel. The Shutdown Risk evaluation reported in DCD Chapter 19.8A provides a figure (Figure 2.8-4) which indicates that two additional sets of HJTC's are imerted through the reactor vessel head to measure the hot leg water level. The l DCD text indicates this level should be measured to an accuracy of plus or minus 1 inch. In l Chapter 19 the tank across the hot leg will be shown for measuring the hot leg water level during

! mid loop operations. The water level will be measured with heated junction thermocouples that l operate in the same manner as the ICCMS HJTC located in the reactor vessel.

l The reasons of the proposed change are as follows:

I 1

1. the instruments are permanently installed and are not affected by refueling activities in l the reactor vessel head area.

l 2. the reliability of the instruments is enhanced by relocating them to a more benign area.

l l 3. there is improved flexibility for plant operations and maintenance for servicing the instruments and installing nozzle dams.

4. the water level measurements are taken closer to the Shutdown Cooling suction line which is the point ofinterest.

Locating the Mid-Loop HJTCs through the reactor vessel head have limitations that the proposed change eliminates. The cable to the head mounted instruments must be disconnected when the reactor vessel head is removed. The disconnection affects the ability of the instruments to be used. Having the instruments permanently connected reduces the potential for human error. j j Since the position of the reactor vessel head (and components attached to it) changes when the i head is detensioned, the instruments should only be used with the head tensioned. The amount  ;

! of position change is a function of the fuel burnup and is therefore not constant from fuel cycle to l fuel cycle or from before refueling to after refueling. Relocating the instruments will reduce human error and increase operational flexibility. The current design reduces the flexibility in TKSdedch. doc 6 6/11/96

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i System 80+ DCD - Potential Design Channes rw & % I \

when the nozzle dams can be installed or removed from the steam generator. The proposed change allows more flexibility and will result in the nozzle dams being installed for a shorter l l period.

The location of the instruments in the reactor vessel means that they are exposed to reactor i pressure, temperature and radiation. These conditions will decrease the operating life of the instruments compared to locating the instrument in the tanks. Instrument replacement in the reactor vessel is more difficult and results in higher personnel exposure. l The Mid-Loop HJTC System uses the proven Heated Junction Thermocouples to measure the i -

water level that are the same design as the ICCMS HJTCs. The HJTC heater controllers are the same as the ICCMS HJTC heater controllers and each controller is connected to a separate power supply. Placing the HJTC instruments in the tank allows a larger instrument diameter which results in higher resolution of the HJTC by using radial distance to accomplish thermal separation of the HJTCs. The instrument operating environment is at atmospheric pressure and i

temperatures of less than 212 F which will increase the operating life and increase reliability.

l The water level is indicated in the Control Room.

There is a fluid connection from the bottom of the hot leg to the bottom of the tank. An air connection from the top of the hot leg to the top of the tank causes an equalization of the water level in the hot leg and tank during mid-loop conditions. There are two (2) isolation valves on

each of the connecting lines. The connecting lines up to and including the second isolation valve 1

are designed RCS conditions using the ASME B & PV Code,Section III, Class 1. The remaining portions of the connecting lines and the tank up to and including the tank drain valve

are designed for reactor operating temperature and pressure in accordance with ASME B & PV

Code,Section VIII. The drain line from the tank connects to the liquid radwaste system.

1 DCD Markuns Attached? Yes

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TKSdedch. doc 7 6/11/96

.Tlvn fs Sev l Syst!m 80+ Drsion Control D?cumint

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5.0 Reactor Coolant System and Connected Systems l 5.1 Summary Description The reactor is a pressurized water reactor (PWR) with two coolant loops. The reactor coolant system (RCS) circulates water in a closed cycle, removing heat from the reactor core and internals and i

transferring it to a secondary system. The steam generators provide the interface between the reactor coolant (primary) system and the main steam (secondary) system. The steam generators are vertical U-4 tube heat exchangers with a'i integral economizer in which heat is transferred from the reactor coolant to the main steam system. Reactor coolant is prevented from mixing with the secondary steam by the

- steam generator tubes and the steam generator tube sheet, making the RCS a closed system thus forming a barrier to the release of radioactive materials from the core of the reactor to the secondary system and containment building.

The arrangement of the RCS is snown in Figures 5.1.31 and 5.1.3-2. The major components of the system are the reactor vessel; two parallel heat transfer loops, each containing one steam generator and 3 two reactor coolant pumps; a pressurizer connected to one of the reactor vessel hot legs; and associated piping. All con.ponents are located inside the containment building.

Table 5.1.1-1 shows the principal pressures, temperatures, and design minimum flowrates of the RCS under normal steady-state, full power operating conditions. Instrumentation provided for operation and control of the system is described in Chapter 7.

System pressure is controlled by the pressurizer, where steam and water are maintained in thermal equilibrium. Steam is formed by energizing immersion heaters in the pressurizer, or is condensed by the pressurizer spray to limit pressure variations caused by contraction or expansion of the reactor coolant.

The average temperature of the reactor coolant varies with power level and the fluid expands or contracts, changing the pressurizer water level.

The charging pumps and letdown control valves in the chemical and volume control system (CVCS) are used to maintain a programmed pressurizer water level. A continuous but variable letdown purification flow is maintained to keep the RCS chemistry within prescribed limits. A charging nozzle and a letdown nozzle are provided on the reactor coolant piping for this operation. The charging flow is also used to alter the baron concentration or correct the chemical content of the reactor coolant.

Other reactor coolant system penetrations are the pressurizer surge line in one hot leg;,pe four direct vessel injection nozzles in the reactor vessel for the safety injection system; two return nozzles to the shutdown cooling system, one in each hot leg; two pressurizergrg" "%"M-nozzles; vent and drain connection and sample and instrument connections /M9 h a**v Overpressure protection for the reactor coolant pressure boundary is prcvided by four spring-loaded ASME Code safety valves connected to the top of the pressurizer. These valves discharge to the in-containment refueling water storage tank, where the steam is released unde.r water to be condensed and cooled. If the steam discharge exceeds the capacity of the in-containment refueling water storage tank, it is vented to the containment atmosphere, i Overpressure protection for the secondary side of the steam generators is provided by spring-loaded ASME Code safety valves located in the main steam system upstream of the steam line isolation valves.

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range DP sensors measure RCS level in the hot leg region. The narrow ranj;e instrumentation includes low and low-low alarms which annunciate in the control room. The wide range DP sensors measure RCS level in the hot leg to the top of the pressurizer. He wide range instrumentation also includes low and low-low alarms which annunciate in the control room. The indication and alarms allow the operator to monitor RCS level from the control room during shutdown operations which require reduced RCS s inventory.

N D. Two redundant of the vessel to the fuel Refueling alignment plate.Water The narrow Level Probes range RWLP (RWLP) assemblies measure provide independent reactor coolant N liquid inventory in the hot leg region during reduced inventory periods when the reactor head is installed.

K Re wide range RWLP assemblies measure reactor coolant liquid inventory in the upper portion of the vessel to the fuel alignment plate. The basic principle of operation is the detection of a temperature difference between one of the heated thermocouples and the imhetM thermocouple at the bottom of the y RWLP assembly. Each RWLP assembly includes multiple HJTC sensors, r.n outer sheath, a seal plug 13 and electrical connectors.

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-4 he RWLP thermal hydraulic operating environment is relatively uncomplicated. He narrow range h RWLP is used only during non-power operation while the reactor vessel head is in place. The probe assembly is housed in a stainless steel structure that protects it from flow loads.

He narrow range RWLP heated junction thermocouple sensors are more closely spaced in the hot leg region to provide improved resolution. The RWLP's provide indiention, high, high-high, low, and low-low alarms in the control room.

n RCS temperature is measured using the existing CET temperatures, HJTC unheated sensor temperatures, and RCS Hot Leg RTD temperatures. De CETs and HJTC imhe*M sensors have high and high-high alarms to annunciate the approach to bulk boiling in the core. The RTDs have a high alarm anmmdstian.

He HJTC unheated sensor temperature is not available when the head is off.

Each train of the SCS has a measurement of SCS flow. This measurement provides indication of return flow to the RCS when either the SCS pump or CS pump is being used for shutdown cooling. Low flow is annunciated in the control room.

To monitor the performance of the SCS and CS pumps, pump suction pressure, discharge pressure and motor current are monitored and annunciated in the control room.

He performance of the SCS heat exchanger is monitored and annunciated by measuring the temperature j in the inlet and return lines. Valve position indication is provides inlication of the system lineup and provides the status of the available flowpaths, 7.7.1.1.16 Steam Generator Tube Rupture Detection Instrumentation System 80+ incorporates N-16 gamma detection with a scintillation detector and microprocessor based I signal conditioning on a header leaving each steam generator. A description and the applicability of using N-16 gammn in detecting steam Generator tube leaks is provided in Section 5.6 of Appendix 5F. The detection system will alert the operator to a SG tube leak condition origimting at power and identify which steam generator is affected.

He addition of N-16 radiation detection and monitoring equipment further enhances the diagnosis of steam generator tube leaks or ruptures and provides the operator with more accurate information to assess Anreved Deeign Meternet - Innven=sentsoen and Conuel Pope 1.719 l

% (,, Rotrj System 80+ DCD - Potential Design Changes Insert in 7.7.1.1.15 l

The first HJTC system displays the output from the two (2) inadequate core cooling probes I located inside the reactor vessel that are availabe when the reactor vessel head is installed. The inadequate core cooling probes are described in Section 7.5.1.1.7. j 1

The second HJTC system measures the water level in the RCS hot leg pipes duiing Mode 5 reduced inventory conditions. This Mid-Loop HJTC system consists of an instrument installed i

in a tank attached to the RCS hot leg pipe. There is a separate tank and instrument attached to each hot leg in the vicinity of the SCS suction connection. There are two (2) connections each tank and corresponding hot leg. One connection is at the bottom of the hot leg to the bottom of 1 the tank. The second connection is at the top of the hot leg to the top of the tank. When the isolation valves on the connections are open, the water level in the tank equalizes with the water level in the tank. The isolation valves are operated from the control room. The position of the j valves is indicated in the control room.

1 The connecting pipe, up to and including the second isolation valve, is designed to the same i conditions as the Reactor Coolant System and is ASME B & PV Code,Section III Class 1. The tank and connecting pipe after the second isolation valve is designed for RCS operating pressure and temperature in accordance with ASME B & PC Code,Section VIII up to and including the tank drain valve.

During Mode 5 reduced inventory conditions, the Mid-Loop HJTC system is connected to the RCS by opening the isolation valves. The RCS mid-loop vent path described in Chapter 19.8A, Section 2.3.3.3 assures the system pressure at near to atmospheric and the fluid temperature at less than 212*F with no boiling.

The Mid-Loop HJTC instrument consists of a vertical array of heated junction and unheated junction thennocouples. The heated junction thermocouples are spaced to obtain the required measurement resolution and provide alarm points for high level (water level approaching the steam generator nozzles) and low level (water level approaching loss of SCS suction). The thermocouple design is described in Section 7.5.1.1.71.2. Since there is no two phase conditions ,

in Mode 5 at the tank, the instrument does not include the phase separator tubes included in the l ICCMS HJTC. The are separate heater controllers for each Mid-Loop HJTC instrument. Each l controller is connected to a separate power supply to mitigate common mode failure.

The water level in the RCS hot legs is displayed in the control room.

l l

TKSdedch. doc 11 6/11/96 l

T k 5 2eul Syst~m 80 + D sign Control Documint

• mitigation planning aimed at the reinitiation of shutdown cooling, delaying the onset of boiling, and delaying core uncovery.

The design goals of the instrumentation package are to provide:

Prevention -

enhanced monbring capabilities for prevention of a complete loss of SCS operation, and Mitigation -

the timely response to a loss of SCS.

These goals have been achieved with the design features of the System 80 - instrumentation described in the following.

2.8.3.2 Instrumentation Description 9 ,

Table 2.8-1 describes the instrumentation package for reduced inventory operations included in the T '

System 80+ design. Additional details are provided below. &

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2.8.3.2.1 level "l  !

i Four unique sets of instruments are provided for the measurement of level during RCS draindown and 4q l reduced inventory operations. These instruments make up the retueling water level indication system. ij i The first set of instruments is a pair of wide-range, dP-based level sensors. These sensors are provided to measure level between the pressurizer and the junction of each SCS suction line with the RCS during g[*t draindown operations. Another pair of dP-based level sensors is utilized to determine RCS water level I*

once it is within the reactor vessel. These narrow-range level sensors function to measure level between N the direct vessel injection (DVI) nozzle and the junction of the SCS suction lines with the RCS.

It One wide-range and one narrow-range dP instrument are connected to each SCS suction line. Separate -

lower level taps are provided for each instrument. See Figure 2.8-2. Because of the location of the Q upper level taps, each of these dP instruments will operate with, or without, the reactor vessel head in place. f3 J TT In addition to the dP-based instruments described above, two heated-junction thermocouple (HJTC) systems @ also be available for reactor vessel level measurement during Mode 5 reduced inventory &g Jp operations. The first system displays the output from the two inadequate core cooling probes which are j located inside the reactor vessel. The range of these probes extends from the reactor vessel head to the = m fuel alignment plate (See Figure 2.8-3). The measurement of RCS water level via these probes is limite only to those periods when the reactor vessel head is installed. N f \ A N

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1. %^ probes spect ical y es wii k e e ustered in the hot leg region (See Figure 2.8-4). The benetit of this design is that it permits very accurate measurement when the reactor vessel water level is N in the hot legs.

The. fan Xs a tc. loca4vt} sm s a m a. 9 en era l o vea. SS k No em jers, The HJTC systems compensate for the flow gradient across the core associated with the operation of only one SCS suction line. The HJTC instruments are located in areas which minimize the effect of the core outlet nozzles. The HJTC sensors have an accuracy and response time consistent with the maximum Approved Design Matenal- Probabinstic Risk Assessenent Pape 19.8A.47

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l The narrow-range level indication for mid-loop operation is made via measurement of the reactor '

l vessel water level in the hot legs. The mid-loop level instrument is located in tanks connected to the hot legs. The tanks are located to place the instruments at the same elevation as the hot legs. I The connections to the hot legs allow equalization of the water level in the hot legs with the I

! water level in the tanks.

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Monitored Instrument Instrument Indication and Parameter Type Function Range Alarm Comments Location  ;

KCS Water Refueling Water Continuous, Wide Range: two Control Room, Highly reliable. Meets I level Level Indication redundant wide mstruments, each with low and NRC requirement for System (dP based range RCS water with a tap at hot low-low level water level measure-design) level indication leg / SCS suction alarms. ment to a point lower during draindown line interface, and than that required for operations. separate reference SCS operation.

legs at the top of the pressurizer.

RCS Water ICCI HJTC Independent Top of the vessel Control Room, Redundant axial strmgs l Level probe assembly contmucus level down to the fuel with low and of thermocouples from I system indication in the alignment plate. Iow-low level the vessel head to the reactor vessel. alarms. fuel alignment plate. j System provides ex-cellent accuracy and continuous measure-ment.

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RCS Water Refueling Water Independent. Tg ' Le ved Control Room, Redundant axial strings i Level Level Indicating continuous, dawne the Nd with low-low, of thermocouples fresh 4 tv l System (clustered narrow-range alignmens-plata low, high and de v=c! hd :c S- .j., p o $ L.f-HJTC design) level indication 7 p o r- k. h high-high level fueblignment.plew.

pD 4g m the remeter- ,, alarms. This instrument is vessel h e i--  % r,o o, c h different from the ICCI bc.Md Leg, o f-lee hg HJTC system discussed 19 4-j g ,

above in that thermocouples are clustered in the hot leg region to provide greater instrument accuracy (< 1").

RCS Water Refueling Water Contmuous, Narrow range: Control Room, Highly reliable for mid Level level Indication redundant narrow two instruments, with low and loop operations. Meets (dP based range level each with a tap at low-low level NRC requirement for design) indication during SCS suction line/ alarms. water level measure-reduced hot leg interface, ment to a point lower inventory opera- reference leg at than required for SCS tions. DVI nozzle. operation.

RCS CETs Measures Optimized for Control Room, Tracks approach to Temperature (thermocouple temperature of SCS and refueling with alanns at boiling. Temperature design) coolani ex.; ting modes. Will high and high- indication provided core. l measure boilmg. high even when head is off

! Approximate temperature. vessel. Not available range 50-250 deg during fuel shuffling.

F. Availability will be maximized.

Approved Desrgn Maten'al- Probabikstic Risk Assessment Page 19.8A-156

SW $ ECV h System 80 + Disign ControlDocuma D

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9 Schematic Representation of One AltenNe Design for Narrow Range Heated Junction Thermocouple Probes Figure 2.84 Approved Desogn Matenal Probabilistic Risk Assessment Page 19.8A.193

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System 80+ DCD - Potential Design Changes Item Number: f(7 1

Summarv

Description:

Reduce Safety Depressurization System (SDS) Line Size Affected DCD Sections: CDM: None ADM: Figure 5.1.2-3 Chapter 6.7 Chapter 5.4 Chance

Description:

During the Design Certification engineering, a preliminary design analysis for the SDS line size was performed that showed 6 (six) inch piping and valves would pass more than the minimum flow to meet the EPRI requirements to mitigate a Total Loss of Feedwater (TLOFW) event.

During detailed engineering, a more detailed system calculation was performed to determine if 4 (four) inch piping and valves would pass adequate flow. The new analyses confirms that a 4 (four) inch system passes more than enough flow to meet the requirements specified in the DCD (Section 6.7) and preserves the validity of the original TLOFW analysis. This change will revise the RCS/SDS P&ID (DCD Figure 5.1.2-3) to revise the SDS piping and valves from 6 to 4 inches. Modifications to the line and valve sizes are also required in Section 5.4 and 6.7.

Using a four inch SDS system for future System 80+ contracts will allow the use of the same gate and globe valves used for Yonggwang 3 & 4. These valves have been tested and qualified for this service, so proven components would be used and retesting would not be required.

DCD Markups Attached? Yes 1

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l Trr% 7 System 80+ oesian controlDocument Table 5.4,10-1 Pressurizer Parameters Property Parameter Design pressure, psia 2500 Design temperature, 'F 700 Normal operating pressure, psia 2250 Normal operating temperature, *F 652.7 Internal free volume, ft' 2400 Normal (full power) operating water volume, ft' 1200 Normal (full power) steam volume, ft' 1234 Installed heater capacity, kW 2400 Hester type immersion Spray flow, mmimum design capacity, gpm 375 Bypass Spiny flow, continuous, gpm 1-6 Nozzles Surge, in. (nommal) 12, schedule 160  ;

Spray, in. (nommat) 4, schedule 160 '

Safety valves, in. (r.ommal) 6, schedule 160 Safety depressurization, in. (nommal) M/ schedule 160 Instrument level, in. (nomins]) 3/4, schedule 160 Temperature, in. (nommal) 1, schedule 160 Pressure, in. (nommal) 3/4, schedule 160 Hester, 0.D., in. 1-1/4 ANwered Desigt. Material- RCS and Carmected Systems Page 5.4 45

. rveu 7 System 80+ Desian controlDocument ,

cylindrical in shape, about six inches in diameter and range in length from fourteen feet to sixteen feet from the junction with the horizontal header. Because of the asymmetry of the IRWST, the spargers on the side with the ICI chase are submerged about ten feet and about twelve feet on the opposite side.

The evaluation of the hydrodynamic loads was performed with methods and codes developed by ABB-Atom and verified through single cell and in-plant measurements. Analyses were completed based on the maximum mass rate of flow for four PSVs during rapid depressurization following the TLOFW event.

Figure 6.7-5 shows typical results (in this case for the header section) of the influence on the loads due to the gas dynamic shock, the water and air clearing phases and finally steam discharge.

The loads on the SDS piping are within the design capability of piping and supports for SRV piping.

Thus, design of the SDS piping supports can utilize standard methods for piping analyses and support, design.

Table 6.7-1 Safety Depressurization System - Active Valve List Power Source Valve Number Type Line Size 125V DC Bus Actuator Reactor Coolant Gas Vent Valves RC-410 Globe 1.5 inch A Solenoid RC-411 Globe 1.5 inch B Solenoid RC-412 Globe 1.5 inch C Solenoid RC-413 Globe 1.5 inch D Solenoid RC-414 t Globe 1 inch B Solenoid RC-415 Globe 1 inch A Solenoid RC-416 Globe 1 inch D Solenoid RC-417 Globe 1 inch C Solenoid RC-418 Globe 1.5 inch A Solenoid RC-419 Globe 1.5 inch B Solenoid Rapid Depressuruation Vnves RC-408 Gate Y /mch B Motod!1 RC-406 . Globe J/ finch D M otorD I  ;

RC-409 Gate 1/ finch A M otor03 RC-407 Globe 1/ finch C MotorDI i

U1 480 VAC motor operator supplied from Class IE 125V DC through inverter and step-up transformer.

Apnwoved Design Material- Engineered SsMy Feetwes Page 6.717

l System 80+ DCD - Potential Design Changes Item 8: Addition of Alternative LBB Evaluation Method DCD Section: 3.6.3. 3.9A. Figure 3.9A-12 Description of Chanee:

The NRC has approved a change for determining leak-before-break (LBB) crack stability criterion on load. The System 80+ certified crack stability criterion on load was thu a pipe with a leakage crack length subject to loads of 42 x (NOP + Maximum Design Load) was required to have significant margin between the material and loading curves. An alternative stability

' criterion on load is 1 x (NOP + Maximum Design Load) when the components of the NOP load as well as the NOP plus Maximum Design loads are combined by the absolute summation method. This criterion was approved by the NRC staff on another ALWR design. The System 80+ DCD will be changed to include this criterion as part of an acceptable alternative method for demonstrating LBB.

Related Changes to DCD DCD Section Insert # Revision 3.6.3.7 3.6.3.7 $ An acceptable alternative method for the margin on loads and margin on crack length e,mluations is to combine each component of the NOP q load and the Maxir. um Design Load absolutely. This method is referred to as 'the absolute summation of loads method'. If this alternative method is used, the margin on load for the leakage crack size is reduced from 42 to 1. The margin on crack length (2 time the leakage crack size) remains the same.

3.6.3.8, 3.6.3.8 Alternatively, cracks of the length that leak at the rate given above can (2nd bullet) withstand the absolute combination of normal operation load components and maximum design load with afactor of1.

App 3.9A, 3.9A.1 5 If the absolute summation of loads method is used to evaluate the i1.9.6.5.1 margin on load and margin on crack length, the PED for the ai analysis is constructed using the formulas Me = (NOP i + SSE i ) and M e

= SFi for the points labeled "1" and Me = (NOP2 + SSE2 ) and Me = SF2 for the points labeled "2" App 3.9A, 3.9A.2 $ If the absolute summation ofloads method is used. the PEDS will be 51.9.6.6 reconstructed, and the piping design, evaluation and reconciliation will be based on the reconstructed PEDS.

Figure 3.9A-12 3.9A.3

• When the absolute summation ofloads method is used to evaluate the margin on load and the margin on crack length. ' Max. Load /V2' .

becomes ' Max. Load'.

)

MeM 8 Systsm 80 + Design control Documrnt which is the largest of the dynamic loads (due to safe shutdown earthquake, thermal stratified flow, rapid valve closure, or other load) included in the crack stability analysis. The J-integral is determined for two different crack lengths for each geometric model. For the margin on loads evaluation, the J-integral for the leakage crack size is evaluated for J2 x (Pressure - NOP - Maximum Design) loads. For the margin on crack length evaluation, the J-integral for 2 times the leakage crack size is evaluated for (Pressure + NOP + Maximum Design) loads.

3.6.3.7 Stability Evaluation -

The stability of the cracked pipes is assessed by comparing the J integral value due to the applied loads on the pipe to the material crack resistance. The stability criterion for ductile crack extension employed is:

l if J-applied < J tc material, and (dJ/ja) ,ppw < (dJ/da) ,mne then crack stability is assured.

l The change in J-integral with crack length "a" is determined by analyzing several crack lengths in the region of interest. For a leakage crack of lengtn "a", erack tengths "a", a-6, and a+6 are analyzed.

Similarly, the change in J-integral with crack length in the region of length "2a" is determined by )

l analyzing cracks with lengths 2a,2a-6, and 2a+6. This method provides the derivative information in l l the two regions of interest. The variation of J with crack length in the region of "a" and "2a" is plotted '

along with the material curve. Evaluation of the plots allows for direct verification of the stability l criteria, I l l l The evaluations are performed for the locations chosen to envelop all limiting cases. The pipes with the l leakage crack lengtn subject to loads of J2 x (P - NOP - Maximum Design Load) and the pipes with l crack length 'twice the leakage crack length with loads of (P - NOP - Maximum Design Load) are demonstrated to have significant margin between the material curve and the loading curve, indicating that fgs;rgy 1 pipe locations satisfy the LBB crack stability criteria. f Lfaa f"?  ? )

, . See Appendix 3.9A, Sections 1.1.9.5.4 and 1.1.9.6 for a discussion of LBB design criteria development and a further discussion of analytical methods.

3.6.3.8 Results The piping listed in Section 3.6.3 and evaluated by the methods described above are shown to meet all the ((criteriaforapplication ofleak-before breaklf according to NUREG 1061, Volume 3. Specifically,

. these criteria (( require thefollowing:}f a Cracks which are assumed to grow through the pipe wall leak significantly while remaining stable. The amount of ((leakage is detectable with a safety margin of at least afactor of10]f unless otherwise justified.

((Cracks of the length that leak at the rate given above can withstand normal operation plus l maximum design load loads with a safetyfactor of at least 42 9 SEK7 .7. A 2

NRC Staff approval is required pnor to implementmg change m this information; see DCD Introduction Section 3.5.

Approved Design Morenal Design of SSC Pope 3.644

. S'TG M b System 80+ Design controloocument l

The LBB piping evaluation plot requires performing two complete LBB evaluations. The evaluations are for two NOP loads which span the typical loadings for tne line under consideration. A comoleted typical j diagram is shown in Figure 3.9A-28. The procedure used for generating that figure is as foiiows:

l 1. Choose NOP = Pressure + NOP i l

j 2. Determine ai

3. Increase the analysis moment until the critical moment is found for ai and 2a,

) 4 Separate the critical analysis moment, M., into the correct addition of SSE and NOPi proportion j for the ai and 2ai evaluations.

] M, = g (NOP +i SSE i) (ai Analysis) 2 M*

SSEi = - NOP i and

! d M = (NOP + SSE ) (2a Analysis)

SSEi = M - NOP

5. Plot SSE values at NOPi for the ai and 2ai analyses, respectively. This corresponds to the points labeled "1" in Figure 3.9A-28.

6. Repeat steps (1) to (5) for NOP:. The results are shown in Figure 3.9A-28, labeled "2" Two stability evaluations are performed for each pipeline under consideration in order to complete the piping evaluation diagram.

When stratified flow (SF) is a critical thermal transient that must be considered in the stability analyses, the PED is constructed using the following relationships for the ai and 2ai evaluations in (4) above:

4. M, = g SF i (ai Analysis)

SFi = M* and 5

M, = SF (2ai Analysis)

SF - M, As in the case for SSE, Step (4) is repeated to determine SF, for the ai and 2ai analyses, and SF is lotted vs. NOP.

\ 2.9.A.*I Approved Desogn Meterial. Design of SSC Page 3.DA42

1 TreM s Syst1m 80 + Drsign Control Docum:nt 1.9.6.5.2 Using an LBB Piping Evaluation Diagram )

l Once the lines marking the acceptable areas of allowable piping loads are plotted as described in the previous section. normal operating piping loads and corresponding maxtmum design loads for the critical piping locations are plotted on the evaluation diagram. The critical locations are selected as the highest j stressed point for each different type of material in the line. Figure 3.9A-29 shows how the plot is used i for a hypotheticalline. In this example, three points failed LBB and one point passed LBB. The reasons for each failure are given in the figure. The piping design can then be revised using the results (e.g.,

lowering the SSE response load by rerouting or by adding a snubber). Further review may result in other options for reducing the loads.

1.9.6.6 Results Piping Evaluation Diagrams (PEDS) for piping systems listed in Section 3.6.3 are shown in Figures 3.9A-30 to 3.9A-36 and provide LBB acceptance criteria for these piping systems. These criteria are based on piping design parameters given in Table 3.9A-2. Analyses of preliminary design of these piping systems have demonstrated that the LBB criteria are met.

Site-specific information will demonstrate that the final detailed design parameters of each piping system are consistent with those given in Table 3.9A-2 and that the final detailed design meets the LBB criteria of Figures 3.9A-30 to 3.9A 36. If design parameters for a piping system are not enveloped by those in  ;

Table 3.9A-2, a new PED for that piping system will be constructed using the methodology given in this i appendix and the piping design will be revised, as necessary, to meet the LBB criteria of the new PED.

If a PED given in Figures 3.9A-30 to 3.9A-36 is applicable to the detailed design of a piping system but the detailed design does not meet the LBB criteria of the PED, the design will be revised to meet the LBB criteria of the PED.

Reconciliation of the as-built piping systems with the final design will be documented in a LBB Evaluation Report. The LBB Evaluation Report shall contain results of the LBB evaluations for as-built I piping. The LBB evaluations shall employ methods described in Section 1.9 of this appendix.

Reconciliation of each as-built piping system qualified for LBB will be made by demonstrating that:

1. the as-built piping system meets the screening criteria of Section 3.6.3,

2. the dimensional and material properties of the as-built piping system are consistent with the parameters used in the development of the final LBB PED (s) for that piping system, and

3. the as-built piping responses meet the ASME Code allowables and the final LBB PED criteria.

See also Sections 3.6 and 3.6.3.8.

/NSBZ7 ,

i 3 9. A 2-l 1

~

l

^

Approved Design Metenof- Design of SSC Pope 3.9A 33

fTeM 8 System 80+ D?ssign Convol Docuwnt SELECT SELECT PlPE SELECT LEAK DETECTION PIPE SIZE SYSTEM CAPABILITY MATERIAL U

10 x LDS CAPA81LITY SELECT If LOW c SELECT

NOP LOADS a HIGH l NOP LOADS l

U Ir l

DETECTABLE i

LEAKAGE DETECTABLE CRACK LEAKAGE (DLC) CRACK 1

1 if U

1f C

2 x DLC UU ifif LOAD h LOAD CURVES J4 CURVES YU lf lf MATERIAL J . dJ/da -

J.a CURVES

~ -

J . dJ/da CURVES CURVES Y If 1f if MAX. LOAD /d (l) LBB PIPING MAXIMUM EVALUATION c LOAD DIAGRAM

NOP LOAD vs.

MAXIMUM DESIGN LDAD N

1.[/ SE2 Y}'-

l i

LBB Design Criteria Development Diagram Figure 3.9A-12 l

• l I

Approved Design Matenal Design of SSC Page 3,9A.61 l

System 80+ DCD - Potential Design Changes 8/Peveion c.u-m Item Number: 77 Summary

Description:

Eliminate Charging Pump Interlock (Add 160 gpm Flow Limit)

Affected DCD Sections: CDM: Section 2.7.16 ADM: Section 9.3.4 l

Change

Description:

Reason for Change: The current System 80+ CVCS design includes an interlock in the charging pump contrn'.s so that both charging pumps cannot be operated at the same time. The interlock was added to limit the maximum charging flow within the upper bound assumed in boron dilution analyses.

1 i During System 80t detailed engineering, a concern was identified related to charging pump I transfer from the operating to the standby pump. Normally, plant operations personnel start a l l

4 standby pump, then shut down the running pump. This method of pump switching ensures continuity of charging flow, especially seal injection flow to the reactor coolant pumps. It also prevents thermal transients on the charging and letdown subsystems. Although the dedicated seal injection pump (DSIP) could be started during charging pump switching, the DSIP does not have sufficient capacity to provide charging flow also. During the transition period without charging flow, letdown flow may be reduced or isolated because the charging stream is not available to I cool the letdown stream in the regenerative heat exchanger.

i

Description:

The interlock that prohibits the operation of both charging pumps would be deleted. In its place, a commitment will be added to Tier 1 that the CVCS will limit charging flow to the RCS to 160 gpm. This flow limit will ensure that the maximum flow assumption in the boron dilution analyses is preserved and that the design flow for CVCS charging line components, typically 200 gpm, is not exceeded. The flow limit will most likely be implemented  ;

by adding a flow indicator controller and an isolation valve, in the combined charging pump discharge, which will close when a high flow rate is measured.

4 DCD Markups Attached? Yes 4

i dedchgjr. doc 4 6/11/96

.ZVem 9 Pov f Syst*m 80 + D sign ContrelDrcum nt l

The letdown line is isolated by a safety injection actuation signal (SIAS). The RCP controlled bleedoff 1 line is isolated by a containment spray actuation signal (CSAS).

l

'.^.a - epcreting et e tinic.

7ntin:cr:ock  ;.;,rovided o is: no OMmore moi&r m onc um Cche$r-ing puni" ;b o rging Rou 9 fAe )

I CucS P Pin 9 cleCl9A hmds .

Inspections, Tests, Analyses, and Acceptance Criteria Table 2.7.16-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the l Chemical and Volume Control System.

l -

l l

l l

Certrhed Design Metenal Page 2.7-56

,T'/m 'l fet/ f Syst*m 80+ Disign Control Documrnt Table 2.7.16-1 Chemical and Volume Control System (Continued)

. 1 Design Commitment inspections Tests Analyses Acceptance Criteria l l

5. Valves with response 5. Testing of loss of motive 5. These valves change I I

positions indicated on power to these valves will position to the position Figure 2.7.16-1 change be performed. indicated on Figure position to that indicated 2.7.16-1 on loss of motive on the Figure upon loss of power.

motive power.

6.a) The letdown line is 6.a) Testing will be performed 6.a) The two CVCS letdown isolated by a safety using a signal simulating isolation valves inside  :

containment close upon l injection actuation signal an SIAS.

l (SIAS). receipt of a signal ,

~

simulating an SIAS.  ;

6.b) The RCP seal controlled 6 b) Testing will be performed 6.b) The RCP seal controlled l bleedoff line is isolated by using a signal simulating a bleedoffline isolation a containment spray CSAS. valves close upon receipt actuation signal (CSAS). of a signal simulating a CSAS.

g, O go '.: !:2:'r' :: pr~" A" 7. Testing will be performed 7. L . " J. .. . p .;

p  :: $2: re -^- '" aa- b ""'! ne' :' J' "ter 9 -

- % gj M{4( .;=pf n;; :rcmer',u-h l pt p a m m _ -- : . - m,- :,m, '

Mb 5 i$ ::kb pum; Ob#

(,9 {

{0[ ~ -- w blM LS ltg3 $hos OvQua) $Dlb0)0N f e') 8. Motor operated valves 8. Testing will be performed 8. Each MOV having an go 6 (MOVs) having an active to open, or close, or open active safety function

• safety function will open, and also close, MOVs opens, or closes, or opens or will close, or will open having an active safety and also closes, and also close, under function under differential pressure or preoperational differential fluid flow conditions and pressure or fluid flow under temperature conditions and under conditions. temperature conditions.

l 9. Check valves shown on 9. Testing will be performed 9. Each check valve shown Figure 2.7.161 will open, to open, or close, or open on Figure 2.7.16-1, or will close, or will open and also close, check opens, or closes, or opens and also close, under valves shown on Figure and also closes.

system pressure, fluid flow 2.7.16-1 under system conditions, or temperature preoperationalpressure, conditions. fluid flow conditions, or temperature conditions.

10. Flow limiting orifices are 10. Inspection of the as-built 10. Each letdown line flow provided in the letdown letdown orifices will be limiting orifice has a line. performed. cross-sectional area not greater than 0.01556 square feet.

s6 P4rahlR bo$45 champ punpr and '

ricans,ng the f/nu

+o a maxmm uMe c Ceti6ed Design Matwini gQ g [$ Page 2.7 59 a + W' Psig.

l Mem 9 Peu)

Syst?m 80+ D sign Contr-1Docum nt  ;

i t

l l

• All three tanks are located within a common dike structure designed to contain the maximum j '

) combined liquid mventory m the tanks. i 1

• The dike structure will be designed to comply with applicable state and local regulations. l l

9.3.4.2 System Description I 9.3.4.2.1 System 1

The normal reactor coolant flow path through the CVCS is indicated by the heavy lines on the flow diagrams (Figure 9.3.4-1, Sheets I through 4). Design parameters for the major components are shown in Table 9.3.4-4. Normal operating parameters for the CVCS are listed in Table 9.3.4-5. Process flow I l data is shown in Table 9.3.4-6. I i

Letdown flow from the RCS passes through the tube side of the regenerative heat exchanger where an initial temperature reduction takes place via heat transfer to cooler charging fhdd ,n the shell side of the heat exchanger. The regenerative heat exchanger is designed to cool lethwn tic,. , less thaa 450 F for all normal operations and to heat the charging flow by a minimum oi " 1 temperature reduction to the purification subsystem operating temperature is made by the icu exchanger.

The letdown heat exchanger is sized to cool inlet water from the maximum regeneratiw exchanger i outlet temperature to 120*F (or lower) for most operating conditions. Both the L n and the l regenerative heat exchangers are designed for full RCS pressure and both are located inside containment.

l Letdown fluid pressure is reduced from RCS pressure to the operating pressure of the purification subsystem in two stages. The first pressure reduction occurs at the letdown orifices and the second occurs at the letdown control valves located downstream of the orifices. The letdown orifices are l located inside containment. The letdown orifices are sized to pass the maximum letdown flow at full RCS I pressure with one control valve full open. The orifice provides the pressure reduction necessary to minimize erosion of the letdown control valve seating surfaces during normal RCS operations. A bypass valve around the orifices is provided for low pressure operations. The process flow is then filtered via l the purification filter purified via a purification ion exchanger, and sprayed into the VCT. An excess hydrogen inventory is maintained in the RCS by keeping a hydrogen overpressure on the VCT contents.

The charging pumps normally take suction from the VCT and discharfe tolhe RCS. Durine normal L, o erations, one char in pump is running and the other is in standb h =!cd it preWM e^ '"'l

- en "n e 7  ;; rr u ^ MM i dWJ a=dn60ne letdown an Gne charg1nfpurn w control valve are normally selected for use. Seal injection water is supplied I to the Reactor Coolant Pumps (RCPs) by diverting a portion of the charging flow just downstream of the charging pumps. This seal flow is then heated in the seal injection heat exchanger to approximately l 125*F before filtering. Once the flow has been filtered, the seal injection fluid is distributed to the four RCPs. The undiverted charging fluid is sent to the regenerative heat exchanger where it is heated before injection into the RCS.

A chemical addition tank and a chemical addition metering pump are used to transfer chemical additives j to the charging line downstream of the seal injection takeoff connection. Sufficient connections exist between the CVCS and the IRWST to allow for purification, inventory adjustments, and boron adjustments to the contents of that tank.

The CUCS Pifi 9 e lesty hmil1 ths. comke0 motimum cAesp'9W

-+c N Ros.

i Approved Design Matenet. AuxiUnry Systems (2/95) Page 9.3-29

l i

k System 80+

.Ih m 9 Pevl l

o-Wan cowd amm i

'^

Analysis ofinadvertent deboration events initiated during operationa; modes 2 through 6 (defined i

in the technical specifications) were performed. For Mode 1 operation, the reactivity addition 4

due to a boron dilution event is less limiting than the CEA withdrawal events. These analyses show that Mode 5 (cold shutdown) in the drained down configuration results in the shortest

available time for detection and termination of the event. Herefore, the initial conditions and

! analysis parameters are chosen for the cold shutdown operational nuxle to minimize the interval from initiation of dilution to the time at which criticality is reached. His results in the least j amount of time between detection and criticality.

The following are the analysis assumptions for Mode 5:

j

^

1. De Technical Specification lower limit on shutdown margin for cold shutdown is i

assumed,5.75% 4

2. The most adverse initial core condition would be for an initial Kg corresponding to 5.75% do suberitical and assuming suberiticality is ma:r ailed by boron concentration j only.

1

{ 3. The cold reactor coolant volume, including only the volumes for Mode 5 dramed, is 3

3,%1 ft . A conservatively low reactor coolant mass was assumed by using the cold )

j RCS internal volume. Assuming the coolant temperature of 210'F, the Technical '

.A Specification upper limit for cold shutdown, the resultin nwntom i

j

4. O= -5;ing p ; i: -----f = h: r--t : it -- ' -

.; , L'^ ,y .2 1; zir b !=ind f;;.c. i: .iCS. "=;r=, fr = .n. :--'--The analysis usedkcharging

1 map. flow rategof 160 gpm,which corresponds to 22.26 lbm/sec.

9NAO

! 5. The critical boron connaarration with all rods in except the largest worth rod stuck out and the inverse boron worth are 814 ppm and 66 pps/%4, myecdvely, including i uncertainties. De initial boron concentration for the cold s;mtdown mode is found by adding the product of the inverse boron worth and the mininmm shutdown margin (i.e.

5.75 percent) to the critical boron concentration. He resulting rninimum initial boron concentration in Mode 5 is 1193 ppm. Thus, the change of boron concentration from 5.75% 4 suberitical to critical is 379 ppm.

He parameters discussed above are summarized in Table 15.4.6-1.

e Results Using the above conservaive parameters in Eriaa (3), the mininmm possible time interval to dilute from 5.75% 4 subcritical to criticality is 67 minutes. Utilizing only the radandant, gnallflad neutron flux alarm, this time period will assure detection of a boron dilution event at least 30 minutes prior to criticality. Boron dilution will then be terminatad before loss of shutdown margin by the operator actions discussed in Section 15.4.6.2.

15.4.6.4 Conclusions O The inadvertent deboration event will result in acceptable consequences. Sufficient time is available for the operator to detect and to terminate an inadvertent deboration event if it occurs. Fuel integrity is not challenged during this event.

Appewear Dem> neeenneet. AneMont A& hoe fa.4-11

+6 - sL A + "g--- - 4 5- 4 I-l w 9 Revl SuffW nf 5SSYmqN97 -hl0 cbG System 80+ D sian control occament Table 15.4,6-1 Assumptions for the Inadvertent Deboration Analysis l

l  !

I l

Parameter Assumptions

)

Cold RCS Volumelll (mid-loop operation), ft8 3,961 l

l RCS Mass (mid loop operation), Ibm 237,185 1 i  !

Volumetric Charging Rate, gpm 160 h Mass Charging Rate, Ibm /sec 22.3 l Dilution Time Constant, r, sec'3 10650 1

Initial Boron Concentration - Co, ppm 1193 Critical Boron Concentration - C , ppm 814 l

O, 1

i i

1 9

[

l3 Includes the reactor vessel up to the mid-plane of the hot legs, half of a single hot leg, half of two cold

discharge legs and a shutdown cooling system.

W Dukn neeeenet AeaMant Aneken noe gg,4 2g

4 Jtem 10: Increased Damping for Response Spectrum Piping Analyses i

DCD Section: 3.7. Table 3.7-1. Figure 3.7-32. Appendix 3.9A Description of Change:

The NRC has approved a change in the maximum allowable value of applied damping for piping analyses in which the uniform envelope response spectrum analysis method is used. The maximum allowable damping in the System 80+ DCD for this type of analysis is currently based on ASME Code Case N-e11-1, which allows the damping to vary from 5% for modes of vibration up to and including 10 Hz to 2% for modes above 20 Hz. The NRC approved change l

allows 5% damping for all modes of vibration when the uniform envelope response spectrum analysis method is used on piping. This change is under review by the NRC staff on another

+

ALWR design.  !

l Damping values from Table 3.7-1, based on piping diameters 5:12 inches and >12 inches, continue to be used when piping is analyzed by time history or multiple support input methods.

l 2

Related Changes to DCD 4 ]

DCD Section Insert # Revision 3.7.1.3 Revise as marked i

Figure 3.7-32 Delete i

Appendix 3.9A Replace 11.4.3.2.1.3 with the following:

11.4.3.2.1.3

(( Damping values areprovidedin))' Section 3.7.1.3 and (( Table

3. 7-1.)) Table 3.7-1 gives an alternative damping of 5% for i piping analysis when the uniform envelope response spectrum t

method is used. No combination of piping damping criteria

given in Table 3.7-1 is used. Composite modal damping is used in piping analysis in accordancew' ith the procedure described in Section 3.7.2.15.

, rre u io i

Table 3.71 Add entry for 5% damping, revise Note [2]. Last 3 entries to Table 3.7.-l and Note [2] should read as follows:

1

(( Piping (diameter s 12 inches) 2.0 Piping (diameter > 12 inches) 3.0 Piping (umform envelope response spectrum 5.0M\') method ofanalysis) (21 Use of 5% damping for uniform envelope response spectrum analysis of piping is subject to the same  ! limitations as those for Code Case N-411-1 specified in Regulatory Guide 1.84. l l 1 l l 1 i i l l l l l I

             -                      --.-.       ~1 -.. - ~- - ... - .-                                        .             . - . . .                                .   -

1 T-rem ao System 80+ Drsion ControlDocument Chapter 3 Figures (Cont'd.) Page 3.7-6 Synthetic Time History H2 Spectra vs. Target Spectra for CMS 2 (1 and 2% Damping) 3.7-7

                                                                                                     .. . . ... ....                                     .      3.7-44 Synthetic Time History H2 Spectra vs. Target Spectra for CMS 2 (5 and 7% Damping) . . . .

3.7-8 . . . . .. .. ....... .. . 3.7-45 Synthetic Vertical Time History Spectra vs. Target Spectra for CMS 2 (1 and 2% Damping) . . . 3.7-9

                                                                                     . . . . . ..                  ....                  ....                   3.7-46 Synthetic Venical Time History Spectra vs. Target Spectra for CMS 2 (5 and 7% Damping) . . . . . . .                                                                                                           1 3.7-10
                                                                                                            .    . . ....                          .            3.7-47 Synthetic Time History H1 Spectra vs. Target Spectra for                                                                                              j CMS 3 (1,2,5 and 7% Damping) . .                                                                                                                      .

3.7-11

                                                                                 . .              . ..          . . . ..                           ...         3.7-48      1 Synthetic Time History H2 Spectra vs. Target Spectra for CMS 3 (1,2,5 and 7% Damping) .

3.7-12

                                                                                                                                        ...                    3.7-49 Synthetic Vertical Time History Spectra vs. Target Spectra for CMS 3 (1,2,5 and 7% Damping) . . .                                                                                                                 ,

3.7-13

                                                                                                 ...            ..                   ...                       3.7-50 Schematic Diagram of Interior Structure, Shield Building, FB,CVCS             ...............

3.7-14 . . . . . . . ..... ... ....... .. 3.7-51 ) Schematic Diagram of Interior Structure, Shield Building, DG - 1, DG -2 . . . . . . . . . . . . . . . . . . . . . . . . 3.7-15

                                                                                                      . ..        .............                                3.7-52 Schematic Diagram of Interior Structure, Shield Building, EFW1 (Horizontal), EFW2 (Horizontal)                              . .

3.7-16

                                                                                                      ..                       ......                          3.7-53 l                    Schematic Diagram of Interior Structure, Shield Building, EFW1 (Vertical), EFW2 (Vertical) . . . . . . .

3.7-17 . . . . . . . . . . . . . . . . 3.7-5 4 Schematic Diagram of Interior Structure, Shield Building, CAA, CAB , . 3.7-55 3.7-18 Finite Element Model of Steel Containment Vessel . . . ... 3.7-56 3.7-19 . . . Schematic of Combined NI Structures (Elevation Looking South) 3.7-57 3.7-20 .. . Schematic of Combined NI Structures (Elevation Looking West) . .. 3.7-58 l 3.7-21 Schematic Diagram of the SASS) Anal l Motions . . . . . . . . . . . . . . . . . . ..ysis Process Using CMS 2 and CMS 3 3.7-22 ................... 3.7-59 Schematic Diagram of the SASSI Analysis Process Using CMSI Motions 3.7-60 3.7-23 Reactor Coolant System Seismic Analysis Model 3.7-24

                                                                                              .          .                .           ...     ..              3.7-61 Pressurizer Seismic Analysis Model . . .

3.7-25

                                                                                                                 ...               . ..                       3.7-62 l

Typical Surge Line Seismic Analysis Model ! 3.7 26 3.7-63 Reactor Internals Horizontal Seismic Analysis Model 3.7-27 3.7-64 3.7-28 Reactor Internals Nonlinear Horizontal Seismic Model . 3.7-65 Core Seismic Model; One Row of 17 Fuel Assemblies 3.7-29 . .. . 3.7 66 Reactor Internals Linear Vertical Seismic Model 3.7-67 3.7-30 Reactor Internals Nonlinear Vertical Seismic Model . .. 3.7-68 3.7-31 .. Core-Support Barrel Upper Flange Finite-Element Model . . . . . 3.7-69 3.7-32 D=p.... t fa En

• n:Jp cf 9 ping . De/, rear. 3.7-70

3.7-33 Propertional Damping . . . . . . . . .. . . ...... ... .. ....

3.7-71

                                                                                                                                                                       /

3.8-1 Containment Details . . .

3.8-44 3.8-2 Category I Structures - Typical Feedwater Penetration 3.8-3 . . 3.8-47 ' Three-Dimensional ANSYS Containment Model 3.8-54 3.8-4 Axisymmetrk ANSYS Containment Model

                                                                                                                      . .                                    3.8-55 l 3.8-5         Nuclear Island Structures . . . . . . . . . . . .
                                                                                                           ...... .........                                 3.8-57 Approved Design Material Design of SSC I2/95)        Page vi l

i

1 rrm io l System 80 4 Dzstan control Document

    ~

3.7.1.2 Design Time History Since the System 80+ Standard Design is designed for generic site conditions, for the time history method of analysis, the generic free-field ground surface time histories are used as control motions in the analyses. In the soil-structure interaction analyses, for each generic site, the corresponding two horizontal and one vertical time histories at the free-field ground surface are used with the SSI model of that site. For the fixed-base analyses, the rock outcrop time histories are directly used as the control time histories. The response spectra at 2,5 and 7% damping of control motion CMS 1, and 1,2,5 and 7% damping of control motions CMS 2 and CMS 3 and the corresponding spectral ordinates of the matching time histories 1 are shown in Figures 3.7-1 to 3.7-12. The Power Spectral Densities of all time histories are included ) in Section 2.5. Each time history that is used in the SSI and rock analyses contains 20.48 seconds. For the SSI analyses. a time step of 0.005 seconds is used. For the Nuclear Island rock analyses, a time step of 0.0025 seconds is used. i

For Category I structures not on the Nuclear Island a time step of 0.005 seconds is used for both SSI and rock analyses. r <

qllgenafive rtgum&& S 3.7.1.3 Critical Damping Values un @ rrri e n'A' p;pmg anaty.cs. f8Sf## N# Damping values used for v s nuclear safety related structures systems and components are based upon Regulatory Guide 1.61 ord :" L L " "' ' lSx T: gar :. ::}. These values are expressed in percent of critical damping and are given in Table 3.7-1. " 9' q = = + ... ... J ..

s. .- a r.-

g. u_ .g ,y y _ g; g 3.y;,

i 3.7.1.4 Supporting Media for Seismic Category I Structures Category I structures are founded directly on rock or competent soil. For the Nuclear Island the foundation embedment depth for System 80+ standard plant is approximately 51 feet (Reference 7). The rock properties and the layering characteristics, including shear wave velocity, shear modulus, and density, are given in Section 2.5. The System 80+ Nuclear Island is designed for the range of soil conditions discussed in Section 2.5 and shown in Appendix 3.7B. 3.7.1.4.1 Soll Structure Interaction (SSI)

 ' Two different types of analysis methodologies are used for the seismic analyses for the Nuclear Island.

For the fixed-base cases, modal superposition time history analyses are performed using the three control motions (CMS 1, CMS 2, and CMS 3) corresponding to rock site conditions. When a structure is supported on soil, the SSI is taken into account by coupling the structural model with the soil medium. To accomplish this, the methodology of the computer program S ASSI(System for Analysis of Soil Structure Interaction, Reference 6) is used. Detailed metnodology and results of the SSI analysis for the Nuclear Island are presented in Appendix 3.7B. The methodology for the soil structure interaction for the non-Nuclear Island structures is presented in Appendix 3.7C. - Aporoved Design Morenal . Design of SSC Pope 3.7 2

zun <a System 80+ D*s/gn ControlDocument l 4

21. EPRI Report No. NP-6695, " Guidelines for Nuclear Plant Response to an Earthquake,"

December 1989.

22. EPRI Report No. TR-100082, " Standardization of Cumulative Absolute Velocity," December 1991.

Table 3.7-1 Damping Values d Stmeture Safe Shutdown Earthquakell! } (Percent of Critica0 l Welded steel structures 4.0 i Bolted steel structures 7.0

Prestressed concrete structures 5.0 l

Reinforced concrete structures 7.0 Equipment (steel assembly) 3.0 llPiping (diameter s;12 inches) 2. Piping (diameter >12 inches) 3. b)*in) (OIbrM &MWbf4 .j', 0 f l'effMISd. Jfte t'Ah1 MAS sf et P 3.sas) CYSC ef$$b A.qll)1

                                                                                            *>t1%~>s. eweA a <arysn.ra                  1 SfdChes> Ann f.ro r sb f              e)oiny Ill J.s J         bef k 'fMd J"A*hed              i Soil material dampings are provided in Section 2.5.2.             l  8/m,'7'adSwJ             45 Masc. dr-w-

r2) ""

                         . ; , ._        ;. . ;' M of ...:   . . ad, d_..., ...; . : _ .m, b had : -4Code Case N-411 1 1

f/sc$4M ",7 " " " d D Regulatory Guide 1.84. (31 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Approved Design Material- Design of SSC Page 3.7-38

                                                                                ',                   su M eo l Syst~m 80 +                                                                                            1 Drsign ControlDocumtnt        l l

l l3 bVe b dM' bis?(2 ))y [$ Ysy k i 6 = l I l l 5 ! I I I I 4 = l h4 l

            #          3    =

l l 2 = I e< l l Jd 1 l l l l 1 l l 1 l l 1 1 l 1 l l I i l I I ' l I I f l, 0 , e 0 10 20 100 FREQUENCY, HZ GENERAL NOTE: DAMPING INDEPENDENT OF PIPE DIAMETER. D_.g:..,, '.'2! ; r '- E;-o f_ g; - .. e p ,, , ,, gpgg i I Approved Design Material Design of SSC Page 3.7 70

                                                                                                      -                prem so Syst:m 80+                                                                              Drsign ControlDocumznt Option 4:

Acc,rding to this option, the site-specific response spectra for a 0.3 g Safe Shutdown Earthquake will be wd for the design of all piping and components. The site-specific analyses will be performed using , tb site soil conditions and properties and the design control motion (CMSI, CMS 2 or CMS 3 anchored j w a 0.3g peak ground acceleration) which envelops the site requirements and is appropriate for the site conditions. l Peak broadening or peak shifting methods are used. The method for peak shifting is detailed in ASME  ! Co6., Section III, Division 1, Appendix N, Section N-1226.3. When this method is used, the entire peak shifting process is performed Mfa J.7-/ independently gives n ah'emarWe. for each of the three di 1 . 3.2.1.3 Damping sf.6~% fr. ppinq 2ns/es-i.s den t/rq umkn.1 dnnhpt n fs p sx fe rP *' fr M

• M ';7 % d ' s " ' e d - l

(( Damping values areprovidedin))1 Section 3.7.1.3 and (( Table 3.7-1.)) 6 : 2n 2!!c-n2t:re to T b!; 2.'- Mering =!u;;, urkb'c &mping "Ses i" rred2":e ~!!h H *epire'"e"" '"d """t2tiens of the ASMC Cui Cex N U 1 1 are uxJ in p; ping unulym. No combination of **-me damping criteria is used. 7.e varkbk &:nping cur;; a p s 'i&d in Tip,e 3.7-E. Composite modal damping is use in gerj the analysis in accordance with the procedure described in Section 3.7.2.15. , , j ,

                                                                                                                                       \

1.4.3.2.1.4 Modal Cutoff and Rigid Range Acceleration Effects [y ven at7I4/e J.7-/ The number of modes included in the analysis is chosen to correspond with the range of seismic excitation frequencies up to the frequency corresponding ta the ZPA. At modal frequencies above the l frequency corresponding to the ZPA, pipe members are considered rigid. The acceleration associated { with these rigid modes is usually small. In certain situations ((the response ofhighfrequency modes can l significantly affect support loads, particularly axial restraints on long runs. To accountfor these efects, a missing mass correction is applied.))' One method used to calculate the missing mass correction is as follows: I For each degree of freedom in the direction of excitation, and for a steady-state 1-g acceleration, . I

1. determine the inertial force for each mode used in the analysis,

2. um the inertial forces in (1) over all modes used,

3. determine the total inertial force based on the total mass at the d.o.f.,

4. calculate the " missing" inertial force, (3)-(2).

Apply the resulting force system, multiplied by the ZPA, in a static analysis. Add the results of that analysis to the modal results using a method at least as conservative as SRSS. Repeat the above procedure for the other two excitation directions. 3 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Approved Design Material- Design of SSC Page .'J.9A 6

T4E A 1l System 80+ oesign controlDocument Table 3.6-3 liigh-Energy Lines Within Containment (Cont'd.) Line Operating Operating item Functional Pressure Tanperature No. System Description (>275 psig) p 100'F) Figure l 38 Reactor Coolant SG No. 2 RCS Loop Yes Yes 5.1.2-1 2B Drain Line to RDT (High-Energy to Isolation Valve RC-332) l 39 Reactor Coolant SG No. 2 RCS Hot Yes Yes 5.1.2-1 Leg Drain Line to RDT (High-Energy to Isoldion Valve RC-215) } 40 6 P $! :: Preswriz Relief

                                                              /

e[ Y 5% Lme #1 to ssunzer /% J SafetyN' ve -200 4 41 / Redr Coolad Prps'surizer Rellef Yes , Yes 5.1.2-3 Sht / ,/ / Ldne #2 to P F5% as ,, / / Safety Valye/ssurizer RC-201 /

                                                                               /

[

            /42      Reac[or Coolan[ Pressunter Relief                Yes              Yd       5.1.I3 g    ,             /          /          Line/3           to Pressurize [r f          /            Safety Valve RC-IO2
                                                                                                           /

43' Reactor Coolant j P' ressunzer hef Yes Yes 5.1.2-3

                                      / Line #4 topressurizer q                                Safety Valve RC-203 i

44 Reactor Coolant Pressunzer Spray Line Yes Yes 5.1.2-1 & l from Cold Leg Loop 5.1.2-3 l 1A to IA Spray Control Valve RC-100E 45 Reactor Coolant Pressurizer Spray Line Yes Yes 5.1.2-1 & from Cold Leg Loop 5.1.2-3 IB to IB Spray Control Valve RC-100F 46 Reactor Coolant Pressurizer Spray Line Yes Yes 5.1.2-3 from Loop 1A Spray Control Valve RC-100E to Pressurizer Spray e Common Header Approved Design Material- Design of SSC Page .1.6-42

[ T2 M ll System 80+ Design controlDocument 4 Table 3.6-3 High-Energy Lines Within Containment (Cont'd.) Line Operating Operating Item Functional Pressure Temperature No. System Description (>275 psig) ( > 200* F) Figure 57 Reactor Coolant Direct Vessel Injection Yes Yes 5.1.2-1 & Connection #4 to SIS 6.3.2-I C Interior Check Valve SI-217 , _ _ { 58 Safety 1_.b Off of 4 ~ ~ \ Yes Yes 5.1.2-3 Depressurizatio  ! Pressurizer Safasy System f '/L "O 2^: i_a l Line to RC-409

                                    -4 Rapid Depress. Line),

' / I ( 59 Safety - ^ ^ " ' b -. Yes Yes 5.1.2-3 l Depressurizatio Pressurizer Seby I j System Valve =RC403 4 team 4 12ne:4to RC-108 l , j '/ Rapid Depress. Line). l 60 CVCS Letdown Line from Yes Yes 9.3.4-1, Imp 2B to 5.1.2-1 Regenerative Hx 61 CVCS Letdown Line from Yes Yes 9.3.4-1 Regenerative Hx to Letdown Hx 62 CVCS I.etdown Line from Yes No 9.3.4-1 Letdown Hx to Containment Penetration 63 CVCS Charging Line from Yes No 9.3.4-1 Contamment Pen to Regenerative Hx 64 CVCS Charging Line from Yes Yes 9.3.4-1 Regenerative Hx to RCS Loop 2A 65 CVCS Auxiliary Spray Line Yes Yes 9.3.4-1, to Pressurizer Spray 5.1.2-3 Common Header

        ' 66    CVCS                  SCS Hx Shutdown                 Yes          Yes         9.3.4-1 Purification Lir.e Cont Pen Check Valve CH-304 to letdown O
   /ppmved Design Material Design of SSC                                                              Page 3 6-44

                                                                                                       .         pre a r>

i System 80+

                                                                    -{< <                                  Desian conteni n::-,,a,..YS s-Table 3.9-2        Loading Combinations,ASME Code Class                              1,2, a%d Cornportertt hvP 4

no a component {l ( i Condition Design Loadingl33 Combination Design PD + DW Level A (Normal)t21 pfow Level B (Upset)t21 pohow Level C (Emergency) POkDWhE f i ( Level D (Faulted) PC +DW + HSEDF Q j V /

                                                                                      \

0 4(QCr&, CUA 48 t i ( [1] legend: [ PD = Design pressure PO = Operating pressure DW = Dead weight i SSE = Safe Shutdown Earthquake DE = Dynamic system loadings associated with the emergency cond !on DF = Dynamic system loadings associated with pipe breaks (not elimmated by leak before break analysis) [2] As required by ASME Code Section III, other loads, such as thermal transient, and thermal gradient require consideration in addition to the primary stress producing loads listed. SSE is considered in equipment fatigue evaluations in accordance with Section 3.7.3.2. [3] For piping, see Tables 3.9-10 and 3.9-11. A

 .,w
                                                                                                         ~

Approved Design Material Design of SSC Pese 3.9-13

['TCM 13 System 80+ Design ControlDocument Table 4.2-3 C-E Poolside Fuel Inspection Program Summary (Reference 70) (Cont'd.) Shutdown Reactor Date/ Cycle Inspection Program Scope [1] Yankee Rowe 1987/18 VE,UT,SRE Millstone-2 1977/1 VE 1982/4 VE St. Lucie-2 1987/3 VE,UT 1989/4 VE,UT,SRE ANO-2 1981/1 VE, DM, SRE on C-E/EPRI Test Bundles 1982/2 VE, DM

 / -l                                     1983/3      VE, DM, SRE on C-E/EPRI and DOE Test Bundles
   -l                                     1985/4      VE, DM, SRE on DOE Test Bundles 1986/5      VE, DM, UT A ^l                   _~

g

                                      .-.lj8/g _ .g. .g_SRE on DOE Test Bundles San Onofre-2                      1984/1      VE, DM 1985/2      VE,DM 1987/3      VE, UT, GS, SRE 1989/4      VE, UT, SRE, DM San Onofre-3                      1985/1      VE,UT 1988/3      VE,UT,SRE 1

Palo Verde-1 1987/1 VE, DM 1989/2 VE, DM Palo Verde-2 1988/l VE, DM Waterford-3 1988/1-2 VE,UT,SRE l tu VE Visual Erammation GS Gamma-Scanmng CS Crud Sampling S Sipping UT Ultrasonic Testing SRE Disassembly and Single Rod Exammations DM Dimensional Measurements

                                                                                     / '

Approved Design Meterial- Reector Page 4.2 76 l l

System 80 + Design ControlDocument Table 5.4.7-2 Shutdown Cooling System Failure Modes and Effects Analysis Symptoms and Local Effects Inherent Including Dependent Compensating Provision Remarks and No. Name Failure Mode Cause Failures Method of Detection Other Effects l} Shutdown Cooling Fails Closed Corrosion, Effective loss of one Low flow indication Parallel redundant Valve is normally Pump Suction mechanical shutdown cooling train F-302, F-305; periodic shutdown cooling path locked open Isolation Valve binding, cooling testing SI-106, SI-107 operator error Fails Open Same as la) No effect on SCS operation Periodic testing Nonc required

2) Shutdown Cooling Fails to start Mechanical Effective loss of one SCS Low flow indication Parallel redundant Pump 1,2 failure, train F-302, F-305; periodic shutdown cooling path electrical testing failure

( 3) Shutdown Cooling [ ails Open Corrosion. Effective loss of one Low flow indication Parallel redundant Valves are kxked i Pump Discharge < Clevd mechanical shutdown cooling pump F-302, F-305; periodic shutdown cooling path open; min. flow line Isolation Valve 7 binding, testing will provide the min. SI-578, SI-579 operator error flow required to protect the pump Fails Open Same as 3a) No effect on SCS operation Periodic testing None required

4) Shutdown Cooling Loss of Cooling insufficient Dimir.ished ability of Iligh temperature Parallel redundant IIcat Exchanger 1,2 component subsystem to provide RCS indication from T-302, shutdown cooling path cooling water her t removal T-305 flow, excessive fouling

5) Shutdown Cooling Fails Closed Corrosion, Delays use of affected SCS Valve position Parallel redundant Same as 3a) llent Exchanger mechanical train indicator; periodic shutdown cooling path Bypass Valve binding, testing SI-312, SI-313 electrical failure Q

                                                                                                                                                                               ,.T

! Fails Open Mechanical Effective loss of one Valve position Parallel redundant l failure, shutdown cooling path indicator; periodic shutdown cooling path electrical testing  % failure Approved Design Metodel- RCS and Connected Systems Page 5.4-40 0 -- - -- 0 - - - - - - - - - _ - 0

h6 M [6 System 80+ Design controlDocument The boron recovery portion of the CVCS accepts letdown flow diverted from the VCT as a result of feed and bleed operations for shutdowns, startups, and boron dilution over core life. The diverted letdown l flow, which has passed through a purification filter t.ul ion exchanger, also passes through the pre-holdup ion exchanger. The pre-holdup ion exchanger retams cesium, lithium, and other ionic radionuclides with high efficiency. The process flow then passes throur,n the gas stripper, where hydrogen and fission gases are removed with high efficiency; thus (1) precluding the buildup of explosive gas mixtures in the holdup tank and (2) minunizing the release of radioactive fission product gases in aerated vents or liquid discharges. The degassed liquid is automatically pumped from the gas stripper to the holdup tank. Reactor coolant quality water from valve and equipment leakoffs, drains, and reliefs within the containment is collected in the Reactor Drain Tank (RDT) and scheduled for batch processing. Recoverable reactor coolant quality water outside the containment from various equipment and valve leakoffs, reliefs, and drains is collected in the Equipment Drain Tank (EDT) and scheduled for batch processing. Reactor coolant collected in either of these tanks is periodically discharged by the reactor drain pumps through the reactor drain filter and pre-holdup ion exchanger, and processed in the same I manner as diverted VCT flow, as described above. This liquid is also pumped to the holdup tank. When a sufficient volume accumulates in the holdup tank, it is pumped by a holdup pump to the boric acid concentrator, where the bottoms are concentrated to within the range of 4000 to 4400 ppm boron. The boric acid concentrator bottoms are continuously monitored for proper boron concentration, and l normally pumped directly to the BAST. In the event that abnormal quantities of radionuclides are I present, the bottoms are discharged to the LWMS. The boric acid concentrator distillate passes through  ! a boric acid condensate ion exchanger, where boric acid carryover is removed. The distillate is collected in the RMWT for reuse in the plant. If recycle is not desired, the distillate is diverted to the LWMS. When the SCS is operational, a flow path through the CVCS can be established for purification. This is accomplished by diverting a portion of the flow from the shutdown cooling heat exchanger to the letdown line upstream of the letdown heat exchanger. The flow then passes through the purification j filter, purification ion exchanger, and" letdown strainerg is remmed-to-the scien of ee shutdown

 -cooling-pumps.        -

• b* N YCC The When continuous degasification of the RCS is desire < gg g gg 4 om the inlet of the VCT to the gas stripper, bypassing the pre-holdup io is processed in the gas stripper and is then returned to the VCT via the -Nu_ Rcs Q the ed, VCT hydrogen overpressure can be used to replace the hydrogen ,chagi p p.. ing process. The charging pumps take suction from the VCT, and rett :S.

A makeup subsystem of the CVCS providrs for chanwa su nw outuu concemrauon. Boron is initially added to the CVCS using the boric acid hatching tank (BABT). Reactor makeup water is added to the BABT via the makeup supply headar, anc the fluid is heated by immersion heaters. Boric acid powder is added to the heated fluid while a mixer agitates the fluid. A boric acid concentration of as high as 12 weight percent can be prepared. Electric immersion heaters maintain the temperature of the solution in the boric acid batching tank high enough to preclude precipitation. The concentrated boric acid solution in the BABT is drawn into the boric acid batching eductor and diluted by fluid being circulated from the BAST via the boric acid makeup pumps. The reactor makeup water pumps can also be used by taking suction from the reactor makeup water tank and pumping the water through the eductor to the BAST. 8 Approved Desigrs Material Auniary Systems Page 9.3 30

Red i69 System 80 + oestan control Document The resulting concentration of the refueling pool and the RCS is between the lower operating boron concentration limitation of the IRWST (4000 ppm) and the maximum operating boron

                                                                                               ~

concentration of the IRWST (4400 ppm). Thus, the contents of the refueling pool can be returned directly to the IRWST prior to plant startup without hindering plant operations. ! During refueling shutdowns, the reactor makeup water supply piping is continuously monitored l via flow switch F-250. An alarm is annunciated if flow is detected in order to prevent dilution of the refueling pool. 9.3.4.3 Design Evaluation l 1 i 9.3.4.3.1 Availability and Reliability i A high degree of functional reliability is assured by providing standby components and by assuring fail-safe responses for the most probable modes of failure. l Redundancy is provided as follows: l 1 Comnonent Redimdancy I Purification and Deborating lon E: changers Three identical components Charging Pumps One operating, one in standby I j .. Charging Pump Flow Control Valves One operating and one parallel, standby valve Letdown Control Valves One operating and one parallel, standby valve Boric Acid Makeup Pumps Two identical pumps in parallel, one operates on demand, one in standby Gas Stripper Package ne gas stripper package includes redundant standby pumps Altars Seal Injection Filters Two identical pwnps in parallel, m- - one ;w@) on4emand, one in standby . Purification Filters RLtus Two identical pumps in parallel, one sh M emismand, one in standby Reactor Makeup Water Pumps Two identical pumps in parallel, one operates

      ,                                                            on demand, one in standby                                          I Boric Acid Concentrator                                  De concentrator package includes redundant standby pumps in addition to component redundancy, it is possible to operate the CVCS in a manner such that some l        components are bypassed. It is possible to transfer boric acid to the charging pump suction header by bypassing the VCT. He letdown filter, and the purification and deborating ion exchangers can be bypassed. Controlled bleedoff flow can be routed to the RDT rather than the VCT.

Independent and redundant gravity feed lines from the BAST to the charging pump suction are provided to assure makeup. He charging pumps also have an alternate source of borated water from the spent fuel pool, which is maintained above 4000 ppm boron. l l Anment Dennen nennennt- Amafnery speanne Pare Aw

[Tc M h System 80 + Design Control Document l Table 9.3.4-4 Principal Component Data Summary (Cont'd.) l Reactor Makeup Water Pumps Quantity 2 Type Centrifugal

Design pressure 200 psig Design temperature 200'F Rated head 300 ft.

Normal flow 180 gpm Normal operating temperature 40-120'F NPSH required 15 ft Material in contact with pumped fluid Austemtic stainless steel Fluid Demmeralized water Code None Holdup Pumps l Quantity 2 Type Centnfugal l Design pressure 100 psig l Design temperature 200*F Rated head 145 ft Normal flow 50 gpm Normal operating temperature 40-120'F l NPSH required 10 ft I Materials in contact with pumped fluid Austenitic stainless steel

Fluid 2.5 wt % boric acid, manmum Code None Reactor Drain Pumps 4

Quantity 2 Type Centnfugal Design pressure 200 psig Design temperature 200*F Rated head 145 ft Normal flow 50 gpm , Normal operating temperature 120*F NPSH required 10 ft Materials in contact with pumped fluid Austenitic stainless steel Fluid 2.5 wt % bonc acid, maximum Code for fluid end ASME Ill, Class 3 , Volume Control Tank Quantity 1 Type Vertical, cylindrical Internal volume 5,800 gallons (approx) Design pressure, internal 75 psig Design pressure, external 15 psig Normal operating temperature 120*F G l Normal operating pressure 209srf $ 2C-Go PWs 1

                                                                     'L_ s

_2 Approved Design Matenal- Auni5ery Systems Page 9.3-74

l f4c.& \ f.) System 80 + oesian controlDocument i s Adequate clearances shall be provided for inservice inspection of the ASME Boiler and Pressure Vessel Code Section III, Class 2 portions of the main steam system piping, in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

• Loop seals are not utilized in safety valve inlets.

10.3.2.3.2 Valves 10.3.2.3.2.1 Main Steam Isolation Valves (MSIVs) and MSIV Bypass Valves The valves are designed so that no damage due to excessive closure force is incurred during closure under design conditions. Backseating of valve stems is provided when the valve is in the full open position. Unrecovered pressure loss from valve inlet to valve outlet at rated flow with the valve full open does not exceed 3 psid. l The Main Steam Isolation Valve (MSIV) in each main steam line is remotely operated and is capable of maintaining tight shutoff under the main steam line pressure, temperature and flow resulting from the transient conditions associated with a pipe break in either direction of the valves.

• The MSIV leak rate through a closed MSIV flow does not exceed 0.001% of nominal flow at P

1200 psia in the forward direction and does not exceed 0.1 % of nominal flow at 1200 psia in the reverse direction.

• The full open to close stroke time of the MSIV's under fully developed steam line break flow is 5 seconds or less upon receipt of a Main Steam Isolation Signal (MSIS). The full open to close stroke time of the MSIV's Bypass Valves is gseconds or less upon receipt of a MSIS. I 6
• The MSIVs are supported such that the valve body and actuator will not be distorted to such a degree that the valve cannot close or be displaced as a result of pipe break thrust loadings.
• The MSIVs and the MSIV bypass valves are designed, fabricated and installed such that the requirements for In-service Testing and Inspection of ASME Section XI, Subsection IWV can be met.
• The provisions of General Design Criteria 57 for containment isolation valves are met.
• The MSIV is a fail close valve; upon receipt of a Main Steam Isolation Signal the MSIV closes automatically.
• The MSIV bypass valve is a fail-close valve; upon receipt of a Main Steam Isolation Signal the MSIV bypass valve closes automatically.
• The MSIVs and their supports and the MSIV bypass valves and their supports are designed to

!I withstand loads arising from the various operating and design bases events as specified in Section

3.9.3.

Approved Design Material Steam and Power Conversion Page 10.3 7

System 80 + Dssign ControlDocument

• To permit testing for pH and the existence of foreign substances, sample connections are provided in the steam line piping between the steam generator nozzles and equalization header.
• During initial startup and during periods of unit shutdown, the tripping mechanisms for the main steam isolation valves are tested for proper operation in accordance with the technical specifications. The valves are periodically in-service tested for leakage and freedom of movement during plant operation in accordance with ASME Code Section XI, Subsection IWV.

The main steam safety valves are tested during initial startup or during shutdown operation by checking the actual lift and closing pressures of the valves in comparison to the required design opening and closing pressures in accordance with ASME Code, Section XI, Subsection IWV.

• ASME Code Section XI, Subsection IWV requirements for in-service testing and inspection of nuclear safety-related valves apply to the atmospheric dump and atmospheric dump isolation valves.
• A test will be conducted to verify MSIV response to a simulated Main Steam Isolation Signal (MSIS).

1. The objective of the test is to verify the function of the MSIVs and to confirm the 5  ;

second closing time required by Section 10.3.2.3.2.1.

2. Since steam pressure is normally required to operate the MSIVs, a supply of steam at O conditions comparable to main steam is a prerequisite, in addition to the completion of V construction activities on the MSIVs and required support systems.

3. The test method consists of the application of a simulated MSIS to the controls of the MSIV under test, the recolding of temperature and pressure parameters upstneh and downstream of the valve seat, and the timing of the closure process from the receipt of signal to the instance of valve closure as indicated by the valve stem travel indicator.

4. Acceptance criteria are that the MSIV operate to shut in 5 seconds or less, in accordance with Section 10.3.2.3. The test must be related by calculation and manufacturer's shop or type testing to the design basis conditions required by the safety function.

• A test will be conducted to verify MSIV bypass valve response to a simulated Main Steam Isolation Signal (MSIS):

1. The objectiveAfthe test is to verify the function of the MSIV bypass valves and to confirm the)03econd closing time required by Section 10.3.2.3.2.1. I

                                   ~5

2. Construction activities on the MSIV bypass valves and their required support systems must be complete as a prerequisite.

3. The test method consists of the application of a simulated MSIS to the controls of the MSIV bypass valve under test, the recording of temperature and pressure parameters upstr" and downstream of the valve seat, and the timing of the closure process from j rece ' signal to the instance of valve closure as indicated by the valve stem travel 1 indic .t.

j AMwoved Design Matenal Steam and Power Conversion Page 10.313

X i G6)' \ System 80+ Design Control Document i l 4. Acceptance criteria are that the MSIV bypass valves operate to shut in idTeconds or less, l in accordance with Section 10.3.2.3. The test must be related bj' calculation and I manufacturer's shop or type testing to the design basis conditions required by the safety function. 10.3.5 Secondary Water Chemistry ' 10.3.5.1 Chemistry Control Basis l Steam generator secondary side water chemistry control is accomplished by:  ; l

• Close control of the feedwater to limit the amount ofimpurities which can be introduced into the steam generator.
• Continuous blowdown of the steam generator to reduce the concentrating effects of the steam i generator.
• Chemical addition to establish and maintain an environment which minimizes system corrosion.
• Pre-operational cleamng of the feedwater system. i e Minimizing feedwater oxygen content prior to entry into the steam generator.

Secondary water chemistry is based on the zero solids treatment method. This method employs the use g of volatile additives to maintain systen pH and to scavenge dissolved oxygen which may be present in the feedwater. W  ! 1 A neutralizing amine is added to establish and mahtain alkaline conditions in the feedtrain. Neutralizing amines which can be used for pH control are ammonia and morpholine. Ammonia should be used in plants employing condensate polishing to avoid resin fouling. Although the amines are volatile and will not concentrate in the steam generator, they will reach an equilibrium level which will establish an alkaline condition in the steam generator. . 1 l Hydrazine is added to scavenge dissolved oxygen which may be present in the feedwater. Hydrazine also tends to promote the formation of a protective oxide layer on metal surfaces by keeping these layers in a reduced chemical state. Both the pH agent and hydrazine can be injected continuously at the discharge headers of the condensate pumps or condensate demineralizer, if installed. These chemicals are added as necessary for chemistry control, and can also be added to the upper steam generator feed line when necessary. Operating chemistry limits for secondary-side steam generator water, feedwater and condensate as extracted from Reference 2 EPRI Report NP-6239, are given in Tables 10.3.5-1,10.3.5-2 and 10.3.5-3. The ihnits stated are divided into three groups: normal, abnormal and immediate shutdown. The limits provide high quality chemistry control and yet permit operating flexibility. The normal chemistry conditions can be maintained by any plant operating with little or no condenser leakage. The abnormal steam generator limits are suggested to permit operations with minor system fault conditions until the affected component can be isolated and/or repaired. The immediate shutdown limits represent chemistry conditions at which continued operation could result in severe steam generator corrosion damage. Approved Design Material- Steam and Power Conversion Page 10.314

i l System 80+ DCD - Potential Design Changes Item Number: 19 Summary

Description:

Revision to Section 19.7 Affected DCD Sections: CDM: None ADM: Sections 19.7.1 through 19.7.5.3; ADM: Tables 19.7.5.1-1 through 19.7.5.4-7 Change

Description:

See Attached. l I l I dedchng4. doc 6/26/96 l

i ! _TrM ta) t l Sections 19.7.1 through 19.7.5 Add sentence to each section which identifies appropriate section where insights from System 80+ analyses are provided, e.g. " Insights from the external event analysis are given in Section 19.15.3." Section 19.7.5.1 Add sentences to refer to retained tables and that HCLPF values used in System 80+ seismic margins assessment provide guideance to the COL applicant for future evaluations.

                          " Tables 19.7.5.1-1 and 19.7.5.1-2 present HCLPF component and structural fragilities expressed in terms of peak ground acceleration. The HCLPF values used in the System 80+ seismic margin assessment provide guidance to the COL applicant for evaluating potential introduction of vulnerabilities in the as-built design (see Table 19.15-1)."

l Section 19.7.5.3 Add sentence to refer to Table 19.7.5.3-1 "HCLPF values for System 80+, sorted by sequence in ascending order, are provided in Table 19.7.5.3-1. " Reword the second paragraph. The inferred COL action is a carryover from the Draft i Safety Evaluation Report. The identified action was completed as a part of the Seismic Margin Analysis performed for System 80+ certification and no additional COL action is required. In fact, if the COL applicant's site conditions fall within the envelope of site l parameters used for System 80+, no additional seismic analysis is required by the COL applicant.

                         "The internal events model included as part of the Seismic Margins Assessment            4 was updated to include structural and passive failures. This completes COL action        l item 19-2 ."

Table 19.7.5.1-1 Delete column providing random (independent) component failure and human error failure probabilities; subsequently, delete all line items which do not provide component HCLPF values for seismic failures. Retained component HCLPF values for seismic failure provides insight to the COL applicant as to the basis for the plant level seismic HCLPF. Table 19.7.5.1-2 l l Retain - Table provides seismic fragilities for System 80+ structures and NSSS

rre a um components which provide COL applicant with insight as to basis for the plant level seismic HCLPF. Table 19.7.5.3-1 Delete column providing random and human error failure probailities and column which identify dominant cutsets. Remainder of table provides summary of seismic margin results. Table 19.7.5.3-2 Delete - Sensitivity analysis results from varying HCLPF of a few selected components which provide expected results; no significant insights gained. Tables 19.7.5.4 1 through 19.7.5.4-7 Delete - Purpose is only to demonstrate that the seismic margin results for a rock site are bounding for the System 80+ design. l l l i l l l l i

Trex ie) eev ot System 80+ oesign controloocument l l 19.7 External Events Analysis 19.7.1 Qualitative External Event Evaluation This Section Intentionally Blank. 3Ms h 5 b

• eMerna ( < ve of an*(pis an y v'e o M Se ck e n a. N 5 19.7.2 Tornado Strike Analysis

This Section !ntentionally Blank. rqds bm % korw2 J. A A ke a ns(gd s ars give s l nw Secken \ e), g g , -=3. g ,

19.7.3 Internal Fire Analysis l 1 This Section Intentionally Blank. Imph N ue 'ml< co I QM. a n a 1 3sis we qEen sa sech on se. ts. 3. z . 19.7.4 Internal Flood Analysis This Section Intentionally Blank. 2nighh b * ' N '""d M ""'IU5h *

• 6' TV\ $ecYe n % s5*heS.

19.7.5 Seismic Margins Assessment Inegw+s .N. h .4uinwA vn4p4 assess ment am gee n in Seck w is.i s.3 4 . 19.7.5.1 Methodology l ((The COL must confirm the use of seismically robust electromechanical relays in the engineered safety

                                                                                        'Ta ble s ia; L5.i-t 4na t *> o 5. i          features preunt actuation act.fv u and
                                  ~s Pocontrol   a sgt4ms    (COL  { tem 9fv-10)))]'.

an6 4 ruc r4 frg;t&ss ex pesssea :n terms .O 19.7.5.2 Seistrac Event Sequences Y' D "# v4Wes u+e d m -% e 5sh

                                                                                               " " d " E *** * ##

N 2 5d e- Wre t 3 l AT ane% ment provic\e e i -te t%1c. C OL A PP icm This Section Intentiot ally Blank. p ey.,y k g 7.,4,d.au'd cu ;4rduc4w e son of wfnerab'ilib in w as.ia.it d e4cy pee n ble m . is- O. 19.7.5.3 Seismic. Margins Results and Insights ((The COL applicant will be required to verify that key assumptions for structures, systems and  ! components considered in the Seismic Margin Assessment (SMA) are valid for the as-built plant conditions [ COL Item 19-4]. This will include evaluation of High Confidence of Low Probability of Failure (HCLPF) values for structures which house non-safety related equipment relied upon in the SMA evaluations such as the combustion gas turbine. The verification process will include a seismic walkdown, including development of detailed procedures [ COL ltem 19-4] to ensure that as-built conditions conform to the assumptions used in the SMA and to assure that proper anchorage for equipment has been provided and that the potential for seismic spatial system interaction does not exist.))' incW,4 as As 1e 3;mmm Jy.. m M...; .; der , i;;; .; ; n;;d :c ;;ga: The internal events modelje $: . l

                                                                                                                                             ='::yn per:S!:, by rp!!d :rdrder cf e rerd -d r9 pr 9 efd!r r *"> re r drded f cr de!::r d 7._.-      _799       n.3 9 r ru i. _ in p i, p er ge _ggg g . d                               ' fr the Seismic Margins            l 955c% meat ~          :rd re--              ede' was updated to include structural and passive failures. This completes COL action item 19-2.'

SCLPF Mt* 9"- Sq$b Ws 4" 4c p act- 'n ' 3'5 ce n ordtr3 am. p .,4,a x m m.v.s.3-s I l

        '             COL information item; see DCD Introduction Sectior. 3.2.

Approved Design Material- ProbabiEstic Risk Assessment (kV95) Page 1921

rres ie; cese oe

                                                                                                                 /

System 80 + Desian controt' Document Table 19.7.5.1-1 Components in Seismic Fault Tree Models , I NAME l DESC l PROB l HCLPF Initiator and Special Events l

     -EQ.MTC-                         miwm Moda.-:ct T. g.as CmfLicm tmiC)                    1.cou-02                             C EQ-PSV                        Primary Safety Valve (PSV) Fails to Resent                      l                     1.35 EQ-PSVO                       Primary Safety Valves (PSVs) Fail to Open                       l                     1.35 EQA'DVS                       Seismically Induced ATWS                                        l                    0.74 EQLOSP                        Seismically Induced Loss of Site Power                          l                    0.12 EQMLOCA                       Seismically Induced Medium LOCA                                 l                    2.59 EQRVR                         Seismically Induced LOCA in Excess of ECCS                                           0.73 Capacity                                          ,

EQSLOCA Seismically Induced Small LOCA l 0.90 EQTRANS Seismically Induced Transient l 0.05 i Emergency Feedwater System (EFWS) l l ABDZEFWPBRKR Seismically Induced Failure of EFW Motor Pump 0.95 1 Circuit Bruker AHFFASCSLOCA Operator-Fails to-Perform-Aggressive Secondary----- 6.40E - l _ Cooldown for Small LOCA l AKPZEFWSCABLE Seismically Induced Failure of EFWS Power Cable l 1.80 ! Motor-Driven Emesescv Pman EFWP-102 Faile in 1 ons 01 APAJFFWP W SM f A?AJEFWP-104 Motor-Driven Emergency Pump EFWP-104 Fails to 3.00E-03 Start j WAKEFWP-102 Motor-Driven Emergency Pump EFWP-102 Fails to 3.60E-03 Ru l j APAKEFWP-104 Motor-Driven Emergency Pump EFWP-104 Fails to 3.6CE-03 i Ru APAVEFWP-102 Subtrain IB Unavailable Due to Maintenance 2.00E-03 APAVEFWP-104 Subtrain 2B Unavailable Due to Maintenance 2.00E-03 APTJEFWP-101 EFW Turbine Dnven Pump EFWP-101 Fails to Start 1.50E-02 APTJEFWP-103 EFW Turbine Driven Pump EFWP-103 Fails to Start 1.5C E-02 APTKEFWP-101 EFW Turbine Driven Pump EFWP-101 Fails to 7.2CE-03 l Operate APTKEFWP-103 EFW Turbine Driven Pump EFWP-103 Fails to 7.20 E-03 Operate . APTVEFWP-101 Subtrain 1A Unavailable Due to Maintenance 5.003-03  ! APTVEFWP-103 Subtrain 2A Unavailable Due to Maintenance 5.00E-03 APTXDP101-103- -Gwarana N > D~n=ad Failure of EFW Turbine _ME-03 , /

 -                                     pump, APTZEFWP                      Seismically Induced Failure of Turbine-Driven EFW                                     2.38 Pump AQPZEFWSPIPE                  Seismically Induced Failure of EFWS Piping                                            0.90 AQWZEFWDGROOM                 Seismically Induced Failure of Wall Separating                                        0.84 EFWS/DG Room

( Anuered Design Material Prebebilstic flish Assessment Pepe 19.7 3

TTEM L ") suc+ o\ System 80+ Design ControlDocument Table 19.7.5.1-1 Cornponents in Seismic Fault Tree Models (Cont'd4 i NAME DESC PROB HCLPF AVCZEF-214 Seismically Induced Failure of EFWS Check Valve i 2.38 EF-214 l gDAFF-LOR Pneum. Diaphragm Operated Valve EF-108 Failsjo_._ __2.96E-03_ -

                                                                                                                                                                     ~

Open i I AVDAEF-109 Pneum. Diaphragm Operated Valve EF-109 Fails to 2.96E-03 Open i AVDAEF-112 Pneum. Diaphragm Operated Valve EF-112 Fails to 2.96E-03 Open f AVDAEF-ll3 Pneum. Diaphrar.n Operated Valve EF-113 Fails to 2.96E-03 Open i l AVMAEF-102 MOV EF-102 Fails to Open 4.00E-03

                           -AVMAEF-103                     kiGV EF-103-Faus to open                                                4.00E-03 AVMZPDSCHGVLVS                 Seismically Induced Failure of EFWP Discharge M-O                                               1.35 Valves AVNZEF-215                     Seismically Induced Failure of EFWS Manual Valve                              j                 2.38 EF-215                                                                       I
                             ^YS^.S." 100 DC Motor valve trPriOO-F8 *a rhwa                                       4.00E-01                      -

m I AVSAEF-101 DC Motor Valve EF-101 Fails to Open 4.00E-03 l AVSDEF-104 DC Motor Valve EF-104 Transfers Closed 1.09E-02

                                                                                                                                            ~

_AVSDEF-105- DC-Motor-Valve-EF:10STiansfers Closed 1.09E-02 MTXZCST Seismically Induced Failure of Condensate Storage 1 0.56 Tank (CST) Component Cooling Water System (CCWS)/ Station Service Water System (SSWS) i CBDZCCWPBRKR Seismically Induced Failure of CCWS Pump Circuit 0.95 Breaker CBDZSSWPBRKR Seismically Induced Failure of SSWS Pump Circuit 0.95 Breaker

                  ^

CHFFCC102-103 Onere'ne Fails to Close Non-E==aa'i=1 Cc = ===+ ESE-03_

                                                         , CCW isolation Valves                                                                                        }     '

CHFFCC202-203 l Operator Fails to Close Non-Essential Component 3.15E-03 CCW Isolation Valves CHFFCCWPIB Operator Fails to Start CCWP IB 1.20E-03 t CHFFCCWP2B Operator Fails to Start CCWP 2B 1.20E-03 g CHFFISOLATECCWS Operator Fails to Isolate CCWS From Spent Fuel 5.60E-02 kb Pool Heat Exchanger CHFFSFPHXI Operator Fails to Isolate SFP HXI 4.5DE-03 CHFFSFPHX2 Operator Fails to Isolate SFP HX2 4.5DE-03

CHFFSSWPIB Operator Fails to Start SSWP IB 1.20E-03 l

} CHFFSSWP2B Operator Fails to Start SSWP 2B 1.10E-03 CHFFSTBHX1B Operator Fails to Open CCW HX IB Isolation Valves 3 75E-03 i CHFFSTBHX2B - Operator-Fail. m um dw itx 2B Isolation Valves 7.T5E-03 i Approved Design Afsterial Probabinstic flish Assessment Page 19.7 4

4 ! I re % i9 rc4N oI z i System 80+ Design ControlDocument Table 19.7.5.1-1 Cotnponents in Seismic Fault Tree Models (Cont'd. 'T a

                                                                                                                                                      \

NAME DESC PROB HCLPF

CHWZCCWS Seismically Induced Failure of CCWS Heat i 1.23 j Exchanger ,

,                          CHXVSF"l!X1                                          SFP-HXrtsolated-for-Maia'aaanca                          J.00E                          C*

s j t CHXNSFPHX2 -SFP HX2 Isolated ~for Maistenance 1.00E C -- l CHXZESFP Seismically Induced Failure of ESF Pump Lube 1.54

Cooler i CHXZSFP Seismically Induced Failure of Spent Fuel Pool Heat 1.10 I Exchanger CKPZCCWSCABLE Seismically Induced Failure of CCWS Power Cable - 1.79

!                          CKPZSSWSCABLE                                        Seismically Induced Failure of SSWS Power Cable                     i           1.79 l                  ~

CPBJCCWPIA --Co y,., Cooling-Water-Pump-CCWP1A-Fails-to- -1,30E y

             "                                                                                                                                                                          l j                                                                                Start                                                              f l                           CPBJCCWPIB                                           Component Cooling Water Pump CCWP1B Fails to              1.30E-03 l                                                                                Start                                                            f a                           CPBJCCWP2A                                           Component Cooling Water Pump CCWP2A Fails to              1.30E-03                                      1

Start l l CPBJCCWP2B Component Cooling Water Pump CCWP2B Fails to 1.30E-03 1 Start j I

CPBVCCWP1B CCWP IB Unavailable Due to Maintenar.ce 1.75E-03 [ i CPBVCCWP2B CCWP 2B Unavailable Due to Maintenance 1.75E-03 1 l CPWJSSWPIA Service Water Pump SSWP 1 A Fails to Start 2.40E-03 CPWJSSWPIB Service Water Pump SSWP IB Fails to Start 2.40E-03 CPWJSSWP2A Service Water Pump SSWP 2A Fails to Start 2.40E-03 l

CPWJSSWP2B Service Water Pump SSWP 2B Fails to Start 2.40E-03 I

4 ) CPWVSSWFIB SSWP IB Unavailable Due to Maintenance 1.7.5E-03

              -{PWVSSWP2B                                                      -3S%7 2B iJnavanaoie uue to Mamtenance                     1.75E-03             N CPWZBLDG                                             Seismically Induced Failure of SSWS Pump Building            l                  1.0g"                   l CQPZCCWSPIPE                                         Seismically Induced Failure of CCWS Piping                                    1.0g*"

i CrXZSURGE Seismically Induced Failure of CCWS Surge Tank { 1.10 CVDBCg-tm . Diaphr=g= np-e~4 Valve CC-102 Fails to 1.55E - CVDBCC-103 Pneum. Diaphragm Operated Valve CC-103 Fails to Close 1.5:iE-03 ( i . , . CVDBCC-112 -Pneum.Eiap_hragm Operated Valve CC-112 Fails to 1.51E-03 , Close  ; i 1 i 4 This HCLPF is based on seismically induced sliding of the CCW Heat Exchanger Building for the most j limiting soil site This component represents failure of the CCW piping between the CCW Heat Exchanger building and the Nuclear Island due to seismically induced sliding of the CCW Heat Exchanger Building The HCLPF is i for the most limiting soil site. Amroved Design Matenal Probabilstic Risk Asseesment Pege 19.7 5

2 tm e fBW c\ System 80+ Design controlDocument Table 19.7.5.1-1 Components in Seismic Fault Tree Models (Cont'd.),} NAME DESC PROB IICLPF CVDBCC-202 gDiaphram Opm'ad Valve-CC-202-Fails-to- -1.55E -- Close 3 CVDBCC-203 Pneum. Diaphragm Operated Valve CC-203 Fails to 1.55E-03 Close CVDBCC-212 Pneum. Diaphragm Operated Valve CC-212 Fails to 1.55E-03

                                       ~Close                                                            <

CVDZAOV Seismically Induced Failure of CCWS Air-Operated  ! 1.35 { Valve I CVDZSFPHXVALVE Seismically laduced Failure of SFPHX Pneumatic  ; 1.35 l 1 solation Valve I i CVMAcc-iv/ MvV CC-107 Fails tu Crn 4-00E43-CVMACC-109 MOV CC-109 Fails to Open 4.00E-03 l CVMACC-207 MOV CC-207 Fails to Open 4.00E-03 l CVMACC-209 MOV CC-209 Fails to Open 4.0CE-03 CVMASW-121 MOV SW-121 Fails to Open 4.00E-03 l CVMASW-123 CVMASW-221 MOV SW-123 Fails to Open MOV SW-221 Fails to Open 4.00E-03 4.00E-03 k ; j CVMASW-223 MOV SW-223 Fails to Open 4.00E-03 j CVMBCC-113 MOV CC-113 Fails to Close 4.42E-03  !

   -     _CYMBCC 213                   -MOV CC;21TFalls to Close~                              4.4ze-u3                "

CVMZMOV Seismically Induced Failure of CCWS 1.35 Motor-Opera:cd Valve { CVMZSFPHXVALVE Seismically Induced Failure of SFPHX M-O Isolation 1.35  ! Valve Atmospheric Dump Valves (ADVs) l DVMASC 17S ADV 5G-178-PaIlstrepen --4.00E ~ DVMASG 179 ADV SG-179 Fails to Open 4.00E-03 DVMASG-184 4VMKSG 185 -- ADV SG-184 Fails to Open ADV SG-185-Failstropen 4.00E-03 4.00E'03

                                                                                                                           ]f     i DVMZADV                       Seismically Induced Failure of Atmospheric Dump                           1.35 Valve (ADV)

Electrical Distribution System l EBDZ125CIE Seismically Induced Failure of 125VDC Class IE 1.06 Circuit Breaker EBDZ125CNIE Seismically Induced Failure of 125VDC Non-lE 0.12 Circuit Breaker 1 EBDZCHGIE Seismically Induced Failure of IE Battery Charger f 1.06 Circuit Breaker l EBDZCHGNIE Seismically Induced Failure of Non-lE Battery i 0.12 Charger Circuit Breaker EBDZDG Seismically Induced Failure of DG Supply Breaker 0.95 EBDZESF Seismically Induced Failure of 4.16KV ESF Breaker . 1.06 Approved Design Material ProbabiUstic Risk Assessment Page 19.7-6

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

1 xTou s c) l Rav c\ l System 80+ oesian controlDocument Table 19.7.5.1-1 Components in Seismic Fault Tree Models (Cont'd4 l i i NAME DESC l PRQB HCLPF EBDZLCIE Seismically Induced Failure of Class IE 480V LC 0.95 Circuit Breaker EBDZLCNIE Seismically Induced Failure of Non-1E 480V LC 0.12 , Circuit Breaker i EBDZPNS Seismically Induced Failure of 4.16KV PNS Breaker l 0.84 l EBDZRATPNS Seismically Induced Failure of Circuit Breaker 0.84 RATPNS I EBDZSBAC Seismically Induced Failure of Circuit Breaker For i 0.84 Standby AC EBDZUATPNS Seismically Induced Failure of Circuit Breaker  ; 0.84

UATPNS' I

! punPR1.R7 1 nan nf Pnw,.gr ,_ p;j,; ,,g_g.ggyg ggggg of

                                                                                                                '"3':60E-01=   =

Power From Switchyard II l

                                                                                                                                             ]

EBGPB2-B1 Loss of Power From Switchyard II Given Loss of 3.60E-01 Power From Switchyard 1  ! EBGPLOOP1 Loss of offsite Power From Switchyard I 1.51E-04 EBGPLOOP2 Loss of Offsite Popwer From Switchyard II 1.51E-04 ECBVPX8801 Battery PX8801 Unavailable Due to Maintenance 1.00E-03 ECBVPY8801 Battery PY8801 Unavailable Due to Maintenance 1.00E-03 \ l ECBVS18801 Battery S18801 Unavailable Due to Maintenance 1.00E-03 ECBVS28801 Battery S28801 Unavailable Due to Maintenance 1.00E-03 i ECBVSA8801 Battery SA8801 Unavailable Due to Maintenance 1.Q0E-03 ECBVSB8801 Battery SB8801 Unavailable Due to Maintenance 1.Q0E-03 ECBVSC8801 Battery SC8801 Unavailable Due to Maintenance 1.00E-03 1ECBVSD880i B.um, 3D;;Gt-U-vaiinoie vue to iviamtenance 1.wE &

                                                                                                                                              )

f - ECBZ125CIE Seismically Induced Failure of 125VDC Class IE 1.58 Battery & Rack ECBZ125CNIE Seisnically Induced Failure of Non-1E Battery & 0.12 Battery Racks ECCZCHG1E Seismically Induced Failure of Class IE Battery 1.06 l Charger ECCZCHONIE Seismically Induced Failure of 125VDC Non-1E 0.12 Battery Charger _ EDDJDGh- -Emergency _Riesel Generator DG A Fails to Start & 1.40E-02_ - Load EDDJDGB Emergency Diesel Generator DG B Fails to Start & 1A E-02 Load EDDKDGA Emergency Diesel Generator DG A Fails to Run 5.16E-02 I (. EDDKDGB Emergency Diesel Generator DG B Fails to Run 5.16E-02 / I EDDVDGA DG A Unavailable Due to Maintenance 6.(CE-03 l 4.EDDVDGB -- _DG B-Unavail.1 Os iv I4-.- - - - - - - _ x &c m s / Aneroveef Design Atatorial* Probabilstic Risk Assessment Page 19.7 7 l l

rre a 6.av o n System 80+ Design ControlDocument

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Table 19.7.5.1-1 Cotnponents in Seismic Fault Tree Models (Cont'd. I NAME DESC PROB HCLPF EDDXRDGA B Common Cause Operatinn Failure of Diesel 4.20E-03 Generators ' ~ EDDZBLDG Seismically Induced Failure of DG Building 1 0.89 EDDZDG Seismically Induced Failure of Diesel Generator l 1.06 EDDZPERIPHERAL Seismically Induced failure of DG Peripherals . I 1.06 EDDZPNL Seismically Induced Failure of DG Control Panel l 1.06 EDGZSBAC Seismically Induced Failure of Combustion Gas 0.36 l Turbine (SBAC) l EDTPT/G Turbine Generator T/G Fails to Provide Adequate 4.80E-03 vutput l

                                                                                                                                                                                             \

EJECDGAE1 Exhaust Damper DGAE1 Fails to Operate 3.Q5E-03 EIECDGAE2 Exhaust Damper DGAE2 Fails to Operate 3,D5E-03 EJECDGASI Supply Damper DGASI Fails to Operate 3/05E-03 EJECDGAS2 EJECDGBEl Supply Damper DGAS2 Fails to Operate 3J05E-03 g Exhaust Damper DGBEl Fails to Operate 3[05E-03 EJECDGBE2 Exhaust Damper DGBE2 Fails to Operate 3;.05E-03 EJECDGBS1 Supply Damper DGBS1 Fails to Operate $.05E-03

                                              -EJECDGBS2                -Supply-Damper-DGBS2 Fails toc              i m.i.                        405E:03--                              -

ELBZ416ESF Seismically laduced Failure of 4.16KV ESF 0.95 Switchgear ELCZ125CIE Seismically Induced Failure of 125VDC Class IE 1.06 Distribution Panel ELCZ125CNIE Seismically Induced Failure of 125VDC Non-lE 0.12 Distribution Panel i ELLZESF Seismically Induced Failure of 480V ESF Load  ! 0.95 Center (LC) ELLZNESF Seismically Induced Failure of 480V Non-1E Load 0.12 Center (LC) ELMZMCCIE Seismically Induced Failure of 480V ESF Motor 0.95 Control Center (MCC) ELMZMCCNIE Seismically Induced Failure of 480V Non lE Motor 0.12 , Control Center

                                   *.         ESXRDeA                    uCtrL-+1 M- F:!!s.en n==a                                                  4.4)E-03                             'N
                                             'ESXRDGB                ~WB -load 4equencer-Faile-to- Operate                                          4.4')E-v1                               N ESXZSEQ                   Seismically Induced Failure of DG Sequencers                                                             1.06 EXLZESF                   Seismically Induced Failure of 480V LC ESF                                                              0.95 Transformer EXLZNESF                   Seismically Induced Failure of 480V Non-1E LC                                                           0.12 Transformer k-RCVRSBAC~                   --
                                                                       -Failure to Start *And' Load Standby ~AC-Power                           -5.0C E                                                                                                                                                                                               (

Approved Design Meterial Prebebristic Risk Assesement Pope 19.7-8

27% i => (20) o\ System 80+ Design ControlDocument l l 1 Table 19.7.5.1-1 Cotuponents in Seismic Fault Tree Models (Cont'd.) g NAME l DESC l PROS l HCLPF l

    -- Le       ;;.J-Safety-Feature-Actuation-System-(ESFAS)                                                      l                       c      )

FMFFCSAS Operator Fails to_ Generate Containment Spray 4.60E-03 l Actuation Signal l FHFFEFWS Operator Fails to Actuate EFWS Components 4.60Ef03 FHFFSIAS Operator Fails to Generate Safety injection Actuation 4.60E'-03 Signal

                                                                                                                                            ~f~

FKPZESFASCABLE Seismically Induced Failure of ESFAS Power Cable l 1.80 FSAXCSAS Common Cause Failure of Contamment Spray 3.02E' 03__ - Actuation Signal l FSERAPS No EFAS Actuation Signal From Alternate Protection 2.60E-02 System l FSEX-EFAS Common Cause Failure, of Emergency Feedwater 2.79E-03 g Actuation Signal T l FSSXSIAS Ca==an ca"- Failure of Safety injection Actuation 3.02E-03 _

       -                                 Signals                                                             l FSXZESFAS                       Seismically Induced Failure of ESFAS Actuation                      !              1.06 System                                                            l Containment Spray System (CSS)/In-Containmaat Refueling Water Storage Tank (IRWST) l GBDZCSSPBRKR                    Seismically Induced Failure of CSS Pump Circuit                                    0.95 Breaker GHCZCSSHX                       Seismically Induced Failure of CSS Heat Exchanger              l                   1.55 i

I GHFFCSS Op ;= F-u ta Manually Actuate Containment . 3AhF-02 .

                                                                                          ~~~

Spray System GHFMCSS11RWST Operator Fails to laitiate Cooling the IRWST Via CSS 3.8]E-03 1 Path l l GHFMCSS21RWST Operator Fails to initiate Cooling the IRWST Via CSS .3.8fE-03 ,,)

         #                               2 Path GKPZCSSCABLE                    Seismically Induced Failure of CSS Power Cables                    l               1.80 nocrceeog                       cm...; = =22      so,,y g . - . cseoi ye ;, 2,,,

3,4.g3 , GPCJCSSP2 Contamment Spray Pump CSSP2 Fails to Start 5.005-03 GPCKCSSP1 Containment Spray Pump CSSP1 Fails to Run 1.20)i-03 g GPCKCSSP2 Caa*=iament Spray Pump CSSP2 Fails to Run 1.2011-03 GPCVCSSP1 Caa*=iament Spray Pump 1 Unavailable Due to 1.7511-03 Testing / Maintenance GPCVCSSP2 Cnataiament Spray Pump 1 Unavailable Due to 1.751I-03 j

       -                                 Testing / Maintenance GQHZCSHDR                       Seismically Induced Failure of Containment Spray                                   0.90 Header GQPZCSSPIPE                     Seismically Induced Failure of CSS-SCS/IRWST                                       0.90 Suction Piping GVCZHXDSCHGVLVS                 Seismically Induced Feilure of CSS HX Discharge                                    2.38 Check Valves ANwered Design Metenal Probablistic Risk Asseesment                                                                Pope 19.7 9

                                                                                                                                   .Tre s ta)

Rm/ ci System 80+ oesian control Document Table 19.7.5.1-1 Components in Seismic Fault Tree Models (Cont'd.{ NAME DESC PROB HCLPF GVCZIRWSTVLVS Seismically Induced Failure of IRWST Return Line l' 2.38 Check Valves GVCZPDSCHGVLVS Seismically Induced Failure of CSS Pump Discharge 1.35 Valves GVCZPSUCVLVS Seismically Induced Failure of CSS Pump Suction 1.35 Valves GVCZSI-100/101 Seismically Induced Fail of IRWST Return Valves 1.35 SI 100/101 nVMCC-114__ _.. MOV-GC-1-14-Fails to Opec 4.00E-03 - , GVMACC-214 MOV CC-214 Fails to Open 4.00E-03 GVMASI-300 MOV SI 300 Fails to Open 4.00E-03 GVMASI-301 MOV SI-301 Fails to Open 4.00E-03 GVMASI-314 MOV SI-314 Fails to Open 4.00E-03 GVMASI-315 MOV SI-315 Fails to Open 4.00E-03 l GVMASI-657 MOV SI-657 Fails to Open 4.00E-03 5 g GVMASI-658 MOV SI-658 Fails to Open 4.00E-03 GVMASI-671 MOV SI-671 Fails to Open 4.00E-03 GVMASI-672 MOV SI-672 Fails to Open 4.Q05-03 GVMASI-686 MOV SI-686 Fails to Open 4.00E-03 GVMASI-688 MOV SI-688 Fails to Open 4.q0E-03 GVMASI-693 MOV SI-693 Fails to Open 4. DOE-03 1

 .     .GVMASl=696                   MOV-SI:696-Faiis m Gm                                            a nnn.m                                      -J        )

GVMZHXDSCHGVLVS Seismically Induced Fail of CSS HX Discharge M-O 1.35 Valves GVMZRTRNVLVS Seismically Induced Failure of CSS /IRWST Return 1.35 M-O Valves GVMZSI-300/301 Seismically Induced Failure of IRWST Return Valves 1.35 SI-300/301 GVNZPSUCVLVS Seismically Induced Failure of CSS /SCS Suction 1.35 Valves Safety injection System (SIS)/ Safety injection Tank (SIT) HBDZSIPBRKR Seismically Induced Failure of SI Pump Circuit 0.95 Breaker HKPZSISCABLE Seismically Induced Failure of SIS Power Cable 1.80

  -     MDWP V                       safety uijwavo Pomi, i Sik te "'M                              -1,0,0E                                  -    -

HPSJSIP2 Safety injection Pump 2 Fails to Start 1.00E-03 HPSJSIP3 Safety Injection Pump 3 Fails to Start 1 Q0E-03 HPSJSIP4 Safety injection Pump 4 Fails to Start 1.00E-03 HPSKSIPl Safety Injection Pump 1 Fails to Run 1.20E-03 HPSKSIP2 Safety Injection Pump 2 Fails to Run 1.20E-03 _HPSKSIE3 _ Afety_ Injection Pump 3 Falk en R9a ..., , . . , I Approved Design Meterial Probabilstic Misk Assessment Pope 19.710

i Tre% i9 Eev on System 80+ Design ControlDocument l Table 19.7.5.1-1 Cornponents in Seismic Fault Tree Models (Cont'df l NAME DESC PROB HCLPF

      ""SKS:74                     Safety-injection Pump 4-Fallrto-Run                       1.20ti-03                         C'
                                                                                                                                  ]

HPSVSIPl SI Pump 1 Unavailable Due to Testing / Maintenance 2.00E-03 HPSVSIP2 SI Pump 2 Unavailable Due to Testing / Maintenance 2.00E-03 SI Pump 3 Unavailable Due to Testing / Maintenance 2.00E-03 HPSYSIP3 j HPSVSIP4- -St-F--y 4 Unavausole vue to Testing / Maintenance T00E6 ' HQPZPSUCPIPE Seismically Induced Failure of SI Pump Suction Pipe l 0.90 HTXZIRWST Seismically Induced Failure of IRWST l 1.08 HVCZDVIVLYS Seismically Induced Failure of DVI Line Check f 2.38  ; Valves , i l HVCZHOTLEGVLVS Seismically Induced Failure of Hot Leg Check Valves j- 2.38 HVCZSIPATHVLVS Seismically Induced Failure of SI Line Path Valves l 1.35

                                                                                                                     ~
                                                                                                                             /
      !!"'i'.S! 321                liot Leg i hi-O "i; SI 321 F.o!. iu Gro                   4.00E        HVMASI-331                 i Hot leg 2 M-O Iso Valve SI-331 Fails to Open              4.00E-03 HVMASI-604                 I Hot Leg 1 M-O Iso Valve SI-604 Fails to Open              4.bOE-03                             j HVMASI-609                   Hot Leg 2 M-O Iso Valve SI409 Fails to Open               4.l00E-03                            i HVMASI-616                   Si Line 4 M-O Iso Valve SI-616 Fails to Open              4.00E-03                          k HVMASI-626                   Si Line 2 M-O Iso Valve SI-626 Fails to Open              4.'00E-03 HVMASI-636                   Si Line 3 M-O Iso Valve SI-636 Fails to Open              4.00E-03 si une i m-v iss-Valve-S!446-Failrty uno                  ,.'vvc-us                 '     J
   - -HYMAsi-040 HVMZHOTLEOVLVS               Seismically Induced Failure of Hot Leg M-O Valves            \               1.35 HVMZPDSCHGVLVS               Seismically Induced Failure of SI Pump M-O                   1               1.35 Discharge Valves LQPZSITPIPE                  Seismically Induced Failure SlT Discharge Piping                             1.91 LTSZSIT                      Seismically Induced Failure of Safety injection Tank           i            2.60 (SIT)                                                          !

LVCZSITVLVS Seismically Induced Failure of SIT Discharge Check l 2.38 Valves i instrument Air System i IKPZIACABLE Seismically Induced Failure of IA System Power 1.80 Cable IQPZIAPIPE Seismically Induced Failure of IA System Piping 0.90 IVNZIAVLVS Seismically Induced Failure of IA System Valves 1.35 Nrre.1n 9 m ; m ; C-I B ";il: A ; --2d)0E - m IWCJC 2A Air Compressor C-2A Fails to Start 2.00E-02 IWCJC-2B Air Compressor C-2B Fails to Start 2.60E-02 4 IWCKC-1A Air Compressor C-1 A Fails to Operate (Run) 2.40E-03 IWCKC 1B Air Compressor C-1B Fails to Operate (Run) 2.4pE-03 IWCKC 2A- Air Compressor C-2B Fails to Operate (Run) 2.4pE-03 IWCKC-2B Air Compressor C-2b Fails to Operate (Run) 2.4QE-03

     -lWCXDCOMr                     Cco o; Caua--D- ad WHue ef 9 Ca.y..22::-                  L%E 03
                                                                                                         \

Anreved Design Matenal Probabidsde Mist Assensmont Pege 19.711

Tras t -) thN di l System 80 + Design ControlDocument

                                                                                                         \

Table 19.7.5.1-1 Components in Seismic Fault Tree Models (Cont'd.Q l NAME l DESC l PROB l HCLPF Shutdown Cooling System (SCS) l JBDZSCSPBRKR Seismically Induced Failure of SCS Pump Circuit 0.95 l Breaker i JHRZSCSHX Seismically Induced Failure of Ses Heat Exchanger l 1.54 JKPZSCSCABLE Seismically Induced Failure of Ses Power Cables l 1.80

    -JPRJSC&iM                    Shutdown Cooling-System-Pump-t-Fails-to@M                  2.30E                   -        -

JPRJSCSP2 Shutdown Cooling System Pump 2 Fails to Start 2.30E-03 JPRVSCSP1 Shutdown Cooling System Pump 1 Unavailable Due 2.00E-03 to Testing / Maintenance l _ JPRVSCSP2 Shutdown Cooling System PumpljJnay=it.hle nue 2.00E-03 3 to lestmg/ Maintenance i j JQPZSCSPIPE Seismically Induced Failure of SCS/RCS Suction O.90 Piping JVCZHXDSCHGVLVS Seismically Induced Fail of SCS HX Discharge Check 1.35 Valves JVCZPDSCHGVLVS Seismically Induced Failure of SCS Pump Discharge 1.35 Valves u TVM A CC-111 CN/SCSMX1 M O Vi/c CC-11-1-Fails te Oy 100E-G3-JVMACC-211 CCW/SCSHX2 M-O Valve CC-211 Fails to Open 4,00E-03 JVMASI-310 SCS HXI Flow Control Valve SI-310 Fails to Open 4,00E-03 JVMASI-311 SCS Hx2 Flow Control Valve SI-311 Fails to Open 4,00E-03 JVMASI-00 SCS Train 2 Discharge ISO Valve SI-600 Fails to 4.00E-03 Open JVMASI-601 SCS Train 1 Discharge ISO Valve SI-601 Fails to 4.D0E-03 Open JVMASI-651 SCS Suction M-O ISO Valve SI-651 Fails to Open 4.%0E-03 JVMASI-652 SCS Suction M-O ISO Valve SI-652 Fails to Open 4.' DOE-03 JVMASI-653 SCS Suction M-O ISO Valve SI-653 Fails to Open 4. X)E-03 JVMASI-654 SCS Suction M-O ISO Valve SI-654 Fails to Open 4. ME-03 JVMASI-655 SCS Suction M-O ISO Valve SI-655 Fails to Open 4.<2E-03

                                                                                                             ~

4VMA5i-ose SS S=.eu M-G isv valve sI-oso Fanis to vpen 4.90E W JVMZHXDSCHGVLVS Seismically Induced Failure of SCS HX Dircharge 1.35 M-O Valves JVMZRTRNVLVS Seismically Induced Failure of SCS/IRWST Retum 1.35 M-O Valves JVMZSUCVLVS Seismic 41y Induced Fail of SCS/RCS Suction M-O 1.35 Valves WXCONN CCF of CSSeo-SCS Cross-Connect Manual Valves to 1.5$E-03 ,c Open  ! JVNZPSUCVLVS Seismically Induced Failure of CSS /SCS Suction 2.38 Manual Valves Anwoved Design Meterial Prebebiistic Risk Assessment Page 19.712

                                                                                                                                                                                                     .y 1
                                                                                                                                                                            .rrm i =>                  I itev ol Systrm 80+                                                                                                                     Design Control Document l

Table 19.7.5.1-1 Components in Seismic Fault Tree Models (Cont'd. i NAME DESC PROB HCLPF JVNZXCONNVLYS Seisutically Induced Failure of CSS /SCS 1.35 Cross-Connect Valves Chemical and Volume Contrcl System (CVCS) UBDZBAMPBRKR feismically Induced Failure of Boric Acid ? .E$p 0.12 Ptucp Circuit Breaker UBDZCHPBRKR Seismically Induced Failure of Charging Pump Circuit 0.12 i Breaker UHFFBORONRCS Operator Fails to initiate Boron Delivery to_RCS Via 3.25E-02 Charging Pump l UKPZCVCSCABLE Seismically Induced Failure of CVCS Power Cables 1.80

          .             IIPMICMP2                                            -Charging-Pump 2 Fails to'Stan                                            2.GOE-03      -                      d UQPZCVCSPIPE                                           Seismically Induced Failure of CVCS Piping                                                     0.12 UTXZBAST                                               Seismically Induced Failure of Boric Acid Storage                                   }          0.12 Tank (BAST)                                                                         '

UVCZBAMPVLVS Seismically Induced Failure of BAMP Flow Path 0.12 Check Valves s UVCZCHPRGHXVLVS Seismically Induced Failure of CHP/ REGEN HX - 0.12 Check Valves  ! l t"'M/.CI-i-J i, Muv t rs-3;4 F S te W 4.00E MOV CH-534 Fails to Open

                                                                                                                                                                                          ]
                       .UVMACH-534                                                                                                                     4.00E-03                            [

UVMACH 536 MOV CH-536 Fails to Open 4.00E-03 ! UVMBCH-501 MOV CH-501 Fails to Close 4.42E-03 _LLVMBCH-jim -MOV-GH-504 Fails to Fina- A Mn m - UVMZCH-514 Seismically Induced Failure of BAMP/CHP Suction 0.12  ! l ! Valve 3 UVMZFEEDVLVS Seismically Induced Failure of Gravity Feed /CHP 1.35 M-O Valves l UVMZVCTOUTVLVS Seismically Induced Failure of VCT Outlet M-O 1 1.35 Valves ( Safety Depressurization System (SDS) or Bleed System { VClZINVERTERS Seismically Induced Failure of Inverter For Bleed l 1.06 l Valve l l VHFFEEEDB1-EED- OpmtorFails-to-Initiate-Feed & Bleed-System 9,16E - c l VKPZSDSCABLE Seismically Induced Failure of SDS Power Cables l 1.80 VQPZSDSPIPE Seismically Induced Failure of SDS Piping l 0.90 "YMARC-406 - ..MOV. RC-406_ Fails to Opec 7:40E _ VVMARC-407 Fails to Open 2 40E-02 l VVMARC-408 MOV RC-408 Fails to Open 2.40E-02 R ! VVMARC-409 MOV RC-409 Fails to Open 2 40E-02

             -      -VVMXBtDV                                                ~ Commori Cause-Failur cf BM. V.!we                            -

4 *nE.03- ~ l VVMZBLEEDVLYS Seismically Induced Failure of Bleed Path Valves { l.35 l ! )* Apnwovent Design Material Probabilstic Itisk Assessment Page 19.713

s esoe,c T. eo rna rw - na g System 80+ nesign controlDocument l Table 19.7.5.1-2 Seismic Fragilities For System 80+ Structures and NSSS Components Structure /NSSS Component CDFM llCLPF (pga) Interior Structure 1.08g Fuel Building 1.35g CVCS 1.Olg Diesel Generator 1 or 2 0.89g l EFW Storage Tank I or 2 0.89g Control Room Area 1.12g Shield Building 1.25g Service Water Pump Building 1.00g Containment Vessel 0.73g Nuclear Island - Sliding 1.00g CCW Heat Exchanger Building-Sliding 1.0g CEDMs 1.35g RCP/ Supports 0.86g Reactor VesseUSupports 1.14g Reactor Internals-Spacer Grid 0.75g Reactor Internals-Fuel Assy 0.87g Reactor Internals-UGS/CSB 0.74g Steam Generator / Supports 0.87g Pressurizer / Supports 2.59g NSSS Piping 1.91g l l i Appmved Design Material Probabilstic Risk Assessment Phe 19.714

                                                                                              .TroA        l')
                                                                                                   /Ecrv oI System 80+                                                                      Design controlDocument Table 19.7.5.3-1         Summary ofII LPFs for Seismic Sequences                              C Sequence      HCLPF        Dominant Cutset        Mixed HCLPF Failure Prob.        Dominant Cutset EQSTR            .73                     -

1.25 2.50E-01 - EQA-15 - - .74 1.00E-02 1 i EQRVR .86 - - - - EQT-7 .89 97 - - 1-SEIS-SBO .89 21 , .12 2.10E-04 100 EQLP-8 .89 20' .89 5.00E-02 ~151 EQA-10 .89 6- .95 1. *19E-03 33 EQLP-7 .90 4 - -' - EQSLO-3 .90 71 - -: - EQSLO-6 .90 7 -

                      .90                    80 EQA-3                                                    -

EQA-4 .90 76 - -!  ;- EQSLO-9 .90 10 - l 1 - EQLP-3 .90 24 - - - EQT-3 .90 80 - - - EQLP-4 .90 l61 - EQA-7 .90 l2 .90 3.25E-02  ! 58 . I EQA-8 .90 l67 .90 3.25E-02 i174 i EQSLO-10 .90 l 17 .90 3.30E-03 I 64 EQT-4 .90 l100 1.10 5.60E-02 l 138 EQT-8 .90 l 76 1.10 5.60E-02 l 142 EQSLO-11 .95 l2 .90 6.40E-02 l 80 EQA-5 1.06 l 13 - f  ! - EQT-5 1.06 39 - -

                                                                                             }     -

EQLP-5 1.06 16 - l- l j EQSLO-4 1.06  ! 48 - j- l - l EQA-9 1.06 6 .74 9.15E-03 { 79 EQLP-9 1.% l1 .39 9.15E-03 l 82 EQT-9 1.06 l1 .89 9.153-03 } 71 EQSLO-7 1.06 l1 .90 9.15E-03 \ 17 EQA-14 1.35 l1 - l- { - EQA-12 .--. 1.35 l2 - l- l - EQA-13 1.35 l1 1.35 5.60E-02 l 27 EQLP-12 1.35 1 1.35 5.60E-02 l 53 EQLP-11 1.35 i 1 1.35 1.55E-03 l 15 - EQML-3 2.59 1 2.59 5.60E-02 l 7.9 EQML-2 2.59 1 2.59 5.60E-02 l 17 l Approved Design Moren.el Probabiistic Misk Assusment Pope 19.716

                                                                     ' M :a 7Ej}e /9. 'Ed3 -2               Saismic Mnegins Senr$>+ h Ak yA " '
                     ?k- 73Ade Inkk24 R&x)                                      ,
   *TL}g /9.754- /             <Cs'm f;,. Ap tand sti h so;/.s/e, rx s l / m r a e x / N fL M f                 75 4 7s A l< Tn16 xv % s f y/ 2 /ss )                          ,
   % A /9%.c4-2                & Ara and /4,he Asss ayaed
                               #clPnr Se Sk and 5;/ Sfes                            :
)                   % 7E44:. Infen#euL& 8fa A                                   l   l 1

7;Aja /9.1:.c /.3 S>cnz9 s( #4Pfs di- Se a w a \ Seguences de Sif thee 2/ 7}& /aAfe .Iidex1%zMs 2/exA \

    %Ib/s. /9XL9'-f' Su.musy & EDAM /j,' Seirinic sapaneas C, 24s so;/ .s/e.

7 bis YiDe YntenurMy.B&n/ 75Afe. /9. Xr.4-5 Sa.m>uq *f HCLP& -Er Sesino'c

                                .Sepe>eces Re soif 5% 82 7 bis 7}.)?c. f ./g,u.fx,g))g p , Q 7 A A A / 9. L C f - 6    ~ 5 m sf # G P F R e S eis*1'*c.

squ.excas Ge .5 '/ s~h 2s.r J 7Nis /TE.Pa. .Z>,len//orusff, MJ ^

    '7Ab}e. J9 7'5./ '7          Sunrmayr

• l E d L P A ;i y C,-. S e i.r n i e sayunce.r Ar Sa.') 2/e M 7NI.c 7 [Afe Zvfen?% z Hy B&ad

I z.7eu 19 i l nev o s System 80+ Design ControlDocument 7 i Table 19.7.5.3-2 Seismic Margins Sensitivity Analyses l

   'N                                                                                         /        1 l                                                                           Min Seq   Plan [HCLPF       '

l Sensitivity Case From To HCLPF / Base Caseh 0.,B[ 0.73

1. Change Piping HCLPF 0.90 0.50 [ 0.50 0.50

2. Change Piping \CLPF 0.90 0 00' O.89 0.73

3. Change ATWS HCL)Fs 0.74 / 0.50 0.89 0.73 j 4. Change ATWS HCLPF \ N74 1.00 0.89 0.73

5. Change MOV HCLPF \ ./ 1.35 0.50 0.50 0.50

6. Change MOV HCLPF b 1.35 2.00 0.89 0.73

7. Change Brealer HCLPF [ \ 0.95 0.50 0.50 0.50

8. Change Breaker HCLPF[ Oh 1.00 0.89 0.73

9. Chenge EFW Tank /DG Room HCLPF 0.89 k50 0.50 0.50

10. Change EFFIank/DG Room HCLPF 0.89 N

1.00 N 0.90 0.73

11. Change' Condensate Storage Tank HCLPF 0.56 1.00 k 0.73  ;

12. dge Condensate Storage Tank HCLPF 0.56 0.10 0.89 \ 0.73 1

f l j I 1 l a Approved Dess e n Morenal- ProbeMstic Risk Assessment Pope 19.716

     .m . _ _ _ _ _ .            -_ . _ _ _ _ _ _ . _ _ _ _ _ . _ _ . _ _ _ . _ _ _ _ _ . . _ . _                                                               _ _ . .       __

1 ,f - ITafA4 19 stev o j l j System 80+ Desian ControlDocument Table 19.7.5.41 Comparison of Component HCLPFs for Rock and Soll Sites f l }  % ) N Rock l B1 Soil Bl.5 Soil B2 Soil B3.5 Soil B4 Soil / j ponent HCLPF HCLPF HCLPF HCLPF HCLPF HCLPF  ! ABDZEFWP8RKR 0.95 1.03 1.10 1.60 2.04 1/1 l } AKPZEFWSCABM* 1.80 1.80 1.80 1.80 1.80 / 1.80 l AITXEFWP* \ 2.38 2.38 2.38 2.38 2.3s / 2.38 I ! AQPZEFWSPIPE* \ 0.90 0.90 0.90 0.90 0.90/ 0.90 ! AQWZEFWDGROOM* \ 0.89 0.89 0.89 0.89 OM9 0.89 f AVCZEF-214* \ 2.38 2.38 2.38 2.38 / 2.38 2.38 l AVMZPDSCHGVLVS* \ 1.35 1.35 1.35 1.35 / 1.35 1.35 l AVNZEF-215* \ 2.38 2.38 2.38 2.3/ 2.38 2.38 l MTXZCST* kS6 0.56 0.56 g56 0.56 0.56 i CBDZCCWPBRKR O.h 1.03 1.10 /1.60 2.04 1.71 ! CBDZS5WPBRKR 0.95 \ 1.03 1.10 / 1.60 2.04 1.71 CHWZCCWS+ 1.23 \ 0.97 0.9/ 0.97 0.97 0.97 l

CHXZESFP+ 1.54 \ 1.29 [ 29 1.29 1.29 1.29 j CHXZSFP* 1.10 \1.10 /1.10 1.10 1.10 1.10 I CKPZCCWSCABLE* 1.80 l'.11 0 / 1.80 1.90 1.80 1.80 CKPXSSWSCABLE* 1.80 1.8h/ 1.80 1.80 1.80 1.80 l

f CPWZBLDG 1.00 1/0 \ 1.50 1.10 1.00 1.00 ! CQPZCCWSPIPE 1.00 /1.50 \ 1.50 1.;0 1.00 1.00 ) CTXZSURGE* 1.10 / 1.10 \ 1.10 1.10 1.10 1.10  ; h CVDZAOV* 1.35/ 1.35 \1.35 1.35 1.35 1.35 j CVDZSFPHXVALVE* 1.)d 1.35 1h5 1.35 1.35 1.35 ) CVMZMOV* /i.35 1.35 1.3% 1.35 1.35 1.35 l CVMZSFPHXVALVE* / 1.35 1.35 1.35 \ 1.35 1.35 1.35 DVMZADV* / 1.35 1.35 1.35 \ 1.35 1.35 1.35 EBDZ125CIE / 1.06 0.89 0.93 \1.29 1.91 1.41 EBDZ125CNIE** / 0.12 0.12 0.12 h.12 0.12 0.12 EBDZCHGIE / 1.06 0.89 0.93 1.b 1.91 1.41 EBDZCHGNIE/ 0.12 0.12 0.12 0.1d\ 0.12 0.12 EBDZDG / 0.95 1.03 1.10 1.60 h 2.04 1.71 EBDZESF/ 1.06 1.03 1.10 1.60 \ 2.04 1.71 EBDZ1.d1E 0.95 1.03 1.10 1.60 \704 1.71 EBldLCNIE** 0.12 0.12 0.12 0.12 k12 0.12 0.84 0.90 0.91 1.47 1.9I 1,76 EJIDZPNS

                         /IBDZRATPNS                                          0.84                0.90                  0.91          1.47           1.92\         1.76
                      / EBDZSBAC                                              0.84                0.99                  0.91          1.47           1.92 \        1.76
                                                                                                                                                              \

Appmeenf Demon A0etenel Prebendeeio niek Asseesment Pope 19.1< 17

l ITGM 19 R91 of System 80+

                                                                 /                             Design controlDocument l

Table 19.7.5.4-1 Comparison of Component HCLPFs for Rock and Soil Sites (Cont'.d) N u B1 Soil Bl.5 Soil B2 Soil B3.5 Soil B4 Soil N Component mPF HCLPF HCLPF HCLPF HCLPF HCLPF EBDZ.UATPNS 0.84 0.90 0.91 1 47 1.92 /1.76 ECBZ125CIE* 1.58 1.58 1.58 1.58 1.58 / 1.58 l ECBZ125CN1E" 0.12 0.12 0.12 0. j 2 0.1/ 0.12 l ECCZCHGIE\ 1.06 0.89 0.93 1.29 g91 1.41 ECCZCHGNIEk 0.12 0.12 0.12 0.12 / 0.12 0.12 EDDZBLDG* \ 0.89 0.89 0.86 0.89 / 0.89 0.89 EDDZDG+ \ 1.06 0.92 0.92 0.92 0.92 0.92 l EDDZPERIPHERAL+ \ 1.06 0.92 0.92 jd.92 0.92 0.92 EDDZPNL \ 1.06 1.02 1.12 / 1.59 2.42 2.04 EDGZSBAC** \ 0.36 0.36 0.36 / 0.36 0.36 0.36 ELBZ416ESF \ 0.95 1.03 I ff'O 1.60 2.04 1.71 . ELCZ125CIE 1 06 0.89 /0.93 1.29 1.91 1.41 ELCZ125CNIEa 0.11 0.12 / 0.12 0.12 0.12 0.12 ELLZESF 0.95 \ 1.03 / 1.10 1.60 2.04 1.71 l ELLZNBSF** 0.12 \ 0.}d 0.12 0.12 0.12 0.12 l ELMZMCCIE 0.95 1 03 1.10 1.60 2.04 1.71 l ELMZMCCNIE** 0.12 /0.1hs 0.12 0.12 0.12 0.12 ESXZSEQ 1.06 / 1.02 \ 1.12 1.59 2.42 2.04 l EXLZESF 0.95/ 1.03 \1.10 1.60 2.04 1.71 l EXLZNESF** 0/2 0.12 h12 0.12 0.12 0.12 FKPZESFASCABLE* /1.80 1.80 1.8h 1.80 1.80 1.80 FSXZESFAS* / 1.06 0.89 0.93 \ 1.29 1.91 1.41 GBDZCSSPBRKR / 0.95 1.03 1.10 \ 1.60 2.04 1.71 GHCZCSSHX+ / 1.55 1.29 1.29 1 29 1.29 1.29 GKPZCSSCABLE* / 1.80 1.80 1.80 1.8q 1.80 1.80 GQHZCSHDR* / 0.90 0.90 0.90 0.90 \ 0.90 0.90 GQPZCSSPIPE* / 0.90 0.90 0.90 0.90 \ 0.90 0.90 GVCZHXDSCH6VLVS* 2.38 2.38 2.38 2.38 4.38 2.38 GVCZIRWSTVLVS* 2.38 2.38 2.38 2.38 2.38 2.38

GVCZPpSCHGVLVS* 1.35 1.35 1.35 1.35 1.35 \ 1.35 GVCZPSUCVLVS* 1.35 1.35 1.35 1.35 1.35 \ 1.35 GYCZSI-100/101* 1.35 1.35 1.35 1.35 1.35 '1.35

~ SVMZHXDSCHGVLVS* 1.35 1.35 1.35 1.35 1.35 1.35

)   /  GVMZRTRNVLYS*                           1.35           1.35          1.35         1.35           1.35          1.35 GVMZSI-300/30 l*                         1.35          1.35          1.35         1.35           1.35          1.35 Approved Design Meterial- ProbaMstic Risk Assessment                                                        Page 19.718 i

i

IYem s c) ruv oI System 80+ Desian controlDocument e/ Table 19.7.5.4-1 Comparison of Component HCLPFs for Rock and Soil Sites (Cont'.d) Rock B1 Soil Bl.5 Soil B2 Soil B3.5 Soil

                                                                                                                               /

B4 Soil /

     \          Component                   HCLPF           HCLPF          HCLPF          HCLPF       HCLPF        HCLPF G yZPSUCVLVS*                              1.35         1.35              1.35       1.35       1.35          / 35 HBDMIPBRKR                               0.95          1.03              1.10       1.60       2.04    / 1.71 HKPZSilllQABLE*                            1.80         1.80              1.80       1.80       1.80 /            1.80 HQPZPSUkIPE*                              0.90         0.90              0.90        0.90       ps0              0.90 HTXZIRWST*\                               1.08         1.08              1.08       1.08   / 1.08                1.'08 HVCZDVIVLVS*\                            2.38         2.38              2.38        2.38/      2.38             2.38 HVCZHOTLEGVL $*                           2.38         2.38              2.38       J.'58       2.38             2.38 HVCZSIPATHVLVS* \                          1.35         1.35              1.35   / 1.35         1.35              1.35 l-HVMZHOTLEGVLYS* \                          1.35         1.35             1.35 /      1.35       1.35              1.35 HVMZPDSCHGVLVS*             \             1.35         1.35              1/5        1.35       1.35              1.35 LQPZSITPIPE*                  \           1,91        1.91       / 1.91             1.91       1.91              1.91    j LTSZSIT*                         \ 2.60               2.60 /            2.60        2.60       2.60             2.60     l l        LVCZSITVLVS*                       \ 2.38             2.3J/             2.38        2.38       2,38             2.38 l

IKPZIACABLE* h80 )[80 1.80 1.80 1.80 1.80 IQPZ1APIPE* 0.dQ /0.90 0.90 0.90 0.90 0.90 l l IVNZIAVLVS* 1.35\/ 1.35 1.35 1.35 1.35 1.35 ) I JBDZSCSPBRKR 0.9/ \ 1.03 1.10 1.60 2.04 1.71 JHRZSCSHX+ /.54 \ 1.29 1.29 1.29 1.29 1.29 JKPZSCSCABLE* / 1.80 \ 80 1.80 1.80 1.80 1.80 l JQPZSCSPIPE* / 0.90 0.% 0.90 0.90 0.90 0.90 JVCZHXDSCHGVLVS* / 1.35 1.35 \ 1.35 1.35 1.35 1.35 JVCZPDSCHGVLVS* / 1.35 1.35 \ 1.35 1.35 1.35 1.35 JVMZHXDSCHGVI)(S* 1.35 1.35 \1.35 1.35 1.35 1.35 JVMZRTRNVLVM 1.35 1.35 \{.35 1.35 1.35 1.35 JVMZSUCVLys* 1.35 1.35 Ik5 1.35 1.35 1.35 JVNZPSUQILVS* 2.38 2.38 2.38\ 2.38 2.38 2.38 JVNZXQdNNVLVS* 1.35 1.35 1.35 \ 1.35 1.35 1.35 UBDKdAMPBRKR** 0.12 0.12 0.12 \0.12 0.12 0.12 l UBdZCHPBRKR** 0.12 0.12 0.12 h12 0.12 0.12 l ,1fKPZCVCSCABLE* 1.80 1.80 1.80 1.8q 1.80 1.80 l UQPZCVCSPIPE** 0.12 0.12 0.12 0.12 \ 0.12 0.12 f UTXZBAST** 0.12 0.12 0.12 0.12 \0.12 0.12 l l UVCZBAMPVLVS** 0.12 0.12 0.12 0.12 h12 0.12

    ' UVCZCHPRGHXVLVS**                           0.12         0.12             0.12        0.12       0.1'2s            0.12 l

I UVMZCH-514** 0.12 0.12 0.12 0.12 0.12 \ 0.12 l l

} l Aaprovent Demon Metenal . hebehtiisaic niek Asseneneont Pope 19.719 1  !

XTcn 19 -

                                                          /                                            Rev os  l System 80+                                                                        Design controlDocument i

Table 19.7.5.4-1 Comparison of Component HCLPFs for Rock and Soil Sites i

      \                   (Cont'.d)                                                                            !
                                                                                       /

Rock B1 Soil Bl.5 Soil B2 Soil' B3.5 Soil B4 Soil Com\ ponen( HCLPF HCLPF HCLPF HCLTF HCLPF HCLPF UVMZFEEDVLVS* \ 1.35 1.35 1.35 / 1.35 1.35 1.35

UVMZCTOUTVLVS* \ 1.35 1.35 1 3F' 1.35 1.35 1.35 VClZINVERTERS N,06 0.89 /0.93 1.29 1,91 1.41 VKPZSDSCABLE* 1.80 \ s 1.80/ 1.80 1.80 1.80 1.80 VQPZSDSPIPE* 0.90 MRO 0.90 0.90 0.90 0.90 VVMZBLEEDVLVS* 1.35 / . 35 \ 1.35 1.35 1.35 1.35 NOTES:

• a, k Site HCLPF governed, value qor all sites y

                  +       =       Governing soil site selected by spectra comparison. HCLPF for governing       i soil site used for all soil sites.
                  **      =       HCLPF value assumed for Non-1E equipment used            I sites.
                  #       =       HCLPF for limiting site used for all sites l

I 1 i a,,,,. s o-a,, u,,.,,,1. e,.a.u,u n,& a,, .,,,.,,, ,,,. run  ! 1

Jrm i9  ; trev o1 i System 80+

                                                                                                /                                                  Design Control Document Table 19.7.5.4-2                               Structure and f40r NSSS Component HCLPFS for Rock and Soil Sites
                                                                                                                                                                                 /

COMPONENT / ROCK B1 SOIL B1.5 SOIL B2 SOIL B3.5 SOIL B4 SOIL / STRUCTURE HCLPF HCLPF HCLPF HCLPF HCLPF HCLPF Interior Structure

• 1.08g 1.08g 1.08g 1.08g 1.08g 1.kg Fuel Building
• 1.35g 1.35g 1.35g 1.35g 1.35g [35g CVCS* 1.01g 1.01g 1.0lg 1.01g 1.01g [1.0!g Diesel Generator 1 or 2* 0.89g 0.89g 0.89g 0.89g 0.89g [ 0.89g EFW Storagehnk I or 2* 0.89g 0.89g 0.89g i 0.89g 0.89[ 0.89g Control Room Ark 1.12g 1.12g 1.12g 1.12g (d2g 1.12g Shield Building * \ '1.25g 1.25g 1.25g 1.25g [1.25g 1.25g Contammant Vessel * \ 0.73g 0.73g 0.73g 0.73g [ 0.73g 0.73g  !

Service Water Pump 00g 1.50g 1.50g 1.50 1.10g 1.10g Building CCW Heat Exchanger 1.00g\ 1.50g 1.50g .50g 1.10g 1.10g < Building - Sliding \ f Nuclear Island - Sliding 1.00g k.30g 1.50g[ 0.90g 1.60g 1.20g CEDMs (Rock governs 1.35g 1.0 1. g 1.35g 1.35g 1.35g , except for B1) RCP/ Supports 0.86g 0.90g \[1.49g 2.58g 1.89g 0.91g Reactor Vessel / Supports 1.14g 1.87g / k 1.75g 3.20g 1.72g 1.13g Reactor Internals - Spacer 0.75g 0.7)g k75g 0.75g 0.75g 0.75g Grid * / \ Reactor Internals - Fuel 0.87g [0.87g 0.873 0.87g 0.87g 0.87g Assy* j Reactor Internals - 0.7 0.74g 0.74g 0.74g 0.74g 0.74g UGS/CSB* Steam Generator / Supports [0.87g 1.38g 1.55g 2.k(g 1.94g 1.38g Pressurizer / Supports * / 2.59g 2.59g 2.59g 2.59g\ 2.59g 2.59g NSSS Piping * [ 1.91g 1.91g 1.91g 1.91g \ 1.91g 1.91g NOTES: = Rock Site HCLPF governed, value used for all sites l

                                                                                                                                                                                                     \

I Aweed Denign Atetend hebehdiraic Riek Assensmont Peye 19.7 21 L - . , , - . - . . . - - - I

i _rrem i9 l trev eI \ System 80+ h Design controlDocument Table 19.7.5.4-3 W Summary of HCLPFS for Seismic Sequences for Soil Case B1 Sequence IICLPF Dominant Cutset Mixed IICLPF Failure Prob. Dominant Cutset EQSTR .73 - 1.25 2.5E-01 - I EQA-15 - -

                                                           .74        1.00E-02                 1, SEIS-S'BQ              .89               16               .12        2.10E-04              1,00          l EQA-9 \                .89                6               .74        9.15E-03            / 79 EQLP-9      \          .89                1               .89        9.15E-03         / 82               l EQLP-8         \       .89               44               .89        5.00E-02      /       151           !

EQT-9 \.89 1 .89 9.15E-03 / 71 EQA-10 .89 1 .97 1.19E-0)' 42 EQRVR .90\ - -

                                                                          -/                  -           I EQLP-7                 .90    'N          1                -
                                                                        /-                    -

EQSLO-3 .90 \ 71 -

                                                                     / -                      -

EQSLO-4 .90 \48 -

                                                                   /      -                   -

EQSLO-6 .90  % -

                                                                 /        -                   -

EQA-3 .90 80 \ -/ - - EQA-4 .90 76 \ -/ - - EQA-5 .90 102 \ /- - - EQSLO-9 .90 10 \/- - - EQLP-3 .90 24 /\- - - EQLP-4 .90 61 / \ - - EQT-3 .90 80 / -\ - - EQT 5 .90 121 / -

                                                                 \        -                   -

EQT-7 .90 7/ - N - - EQLP-5 .90 / 47 1 -

                                                                     '\ -                     -
                                 ~

EQSLd-/ .90 / 1 .90 9.15E-03 17 l EQA-7 .90 / 2 .90 3.25EM2 58 EQA-8 .90 / 67 .90 3.25E-725 174 EQSLO-10 .90/ 17 .90 3.30f-03 x 64 EQSLO-11 7 90 3 .90 6.40E 02 \ 80 l ? EQT-8 / .90 56 1.10 5.60E 02 \(42 EQT-4 / .90 100 1.10 5.60E-02 138N EQA-14 / 1.35 1 - - - N EQA-12 / 1.35 2 - - - EQA-13 / 1.35 2 1.35 5.60E-02 27 EQLP 17' 1.35 1 1.35 5.60E-02 53 EQLP/11 1.35 1 1.35 1.55E-03 15 EQML-3 2.59 1 2.59 5.60E-02 29 EQML-2 2.59 1 2.59 5.60E-02 17 b Approved Design Metenet - ProbeMstic Risk Assessment Pope 19.7-22

Trw I c) taev o I System 80+ 0 / Design ControlDocument U Table 19.7.5.4-4 Summary of HCLPFs for Seismic Sequences for B1.5 Soll Site Sequence HCLPF Dominant Cutset Mixed HCLPF Failure Prob. Dommant Cutset l EQSTR .73 - 1.25 2.50E-01 - EQA-15 - -

                                                          .74               1.00E-02                1 /

! SEIS-SBOs .89 26 .12 2.10E-04 100' EQLP-8 \ .89 44 .89 5.00E-04 /51 l EQA-10 \.89 6 .97 1.19E-03 / 42 EQLP-4 .9Q 61 - -

                                                                                           /       -

EQLP-7 .90 \ 4 - -

                                                                                        /          -

EQSLO-3 .90 x 71 - -

                                                                                      /            -

EQSLO-6 .90 \ 7 -

                                                                                -/                 -

l EQA-3 .90 \ 80 -

                                                                                /                  -

EQA-4 .90 \ 76 -

                                                                            /-                     -

EQSLO-9 .90 1%

                                                                         /      -                  -

l EQLP-3 .90 24\ -

                                                                      /         -                  -

EQT-3 .90 80 \ -

                                                                   /            -                  -

EQT-7 .90 3 \ -

                                                                /               -                  -

EQA-7 .90 2 \ .90/ 3.25E-02 58 EQA-8 .90 67 \ M 3.25E-02 174 EQSLO-10 .90 17 \/.90 3.30E-03 64 EQT-4 .90 100 / \J.10 5.60E-02 138 EQT-8 .90 76 / 1h0 5.60E-02 142 EQA-5 .93 102 / \ - - l EQT-5 .93 121 / -

                                                                \               -                  -

EQLP-5 .93 147 / -

                                                                  \              -                 -

EQSLO-4 .93 48/ -

                                                                    \           -                   -

EQA-9 .93 / .74 \ 9.15E-03 79 EQLP-9 .93 /1 .89 \ 9.15E-03 82 EQT-9 .93 /1 .89 N.15E-03 71 EQSLO-11 .93 / 3 .90 6.ME-02 80 EQSLO-7 .93 / 1 .90 9.15K-03 17 i EQa-14 1.35 / 1 -

                                                                                 -\                 -

( EQA-12 1.35 / 2 - -

                                                                                      \             -

EQA-13 1.% 2 1.35 5.60E-02 \ 27 $ EQLP-12 4.35 1 1.35 5.60E-02 \ 53 EQLP-11 1.35 1 1.35 1.55E-03 \ 15 EQRVR 1.49 - - -

                                                                                              \-

1 EQML-3 2.59 1 2.59 5.60E-02 '29

. EQML-2         2.59                  1                2.59               5.60E-02               17 l

Amved Denign Metenal- Probabi6sa'c Risk Assessment Pope 19.7 23 a +

                                                                                                        .C7PM (9 System 80+                                        f l                      /
                                                                                                         /26Y OI Design ControlDocument Table 19.7.5.4-5            Summary of HCLPFS for Seismic Sequences for Soll Site B2 I

l l Sequence HCLPF Dominant Cutset Mixed HCLPF Failure Prob. Dominant Cutset l xEQSTR .73 - 1.25 2.50E-01 - EQA-15 - -

                                                                .74              1.00E-02                _1 SEIS'-SBO           .89               26                   .12             2.10E-04                 A00 EQLP-8\            .89                44                   .89             5.00E-02              / 151 EQA-10 \           .89                 6                   .97             1.19E-03            /

i2 l EQLP-4 x .90 61 - -

                                                                                                 /          -

l i EQLP-7 \.90 4 - - 1 EQSLO-3 .90 71 - -

                                                                                          /                 -

EQSLO-6 .90 \ 7 -

                                                                                     - /                    -

EQA-3 .90 s 80 -

                                                                                     -/                     -

EQA-4 .90 \ 76 -

                                                                                   /-                       -

EQSLO-9 .90 \ 10 -

                                                                                / -                         -

EQLP-3 .90 \24 -

                                                                           /         -                      -

l EQT-3 .90 84 -

                                                                         /          -                       -

EQT-7 .90 3 \ -

                                                                       /            -                      -

EQA-7 .90 2 \ .90 / 3.25E-02 58 EQA-8 .90 67 \ )fV 3.25E-02 174 l i EQSLO-10 .90 17 \ /)0 3.30E-03 64 EQT-4 .90 100 \/1.10 5.60E-02 138 EQT-8 .90 76 / \1.10 5.60E-02 142 l EQA-5 1.29 93 / \ - - l EQT-5 1.29 66 / -\ - - l EQLP-5 1.29 38 / -

                                                                     \              -                      -

l -E

     ,.O. ..S LO-4     1.29               22/                    -
                                                                       \            -                      -

M \-9 1.29 A .74 \ 9.15E-03 79 j EQLP-9 1.29 /1 .89 '9.15E-03 82 EQT.9 1.29 / 1 .89 9.15E-03 71 EQSLO-11 1.29 / 3 .90 6.40E-02 80 EQSLO-7 1.29 / 1 .90 9.15E-03\ 17 EQA 14 1.35' 1 - - - EQA-12 135 2 - -

                                                                                                 '\        -

1 EQA-13 /1.35 1 1.35 5.60E-02 'N 27 i EOLP-12 / 1.35 1 1.35 5.60E-02 53 ) EOLP-I V 1.35 1 1.35 1.55E-03 15 \ EQRyR 1.91 - - - - EQML-3 2.59 1 2.59 5.60E-02 29 EQML-2 2.59 1 2.59 5.60E-02 17 I Approved Design Meterial ProbabiEstic Risk Assessment Pope 19.724

I i l

                                                                                            .rrem 19           ,

ituv of l System 80+ ) / Desian contrat Document Table 19.7.5.4-6 Summary of HCLPFS for Seismic Sequences for Soll Site B3.5 Sequence HCLPF Dominant Cutset Mixed HCLPF Failure Prob. Dominant Cutset EQSTR .73 - 1.25 2.50E-01 - 1 EQA-15 - -

                                                              .74            1.00E-02              1      /

SEIS-SBO N .89 26 .12 2.10E-04 100 / EQLP-8 \ .89 44 .89 5.00E-02 151/ EQA 10 \ .89 6 .97 1.19E-03 f2 EQLP-4 N.90 61 - -

                                                                                              /-

EQLP-7 .90 4

                                                                                           /       -

l EQSLO-3 .90 '\ 71 - -

                                                                                         /         -           1 EQSLO-6              .90    \         7                 -                 -
                                                                                     /             -

l EQA-3 .90 \ 80 -

                                                                                 -/                -

EQA-4 .90 ' N6 -

                                                                               /-                  -

l t EQSLO-9 .90 Ids -

                                                                            /-                     -

( EQLP-3 .90 24 \ -

                                                                         /       -                 -

! EQT-3 .90 80 \ -

                                                                      /          -                 -
                                                      \

EQT-7 .90 3 -

                                                                  /              -                 -

l EQA-7 .90 2 \ .9/ 3.25E-02 58 l EQA-8 .90 67 \/>0 3.25E-02 174 I EQSLO-10 .90 17 / \90 3.30E-03 64 EQT-4 .90 100 / 1.N 5.60E-02 138 j EQT-8 .90 76 / 1.10 \ 5.60E-02 142 l EQLP-5 1.35 4 / -

                                                                    \            -                 -

l l EQSLO-4 1.35 6/ -

                                                                      \          -                 -

I l EQA-5 1.35 2)"'

                                                                         \       -                 -

EQT-5 1.35 ./5 -

                                                                           \     -                 -

EQA-12 1.35 / 2 -

                                                                             \-                    -

EQA-14 1.35 / 1 -

                                                                                \                  -

EQA-9 1.35 / 3 .74 9.15L43 79 EQLP-9 1.35 / 5 .89 9.15E-03s 82 EQT-9 1.3/ 6 .89 9.15E-03 \ 71 ! EQSLO-11 1/35 11 .90 6.40E-02 \ 80 EQSLO-7 /1.35 3 .90 9.15E-03 \ 17 EQA-13 / 1.35 1 1.35 5.60E-02 ' 27 EQLP-12 / 1.35 1 1.35 5.60E-02 53 EQLP-11 / 1.35 1 1.35 1.55E-03 15 EQRVR 1.72 - - - - EQML-3 2.59 1 2.59 5.60E-02 29 I i EQML-2 2.59 1 2.59 5.60E-02 17 ,I i Amved Design Matanal- PmbaMstic Risk Assessment Page 19.7-25 i

                                                                                    \

I r' m 9

                                                                                                                                               /%ev oI System 80 +                                                                        /                            Design ControlDocument Table 19.7.5.4-7             Summary of HCLPFs for Seismic Sequences for Soil Site B4 C

Sequence HCLPF Dominant Cutset Mixed HCLPF Failure Prob. Dominant Cutset EQSTR\ .73 - 1.25 2.50E-01 - EQA-15 \ - -

                                                                             .74                  1.00E-02                                  1 /

SEIS-SBO s .89 26 .12 2.10E-04 100' EQLP-8 \.89 44 .89 5.00E-02 )$1 EQA-10 . 89, 6 .97 1.19E-03 /42 EQLP-4 .90 \ 61 - -

                                                                                                                                     /      -

EQLP-7 .90 \ 4 - -

                                                                                                                            /               -

EQSLO-3 .90 \ 71 - - ' EQSLO-6 .90 \7 - -

                                                                                                                   /                        -

EQA-3 .90 \80 -

                                                                                                        -/                                  -

EQA-4 .90 ' 74 -

                                                                                                        -/                                  -

EQSLO-9 .90 10 \ -

                                                                                                    /-                                     -

EQLP-3 .90 24 \ -

                                                                                               /-                                           -

EQT-3 .90 80 \ -

                                                                                          /             -                                   -

EQT-7 .90 3 ,

                                                                                       /                -                                  -

EQA-7 .90 2 \ / 3.25E-02 58 EQA-8 .90 67 \ / 3.25E-02 174 EQSLO-10 .90 17 \/ 3.30E-03 64 EQT-4 .90 100 /\ 5.60E-02 138 EQT-8 .90 76 / \ 5.60E-02 142 EQRVR .91 -

                                                                       /      -
                                                                                    \                   -                                  -

EQLP-5 1.35 4 / -

                                                                                       \                -                                  -

EQSLO-4 1.35 6 / - N - - EQA-5 1.35 21 / -

                                                                                              \         -                                  -

EQT-5 1.35 5 / - N- - EQA-12 1.35 2/ -

                                                                                                        -\                                 -

EQA-14 1.35 /f --

                                                                                                        -\                                 -

EQA-9 1.35 /3 9.15E-03 \ 79 EQLP-9 1.35 / 5 9.15E-03 \ 82 1 EQT-9 1.35 / 6 9.15E-03 \ 71 EQSLO-11 1.35 / 11 6.40E-02 ) 80 1 EQSLO-7 1.35/ 3 9.15E-03 17 EQA-13 1.35 1 5.60E-02 27 EQLP-12 /1.35 1 5.60E-02 53 EQLP-11 1.35 1 1.55E-03 15 l EQML-3 / 2.59 1 5.60E-02 29 EQML-2# 2.59 1 5.60E-02 17 Approvent Deaign Meterial. Prebehmseic Misk Asseesment Page 19.7 26

i System 80+ DCD - Potential Design Changes item Number: 20 Summary

Description:

Clarify time for core uncovery following Intersystem LOCA l ! l Affected DCD Sections: CDM: None ' ADM: Section 19.I1.5.4.6.1 Channe

Description:

l Section 19.11.5.4.6.1 is revised to read as follows: l l  ! 19.11.5.4.6.1 RCS Response Characteristics l i The intersystem LOCA represents a large LOCA initiated outside of containment. Consequently, the  : RCS response is similar to that of the large LOCA discussed in Section 19.11.5.4.2. In this case the l Shutdown Cooling System line break is equivalent to 1.4 square feet. This larger failure area results in a ! i more rapid RCS response. The ECCS maintains the RCS covered until the IRWST is depleted and  ! suction is lost to the ECCS pumps. In this event a sustained core uncovery begins at 7700 seconds (see ' Figure 19.11.5.4.6.1-2 and Table 19.11.5.4.6-1). Core support plate failure occurs at 13,800 seconds and failure of the reactor vessel is predicted to occur shortly thereafter. l The large failure area results in a rapid system depressurization to near atmospheric pressure which is l sustained for the duration of the transient (see Figure 19.11.5.4.6.1-1). A summary of key transient parameters is provided in Table 19.11.5.4.6-2. l l l i i r j dedchng4. doc 6/26/96 i

Trea z O System 80+ oesian controlDocument Figure 19.11.5.4.5.1-6 shows that the cavity basemat erosion is insignificant (less than 1 inch) during this transient. This is due to adequate quenching of the core debris in the reactor cavity. 19.11.5.4.5.3 Fission Product Releases A summary of fission product group concentrations in'the containment atmosphere at 24 hours after vessel breach is provided in Table 19.11.5.4.5.1-3. 19.11.5.4.6 V Sequence The dominant System 80+ V Sequence consists of an intersystem LOCA (ISL) initiated from a full shear break in the 16" diameter SCS line occurring within the containment building subsphere. This l l event is identified in the PRA as PDS 17. In this event all ECCS systems are operable. The failure of the SCS pipe outside of containment results in a gradual transfer of ECCS inventory from the containment to the subsphere. This ultimately results in failure of the ECCS function due to the unavailability of a water source. Details of this transient are discussed below. 19.11.5.4.6.1 RCS Response Characteristics wn se u nom The ISL represents a large LOCA initiated outside of containment. Consequently the RCS response is similar to that of the large OCA discussed in Section 19.11.5.4.2. In this case the SCS line break is equivalent to 1.4 s feet. The larger failure area resul a more rapid RCS response, inthis event the core itially uncovers in 76 seconds (S r example Figure 19.11.5. .- and Table 19.11.5.4 - ). SIT discharge rapidly tempo recovers the core. l ECCS maintains the RCS ered until the IRWST is pumps. A second sustain re uncovery begins at 2 hours and suction is lost to ECCS upport plate failure occursp,800  ; seconds and RV fail is predicted to occur shortly there r. J The larg ilure area results in a rapid system ressurization to near atmosph se pressure which is sustained for the duration of the transient (Fjgife 19.11.5.4.6.1-1). l

                                                                                       ~

A summary 15f key transient par vided in Table 49T1.5.4.6-2. l 19.11.5.4.6.2 Containment Response Characteristics The ISLOCA releases all the RCS and conuinment liquid inventory into the building subsphere. Once the RV fails the corium is assumed to fully drop into the dry reactor cavity. Core concrete attack begins immediately. Concretes erosion will ultimately lead to a basemat failure. However, the bypass pathway provides a more direct means for releasing fission products to the environment. These MAAP analyses do not credit the water accumulation expected in the subsphere ECCS rooms to scrub fission products leaving the RCS. Furthermore, detailed revolatilization models including the large length of SCS piping are likewise not considered in this demonstration. . 19.11.5.4.6.3 Fission Product Releases MAAP predicted fission product releases for the V sequence are summarized in Table 19.11.5.4.6-3. l Annreved Design Meterial- ProbaMstic Misk Assessment (2/96) Pope 19.11 145

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