ML20101G217

Forwards Intersystem LOCA Evaluation

ML20101G217
Person / Time
Site:	05200001
Issue date:	06/05/1992
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C, Rubin M, George Thomas NRC
References
NUDOCS 9206250344
Download: ML20101G217 (31)

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f-GE Nuclear Energy ADVANCED REACTOR PROGRAMS Mechanical Systems Design i

i June 5,1992 TO:

J.N. Fox cc: A. J. James FT,G t W. B. Taft and A. Sallman

SUBJECT:

INTRASYSTEM LOCA EVALUATION RITERENCE:

Dino Scaletti, NRC to Patrick Marriott, OB

" Identification of New Issues for the General Electric Company Advanced Bolling Water Ret.ctor Review,"

September 6,1991 The following material provides supplemental background and basis

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information to the SSAR secdon 19B.2.15 on intersystem LOCA.

l Attachments 4 and 5 provide revised text and a new table respectively, i

that will be included in the SSAR.

An intersystem LOCA is postulated to occur when a series of failures or inadvenent actions ccur that allow the high pressure { rom one system to be applied to the low pressure design of another system, which could potentially rupture the pipe and loor, coolant from the reactor system pressure boundary.

This may also occur within the high pressure and low pressure piping of one system. Future ALWR dulgns like' the ABWR l

should reduce the possibility of a LOCA outside the containment by desig71ng to-tho' extent practicable all systems and subsystems connected to the RCS to ar.11 tim:.te rupture strength (URS)-at least equal to the full RCS pressure.

'f7.e ultimate rupthre strength (URS) criteria was recommended by the reference letter to determine low pressure piping's design pressure to contain intersystem LOCAs.

The reference to URS occurs in a section on intersystem LOCA in the: second paragraph.

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6/5/92 3

Attachment I shows the equation and ratios basis used to relate the ultimate stress to the design pressure.

The first equation on Attachment i for the minimum wall thickness of a pipe is obtained from a : etion of the

• Ud" Tbde shown on Attachment 2 labeled as " Pipes."

The remainder of Atthenment 2 shows that the general membrane stressef for vessels, pumps and valves is equal or less than the allowable stress in all cases.

This means that the basic ratio developed in Attachmenf 1 also applies to vessels, pumps and valves as shown by Attachment 2, I

The ASME Code materials for the various components of the enginected i

safety features systems can be found in Table 6.11 df the SSAR. With the -

code in materials known, ASME Code Table I-7.1 can be used to find the particular material and its allowable stress and ulthw:' stress.

With these two stress values, a design pressure can then be deteor.aed by the ratio indicated in Attachment 1, For the k. ytem, a dr. sign pressure of 300 psig was selected acceptable from the calculated 256 psig value based on i

j the carbon steel materials SA333 Or 6 or SA672 Or 70. The 300 psig j

design pressure is also applicable for the other carbon steel piping systems evaluated.

Documentation that indicates that the URS criteria also applies to associated flangen, connectors, packings, valve stem seals, pump seals, heat exchanger tubes and valve bonnets, as required by the reference letter, is given by the following documents:

Document 1:

"American National Standard Forged Steel Fit;ings, Sockets - Welding and 'Ihreaded," ANSI B16.11-1973, Sections 6.2 and 9.2.3.

Document 2:

"BWR Owners Group Assessment of Emergency Core Cooling System Pressurization in Boiling Water Reactors," NEDC-31339.

November 1986, Sections 3.2.2 and 3.2.3, on valve integrity and heat integrity, respectively.

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Ten systems were evaluated as follows:

Residual Heat Removal (RHR) System High Pressure Core Flooder (IIPCF) System Reactor Core Isolation Cooling (RCIC) System Control Rod Drive (CRD) System Standby Liquid Control (SLC) System Reactor Water Cleanup (CUW) System Nuclear Boiler (NB) System Reactor Recirculation (RRS) System Makeup Water (Condensate) (MUWC) System, and Maksup Water (Purified) (MUWP) System.

A simplified P&ID diagrams of the various systems are provided by.

The P&lDs contain the following features:

a.

Boundary symbols for design pressures b.

Valve numbers at the boundary limits c.

Pipe diameters d.

Interfacing systems The clouded marked regions on the figures show the piping and equipment that were upgraded. indicates the revisions to the SSAR text portions in section 19B.2.15. is a new Table 19B.2 2 for the SSAR, which defines the boundary limit of the URS design for the systems.

This table contains the primary detailed information identifying the extent of protection to postulated intersystem LOCA conditions.

Key to understanding the formulation of the table is the assumption as stated in tl~ SSAR text (section 19B.2.15, new item (7)) that, "Only static pressure conditions are considered by assuming that the valve adjacent to a low pressure sink remains closed.

Thus, the violent high flow transients are precluded that would occur if the full RCS nressure was connected directly to the low

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6/5/92 pressure sink."

The components considered to be low pressure sinks in this evaluation are the:

suppression pool, condensate storage tank, SLC main tank, SLC test tank, I

LCW funnels, and LCW storage tank.

The following sections present an evaluation and proposed changes for the ten systems interfacing directly or indirectly with the RCS, f

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1,0 High Pressure Cora Floodct (HPCF) Syster Figure 1 (Attachment 3) gives a schematic of the HPCF system showin6 the j

various system design pressures and the interfacing systems.

The design improvsmant proposed is to upgrade the system design pressure from 14 ats (200 to 21 atg (300 pai ) for the fo11owin5 lines and equipment within these psig) 5 lines:

(a) Pumps 122-C001B&C suction lines starting from the intet of the valves E22 F006B&c, and upto the pump inlet, including valves E22 T007s&C and the test lines, h dasign pressure of piping upstream of talves E22 F006B&C is not changed 1,s., kept at 3.16 atg (45 poig), because it opens in suppression pool.

(b) Relief valves E22 r020B&C including their inlet lines. The desi n 5

pressure of relief valve dischar5e piping is not changed i.e., kept at 3.16 at5 (45 psig) because it discharges into the suppression posi.

(c) Pump suction lines from the condensate storage tank starting from the inistofvalvesE22F001B&C,andextendingupto-thetesinithsauctionline from the suppression pool. The design pressure of the piping upstream from the valves F001B&C upto the condensate storage tank is not changed i.e., kept 3

at 14 ats (200 psig) because it is the last valve to be closed in the flow path of the reactor coolant to the condensate storage tank.

2.0 Reactor Core Isolet',on Cooling (ROIO) system Figure 2 (Attach =ent 3) gives a schematic of the RCIC system showing the

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various system design pressures and the interfacing systems. The design improvenant proposed is to upgrade the system design pressure from 14 atg (200 psig) to 21 at5 (300 psig) for the following lines including the equipment that is present in these lines:

(a) Pump KSI 0001 suction line starting fr-m the inlet of the valve E22-7006 l

and upto the pump inlet, including valve E51 F007 and'the test line. The design pressure of piping upstream of valve E51 F006 is not changed i.e., kept et 3.16 ats (45 peig), becausa it opens in suppression pool.

(b) Relief valves E51 T017 including their inist lines, h design pressure of relief valve discharge pipin5 is not changed i.e., kept at 3.16 ats (45 psig) because it discharges into the suppression pool.

(c) Pump suction linas starting from the inlet of valve E51 F001, and extending upto the tee in the euction line from the suppression pool. The Jesign pressure of the HPCF piping upstream from the valves 151 F001 is not changed i.e.

kept at 14 aeg (200 psig) because it is the last valve to be closed in tne flow path of the reactor coolant to the condensate storage tank.

(di condensate pump, its discharge line, branch line, and all components in tne lines.

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3.0 Control Rad Drive (c1D) System Figure 3 (Attachment 3) gives a schematio of the CRD system showing the various system design pressure and the interfacing systems. Listed below era the features of the system that will prevent an intersystem IDCA:

1.

A single CRD pump.is continuously operating during normal plant operation et a pressure higher than the reactor pressure, and unjecting pur5e flow into the vessel throu6h the RIP's.

1 2.

Three check valves are to fail open before high pressure reactor coolant can lead into the low pressure piping (suction piping of the pump), provided both CRD pumps stop. However at least one pump is alwaye operating during normal reactor operation, preventin5 reverse flow of reactor coolant into tne CRD system pipin5' Howevar, with of the presence of the above features, the low pressure piping (1) at the upstream side of the CRD pumps, (2) at the downstream side of valve C12 F017 and (3) at interf aces with MWC system (at valves C12 F022 &

C12 F023) which has a design pressure of 14 ats (200 psig) is retained considering the multiple 1bvel of protection provided, cnd the open path to the condensate storage tank frob the low pressure region.

Therefore no system design pressure, changes from the current design are considered necessary for the CRD systa=.

4.0 St odby Liquid control (sLC) system Figure 4 (Attachment 3) gives a schematic of the SLC system showing the varioussystemdesignprosjuresandtheinterfacingsystems. The design improvemant proposed is to uP5rade the system design pressure from 14 a*;g (200 psig) to 21 atg (300 pois) for the following lines including the equipment that is present in these lines:

(a) Pumps C41 C001A2 suction lines starting from the inlet of the valve C41 T001A&B, upto the pump inist. including valves C41 F002A53, and the interconnecting line between the two suotion lines. The design pressure of piping upstream of valves C41.F001A&B is not changed i.e., kept at static head in the P C S tank, because it opone in SLCS cank.

(b) Ralief valves C41-F003A&B discharge lines.

(c) Valve C41 T012 and its downstream piping, relief valve C41 F026 and 1,:s inlet line, and branch line containing valves C41 F014, C41 F020, and C41 F013 upto the interfees with HWP system, branch line containing valve C41 F015 upto the downstream end of valve C41-F023.

5.0 Residual Heat Removal (RER) System i

Figure 5 (Attachment 3) gives a sche =atic of the RHR system showing the various system design pressures and the interfacing systems. The design improvemant proposed is to upgrade the system design pressure from 14 ats (200 psig) to 21 at5 (300 peig) for the following lines and equipment within these lines:

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6/5/92 (a) Pumps Ell COOLA,B&C suction lines starting from the inlet of the valves E11-F001A,5&C upto the pumps inlet; pumps E11.C002A B&C suction line; and the take of line connecting the high conductivity waste (HCV) system including valve E11.F026A,5&C. The design pressure of piping upstream of valves E11 F001A,B&C is not changed 1.e., kept at 3.16 ats (45 psig), because it opens in suppression pool.

i (b) Relief valves 111.F042A,5&C including their inlet lines.

The design i

pressure of relief valves E11 F042A,5&C and E11.F051A,B&C discharge piping is not changed i.e., kept=at 3.16 ats (45 psig) because they discharge into the suppression pool.

(c) Shutdown cooling suction lines downstream of the outboard isolation valves E11 F011A 5&C upto the tee connection to the suction lineo from suppression pool (valves E11 F012A,B&C included); suction lines from the fuel pool cooling and cleanup (FPC; system valves (valves G41-F029, G41.F031, E11-F0165&C included); lines from the valves E11.F040A,560 (valves included)

(d) Fire protnotion system interfacing check volve E11 F100, and check valve E11 F104 in the piping from outdoor fire truck connection; downstream piping upto inist of valve E11 F101. Note - the design pressure of these lines and valves is being changed from 16 ats (228 psig) to 21 atg (300 psig),

Reaccor Vater cleanup (CUV) gystem Refer to CUV system P&ID in ASVR SAR Fi6ure 5. -12 The CW is a hi h 5

pressure system having a design pressure of reactor pressure or hi her, with 6

the exception of the following lines:

(e) Downstream pipt"5 from valve G31-F025 having a design pressure of 10 at5 (143 psig). TM s line provides an open path to the suppression pool.

(b) Downstream piping from valve C31.F023 having a design pressure of 10 atg (143 psig). This line providen an open path to the LCV collector tank.

(c) Downstresa piping from the relief valve 031 7020 having a design pressure of 10 ats (143 781 ),

This line provides an open path to the LCW collector 5

tank.

(d) Portions of drain and vent piping downstream of the last drain / vent valve that disharge into the LCW funnels.

i No changes in the CW syscam design pressures are considered necessary.

Nuclear Boilar (NB) System Refer to N3 system P&ID in A3'#R SAR Figure 5.13.

The NB is a high pressure system having the steam / water lines design pressure higher than the reactor Therefore no changes in the system desi n pressures are considered 5

pressure.

necessary.

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Esactor Recirculation (LLS) System Refer to LRS syste: P&ID in A3VR SAR Figure 5.4-4.

The RRS system piping connected to the reactor have a design pressure of 87.9sts (1250 pois) or higher with the exception of the following:

(a) Pipin5 d wnstratm of vent / drain valves B31 F501A H, 831-r503A H, desL n pressure of 0 atg 531-F506A-H which open into LCV funnels have 5

(atmosphario).

No chan5es in the system desi n pressures are considered necessary.

5 Maksup Vater (Condensate) (WUVC) System &

Makeup Vater (Purified) (ENP) system Refer to MWC system P&ID in A57R SAR Tigure 9.2 4, and E'VP systa= P&ID in A3VR SAR Figure 9.2 5.

These systa=a do not interface directly with the reactor. The MUVC system interfaces with HPCF, RCIC, RK1. CRD, and CW systems. The.MWP system interfaces with RRS and SLC systems. The desi n 5

pressure of MUVC 6 MWP systems at these interfaces is 14 atg (200 psig). In order to upgrade the system design pressure from 14 at5 (200 psi )

to 21 *C8 8

(300 peig) to bring it upto the design pressure corresponding to the ruptura strength for piping pressurized to reactor pressure, it will require piping / valve pressure rating change from ANSI 150 lb to ANSI,300 lb.

An approximate cost impact of $300X has been estimated for upgrading the pressure rating of both systems.

In considaration of this high coat impact versue the benefit achieved, it is not practicable to reduce the pressure challange any further i.e., into the MWC and MWP systems. Therefore no changes vill be made in the MWC and M'Jw'P system design pressures.

4 9

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6/5/92 i

i NRC REACTOR SYSTEMS BRANCH HEETING SAN Jose, Nov. 20-21, 1991 RHR SYSTEM tm =

PDo

+ A 2(S+Py) tm = (Desian Pressure)(Do1 + A = (Hich Pressure)(Dol + A 2(AllowableStress) 2(UltimateStress)

High Pressure = (Ultimate Strats)(Desion Pressure) i (Allowable Stress) 800psig=f.QX(200psig)

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Design Pressure = (Allowable StressI(Hich Pre 15EA1 (Ultimate Stress) 256 psig = liiX (1025 psig) 1 60K l

Round up; Consider 300 psig minimum desine pressure l

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ASME Code References i

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l NRC REACTOR SY5TEMS BRANCH HEETING v.o SAN Jose, Nov. 20-21, 1991 P4 Il RHR SY5 TEM

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wher r.= minimum required wall.thi:kness 15 If pipe NC.3640 PRES 8URE DESIGN OF PIPING PRODUCTS la ordered by its nominal wall thickneu, the NC-3H1 Straight Pipe r= internal Dwign Pnasure. W D,= outside diameter of prpe, in. For design cal.

NC 3H1.1 Straight Pipe Under Internal Pressure.

The minimum thickness of pipe wall required for De-S= $.aximum allowable stress for the material at sign Pressures and for temperatures not neeeding the Design Temperatuts, pai (Tables I 7.0) these for the varsous materish listed in Tables 17,0, A = an additional thl kness to provide for material including anewances for mecha,nical strength, thall not rarnoved in threading, corteston or erosion be less than that determined by Eq (3) u follows:

allowance, and material required for strue.

PIPES fo, y-a cosmeient having i nlue of 0.4, uespi that,

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for pipe with a D,/r., ratio less than 6, the value of y shall be taken u (6) 7"g p,

I c'." seetfal m8mbftn8 8 Tee 8. Psi. This stress in NC 3321 198f 5ECTION III, VES SEL equal to the averise stress across the solid nu conddersh It aseW h TABLE NC D211 continuitim and concentrations and is pro-STRESS L!tlITS FOR ORIIGN AND SERViCK LOADING 5s duced only by pnasure and other mechanical swese unvt stmi t)wu (Neu (2))

g,,,,,g, ',,*8y#

S= allbwable strene value given in Tables I 7.0, pal The allowable strees shall correspond to the highest metal temperature at the section unds conalderation during the loading under consideration.

NC4418 NC 3000 - D1110N NC 3423 PLLMP S TABLE NC-34161 KTRES$ AND PRE 53URE LIMITS FOR DE31GN AND 5ERVICE LOADING 3 SeMce Stresa Undts F.,

UpWt (Neu (1))

(Meu t131 Level A r, g 5 1.0 tr. e ed + r. s 1.33 NC 3830 1949 3RcrION III. DIVISION 1 - NC NC-3831A I

TABLE NC-35211 LEVEL A 4, C. AND D SERVICE UMIT5 SeMte Stress Umha 7,,,

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20 19B.2.15 High/ Low Pressure Interface Design RHR Sucion Valve Tuting [M]

RCS/RHR Suetico Lino Valve Interlock on PWRs l99}

Interfacing Sptam LOCA at BWRa (105)

OrJ!ns Tutabillry of Protection Sptsms p20) l Iuus Summary In all currently operatles light water reactors, there are a number of hi h/lew preuurs Interfaces between l

the reactor coolant pressure boundary and connected l

systems. This leads to the situation that systems in both BWRa and PWR un dealgned for a ptsuure tour than that of the primary system. For sxample, i

the PWR primary system operatas at about 2150 psig while the ruldval heat removal (RHR) system and related piping is dealped for an operating preuure of 600 pal or less. I:el3tica valves, at leut two, and l

piping to the primary system are designed for 2500 psig. As taother example, the BWR primary system operates at about 1000 paig wbIle the RHR ayatem can operate at pressures up to.500 pelg. Isolation valves, at lesst two, and piping to the primary system are des!gned for about 1250 ps!g. The discharge of tha BWR RHR system, which also functiona u a low

- preuurs lajection system, panes through testable check valvva prior to returning to the reactor coolant system.

The common concern In the above issues la that l

either isits that require Wys actuadon, vslve leakage, or multiple valve failures could result in a system j

pressure which exceeds the design preuurs oflow preuure emergency coollag or decay hest removal Amnomua 11 IfBM

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ABM PacPRIETARYlh70RMAT10N matocAR Standard Nant cas m na, A systema, causing them to fall from onrpreuure. For piled for d valves and piping from and Includ.

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the decay heat removal system (RHR system), the

. sg the outboard ! solation valve to the veuel, dealgn of the h!gh/ low pecuurs interface is mads Pressure laterlocka to protest low pressure more complax because an actuation of isolation piping during taulag and operadon shall also be system (ludverten vsjve closures) may cause louss und. (Su Appendix B, Section BJ.)

of decay host removal capability which causa additional plant chdanges.

(2) Overprouurs Protecdon; Chapter 5, Secdon 4J, 3.3.2,Rev.0 hk calculadens on existing' plants suggests there may be a need for improwd protection againt The portions of the system downstream of the the potential for overpreuerlzition of some emer.

pump sucdos shall be dulgned for a preuurs of gcccy cooling and decay heat removal systsma.

500 psig at a tampersture of 360 P. Thou por.

t!ans of the system upstream of tbs pump suc.

ALWR Resolutica gumtmary tics and those pordons of the s) stem which have a free discharge which cannot be blockad by a-The primary pretsction of the high/ low pres.

valve shd be dealsned for 200 psig at a temper.

sure interface of the ALWR is through the selsstion sture of 360 F.

and use of high qudryisolation valves. System later.

locks and permiuivsa have been selseted consistent Enginsering Rationale with the system safety functions. la addition, a sps, cific requirement has been included for both BWRs This design pressure provides a means of avoid, and PWRs to assure that the ultimate rupture lag structural failure in the ualiksly event the strength of the spum will not be escoeded syss if it systse la laadvertently supond to the primary is exposed to full remor coolant system operating system when at operatlag prsuurs. (See Ap.

preuurs. Busd on liase requirements, thus iussa pendir E, Section BJ.)

should be considered resolved for the ALWR.

(3) Interlocks; Chapter 5. Sections 4.5,3.5.4.1, i

BWR Requirements in the EPRI.ALWR Asquire.

4.3.3.3.4.2 and 4.5.

4J, Rev. 0 ments Documsat A two.way interlock shd be provided so that it (1) High4.ow Ptsuure Interface; Chapter 5, Ssc-la not poulble to havs sa open shutdown con.

l tion 4.2.3J Rev 0 nettien to the vmust la any given loep wbsesvar the poo! suction, pool discharge valvs, or con.

An inboard testable check valve and an out.

tainment spray valves are open in the same loop.

I board motor operated valve shall be used on j

the DHR injection llass. H1 h rascurs piping Engineering Radonals 6F and valves shall be used up to the outboard iso-

!ation valvs. Freucts Intsrlocks shall protscs To prewet drawing down ramctor veuel level by low precaurs p! ping upstream of the outboard flow to the suppresalon pool. (See Appendix 8, Isoladon valve.

Seedon BJ.)

Englaunng Radannt.

Secdon 4.5.3J,4.2 i

Two motor operated valvss ars isnardy pro.

Redundant latarlocks shall prevent opsabg As i

vidad for containment isolation To improve aburdown connections to and from the vens!

i system tallability for injection and to previde whenever the pressurs is above ths shutdown automade and continuous protserlan against range, Increasing prsuure trip shall causs clo.

ovstprenure of low prsuure interfacing sys.

sure of thsas valvsa.

tenu by inadvertent openbg of motor operated valves, an inboard testable check valve can be Er.gineering Rationals i

l used. For low pressure systems consseted to the vanel, a vessel oresure rating shall be sp.

To prever: exceeding design prusure. (See Ap-pendix B, Section B.S.)

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Section 4JJ.5,4J W ; ' ';;..z' _ r ;_2.. i ? d S :

j Redundant laterlocks sha!! prevent opealms or j

will elone the shutdown sucelos consecsleas to (3)

Isolation of the high/ low pressute systems la j

the vessel la any shea loop whenever a low reae.

=la'el=d during vulw tasting.

tor level alsnalis pressat.

1 laolation of the kl h/le* Pressors systems is (4)

S i

Engineering Radona!s malatained under the ecadition of as laadvert.

ent opening of a valve due to na electrical fail.

As a backup to prevent unplanned draw den of ure.

the reactor vesselwater. (See Append!x B, See.

don BJ.)

(5) A!.WS requirements imposed on ABWR for j

high/ low pressure laterface design for AHR, (s) Number, Type and Locados cf Isolation Valmst MPCF,5LCS and RCIC systems we ast.

1 Chapter 5, Secdon 6.2.2.L1, Rev. 0 j

(6) The system dulga pressures requirements (m.

Acceptable a:.s s for location provisions la lines posed by ALWR,l.a.,500 psig for low pressure which penstrats the containment boundary shall piping in RHR system and 200 psig on auction be in accordance with ANS 56.2, Section 3.6, A\\ side of the pump for all four systems, us met.

'Other Defined Bases

• and Regulatory 0ulde V /==== Sg g /ptg47 L14L

,pT)* The over A conclualon is that the con

(,g tified in OSI 105 *!sterfacias System LOCA at Engissering Radonale BWRa,' for ABWR are resolnd.

l GDC 53,66 and 57 idsadfy specific requiremania 198.2.16 Design of ABWR Water Lewi for lines which pesstrate the contalement Instrismentallon boundary and arc (1) part of the reactor coolant j

pressure boundary, (2) connested diredy to the Break Plus Clagle Failure in BWR Water LevelInstru.

J containmnnt atmosphere, or (3) sanaaeted to a mentadon [101)

I closed system inalda contalassat, respectively.

(See ANS 36.2 for dallaldeas,) The ODC permit Inue Summary alternative approaches if justifled. ANS 56.2 has been found acceptable to the NRC. It defines A NRC concern has been Idsstifled for BWR alternative isolation bases for various categories about a postulated break in an instrument lins in con.

of !! ass such as lustrument lines and !!nes of junction with the worst alngle failure. The concern is ensincered safety.cquipment. (See append!x B, that if one of two reference columns breaks a singts 5scden B.J.)

fa!!ure assoelated with the other refersace column sov!d comp!stely defeat mitigation systems for the ABWR Resolut!os resulting tranalent.

Based on the RHR, HPCF, SLC5 and RCIC NRC Resoludes Summary systes dulga and tutl*5 procedure evaluation from the point of view of laterfacing system LOCA and This !sste was resolved by'the retsase of Ocneric overpressurisation of low preuurs systems, the fol-Letter 4911 on June 30,1989. No licenses actions lowing conclusiona ar6 reached:

were warranted.

(1) b low preuuro pordens of the system are ade.

The letter states la part that the staff bellevos that quately protected from high pressure during smargency prendures for the operator to ideadly and normal plant operation.

mitigste the consequenen of lastrument !!nc breaks

$35 Myypppy exists si all plaats and that the reactor operators are (p,-7....u.~.6.." r; M i.

being trained to ach!sve safe shutdown,11 undsd. The

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design basis for this conclusion h documented in NUREO/CR.3112 pub!1shed in March,1989.

Amendment il 19BJ4

c-p Fox 23 6/5/92 i

INTERSYSTEM LOCA DRAFT SUBMTITAL i

4 Revision for SSAR section 19B.2.15, ABWR Resolution, (2) t l

(2)

Proper interlocks for the RHR injection valves (E11-F005) are l

provided that allow operability testing of-the testable check valves f

(El: F006) during normal plant operation or under cold shutdown-l conditions.

Operability testing cf the RHR injection valves will be.

l performed only during plant shutdown conditions.

1 INTERSYSTEM LOCA DRAFT SUBMTITAL Insert into SSAR section 19B.2.15, ABWR Resolution, (7), (now material)

I (7)

Furthermore, the pump suction _ piping design pressure --exceeds the o

ALWR 200 psig requirement-and is established at-300 pois to assure the piping can withstand the full reactor. operating pressure based on ultimate rupture strength (URS) design criteria.

The 300 psig design 3

i pressure.is established to reduce the_ possibility of a LOCA outside containment by. designing to the extent practicable to' an URS at least equal to full reactor coolant system-(RCS) pressure. : This is _ achieved -

4 by assuring that the design pressure exceeds the ratio of-ASME Code-allowable. stress' to the L ASME Code ultimate -(rupture)- stress multiplied by the full RCS pressure.(1025 psig).

The design margin-represented by this ratio also provides' an-adequate margin to-

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sonu an w e,titi 06-09-92.09: 44 AM PO4

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D Fox 24 6/5/92 rupture for associated components, such as the following examples, valve bodies, pump casings, flanges, packings, valve stem seals, pump seals and valve bonnets.

Only static pressure conditions are considered by assuming that the last valve adje. cent to a low pressure sink remains closed.

Thus, the violent high flow transients 1

are precluded that would occur if the full RCS pressure was connected directly to the low pressure sink.

The extent to which the URS design criteria is applied to the systems that interface with the RCS and the supporting methodology are identified by Table 19B.2 2.

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) DESIGN

. INTERSYSTEM IDCA BOUNDARY LIMITS FIME ULTIMATE RUITURE STE9eCTH -(IRS Table 195.2-2 Limit' of IRS BAndser of valves Pressure Isolation Systems with direct Azij acent interface to the interfacing practicality; last in pathway to valves, and valve with IRS limit covered by criteria satisfied reacter coolant systems.

TRS design (1) application system (RCS)

.4 2(a,h,c).4(a)

RER N/A Ell-FOOI.

(3)

RER.

N/A E11-F008, (3) 3 2(a,b,c),4(a)

RIM N/A E11-M31, (3) 4 2(a,b.c),4*a)

RIE N/A

'R11-F028, (3) 3 2(a,b,c),4(a)

, j RIE N/A E11-M51, (3) 3 2(a,b.c),4(a)

N RIR -

N/A E11-F042, (3) 3 2(a,b.c),4(a) 4 2(a,b,c),4(a)

- RHR.

NUUC K11-F032, (5)

RER NUWC E11-F040, (5) 4 2(a,b.c),4(a) 4 2(a,b,c),4(a)

.RHR FFC Ell-M15,

'(5)

ElR FFC-G41-F029 (5) 4 2(a,b,c),4(a) l FFC G41-F031,.

(S) 4 2(a,b c),4(a)

-]

. RIR RHR '.

FCS T49-F010, (5) 4 2(a,b,c),4(a) h RHR

'FCS

,T49-F011, (5) 5 2(a,b c),4(a) o, I'

.=RHR FFS Ell-FICO, (5) 5 2(a,b,c),4(a) e i

RER FPS E11-F104, (5).

5 2(a,b,c),4(a) 4 2(a,b c),4(a)

RHR SAM E11-F044, (5).

K17-FOO6, (5) 5 2(a,b.c),4(a)

IJCW.

-RHR RHR HCW K17-Ist valve, (5) 5 2(a,b.c),4(a)

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h INTFRSYSTDt IDCA DOUNDARY LIKITS FDR ULTIMATE RUPTURE STRENGTH (URS) DESIGN Table 19.B.7-2 (Continued)

Systems with direct Adj acent Limit of URS Number of valves Pressure Isolation interface to the interfacing practicality; In patJmay to valves, and lase valve with LES limit covered by criteria satisfied reactor coolant syst -

IRS desirp (1) '

application system (RCS) l HPCF N/A E22-F001A6B (3) 5 2(a,b c),4(b)

IIPCF N/A E22-FD06A6B (3) 5 2(a,b,c),4(b) i

[

HPCF N/A E22-1909A&B (3) 4 2(a,b),4(b)

HPCF N/A E22-F910AAB (3) 4 2(a,b.c),4(b)

N

-J E22-FD12A&B (5) 4 2(a,b,c),4(b) i HPCF HOW E22-F013A&B (5) 4 2(a,b),4(b)

HPCF NJWC 11PCF NURC E22-IV16A&R (5) 5 2(a,b),4(b)

HPCF N/A E22-IV20A&B (3) 4 2(a,b,c),4(b)

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

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t INTFRSYSTD1 -1M 80UNIMRY LIMITS WR ULTIMATE RUPTURE STRENCTH (URS) DESICH Table 19 3.2-2 (Continued)

Systems with direct Adj acent Limit of (RS Number of valves Pressure Isolac. ion l

Enterface to the interfacing practicality; in patinray to valves, armi j

last valve with iJRS limit covered by criteria satisfied reactor coolant syst-a 1

system (RCS) application URS design (1) 1 i -

~

KCIC HFCF Dl-FOO1 (3) 7 2(a,b,c),4(c) i.

I -l-RCIC N/A E51-FOO6 (3) 7 2(a,b,c),4(c)

,' c RCIC N/A E51-FVG9 (3) 7 2(a,b),4(c) 3 RCIC N/A E51-F011 (3) 7 2(a,b c),4(c)

{

-3 RCIC N/A E51-fV17 (3) 6 2(a,b,c),4(c)

N l

00 RCIC MUWC E51-F021 (5) 6 2(a,b),4(c) s RCIC 10UWC E51-IV33 (5) 6 2(a,b),4(c)

RCIC N/A E51-D005 (3) 6 2(a.h.c),4(c) i l

RCIC M/A ~

Condensate pump (3) 7 2(m.b c),4(c) i i

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-t (URS) DESIGN INTERSYSTDi 1DCA BOT @ARY LIKITS FOR ULTIMATE EUETJRE STREttGnt Tabic 19.E.2-2 (comittrmed)

Limit of URS Numsber of valve.

Pressure Isolation Systems with direct.

I.dj acent in patJamey to valves, ased j

l interface to the interfacing practicality; last valve with Uts limit covered by criteria satistled rmactor coolant systems IRS design (1) ayy11 cation systam (ECS) 4 2(a,b).4(d)

C41-IV01AEB ('3)

SLC N/A 5

2(a,b),4(d)

C41-F011 (3)

SZr N/A.

l 5

2(e.b),4(d)

C41-F012 (3)

SIE N/A 7

2(a,b) 4(d)

C41-F023 (3)

- j S14 N/A i

C41-F026 (3) 5 2(a,b),4(d) g 51E N/A 6

2(a,b),4(d)

F11-FOl9 (5)

SIE MUWP E

3 -

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INTMSYSTEM IDCA BOUNDARY llMITS RMt ULTIMATE RIIrTtRE STED8CTH (IRS) DESIM Table 19.B.7-2 (centirmed)

PTXTIMYTES :

3.

Limit established by assuming that at

1. ' Also count the last valve.

least one valve remains closed in the nath to low pressu.1; it then follows 2.

The following thme criteria per that the v.1ve adjacent to a low presssa

?

SECY-90-016 are applicable for cases region (e.g.,a pool or a tank) detersiws where it is lepractical to extend the the URS boundary limit.

a S

URS design pressure region further.

3 (a)

The capability for leak testing 4.

Pressure isolation valves:

of the pressure isolation j

(a) E11-F005A,B&C Ell-FOIGA,B&C

valves, E11-F006A,8&C E11-FCt1A,B&C W

O (b)

Valve position indication that is available in the contrnl (b) E22-F00 N E22-FDO4A3a room when isolation valve operators are deenergized, (c) E51-F004 E51-F005 (c)

High-pressure alaruts to warm (d) C41-F006A C41-F006d contrcl room operators ther, Practical protection is achieved by the rising RCS pressure approaches 5.

number of wormally closed valves; the design pressure of attached extending the URS design pressure further low-pressure systems and both impacts a large region of interfacing o

T isolation valves are nct system.

closed.

?l 2

y 3.

S 4

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