Text

Rev 1 to M97-0097, Low Pressure Auxiliary Spray Flow Rate for Boron Precipitation. W/Simulator Exercise ROT-9-200, Boron Precipitation Control
	ML20202E378
Person / Time
Site:	Crystal River
Issue date:	11/01/1997
From:	Chris Miller, Powers T, Smalec L; FLORIDA POWER CORP.
To:	;
Shared Package
ML20202E372	List: 3F1297-12, Provides Addl Info Re Post LOCA Boron Precipitation Prevention Plan Submittted W/Lar 223.Info Includes Results of Recent Calculations & Addl Details as Discussed W/W Lyon During Week of 971103 & on 971120; ML20202E378;
References
	M97-0097, M97-0097-R01, M97-97, M97-97-R1, NUDOCS 9712080015
	Download: ML20202E378 (27)
	v • d • e

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U.S. Nuclear R:gulat:ry Commission 3F1297-12 Enclosure 1 FPC Calculation M97-0097

" Low Pressure Auxiliary Spray Flow Rate for Boron Precipitation" (Computer output not included)

[d* EE Mih2 ,

. :lorida INTEROFFICE CORRESPONDENCE Nwer A c -' -

Nuclear Engineering NT5A 240-1628 Of fice MAC Telephone

SUBJECT:

Crystal River Unit 3 Quality Document Transmittal Analysis / Calculation A.

TO: Records Management NR2A -

The following analysis / calculation package is submitted as the OA Record copy:

DoCNO FPC DOCUMENT IDENTiric ATcN NuMetsu Rtv sygT(Msi TOT At FAats TRANswirtD M97 0097 TITLE 1 DH,MU 366 Low Pressure Auxiliary Spray Flow Rate for Boron Prccipitation

)

kWDS ttDINTIFY REYWoRDS FOR LATER RETRt!VAu LOCA, Boron Precipitetior., Aux Spray, PIPF-PC, hydraulic model DRREP (REFERENCES OR FILES . '.1$f > RIM ARY Fi(t $6RST' M910092, M90-0021 314M1, E2 M1 M94-0047 l M97-0088 l vryo n ENeoA NAMn veccR DccuvtNt NuMnN iciatri """5 Lot = "L* *" '

FPC NA M97-0097 Rev 0 TAG DHV-91, DHV 92, DHV-93 l BSP 1 A/B RCV 53. RCV 12 DHP-1 A/B MUP 1 A/S/C 1

COMMENTS (USAGE RESTRICTtoNS. FROPRt(T AR f. (T C 1 PIPF PC Attachments are retained only in the Records Management copy of calculation NOTE:

Use Tag number only for valid tag numbers (i.e., RCV-8, SWV-34, DCH-99), otherwise: use Part number fic1d ti.e., CSC14599. AC1459), if more space is required, write "See Attachment" and list on separate sheet.

DESIGN ENGNtER DATf VIMIC ATON ENGINF P ^ATI $UPERVtSOM. NVcLE AN EN CATE T. R. Powers M L. M. Smalec C. L. M.ller ///[ p cc: Nuclear Proects (if MAR /CGWR/PEERE Calculation Review form Part ll! actions required Return to Service Related)

@Yes O No

@ Yes @ No (If Ycs, send copy of the form to Nuclear Regulatory Assurance and a Supervisor, Conte Mat. Info, copy of the Catculation to the Responsible Organization (s) idertified in Mgr,, Nucl. Operations Eng. (Original) w/ attach Part 111 on the Calculation Review form.)

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Rev 3/97

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M ANALYSIS /CA_LCULATION .

SUMMARY

i oiscrPUNE CONTROL NO REVIS10N LtVtt i

DOCUMENT IDENTIFICATION NUMBER M 97-0097 1 1 tritt classer 6 cation Icsten oNr>

d @ Safety Related Low Pressure Auxiliary Spray Flow Rate ;or Boron Precipitation O Non Safety Related M AR SPSGWRttf 41 NUMBER 1 NA l

VENDOR 00cVMENT NUMBER I NA e

i APPROVAL PRINTED j SIGNATURES NAME ,

1 Design Engineer [MMgg T. R. Powers

(. Date 10/21/97 i Verification Engineer. h,$ggh I'/UNf7 L. M. Smalec I Date ,

. Supervisor ((MM C. L. Miller

. Date //l/f*7

o tvs Revisto 4

i i Entito calculation, pages 15, Replaced all attachments with new attachments 123 i,

VJ C LE $VML1 Ae4 f This calculation determines the flowrate to tne Pressurizer via the LPI Auxiliary Spray Line with RCS l oressures between 0 - 120 psi. LPI has an assumed flowrate of 1600 gpm to the core, HPl :s t

at 600 gpm in a piggyback mode. A case to determine LPI and Auxiliary Spray Flow at 0 RCS pressura [

i and LPI pump flow at 3056 gpm usir J _DHP 1 A is also performed.

AtSULT $ $UMMARY '

Using DHP-1 A, Aux Spray Line flow is between 125.5 - 30.7 gpm with RCS pressure O - 120 psi @ LPl = 1600 i

i Using DHP-18, Aux Spray Line flow is between 114.5 O gpm with RCS pressure 0 - 120 psi @ LPI = 1600

! The differences in flow rates with different LPI pumps is due to the piping configuration of the Au;;iliary i

Spray Line piping, which taps off of the "A" train LPI pump.

Using DHP-1 A, Aux Spray Line flow is between 116.8 - 9.5 gpm with RCS pressure 0 120 psi @ LPI =2240 .

j Eg DHP 18, Aux Spray Line riow is between 100.4 O gpm with RCS pressure 0 - 120 psi @ LPI = 2240 l-

} .- - At 0 RCS pressure, HPl in piggyback at 600 gpm, and LPI pump flow at 3056 gpm, LPI flow te '5e core was calculated to be 2240.7 gpm and Auxihary Spray Flow c! 116 9 gpm using DHP-1 A to provide Aux Spray flow, a \

i i

hev. 30

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

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CALCULATION cuenvrw REVIEW Page 1 of 2 sj cuevtationwomv.

fj M97 0097, Rev.1 PART I - DESIGN ASSUMPTION / INPUT REVIEW: APPLICAF @ Yes O No The following organizations have reviewed and cc .ath the design assumptions and inputs identified for this calculation:

Nuclear Plant Technical Support i A. ID 0 System Engr

$"*' S*' /

Nuclear Plant Operations /# 0 7 OTHtRISt b3"0'~D"' M EOP Groun -

8 2 Egnats. Sat. #

E.gnatur. tat.

PART11- RESULTS REVIEW: APPLICABLE @ Yes O No The following organizations have reviewed and concur with the results of this calculation and understand the actions which the organizations must take to impler en he results.

Nuclear Plant Technical Support w 10!7d/93 System En0r W"'"***

g Nuclear Plant Operations ##

,_,. . .... y Nuclear Plant Maintenance O Yes @ N/A Nuclear Licensed Operator Training O Yu @ NM Manager, Site Nuclear Services O Yes @ NM Sr. Radiation Protection Engineer O Yes B NM OTHERS: /

UD EOP Group- '

r,#. . ... < <

O ......t.

Pev. 3'S? -

m

) M. ..... .o.. CALCULATION REVIEW -

C ALCut.AfloN No 4ttV.

M97-0097, Rev.1

)

' PART lli . CONFIGURATION CONTROL: APPLICABLE @ Y.s O No l

The following is a list of Plant procedures / lesson plans /other documents and Nuclear Engineering calculations which require updating based on calculation results review:

I Qf utrntal Qpte Reauired Responsible Ornanization EOP 14 12/15/97 Operations - G. Becker Upon completion, forward a copy to the Manager. Nuclear Regulatory AssurancIGroup for tracking of actions if any items are identified in Part 111.

PART IV - NUCLEAR ENGINEERING DOCUMENTATION REVIEW The responsible Design Engineer must thoroughly review the below listed documents to assess if the calculation requires revision to these documents. If "Yes," the change authorizations must be listed below and issued concurrently with the calculation.

Enhanced Design Bases Document O Yes @ No FC'l _

Vendor Quahfication Package iV0P81 FSAR O Yes S No E*"*'l Topical Design Basis Doc. O Yes @ No FC88 Improved Tech. Specification O Yes @ No L'"*88 E/SOPM O Yes O NoFC'i improved Tech Spec. Bases O Yes @ No A*"**) Other Documents reviewed:

Config. Mgmt. Info. Systern O Yes Q No ICo' OYesO/Jo smNc.t poc =>amws, Analysis Basis Document O Yes @ No UC O Yes O No (CmNGE (X>C 8ttf tfitwEl Design Basis Document O Yes B No FCa O Yes O No icm cacoc mfim w o Appendix R Fire Study 0 Yes @ No Fes; O Yes O No (CMANGL fA$ PL7iRLhCD

- Fire Hazardous Analysis - O Yes @ No FC O Yes O No icmNs. occ mess 8.cs.

NFPA Code Confcymance Document O Yes @ No UC O Yes O No ICm8NE DOC REf t8ttNCli PART V PLANT REVIEWS / APPROVALS FOR INSTRUMENT SETPOINT CHANGE PRC/DNPO approvalis required if a setpoint is to be physically changed in the plant through the NEP 213 process.

PRC Review Required Yes @ No

- =

PRC Chairman /Date ONPO Review Required Yes @ No DNPO /Date DESaGN ENGINEER /DATI QEMGN ENGINLIR PRINf(D NAME h ~

/e///[9)

/

T. R. Powers mev317

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\. ...........

CALCULATION VERIFICATION REPORT Crystal River Unit 3

%# CALvimAP Fmu Page 1 of 1 CALCVLAt4ON NuMetM M97 0097, Rev.1 '

emostct mits Low Pressure Auxiliary Spray Flow Rate for Boron Precipitation YES NO N/A

1. M O O Are inputs, including codes, standards, regulatory requirements, procedures, data, and Engineering methodology correctly selected and applied?

J O Have assumptions been identified? Are they reasonable and justified? (See NEP 101, V.c,

2. M O for discussion on references).

3. X 0 0 Are references properly identified, correct, and complete? (See NEP 101, V.c., for discussion on assumptions and ju stification.)

4. O O M Have applicable construction and operating experiences been considered?

5. O O Was an appropriate Design Analysis / Calculation method used?

6. ( 0 0 in cases where computer software was used, has the program been verified or reverified in accordance with NEP 135 for safety related design applications and/or are inputs, formulas, and outputs associated with spreadsheets accurate?

7. O O Is the output reasonable compared to inputs?

8. O O M Has technical design information provided via letter, REA, IOC or telecon by other disciplines or programs been verified by that discipline or program?

9. O O ( Has technical design information provided via letter or telecon from an external Engineering Organization or vendor been confirmed and accepted by FPC?

10. -0 M O Do the calculation results indicate a non conforming condition exists? If "Yes,"

immediately notify the responsible Supervisor. '

11. O O Do the results require a change to other Engineering documents? If "Yes," have these documents been identified for revision on the Calculation Review Form?

I have performed a verification on the subject calculation package and find the results acceptable.

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. A 3 37

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

e d' % 9orida INTEROFFICE CORRESPONDENCE sc~u~

1 M.-f.%. . . .w. . e, i e .r Nuclear Engineering NT5A 240 1628 OHice MAC Telephone SUBJCCT: Crystal River Unit 3 Quality Document Transmittal Analysis / Calculation

! To: Records Management NR2A The following analysis /calcult;hn package is submitted as the OA Record copy:

DoCNo (FPC DOCUMENT IDENTIFIC ATION NUMerms T*V SYst:4sl Tot AL PAGt t TRANSMIT 1tb M97 0097 0 DH,MU /92 TITLE Low Pressure .4uxiliary Spray Flow . tate for Boron Precipitation kwDs CDtNflFY RErwoRDs roR LATER RETRttvAU LOCA, Doron Precipitation, Aux Spray, PlPF-PC, hydraulic model exRtF minRtNets oR ntts . ust ea>MARv nLe nast' 310-641, 302 641 l M91-0092, M90 0021 M94-0047 goP-l([

M97 0088 VtNO IVENDOR N AME) VENDDR 000UMINT NUMBER lONREfl $UPER$tDED DOCvMcNis (DAREFI FPC NA NA DHV 91, DHV 92, DHV 93 l l RCV 53, RCV-12 l l DHP-1 A/B l MUP 1C l l t COMMENTS fusACE RESTRicilONs, PROPR;ET ARY, ETC )

PIPF PC Attachments are retained only in the Records Management copy of calculation NOTE:

Use Tag number only for valid tag numbers (i.e., RCV 8, SWV-34, DCH 99), otherwise; use Part number field (i.e., CSC14599, AC1459), if more space is required, write "See Attachment" and list on seperata 1,heet.

OtssGN ENGINEtR DAtt ICATION ENGIN[tR DATt J # ./;6t#1- DA WO j

s176mJkL.m.ukea SUPhERVISC-//pt AR/oAh ENGd 4, cc: Nuclear Projects (if MAR /CGWR!PEERE Calculation Review form Part lit actions required $Yes o Return to Service Related) O ves S No (if Yes send copy of the form to Nuclear Regulatory Assurance and a /

Supervisor, Config. Mgt. Info. copy of the Calculation to the Responsible Organization (s) identified in A Mgr., Nucl. Operations Eng. (Original) w/ attach Part lli on the Calculation Review forrn.)

~

4. Becm. A. cat 4P/h% (UpW) 4 AtTAmtw9ts

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4

. ANALYSIS / CALCULATION

SUMMARY

see e e ei'e, A C SUM FRM a

I

.. D'5C PLtNE CONinot No. Afv1SION LtVit DOCUMENT IDENTIF'ICATioN NUMBER M 97 0097 0 flTLt C.At$elCAY10N (CHECK ONtl

@ Safety Related

. Low Pressure Auxiliary Spray Flow Rate for Boron Precipitation O Non Safety Related M A%P!CGWR4'ith! NUMBER NA VENDOM DOCUMLNT NUVOLM NA APPROVAL PRINTED SIGNATURES NAME Design Engineer T. R. Powers l Date , 9/11/97 Verification Engineer 87Mgh L. M. Smalec Date 9/11/97 Supervlsor [Wffff f f l. M itt.gn Date /4/s/99 i1 EMS PtVikt0 i

Initialissuo ef Calculation i

D; ,, .

-This calculation determines the flowrate to the Pressurizer via the LPI Auxiliary Spray Line with RCS

pressures between 0 120 psi. LPI has an assumed flowrate of 1400 gpm to the core, HPIis it 0 gpm or i .=

at 500 gpm in a piggyback mode, A case to determine LPI and Auxiliary Spray Flow at O RCS pressure l

and LPl pump flow at 3056 gpm is also performed, au ULtsi,Uuutav ,

a With 0 gpm HPl flow, Auxiliary Spray Line flow is between 133.1 52.1 gpm with RCS pressure 0 120 psi i

j With 500 gpm HPl flow, the Spray Line flow is between 128.9 42.5 gpm with RCS pressure 0 120 psi.

At 0 RCS pressure, HPl in piggyback at 600 gpm, and LPI pump flow at 3056 gpm, LPI flow to the core was 1

calculated to be 2240.7 gpm and Auxiliary Spray Flow at 116.9_ gpm. !

1 6

t Rev 31?

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

Os.=

ca'co'aso~ eev'e* Page '. of 2 MD7 0097 Rev. O PART1+ DESIGN ASSUMPTION / INPUT REVIEW: APPLICABLE @ Yes O No lhe following otganizations have reviewed and concur with the design assumptions and inputs identified for this calculation:

Nuclear Plant Technical Support Harvey N .Nin 9Y System Engr 5***

5 Nuclear Plant Operations 9# I7 e,. ......~.

EOP Groun s,ng.. c.i.

/

(

}

s.w <a. v...

PART 11 - RESULTS REVIEW: APPLICABLE @ Yes O No The following organizations have reviewed and concur with the results of this calculation and understand the actions which the organizations must take to imp'ement the results.

Nuclear Plant Technical Support SPFU N *ble f $ 0 Th 5 /

System Engr 5"a* 0W

/

Nuclear Plant Operations /

54n.1,4 c.i.

/ Vb

( /

Nuclear Plant Maintenance O Yes @ N/A Nuclear Licensed Operator Training

"~'""

O Yes @ NM Manager, Site Nuclear Services Y @ NM Sr. Radiation Protection Engineer O Yes @ N/A -

f// f')

? '

OTHERS: ,

EOP Group O 5,_.....

hv. 3/97

_ .m_._. ______.m_.~.__..m.- .__ _ ._ _ _._ _ _. _ _ _ _ _ _ .__

CALCULATION REVIEW

[v)N--

Pa,e , o, ,

C A1.CULAtacN NoJntv.

M97-0097 ,

e V PARTlli CONFIGURATION CONTROL: APPLICABLE % Yes No (qpjO I

The following is a list of Plant procedures / lesson pia s/other documents and Nuclear Faqineering calculations which require updating based on calculation results review:

Document Date Reauired Resoensible Ornanization

EsoP- l'f I:LhrI17 Be

'1

-r 4

4 4

Upon completion, forward a copy to the Manager, Nuclear Regulatory Assurance Group for tracking of actions if 1

any items are identified in Part 111.

PART IV - NUCLEAR ENGINEERING DOCUMENTATION REVIEW

\ The responsible Design Engineer rnust thoroughiv raview the below listed documents to assess if the r calculation requires revision to these documents, if "Yes," the change authorizations must be listed

] below and issued concurrently with the calculation.

Enhanced Design Basis Document O ves @ No UC Vendor Quahfication Package NOP#)

j FSAR O ve S No Iw"*n Topical Design Basis Doc. O ves 2 Nc.'TC'i Improved Tech. Specification O ves G No R*nwo E/SOPM O ves O Noaco l

~

improved Tech, Spec. Bases O ves E No A*"*M Other Documents reviewed:

Config. Mgmt. 'nfo. System O vos E No ICloP') C ves 0 ,No g IcHAN64 DOC MetNENG4)

Analysis Dasis Document O ves G No "C'> 0 ves O No scau.at Doc M*ia Nct.

Design Basia Document O ves E No UCa O ves O No scnu.cs doc MunNcn>

Appendix R Fire Study O ves S No UCa O ves O No (CMANGE DOC. Mf L4LNCil j Fire Harardous Ana!ysis O ves S '4o "ca O ves O No j ICHANGL DOC. MF(MNCil i NFPA Code Conformance Document O ves @ No "Ca _

O ves O No (CHANGE DOC MFtMNCil PART V PLANT REVIEWS / APPROVALS FOR INSTRUMENT SETPOINT CHANGE

-_PRC/DNPO approvalis required if a setpoint is to be physically changed in the plant through the NEP 013 l process.

PRC Review Required O Yes No _ _ _

' PRC Chairman /Date 4

i DNPO Review Required O Yes @ No J DNPO /Date DESIGN P NNEthioAT DEstGN ENGINEEM PRtNilo NAME f b II 7 T. R. Powers Rev. 317

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

Ng ...........

CALCULATION VERIFICATION REPORT Crystal River Unit 3

- V C ALvtRRP.f RNI e

[ Page 1 of 1 TEcutAtioN uvueen M97-0097 PROJECI/IIILI t.ow Pressure Auxiliary Spray Flow Rate for Baron Precipitation YES No N/A

1. @ O O Are inputs, including codes, standards, regulatory requirements, procedures, data, and Engineering methodology correctly selected and applied?

2. M O O Have assumptions been identified? Are they reasonable and justified? (See NEP 101, V.c, for discussion on references).

3. E O O Are references properly identified, correct, and complete? (See NEP 101, V.c., for discussion on assumptions and justification.)

4. O O E Have applicable construction and operating experiences been considered?

5. R O O Was an appropriate Design Analysis / Calculation method used?

6. M O O In cases ovhere computer sof tware was used, has the program been verified or reverified in accordance with NEP 135 for safety related design applications and/or are inputs, formulas, and outputs associated with spreadshetts accurate?

7. E O O is the output reasonable compared to inputs?

8. O O % Has technical design information provided via letter, REA, IOC or telecon by other disciplines or programs been verified by that discipline or program?

9. O O M Has technical design information provided via letter or telecon from an external Engineering Organization or vendor been confirmed and accepted by FPC?

10. O E D Do the calculation results indicate a non conforming condition exists? If "Yes,"

immediately notify the responsible Supervisor. '

~

O Do the results require a change to other Engineering documents? If "Yes," have these

11. $

d ") documents .Seen identifiad for revision on the Calculation Review Form?

/

I have performed a urification on the subject calculation package and find the results acceptable.

VERlHC AllON ENGMLP OATE $.UPERv1SDR. NUCL si g yNggm.Ng DAYt

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Florida Power DESIGN ANALYSIS / CALCULATION Crystal River Unit 3 DESA C F) M 6 DOCUMENT $$NTIFIC ATION NO.

Page 1 PEVISION of S ,

M97-0097 1

1. PURPOSE This calculatiori determines the flow in the Low Pressure (Decay Heat) Auxiliary Spray Line to the "

Pressurizer during LOCA conditions with RCS pressure bt. tween 0 and 120 psig. Two PIPF-PC hydraulic models of the Decay Heat System, modeled with ee low pressure spray line, are developed for this !

calculation to determine the flow through the Low Pressure Auxiliary Spray Line. Two models ar6 required because the Low i'ressure Auxiliary Spray Line is off of the *A" Decay Heat Pump discharge and there are ,

e significant hydraulic differences when DHP-1B is used to supply the Low Pressure Auxiliary Spray Line. l This calc alation also determines the Auxiliary Spray Line flow and LPI injection flow for the condition of RCS pressure = 0 psig and the maximum amount of flow (3086 gpm) going through the operating 6.Pl pump.

This calculation has been performed to support alternatives for prnviding Boron Precipitation m;tigation in the reactor vessel following a Loss of Coolant Accident.

II. RESULTS/ CONCLUSIONS Based upon PIPF-PC model runs, the LPl Auxiliary Sp.ay Line will deliver the following flow to the Pressurizer:

DHP-1 A Supplying Flow to the Low Pressure Auxiliary Spray Line LPI flow = 1600 nom RCS Pressure Aux Spray Flav 3p_m_1 Attachmerg 0 125.5 AUXPZRft1 30 109.4 AUXPZRR2 60 90.7 AUXPZRR3 90 67.2 AUXPZRR4 120 30.7 AUXPZRR5 l

LPI fjaw = 2240 nom RCS Pressure Aux Spray Flow (nom) f,ttacumem 0 116.8 AUXPZRS1 30 99.6 AUXPZRS2 60 78.7 AvXPZRS3 90 60.6 AUXPZRS4 120 9.5 AUXPZRS5**

RCS pressure does not allow LPIinjection flow to reach 2240 gpm Rev 495 RET: ble of Ptarit litSF NuCfes E4neerarg m .

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DESIGN ANALYSIS / CALCULATION Crystal River Unit 3 cesAcrno Page of 5 wb 2 DOCUMENT 6DENTIFICAfION NO. REVISION M97-0097 1 l DHP-1B Supplying Flow to the Low Pressure Auxiliary Spray Lina LPI flow = 1600 com 7 r

RCS Pressure Aux Sprav Flowfoom) Attachment i i

0 114.5 AXSPYCOO f' 30 97.0 AXSPYC30 60 75.8 AXSPYC60 90 46.3 AXSPYC90 120 C' AXSPYC12 l The LPI pump is no longer able to overcome the RCS backpressure and elevation change (=85 ft) change !

between Node 39 and the Auxiliary Spray Line (Node 800), i HPl flow = 2240 com RCS Pressure Aux Sorav Flow (oom) Attachment E r

A O 100.4 AXSPYB00 l 30 80.2 t] 60 53.5 AXSPYB30 AXSPYB60 90 0* AXSPYD90 120 0' AXSPYB12**

The LPI pump is no longer able to overcome the RCS backp essure ar'd elevation change (=85 ft) change j between Node 39 and the Auxiliary Spray Line (Node 800). !

RCS pressure does not allow LPIinjection f!ow to reach 2240 gpm. ,

Maximum LPl Pumo Flow Results This model run shows that with a DHP-1 A pump flowrate of 3086 epm, the flow through the LPI injection path is 2270.8 gpm and the flow through the Aniliary Spray Line is 116.8 gpm. Piggyback HPI flow is assumed to be 600 gpm. These flow values are not corrected for any indication errors. This model duplicates the model using DHP-1 A with a fixed LPI flow of 2140 gpm.

Ill. DESIGN INPUTS I

1. The decay heat model developed in M94-0047, as shown on Ref.1, was modified for use in this analysis. I A diagram of the modified models is shown on Attachment 1.

2. Maximum flow through the LPI pump is 3086 gpm. This is based on NPSH considerations determined in Ref. 5.

3 L/D data for the Auxiliary Spray Line is per Attachment H of Ref. 4.

w) .

IV. ASSUMPTIONS

1. LPI flow to the core, via the LPI core injection path, is 1600 or 2240 gpm actual flow.

\

Rev. 9;SS RET: Ofe of Plant RESP: Nuclear ErgtrwertN

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DESIGN ANALYSIS /CALCUl.ATION Crystal River Unit 3 oEsac mu Page 3 of 5 l DOCUMENT 10ENTIFIC AT10N NO. REVISION M97 0097 1

! 2. An HPl flows of 600 gpm, in piggyback mode, will be used in determining Auxiliary Spray flows.

3. A single Building Spray train is assumed to be operating at a flowrate of 1326 gpm. This BS train takes suction from the RB sump on the same pipiag as the forward flowing LPI pump and must be considered when evaluating Decay Heat Pump NPSH A . Since the characteristics of the discharge of this BS pump are not significant for the analysis being performed, an LPI pump curve has been used to obtain the 1326 gpm flow.

4. Reactor Building pressure shall be modeled at 0 psi.

5. The RCS pressure used in this evaluation shall be 0 - 120 psig. This is the same pressure range used in Ref. 6 to evaluate RCS hot leg injection by backflowing through an idle LPl pump.

6. No flow va8ues used in this analysis are corrected for instrument error.

7. The pressurizer spray nozzle can be modeled in PIPF-PC based upor, vendor supplied capacity vs. pressure information and the equations provided in Ref. 8. From Raf. 4, the spray nozzle is a 4" CRC 150-45' nozzle and capacity vs. AP data is provided for_ AP's between 3 and 100 psi.

p, Reviewing the data in Ref. 4, the nozzle can be modeled within PIPF-PC as a component with an equipment Q flow of 56.6 gpm and corresponding pressure drop of 1,0 psi. These values will be used in each of the PIPF-PC model ru%.

9. For the case in which ;il pump flow is to be equal to 3086 gpm. HPl piggyback flow is assumed to be 600 gpm.

9. When modeling DHP-1B supplying the Auxiliary Spray Line, it is assumed that v11ves are capable of isolating the DHP-1 A suction piping so that there is no reverse flow through DHP-1 A or the pump bypass.

9 V. REFERENCES

1. FPC Dwg 310 641, Rev.1, Decay Heat Removal

2. FPC Dwg 302-641, Sheet 3, Rev. 40, Decay Heat Removal

3. M94-0047, Rev. 2, "CR3 Decay Heat Removal System Hydraulic Studies"

4. M91-0092, Rav.1, "DH System L/D Values"

5. M90-0021, Rev. 9, " Building Spray and Decay Heat Pump NPSH A/R"

6. M 97-0088, Rev. O, " Hydraulic Analysis for LPI Hotleg injection to RCS"

7. Computer Data (Revision 1) Computer Software Number CS-97-005, PC Serial Number 7NH6V, verified 1/23/97.

VVi. DETAILED CALCULATIONS OR ANALYSES The model developed in M94-0047 was used as the starting point for the Auxiliary Spray Line model. Loops were created for the Auxiliary Spray Line using the L/D data contained in Raf. 4.

Rev 6.95 RET: Ofe of Plant RESP; Nucleat Engs,wenst

s~.

)"M DESIGN ANALYSIS / CALCULATION

.j Crystal River Unit 3 otsAc rioa Page 4 of S oocwrwt ietwTrication No. REVIStoN M97 0097 1 l

A flow to a makeup p.mp, in piggyback configuration, and a building spray pump were also added to the model and appropriate loops generated for these pumps. Dammy nodas were created on the discharge of these nodes so that Decay Heat system piping pressures could be accurately calculated.

Ths resultant models are shown pictorially on Attachment 1.

RCS pressures of O, 30, 60, 00, and 120 psig were arbitrarily selected for determining Spray line flow. The ,

models were used with an HPl flowrate of 600 gpm.

The results of these model runs are summarized as follows:

DHP-1 A Suppi,.ng Flow to the Low Pressure Auxiliary Spray Line j LPI flow = 1600 anm RCS Pressure Aux Sprav Flow (com) Attachment 0 125.5 AUXPZRR1 30 109.4 AUXPZRR2 60 90.7 AUXPZRR3 O EO 120 67.2 30.7 AU)"P?RR4 AUXPZRR5 LPI flow = 2240 com

}

RCS Pressure Aux Sprav Flow (anm) Attachment O 116.8 AUXPZRS1 ,

30 99.6 AUXPZRS2 60 78.7 AUXPZRS3 90 50.6 AUXPZRS4 120 9.5 AUXPZRSS DHP-1B Supplying Flow to the Low Pressure Auxiliary Spray Line LPI flow = 1600 anm RCS Pressure Aux Spray Flow (c. u sitachment 0 114.5 AXSPYC00 30 97.0 AXSPYC30 60 75.8 AXSPYC60 90 46.3 AXSPYC90 120 0' AXSPYC12 s.

The LPI pump is no longer able to overcome the RCS backpressure and elevation change (=85 ft) change between Node 39 and the Auxiliary Spray Line (Node 800).

Rev. 8.15 RET: Lafe of Plant RESP: Nuclea, Engmeereg

M%

(Ci)M " " * " " "

DESIGN ANALYSIS / CALCULATION Crystal River Unit 3 resAcau Page 5 of 5 ,,

s -

4 DOCUMENT 4DENTrtCAfDN NO. REVlW N M97-0097 1 HPl flow = 2240 nom RCS Pressure Aux Spray Flow (opm) Attachment 0 100.4 AXSPYB00 30 80.2 AXSPYB30 60 53.5 AXSPYB60 l

90 0* AXSPYB90 '

120 0' AXSPYB12 I

The LPI pump is no longer able to overcome the RCS backpressure and elevation change (=85 f t) change l between Node 39 and the Auxiliary Spray Line (Node 800).

These results are displayed graphically in Attachment 2.

MAXIMUM LPI FLOW ASSESSMENT The model was adjusted to reflect a piggyback HPI flow rate of 600 gpm and then run to obtain an LP) pump flow rate of 3086 gpm at an RCS pressure of 0 psi. A Building Spray pump was assumed to continue to operate at

.1326 gpm. The results of this model run are included in Attachment file AUXMAX1.

I

- This model run shows that at an LPI pump flowrate of 3086 gpm, the how through the LPI injection path is 2270.8 gpm and the flow through the Auxiiiary Spray Line is 116.8 gpm. These flow values are not corrected for any indication errors.

ATTACHMENTS

1. PlPF-PC model for LPI Auxiliary Spray Flow to Pressurizer

2. Graph of PIPF-PC model results ,

3. Attachment file AUXPZRR1,600 gpm HPl flow, RCS pressure = 0 psig

4. Attachment file AUXPZRR2,600 gpm HPl flow, RCS pressure = 30 psig I

5. Attachment file AUXPZRR3,600 gpm HPl flow, RCS pressure = 60 psig

6. Attachment file AUXPZRR4,600 gpm HPI flow, RCS pressure = 90 psig

7. Attachment file AUXPZRR5,600 gpm HPl flow, RCS pressure = 120 psig

8. Attachment f!!e AUXPZRS1,600 gpm HPI flow, RCS pressure = 0 psig

9. Attachment file AUXPZRS2,600 gpm HPI flow, RCS pressure = 30 psig

10. Attachment file AUXPZRS3,600 gpm HPl flow, RCS pressure = 60 psig

11. Attachment file AUXPZRS4,600 gpm HPl flow, RCS pressure = 90 psig

12. Attachment file AUXPZRSS,600 gpm HPl flow, RCS pressure = 120 psig

13. Attachment file AUXMAX1,600 gpm HPl flow, RCS pressure = 0 psig

14. Attachment file AXSPYC00,600 HPl flow, RCS pressure = 0 psig

15. Attachment file AXSPYC30,600 HPI flow, RCS pressure = 30 psig

16. Attachment file AXSPYC60,600 HPl flow, RCS pressure = 60 psig

17. Attachment file AXSPYC90,600 HPI flow, RCS pressure = 90 psig

18. Attachment file AXSPYC12,600 HPl flow, RCS pressure = 120 psig

( 39. Attachment file AXSPYB00,600 HPI flow, RCS pressure = 0 psig

20. Attachment file AXSPYB30,600 HPl flow, RCS pressure = 30 psig

21. Attachrr. int file AXSPYS60,600 HPI flow, RCS pressure = 60 psig

22. Attachment file AXSPYB90,600 HPI flow, RCS pressure = 90 psig

23. Attachment file AXSFYB12,600 HPl flow, RCS pressure = 120 psig Rev. C195 RET: Ofe of Plant RESP:Nucisar Engeswenng

[, ~._,\

(

P

)Q*ower DESIGN ANALYSIS / CALCULATION

%s Crystal River Unit 3 DESACFRM f)

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Page 1 of 2 f DOCUMENT :DENTIFIC ATION 00.

REVISION M 97-0097, Attachment 1 1 Modified Model for Boron Precipitation Mitigation LPI Auxiliary Spray Line

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Page 2 of 2 DOCUMENT IDENTIFICATION NO.

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PlPF-PC Output Results LPI flow = 1600 gpm ;

l~

1 Auxiliary Spray Flow 160 120 -

C/ _ DHP-1A

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DESIGN ANALYSIS / CALCULATION Crystal River Unit 3 otsacinu Page 2 of 2 J

DCCUMENT IOENTFIC AtCN NO. * $ "

M97-0097, Attachment 2 1 PIPF PC Output Results LPI flow = 2240 gpm Auxiliary Spray Flow 120 N

DHP-1A 80 E

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U.S. Nucletr Rrgulatory Commission 3F1297-12 Enclosure 2 i

i i

Simulator Exercise Number ROT-9-200

(

" Boron Precipitation Control"

. . __ _J

l EXERCISE NUMBER ROT-9 200 - REQUALIFICATION CYCLE 1997 / Cycle 6 Year / Cycle l- DESCRIFTION: Boron PreciDitation Control -

REFERENCES:

EOP-14 Enclosure #20 l

1. BASIS FOR DEVELOPMENT: The purpose of this exercise is to allow the optrating crews to practice the boron precipitation control enclosure of EOP-14.

1

2. SCENARIO OVERVIEW: This exercise will take the form of a practice

/ demonstration. The instructor will explain the problems associated with the .

boron-precipitation and the conditions that cause it to occur. The crew will practice both of the main control methods:

. Decay Heat Drop Line.

Auxiliary Spray from LPI.

3. SCENARIO OBJECTIVES:

At the conclusion of this exercise the student will be able to:

a. Explain tha basic steps to control boron precipitation in the reactor core,

b. Identify the conditions that cause boron precipitation,

c. Discuss the indications that boron precipitation is occurring in the Core.

4. PLANT MANIPULATION COVERED:

NONE Prepared By:

Nuclear Operations Instructor /Date-Reviewed By:

Nuclear Training Supervisor /Date

5. EXERCISE SETOP A. Initialize the simulator to IC # 49, this IC is post LoCA after the pump suctions have bein transferred to the RB sump.

B. Freeze the simulator and perform the following initial condition modifications:

1) None.

C. En'.ure-the crew has copies of EOP-14.

6. PRE-SCSSION BRIEFING A. Provide the crew with the shift turnover sheet.

B. Review learning objectives.

C. Discuss with the crew that boron precipitation can occur when:

the RCS is in a saturated or superheated condition.

. a cold leg break exists i.. the RCS.

D. Discuss with the crew that the primary indication that boron precipitation is occurring is when the difference in boron concentration in the RB Sump and the boron concentration in the BwST starts to increase, i.e. the boron is being distilled in the core and demin water is going into the RB sump.

7. EXERCISE OUTLINE A. Unfreeze Scenario A. the simulator and Provide the crew with Turnover sheet for

1) Allow step # 20.17.

the crew to work through E0P-14 Enclosure #20 until they reach

2) Review the flow path of through the core to the sump with the crew.

R. Restore the simulator to IC #49. Unfreeze the simulator and shutdown MUP- -

1B Allow the simulator to stabilize.close DHV-12, and trip DHP-1A with a breaker fail open malfunction.

Scenario 8. Provide the crew with Turnover sheet for

1) Allow the crew to work through EOP-14 Enclosure #20 starting at step #

20.18 until they reach step # 20.32.

2) Review the flow path of through the core to the sump with the crew.

._ 8.: POST-EXERCISE REVIEW A. Review of Session

1) Review the exercise objectives.

2)-Ask if anyone has questions, or comments.

Record appropriate crew comments and transmit to EOP group.

End of Exercise 1

2. Review the class attendance record for-completeness.

__hN

g 9, SHIFT TURNOVER for SCENARIO "A" A. The following are the initial plant conditions:

1 Time in core life - 300 EFPD 2 Rx power and power history - 0% fcr last 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3 Boron concentration - 2400 PPMB.

4) Xenon - Decreasing.

5) RCS Activity - See status board.

B. Tech. Spec. action renuirement(s) in effect: Many.

l C. clearances in effect:.None.

l D. Significant problems / abnormalities: 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ago the plant experienced a large l

' break'LOCA. The boron concentration of the RB sump has been slowly decreasing. The TSC believes that the LOCA is on one of the Cold Legs.

E. Evolutions / maintenance for the on-coming shift:perforts The Tsc re E0P-14 Enclosure #20 Baron Prec F. Units 1 and 2 status: on line.

G. Units 4 and 5 status: on line.

H. SSOD - Instruct with the following thedata:

ROS to walk down the main control board and provide you

1. RCS Average Temperature 4. Make-up T&nk Level

2. RCS Pressure 5. Turbine Load

3. Pressurizer Level 6. Turbine Reference I. Required Emergency Plan Implementation Full Implementation, inclOding all required notifications.

l l Initial / upgrade classifications Internal- Notifications.

,XXX. None.

3. The STA should start this exercise 18 the control Room.

K. The CNO should start this exercise IN the control Room.

L. The SSoD should start this exercise IN the contro1~ Room.

9. SHIFT TURNOVER for SCENARIO "B" A. The following are the initial plant conditions:

1 Time in core life - 300 EFPD 2 Rx power and power history - 0% for last 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3 Boron concentration - 2400 PPMB.

4 Xenon - Decreasing.

5 RCS Activity - See' status board.

B. Tech. Spec action requirement (s) in effect: Many, c.' clearances-in effect: None.

D. Significant problems / abnormalities:

1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ago the plant experienced a large break LOCA.

2 The boron concentration of the RB sump has been slowly decreasing.

3 The TSC believes that the LOCA is on one of- the Cold Legs.

4) DHP-1A tripped due to a motor fault. MUP-18 was tripped by the operator.

E. Evolutions / maintenance for the on-coming The TSc re shift: performusingEOP-14 Enclosure #20 Boron the Auxiliary Spray method.

F. Units 1 and 2 status: on line.

G. Units 4 and 5 status: on line.

H. SSOD - Instruct the Ros to walk down the main control board and provide you with the following data:

1. RCS Average Temperature 4. Make-up Tark Level

2. RCS Pressure 5. Turbine Load

3. Pressurizer Level 6. Turbine Reference I. Required Emergency Plan Implementation

, Full Implementation, including all required notifications.

, l Initial / upgrade classifications Internal Notifications.

,XXX. None.

3. The STA should start this exercise IN the-control Room.

K. The CNO should start this exercise IN the control Room.

L. The SSOD should start this exercise IN the control Room.

q- -

U.S. Nuclear R:gul: tory Commission 3F1297 l l Enclosure 3 i

CH-632D

" Post Accident Sampling and Analysis of Reactor Building Sump" hOh

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

Rev. 0 Effective Date II Al 91 EMERGENCY PLAN IMPLEMENTING PROCEDURE CH-632D FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 POST ACCIDENT SAMPLING AND ANALYSIS OF THE REACTOR BUILDING SUMP APPROVED BY: Interpretation Contact 0D .{n (SIGNATURE ON FILE)

DATE: ll l ') $~l Interpretation

Contact:

ChemRad Specialist II

TABLE OF CONTENTS SECTION PAGE 1

l l

1.0 PURP0SE................................................................. 1

2.0 REFERENCES

-............................................................. 1 2.1 IMPLEMENTING REFERENCES............................................. 1 2.2 DEVELOPMENTAL REFERENCES..................................

......... 1 2.3 CMIS REFERENCES...........................................

......... 2 3.0 PERSONNEL 3.1 INDOCTRINATION................................................ 2 DESCRIPTION......................................................... 2 3 . 2 LIMITS & PRE CAUTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4.0 INSTRUCTIONS............................................................ 3 4.1 SAMPLE TEAM CHECKLIST.................................. 3 4.2 SAMPLE LINE-UP...................................................... ............. 6 4.3 GAMMA ANALYSIS........................................ ............. 9 4.4 BORON ANALYSIS........................................ ............ 12

4. 5 GRAB S AMPLE COLLECTION AT CASB- 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.6 QEMINERALIZED WATER FLUSH....................................

20 4 7 SYSTEM RESTORATION................................................. ...... 23 5.0 CONTINGENCIES..........................................................

5.1 CAT-8 HI-HI LEVEL ALARM........................................... 24 24 5.2 NOTIFICATIONS AND SHIPMENT.......................................... 25 ENCLOSURES 1 Techni cal Support Center Data Sheet . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 2

Assessment of Core Damage Based on Reactor Building Sump Sample ... 27 l

l l

CH-632D Rev. O Page i

O 1.0 PURPOSE This procedure provides instructions for sampling the Redctor Building Sump under accident conditions for Gamma Isotopic and Boron analyses using the Post Accident Sampling System.

2.0 REFERENCFC 2.1 IMPLEMENTING REFERENCES 2.1.1 EM-104, Operation Of The Operational Support Center r

2.1.2 CH-234 Energy, FWHM, And Efficiency Calibration Of The Post Accident Sampling System Hign Purity Germanium Detectors l= 2.1.3 CRI-101, Calibration Of The ABB/CE PASS Boronometer 2.2 DEVELOPMENTAL REFERENQ1 2.2.1- Radiological Emergency Response Plan-2.2.2 NUREG 0737, Post-TMI Requirements 2.2.3 Regulatory Guide 1.97,_ Instrumentation For Light-Water-Couled Nuclear Power Plants To Assess Plant And Environs Conditions During And Following An Accident 2.2.4- RSP-600, - ALARA Progran' 2.2.5- PASS Users Manual-Volumes A through C, trystal River Installation 2.2.6 APEX Technologies Post Accident Sample System Modules Manual. . FPC Manual #2034 2.2.7 -6059-S-002, APEX Technologies PASS Process Flow Diagrams 2.2.8 FD-302-700, Post Accident Sampling System 2.2.9 Nuclear Regulatory Commission RTM-96, Response Technical Manual

.CH-632D Rev. O Page 1

_ J

2 2.3 CMIS REFERENCES DPDP-5A BREAKER 27, DPDP-5B BREAKER 8, CACP-1, CAV-126, CAV-1. CAV-3, CAV-431, CAV-432, CAV-429, CAV-430, CAV-626, CAV-627, CAV-484,.

CAV-439, CAV-636, CAV-519, CAV-447 CAV-437, CAV-448, CAV-623, CAV .

625, CAP-10, CAP-14, CAV-436, CAV-434, CAV-500. CAV-624, CA-74-FI, CA-56-CI, CASB-5, AHF-55, CAV-492, CAV-493, CAV-445, CAV-446, CAV-471, DWV-337, CAP-8, CAT-8,.CAV-470, CAV-433,-CAV-435 3.O PERSONNEL INDOCTRINATION

3.1 DESCRIPTION

The Post Accident Sampling System (PASS) is an on-line system designed to sample r.nd evaluate various liq: id and gaseous sample streams during an accident, including the Reactor Building Sump.

The liquid PASS Automated Isotopic And Chemical Measurement System (AIMS) consists of the subassembly used to perform Gamma Isotopic and Boren analyses of the Reactor Building Sump.

3.2 LIMITS & PRECAUTIONS 3.2.1 Performance of all or part of this procedure will be done by direction of the Emergency Cocrdinator.

3.2.2 Entries into the controlled access areas must have Radiation Monitoring Team preplanning, concurrence, and coverage as-outlined in EM-209, Re-Entry Procedure.

3.2.3 During post-accident sampling, extremely high radiation exposure levels could be experienced. The ability to perform this procedure and~ stay within exposure limits will require ALARA' pre-planning.

3.2.4 Return to the Lab if the dose rate at places requiring work is >

15R/Hr. The 4 REM Total Effective Dose Equivalent (TEDE) whole body exposure limit will be exceeded in 16 minutes at ISR/Hr.

3.2.5 All sampling actions are performed on the Main Control Board by Operations, or in the Count Room either on the VAX Computer or from PASS CACP-1 and Nuclear Data Mimic Panels unless otherwise noted.

3.2,6 Section 4.1 must be completed prior to any sample team re-entries.

3.2.7 Sections 4.3, Gamma Isotopic Analysis and section 4.4, Boron Analysis can be performed simultaneously.

CH-632D Lav. O Page 2

\

=

4.0 INSTRUCTIONS NOTE: Section 4.1 mst be conpleted prior to any sanple team re-entries.

4.1 SAMPLE TEAM CHECKLIST ACTIONS DETAILS l

4.1.1 ASSEMBLE Sample Team LIST personnel performing entry and their dose margins:

Name Dose Marain -

1.

2.

3.

4, 5.

/

Initial /Date 4.1.2 DETERMINE Reactor Building Sump 1. SELECT sample lineup to sample lineup and analyses to perform perform a. Primary Sample Path

b. Alternate Sample Path

2. LIST analyses to perform:

/

Initial /Date CH-632D Rev. O Page 3

I 4.1 SAMPLE TEAM CHECKLIST (Cont'd)

ACTIONS DETAILS l

4.1.3 REVIEW applicable procedures - REVIEW the following procedures:

CH-632D, Post Accident

, Sampling And Analysis Of-The Reactor Building Sump EM-104, Operation Of The

. Operational Support Center

/

Initial /Date 4.1.4 DISCUSS access and exit routes, DISCUSS Sample collection communication techniques and access and exit routes radiological- conditions DISCUSS Dose limits /

radiological condit!ons for sample point access route DISCUSS communication techniques TSC phone number PERFORM-radio check on channel when outside TSC

.IE obtaining CASB-2 grab s ampl e ,

IhEN ENSURE the following:

Allen wrench, or equivalent, as determined by

= Chemistry supervision, - - -

for removing T-Handle from grab sampler and attaching to new grab sampler Knife,-or equivalent, as determined by Chemistry supervision, to cut transit cover strap from lifting eye New tie-wrap, or equivalent as determined by Chemistry supervision to attach transit cover to new grab sampler lifting eye

/

Initial /Date CH-632D Rev. O Page 4

4.1 SAMPLE TEAM CHECKLISI (Cont'd)

ACTIONS DETAILS l

NOTE: The following breakers are normally in the locked open (Off) position by Operations due to not having automatic ES closure functions -

4.1.5 ALIGN electrical power REQUEST Operations ENSURE CLOSED supplies the following breakers:

1. DPDP-5A, Brk. No. 27

2. DPDP-58, Brk, No. 8

/

Initial /Date i

CH-632D Rev. O Page 5

, ________I

i 3

4.2 SAMPLE LINE-UP ACTIONS - '

DETAILS l

4.2.1 PERFORM valve lineup to ENSURE CLOSED the following sample RB Sump valves:

1. CAV-126

- 2. CAV 3. CAV-3

4. CAV-431 -

5. CAV-432

6. CAV-429

7. CAV-430-

8. CAV-626 .

l_ 9. CAV-627 p 10. CAV-484 -

l 11. CA*!-439

12. CAV-636 ENSURE OPEN the following*

13. CAV-519 14.

CAV-447

15. CAV-437

16. CAV-448 ALIGN the following:

17. CAV-623 to SAMPLE

18. CAV-625 to SAMPLE

-19. CAV-626 to DRAIN TANK

20. CAP-10 to-AUTO

21. CAP-10 Flow Control

. Switch to FULL CLOCKWISE

22. CAP-14 to ON

/

Initial /Date CH-632D Rev. O Page 6

. _J

O 4-9 4.2 SAMPLE LINE-UP_(Cont'd)

ACTIONS OETAILS l

i 4.2.2 OPEN Containment Isolation IE using the Primary Sample Path Valves IREN REQUEST Operations OPEN the following:

1. CAV-436

. 2. CAV-434 NOTE: CAV-500 is in the Intermediate Building.

IE using tne Alternate Sample

Path, IEEB:

1. OPEN CAV-500

2. OPEN CAV-439 .

3. CLOSE CAV-448 I

Initial /Date 4.2.3 INITIATE demineralized OPEN the following:

water flow to PRIME 1. DWV-337 CAP-8 2. CAV-471

3. START CAP-8

4. THROTTLE CAV-624 to obtain flow rate between-0.35-0.50 gpm on CA-74-FI

/

Initial /Date CH-6320 Rev. O Page 7

o

.c.

i 4.2 SAMPLE LINE-_UP (Cont'd)

AP' IONS DETAILS l

4.2.4 OPEN Containment Isolation IE using the Primary Sample Valves Path, IliEN REQUEST Operations OPEN the following:

l 1. CAV-433

2. CAV-435 If using the Alternate Sample Path, IliEN REQUEST Operations OPEN the following: *

1. _ CAV-434

2. CAV-436

/

Initial /Date 4.2.5 SECURE the demineralized MIEN sample flow has stabilized priming water on CA-74-FI (approximately 5 minutes)

IEEN CLOSE:

CAV-471 OW-337

/

Initial /Date NOTE:

Refer to section 5.0 if a HI-HI alarm occurs at CAT-8.

4.2.6 ADJUST sample flow for THROTTLE CAV-624 to obtain flow Gamma Isotopic, Boron, rate between 0.35-0.50 gpm on or Grab Sample CA-74-FI

/

Initial /Date CH-632D Rev. O Page 8

s 4.3- GAMMA ANALYSIS ACTIONS DETAILS l .

4.3.1 FLUSH RB Sump Sample 1. ENSURE Section 4.2 SAMPLE LINE-UP. performed-f-

NOTE: While flushing you may continue with step 4.3.2 and 4.3.3.

2. FLUSH for at least 35 minutes

/

Initial /Dat '

4.3.2 PERFORM pre-analysis PASS 1. VERIFY greater than 50 detector checks pounds of liquid nitrogen at ,

PASS liquid nitrogen monitor CAUTION: Do not reset liquid nitrogen monitor until high-voltage bias has been lowered to zero.

- - - - y***************

2. ENSURE high voltage _ applied to PASS detector at value specified in PASS AND RANCE equipment logbook

3. ENSURE a weekly calibration check has been performed within the past 7 days as indicated on weekly.

countroom QC log sheet in Count Room Task logbook '

/

Initial /Dat CH-632D Rev. O Page 9

O l

l 4.3 GAMM ANM.1115 (Cont'd) t l

ACTIONS DETAILS l

4.3.3 PERFORM Gamma Isotopic _ 1. LOG ON VAX computer as Ar.alysi s Username: PASS

2. ___ SELECT PASS MENU

3. ENTER NO to prompt DO YOU WANT A SPECTML DISPLAY l WINDOW 7(Default) 4 SELECT LIQUID SAMPLINC

5. SELECT Reactor Building Sump Sample 6.

EITHER:~

a. ENTER Q to quit MUX display and continue with procedure,

b. QB RETURN to update MUX values

7. ENTER NO to abort sample (Default value)

8. . _. UPDATE sample parameters

9. SELECT ACCEPT

10. ___ SELECT QUIT key to exit

11. GNTER LO to log off VAX computer

12. ATTACH gamma scan to this procedure

13. REPQRT results to Sample Team leader or his designee Gamma Scan ID number:

/ /

Initial / Date / Time Gamma Scan ID number: _

/ /

Initial / Date / Time Gamma Scan ID number:

/ /

Initial / Date / Time

14. IE Additional Gamma Isotopi; Analysis are required, IHEN REPEAT steps 1 through 13

15. If all analyses are complete IdEN PERFORM DemineralizedWater Flush per Section 4.6

/

Initial /Date CH-Ci Rev. O Page 10

, y w ..u -

l 4.3 GAMMA ANALYSE (Cont'd)

ACTIONS DETAILS l

4.3.4 PERFORM Core Damage 1. ._.,_ Sample Team Leader or Assessment designee PERFORM Core damage assessment per Enclosure 2

/

Initial /Date 1

CH-632D Rev. O Page 11

~

i 4.4 BORON ANALYSIS ACTIONS DETAILS j'

4.4.1 PERFORM Boron analysis 1. ENSURE Section 4.2 SAMPLE LINE-UP performed

! . 2. FLUSH sample through tha l

Boronometer for at least i one hour i

e Flush Start Time 4

i NOTE: The Boron concentration of the sample will be displayed i

at the readout (CA-56-CI) -

located on PASS Analyzer i

Panel (CACP-1) in countroom.

! l Boron PPM

3. NOTIFY Sample Team leader

l. cr designee of results

/ /

, Initial / Date / Time

4. IE all analyses are complete, IBIN PERFORM i Demineralized Water Flush per Section 4.6

a. ,

Initial /Date lCH-6320 Rev. O Page 1212

._ _ __ v _ .

t 4.5 GRAB SAMPLE COLLECTION AT CASB-5 ACTIONS DETAILS l

4.5.1 PREPARE CASB-5 (Grab Sampler) Sample Station NOTE: CASB-5 exhaust fan (AHF-55) for Sample collection switch is located to the right of the Intermediate Building door (across from RM-A7)

START CASB-5 exhaust fan

/

Initial /Date * '

I 4.5.2 PERFORM Valve Alignment 1. ENSURE Section 4.2 SAMPLE LINE-UP performed

2. OPEN CAV-445

3. OPEN CAV-446

4. CLOSE CAV-447

5. FLUSH for at least 15 minutes

/

Initial /Dat 4.5.3 ISOLATE Grab sample NOTE: The T-handle operator for CAV-492 and CAV-493 is attached to CASB- 5,

1. CLOSE CAV-492 using T-handle 1

2. CLOSE CAV-493 using T-handle

/

Initial /Date 4.5.4 ISOLATE CASB-5 1. OPEN CAV-447

2. CLOSE CAV-445

3. CLOSE CAV-446

4. STOP CAP-8

/

Initial /Date CH-6320 Rev. O Page 13

4.5 GRA8.SAMPLF M ECTION AT CASB-5 (Cont'd)

ACTIONS DETAILS l

4.5.5 CLOSE Sample Isolation IE using the Primary Sample Path Valves IHEN REQUEST Operations CLOSE the following:

1. CAV-433

2. CAV-435 IE using the Alternate Sample Path, IREN REQUEST Operations CLOSE the following:

1. CAV-434 -

2. CAV-436 NOTE: CAV-500 is in the Intermediate Building.

CLOSE CAV-500 l

Initial /Date NOTE:

Refer to Section 5.0 if a HI-HI alarm occurs at CAT-8.

4.5.6 ESTABLISH Demineralized 1. CLOSE CAV-624 Water Flush 2. OPEN DWV-337

3. OPEN CAV-471-

4. START CAP-8

5. THROTTLE CAV-624 to obtain a flow rate betwetn 0.35-0.50 gpm on CA-74-FI NOTE: While flushing you nay continue with steps 4.5.7 and step 4.5.8.

6. FLUSH for at least 10 minutes

/

Initial /Date CH-632D Rev. O Page 14

_ . _ ~ _ _ _ . _ _ _ . _ _ . _ _ . _ . _ . _ _ _ _ . _ _ _ . _ . ._ _ __. _ . _ . . _ _ _____

4.5 GRAB. SAMPLE CQLLECTION AT CASB-5 (Cont'd)

ACTIONS DETAILS l

4.5.7 A.I.M.S. Flushing Pre- 1. VERIFY greater than 50 Requisites pounds of liquid nitrogen at PASS liquid nitrogen monitor

2. ENSURE high voltage applied to the PASS detector at value specified in PASS and RANCE AIMS Equipment logbook ee...eeeeee***eeeeee**** e***ee .

CAUTION: Do not reset liquid ,

nitrogen monitor until high voltage bias has been lowered to zero.

.......... ****** e*** ***eeee***

3. ___ EN5URE weekly calibration check performed within past seven days per CH-234 as indicated on weekly Count Room QC log sheet in Count Room Task Logbook

/

Initial /Date 4.5.8 PERFORM A.I.M.S. Flush 1. LOG ON the VAX computer as Username: PASS

2. SELECT PASS MENU

3. ENTER N0 to DO YOU WANT A SPECTRAL DISPLAY WINDOW? (Default)

4. SELECT FLUSH SAMPLE LINES -

S. SELECT SUMP DEMIN FLUSH

6. HAXIM!ZE MCA Display 1 and toggle through ADC's until RCS CONFIGURATION shown

7. SELECT the ERASE function on MCA-Display to re-acquire spectrum

8. .__ When a low stable count rate is indicated MINIMIZE MCA Cisplay 1

9. SELECT RETURN

10. DEPRESS PF4 to QUIT

11. ENTER LO to log off

/

Initial /Cate--

CH-632D Rev. O Page 15

.7

(

4.5 GRAB SAMPLE COLLECTION AT CAS8-5 (Cont'd)

ACTIONS OETAILS l

4.5.9 FLUSH CASB-5 -

1. ,__.- OPEN CAV-445

2. OPEN CAV-446-

3. CLOSE CAV-447-

4. FLUSH for at least 5 minutes

/

I Initial /Dat 4.5.10 ISOLATE CASB-5 1. OPEN CAV-447

2. CLOSE CAV-445

3. CLOSE CAV-446

/

Initial /Date 4.5.11 SECURE Demineralized water 1. STOP CAP-8 flush after grab sampling 2. CLOSE DWV-337

3. CLOSE CAV-471 CLOSE the following:

4. CAV-519

5. CAV-447

6. CAV-623

7. CAV-624

8. CAV-625

9. CAV-626

/

Initial /Date i

= .. -

CH-632D Rev. O Page 16

o o

k 4.5 GRAB SAMPLE COLLECTION AT CASB-5 (Cont'd)

ACTIONS DETAILS l

4.5.12 CLOSE Containmeit Isolation REQUEST Operations CLOSE the following:

Valves

1. CAV-436-

2. .___ CAV-434 NOTE: CAV-500 is in the Intennediate Building

3. IE Alternate Sample Path was used, Ili@ CLOSE CAV-439

4. ENSURE CLOSED:

CAV-448

.__ CAV-500

/

Initial /Date 4.5.13 REMOVE CASB-5 (CRAB SAMPLER) 1. __. OBTAIN 3/4" wrench from Primary Chemistry lab key locker *

2. _ PROCEED to CAS* M Scation, 95' elevation k y building

3. REMOVE tl, .. s. mpler ramp from stor. . . ion

4. _ INSTALL the Sampler ramp in front of samp station

5. DISCONNECT CASB-5 from the

-sample station:

a. __ SQUEEZE disengagement lever

b. _ PUSH the engagenent handle to its rearmost position

c. _ PULL UP on cart liandle locking nachanism to release the cart

d. REIO/E CASB-5 cart frcm sample station

6. INSTALL the transit cover over the quick-connects

7. REMOVE the cart and move to the Turbine Building crane well

8. __ UNBOLT CASB-5 from the cart using 3/4" wrench

9. ____= -RE!O/E-T-handle operator --

10. _ GO TO section 5.0 to prepare CASB-5 for shipment off-site

/

Initial /Dat CH-632D Rev. O Page 17

.a 4.5 GRAB SAMPLE COLLECTION AT CASB-5 (Cont'd)

ACTIONS DETAILS 4.5.14 INSTALL new Grab 1. BOLT new Grab Sampler onto cart Sampler 2. REMOVE transit cover

3. ATTACH transit cover to lifting ring on grab sampler l 4. ATTACH T-handle operator to grab sampler

5. OPEN CAV-492 using T-handle

6. .__ OPEN CAV-493 using T-handle

7. PROCEED to sample station CAUTION: When connecting CASB-5. force should NEVER be used. Damage to quick connects will result from forcing connection.

NOTE: Repeated attempts may be necessary to successfully align CASB-5.

8. ENCAGE Grab Sampler CASB-5

a. One person GUIDE CASB-5

b. Another person PUSH CASB-5 UP Ramp AND onto Platform,

c. HALT CASB-5 several inches from connection points NOTE: WHEN positioned correctly, front of CASB-5 will make metal to metal contact with curved face of sample station, i

d. SLOWLY PUSH CASB-5 into Sample Station

e. ENCACE Cart to Station Locking Mechanism

f. PUSH Locking Mechanit, handle completely down.

DITVING lock bolt through hole in cart

- _ =

/

Initial /Date (Continued on next page)

CH-6320 Rev. O Page 18

l l -

4.5 GRAB SAMPLE COLLECTION AT CAS8-5 (Cont'd) l

ACTIONS OETAILS l

4.5.14 INSTA!.L new Grab ********************************************

Sampler CAUTION: When engaging handle, force should (Cont'd) NEVER be used. Damage to quick connects will result from forcing connection.

eeeeeeeeeeeeeeeee ***ee.....................

NOTE: Due to environmental conditions, the click may not be heard,

g. CENTLY EULL Engagement Handle +

forward until a distinct " click" is heard. This signifies that quick connect couplings have engaged I h. ENSURE engagement:

UNLOCK Cart from station by pulling up on cart handle locking mechanism

__ .MQy1 engagement handle back and forth If properly connected, Cart will move back and forth

1. RE-LOCK Cart to Station by pushing locking mechanism handle completely down, driving lock l bolt through hole in cart

/

Initial /Date i

l t

CH-632D Rev. O Page 19

e t .,

4.6 DEMINERALIZED WATER FLUSH ACTIONS DETAILS 4.6.1 CLOSE Sampit Isolation IE using the Primary Sample Path Valves IHEN REQUEST Operations CLOSE l

i the following:

' . 1. CAV-433

2. CAV-435 IE using the Alternate Sample Path, IHEN REQUEST Operations CLOSE the following:

1. CAV-434

2. CAV-436 '

NOTE: CAV-500 is in the Intermediate iuilding.

CLOSE CAV-500

/

Initial /Date NOTE:

Refer to section 5.0 if a HI-HI alarm occurs at CAT-8.

4.6.2 ESTABLISH Demineralized 1. CLOSE CAV-624 Water Flow 2. OPEN DWV-337

3. OFEN CAV-471

4. START CAP-8

5. THROTTLE CAV-624 to obtain flow, between 0.35-0.50 gpn on CA-74-FI NOTE: While flushing you may continue with steps 4.6.3 and 4.6.4.

FLUSH system for at least 10 minutes

/

Initial /Date CH-632D Rev. O Page 20

. u

____c___ . _ _ _ _ _ _ _ _ _ _

I 1

4.6 DEMINERALIZED WATER FLUSH (Cont'd)

ACi10NS OETAILS l

4.6.3 A.I.M.S. Flushing Pre- 1. VERIFY greater than 50 pounds Requisites of liquid nitrogen at PASS liquid nitrogen monitor

. 2. ENSURE high voltage applied to the PASS detector at value specified in PASS AND RANCE AIMS Equipment Logbook CAUTION: Do not reset liquid nitrogen monitor until high voltage bias has been lowered to zero. -

3. ENSURE weekly calibration check performed within past seven days per CH-234 as indicated on weekly Count Room QC log sheet in Count Room Task Logbook

/

Initial /Date 3 NOTE: ERASE cannot be perforned from a remote terminal 4.6.4 PERFORM A.I.M.S. Flush 1. LOG ON the VAX computer as Username: PASS

2. SELECT PASS MENU

3. ENTER NO to DO YOU WANT A SPECTRAL DISPLAY WINDOW 7 (Default)

4. SELECT FLUSH SAMPLE LINES

5. SELECT SUMP DEMIN FLUSH

6. MAXIMIZE MCA Display 1 and toggle through ADC's until RCS CONFIGURATION shown

7. SELECT the ERASE function on MCA Display to re-acquire spectrum

8. ___ When a low stable count rate is indicated MINIMIZE

_MCA Display 1

9. SELECT RETURN

10. DEPRESS PF4 to QUIT

11. ___ ENTER LO to log off

/

Initial /Date CH-6320 Rev. O Page 21 e

r 4.6 DEMINERALIZED WATER FLUSH (Cont'd)

ACTIONS DETAILS 4.6.5 SECURE-Damineralized Water 1. STOP CAP-8 Flush -

2. CLOSE DW-337

3. CLOSE CAV-471 ENSURE CLOSED the following:

4. CAV-519

5. CAV-447 6.

t CAV-623

(

' 7. CAV-624

8. CA't-625

9. C,f-626

/

Initial /Date 4.6.6 CLOSE Containment REQUEST Operations CLOSE the Isolation Valves following:

1. CAV-436

2. CAV-434 NOTE: CAV-500 is in the Intermediate Building

3. LE Alternate Sample Path was used, IliEN CLOSE CAV-439

4. ENSURE CLOSED:

CAV-448 CAV-500

/

Initial /Dste

= _ = . = _ = = _ _

_CH-6320 Rev. 0. Page 22

U

. i 1

4.7 SYSTEM RESTORATION ACTIONS- DETAILS l

4.7.1 SECURE flow . ENSURE CLOSED the following:: i

1. CAV-471 .

2. CAV-447 !

3. _ CAV-448

4. CAV-484

5. CAV-623 .

6. CAV-624

7. _. CAV-625 -

8. CAV-626

9. CAV-627-

10. CAV-519

11. CAV-500

/

Initial /Date y

f i

r

{.

=

I .. A

~

l_

l~

CH-632D Rev..0 Page 23 ,

s 5.0 CONTINGENCIES l

l 5.1 CAT-3 HI-HI LEVEL ALARM l

- ACTIONS. DETAILS l

5.1.1 PERFORM lineup ENSURE the following:

1. CAP-10 0FF

__ CAV-623 CLOSED

s. CAV-627 CLOSED l 4. CONCURRENTLY PLRFORM the HI-HI following entil CAT-8 HI level alarm light clears:

o DEPRESS and hold RESET '

button on Orain Tank level indicator o SELECT CAP-10 to ON

5. OPEN CAV-623

6. SELECT CAP-10 to AUTO 7.

RETURN to the step in the procedure which was in progress when the CAT-8 HI-HI level alarm occurred

/

Initial /Date CH-632D- Rev. O Page 24

. _ I

5.0 CONTINGENCIES (Cont'd) 5.2 NOTIFICATIONS AND SHIPMENT ACTIONS DETAILS l

NOTE: The Emergency 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months access phone number is (804) 522-5833. l 5.2.1

PERFORM notifications Notify the Manager, Nuclear Operations Materials Controls that a grab sample has i been taken and to initiate acquisition process for shielded sample cask Notify the Framatone Technologies Emergency Sample Coordinator when a grab '

sample has been collected that will

. require offsite analysis Required information to ba made available:

o Utility and plant name o Name and phone of ChemRad Specialist to whom follow-up communication should be addressed o Number and type of samples to be shipped, (i.e. li iodine cartridge) quid, gaseous, or o Measured radiation levels at the surface and three feet from the shipping container o Estimated shipping time, noce of transportation, carrier, and estimated arrival at Frama* 'ne Technologies site in Lynchburg, VA Shipping Address:

Framatone Technologies Emergency Sample Analysis Coordinator Nuclear Environmental Services Lynchburg Technology Center Route 746, Mt. Athos Road Lynchburg, VA :24506 All data accumulated per this procedure is to be summarized on Enclosure 1 and forwarded to the Emergency Coordinator via Chemistry Supervision

' /

Initial /Date CH-6320 Rev. O Page 25

MCLOSURE 1 TECHNICAL SUPPORT CENTER DATA SHEET RL& TOR BUILDING SUMP _ 4 Cama Isotonic ard/or Baron Analysjs Results s .

Boron ppm / /

Initial / Date / Time Boron ppm / /

Initial / Date / Time Boron ppm /- /

Initial / Date / Time Total Activity uCi/cc Maior Con.tf_ihytjaa Isotoon uCi/cc uti/cc uti/cc uCi/cc uti/cc 9 uCi/cc uCi/cc uti/cc uCi/cc uCi/cc uti/cc

/ /

Initial / Date / Time CH-6320 Rev. O Page 26

U 4

ENCLOSURE 2 ASSESSMENT OF CORE DAMAGE BASED ON REACTOR BUILDING SUMP SAMPLE

). This method of confirming core damage. ssumes that releases from the core are uniformly mixed in the Reactor Building Sump Sample,

2. The baseline coolant concentrations in Table 1 are for 0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after shutdown of a core that has been through at least one refueling cycle.

3.

~

The half-life of the fission products should be considered in analyzing samples.

4. Estimate a dilution factor based on the dilution volume from injection the Reactor Building Sump sample has been subjected to. .

estimated dilution factor

$. Multiply the PASS Reactor Building Sump sample activities from Enclosure 1 by the estimated dilution factor from Step 4.

6. Compare these adjusted activities with the baseline coolant concentrations in Table 1. This table overestimates the concentration of the long-lived fission products (Cs and Sr) in a new core.

7. Determine the extent of core damage as indicated by Table 1 (i.e.,

normal, gas gap, core melt.

TA8LE 1 BASELINE REACTOR COOLANT CONCENTRATION

~~

Normal Concentration Concentration TMI Nuclida Concentratior. After Cap After Core Concentration (uCi/g) Release Melt (uC1/g) + 48 Hours (uCi/g) (uCi/g)

I-131 4E-2 2E4 1ES 1.3E4 I-133 1E-1 3E4 2E5 6.5E3 I-135 2E-1 3E4 2E5 No Data Cs-134 7E-3 2E3 8E3 6.3E1 Cs-137 9E-3 9E2 SE3 2.8E2 Ba-140 No Data No Data 3E4 No Data Sr-90 1E-5 No Data 1E4 5.3

8. Report determination to Dose Assessment Coordinator.

/

Initial /Date I

i CH-632D Rev. O Page 27 (LAST PAGE)

O s

U.s. Nucle:r Regul;t:ry C:mmission 3F1297-12 j i

Enclosure 4 i

Boron Precipitation Risk Analysis

c . .

o Boron Precipitation Risk Analysis RCS Cold-Leg LOCA Frequencies EPRI TR 102266 The initiating event of interest for boron precipitation is a break in one of the RCS cold legs between the reador coolant pump discharge to the reactor vessel nozzle. There are two different size breaks for which frequencies are needed: 0.05 to 0.10 ft', and greater than 0.10 ft'. These ranges of break sizes corresponds to a range of break diameters of 3.0 to 4.3 inches, and greater than 4.3 inches.

EPRI TR 102266,

Pipe Failure Study Update," April 1993, gives a frequency for PWR RCS pipe breaks for pipes with an inside diameter greater than 6 inches (RCS cold leg diameter = 28 inches) of 2.87x10' per pipe section per hour (see Table 4J in the EPRI report). The piping of interest consists of the cold legs from the reactor coo! alt pump discharge to the reactor vessel nozzle. In the EPRI study, a pipe section is defined as a segment of piping, between major discontinuities such j as valves, pumps, reducers, tees, etc. (WASH 1400). A pipo section is typically 10 to 100 feet long,-

and contains four to eight welds. Each section can also contain several elbows and flanges. In this analysis, each of the cold legs from the reactor coolant pump discharge to the reactor vessel nozzle is considered one pipe section. Therefore, the frequency of a pipe break of any size in these pipe sectiona is:

f.noe. b, a

= (2.87x10'" per section - hour)(4 sections)(8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months per year)

= 4 1.01x10 per year The pipe break frequencies of interest for this study are for breaks in the range of 3.0 to 4,3 inches in diameter, and greater than 4.3 inches in diameter. In the EPRI study, the first of these falls in Size Group 2 (2"<lD<6"). The conditional probability of having a pipe break of this size given that a pipe break has occurred in a pipe with an inside diameter greater than 6 inches is 0.25 (see Table 4-5 in the EPRI study). Therefore, the frequency of a pipe break between 3.0 and 4.3 inches in diameter in one of the RCS cold legs between the RCP discharge and the reactor vessel nozzle is:

fse, o,oup a b,..a = (1.01x104per year)(0.25)

= 2.5x10 per year The second of these fall:5 into Size Groups 2 and 3 (2"<lD<6", and >6"). The conditional prcbability of having a pipe break of this size g'ven that a pipe break has occurred in a pipe with an inside diameter greater than 6 inches is 0.25 + 0.5 = 0.75 (see Table 4 5 in the EPRI study). This is somewhat conservative since the break size falls only partially into Size Group 2. Therefore, the frequency of a pipe break greater than 4.3 inches in diameter in one of the RCS cold legs between the RCP discharge and the reactor vessel nozzle is:

4

/se. oroup r. b, . = (1,01x10 per year)(0.75)

= 7.6x10 per year 1

l' 4 NUREG/CR 2189 Framatome used NUREG/CR 2189 " Probability of Pipe Fracture in the Pnmary Coolant Loop of a l

PWR Plant" Volume 5. September 1981 to come up with an RCS cold leg LOCA frequency of 3x10'

' per year., This value was calculated using a weld failure probability ' rom the NUREG 4 of 7.6x10 l

per year. There are 3 welds per cold leg loop in the pipe section of interest. Additionally, the break is assumed to occurin the bottom third of the pipe. This yields:

/soron p,.uocA = (

(7.6x10'" per weld per year)(3 welds per cold leg)(4 cold legs)(1/3)

= 3.0x104 per year Resolution of the Two Approaches There were soine conservatisms in the approach using tha EPRI pipe failure study. First of all, no adjustment was made for the fact that the break is assumed to occur in the bottom third of the piping. Second, the EPHI approach assumed the pipe section was from the RCP discharge to the reactor vessel nozzl .- Since the smallest unit af this approach was a pipe section, no adjustment was made for the section of interest possibly being a fraction of a pipe section. The size of the first break was from 3.0 to 4.3 inches in diameter. The size group from which the frequency was extracted was from 2 to 6 inches in diameter, if an adjustment was made for the fact that the break size range was only a fraction of the size group, the frequency could be reduced by a factor of (4.3 -

3.0)/(6 - 2) = 0.325. If this adjustment and the adjustment for the bottom third of the piping were made, the EPRI frequency of a break 3.0 to 4.3 inches in diameter wculd be:

Isa. croup s b, = 4 n (2.5x10 peryear)(0.325)(1/3)

= 2.7x17' per year .

Making these adjustments to the EPRI approach reduces the difference between the approaches to one order of magnitude. In the realm of probabilities of ruptures of large pipes, this is not such a great difference.

Applying these 2 same factors to the EPRI approach for calculating the frequency of a break greater than 0.10 ft , the result is:

Isa, o,oup 2.3 d,... 4

= (1.01x10# per year){(0.25)(0.325) +0.5)(1/3)

= 2.0x10 per year Modeling Fault trees for the various methods of mitigating boron precipitation (dump-to-sump, hot-leg injection, and auxiliary pressurizer spray) were developed and quantified. The failure probabilities of each indiv', dual method and relevant combination of methods are given below along with the LOCA frequencies.

2

. g . ___-_

ls Eoron Precipitation Events of Interest i

Event - Probability / Frequency Comments 2.5x104 peryear This is an estimate using the EPRI RCS Coldft Leg)LOCA (0.05 0.10 a; rcoach with some conservatisms.

Removing these consentatisms results in a frequency of 2.7x10* per year. Framatome, using NUREG/CR-2189, calculated a frequency of 3.0x10* per year.

RCS Cold-Leg LOCA 7.6x10" per year This is an estimate using the EPRI

(>0.10 ft') approach with some conservatisms.

Removing these conservatisms results in a frequency of 2.0x10^' per year.

Failure of Hot Leg injection 9.9x10 per demand Failure of Dump-to Sump 5.7:t10 per demand I Failure of Auxiliary 5.3x10 ' per demand Pressurizer Spray Failure of hot Leg injection 4.5x10 per demand and Dump-to-Sump Failure of Hot Leg injection 1.8x10 per demand and Auxiliary Pressurizer Spray Baron Precipitation Core Damage Sequences If the RCS cold-leg LOCA is greater than 0.10 ft*, then pressure and temperature fall quickly, and the operators will use hot-leg injection or dump-to-sump to control boron precipitation. The frequency of such a LOCA accompanied by the failure of both active means of boron precipitation mitigation is:

ft ocAumHuors = (7.6x10 per year)(4.5x10 2)

= 3.4x104per year 2

if the RCS cold-leg LOCA is between 0.05 and 0.10 ft , then pressure and temperature fall slowly, and the operators will use hot-leg injection or, after several hours when decay heat has decreased sufficiently, auxiliary pressurizer spray to control boron precipitation. The frequency of such a LOCA accompanied by the failure of both active means of boron precipitation mitigation is:

ftoc42 %sasgn,y= (2.5x10 per year)(1.8x10)

= 4.5x104per year if the Framatome estimate for initiating event frequency is used, the frequency is:

ft oc42 %nasr= (3.0x10* per year)(1.8x10)

= 5.4x10'" per year Notes:

1. Hot leg nozzle gaps were not considered in this analysis. The nozzle gaps will adequately dilute 6

core boron concentration in the event of the failure of all available active methods until an active method can be initiated.

3

_____g.__-____________________- - _ - _

~

2. Operating crew failure to recognize the onset of a boron precipitation concern and successfully execute an active means of mitigation was not considered in this analysis.

3. If, when attempting to implement dump-to-sump or hot-leg injection using a particular decay heat train, a hardware failure of the train occurs, no credit is taken for using the other train for dump-to sump or hot-leg injection. This is due to the fact that one train is needed for normalinjection to the RCS.

4

..

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