ML18032A232

Rev 1 to BFN-OSG3-048, Safe Shutdown Analysis.

ML18032A232
Person / Time
Site:	Browns Ferry
Issue date:	02/26/1987
From:	Eckert E, Carleen Parker TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18032A233	List: ML18032A232
References
BFN-OSG3-048, BFN-OSG3-048-R01, BFN-OSG3-48, BFN-OSG3-48-R1, NUDOCS 8705270557
Download: ML18032A232 (693)

Text

TVA 10697 (ONE%46) ONE CALCULATtONS

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TITLE PLANTIUNIT Safe Shutdown Analysis BFNP, Units 1, 2, 3 PREPARING ORGANI2ATION KEY NOLJNS (Consult RIMS QESCRIPTQRS LIST)

NEB Safe Shutdown Analyss.s, transient, Accident, System BRANCHIPROJECT IOENTIFIERS Each rime these calculations ere Issued, preparers must ensure that the original lRO) RIMs accession number is filled in.

BFNWSG3-048 Rev (for RIMS'se) R (MS accession number RO 822 '860628 010 APPLICABLE DESIGN DOCUMENT(S)

SAR SECTION(S) UNIO SYSTEM(S)

Revision 0 Rt Safetywe(atedF Yes fp No Q ECN No. for indicate Not Applicabla)

N/A N/A Statement of Problem Prepared The BFNP system bases have not E. C. Eckert been systematically reviewed or Checked documented since the original C. M. Parker license from the viewpoint of Reviewed requirements for safe shutdo~

D. A. Walker of transients, accid qtsr and Approved special events. This

. T. G. Chapman Calculation provides such a Date systematic analysis of system safet List all pages added by this revision.

E'9 a List all pages deleted p,s )

ne P~ by this revision.

>o I( List all pages changed by this revision.

Abstract I

These calculations contain an ungvr ified assumption(s) that must be verified later. s Rk This Calculation serves as a formal documentation of the, system safety actions for which credit has been taken in PSAR, Reload Submi,ttal Analyses, and other licensing communications concerning transient, accident, and special events. It serves as an input to the BFNP Baseline Program evaluation of each of the systems necessary for restart of Unit 2, assuming Units 1 and 3 are shutdown as defined. It describes how each identified system was utilized'in the documented licensing analysis of each applicable event, and summarizes, the functions expected of the systems that have bee identified. <

This Calculation was developed from .the original BFNP PSAR (including the Questions/Answers contained in Volume 7), the Reload Licensing Submittal for Unit 2, Cycle 6 and commitments which have been documented since the original startup. The final result is a comprehensive link between each identified system and the safety actions associated with safe shutdown of BFNP as documented in licensing (0 FDR P

70557 870522 AD 0500025~

PDR (Continued) analyses'oCW D Microlilm and store calculations in R IMS Service Center. Miuofilmend destroy, Q Qr 1'4 Microlilmand return calculations to: D. A 'i~a Address.Wl 1 C72 C-K

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~p ADDENDUM TO THE ABSTRACT USE OP THIS CALCULATION This calculation contains a very wide range of information since, it contains the results of a "front cad" effort to begin the total process of system evaluation. It provides thc 'information accessary to sort out essential elements from those of less importance, but it does not, in itself, perform that task (to bc done ia subsequent Baseline Program activities). Several uses of. the information ares (1) Definitions of several key items are derived from the refe'rences or specially defined as presented ia the following sections:

- Section 6 - Safety Cri,teria (Unacceptable Results) used in analyses of events

- Sections 7.1 to 7.3 - Standardised Safety Actions used in mitigation (or limitation) of events.

- Section 7.4 - A coded. set of categories that are used to further describe system safety actions.

- Section 7.S Reactor Operating States from which each event or safety action occurs.

(2) Shutdown is defined to mean that:

.1.'he reactivity of the reactor is kept to a margin below

2. Core decay heat is 've criticality consistent with the unit Technical S cifications, core or reseCor coolant at'.controlled spot'ep

]

te sufficient esign .limits om being 3~ C t a d s"76,'teed+'cjec~ha 'gal 'nM'&a"Hase onditions are 0 e a ia'g iwithin. their 'de'sign'imfts.

Most e ~l establ shment of the s~~

'C'tilixes the sasutdown same extent conditions'or of special events, shutdown as the original

~

design basis.

(3) . Each applicable BPNP 'event is addressed ia a separate appendix (1A through 33). The documented analyses are utilixed .to identify'hich safety actions werc shown to provide event mitigation and.shutdowa-within the criteria that are appropriate for the category, of the event. Each a'ppendix provides a focus for individual event discussion. Most event ippendiccs contain a tabulated list of safety actions coupled to the systems. which perform (part of all of) the safety action. As much as, possible, support system actions are included along with all primary system actions; so.they too can be sorted out by event category aad reactor operating state ia subsequent Baseline evaluations.

'I

ADDENDUM TO THE ABSTRACT: USE OF THIS CALCULATION (Continued)

(4) System by system use of the document is done by using the sorted lists given in section 8.4. In most respects, these pages provide the primary output of this calculation. Subsequent evaluation, of each system action shown must carefully account for the category (transient, accident, special) of the event from which the action is derived. Actions which occur, but are not required must also be noted before subsequent evaluation steps are established.

(5) Manual actions are also provided in some. events. Certain ones are part of safety shutdown (e.g., initiation of suppression pool cooling)p but many are actions to restore the plant to normal with no immediate challenge to safety criteria. They are summarized in section 8.5, but must be sorted by event category .and actual action performed before subsequent requirements (if any) are established.

Many of these items s'imply provide information related to plant preeedures, net syseen/equtpnent requirsnents.

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Prepared:

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BPN-OSG3-048 (Rev l) aaaaaaaaaaaaaaaaaaaaaaa C. M. Parker TABLE OF CONTENTS

~ 1 e0 PURPOSE 2.0 SCOPE AND METHODOLOGY 3~0 ASSUMPTIONS 4.0 SOURCES OP INPUT INFORMATION 4.1 References 4.2 Review of PSAR Questions> Volume 7 4'3 Review of Licensing Conmitments/Requirements 5 ' DOCUMENTATION OP ASSUMPTIOHS FOR ANALYSIS 6' SAFETY CRITERIA FOR ANALYSIS 7.0 SAFETY ACTIONS AND DEFINITIONS POR ANALYSIS 7.1 Safety Actions for Abnormal Operational Transients 7.2, Safety Actions for Accidents 7.3 Safety Actions for Special Events

~ C 7.4 Safety Pun'ction Codes 7.5 Reactor Operating States 7.6, Standard uences

8. 0 ANALYSIS 'If'(19) II ~

8. 1 Summary Re Aa 'f 'a~J'I 'iI~" ~ \

8.2 ,Secondary ontaiument Analysis.

8.3 Reactor Sta valuation 8.4 Required System Safety Actions

8. 5 Manual Shutdown Actions

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~E. c Date: ~Mz BFN-OSG3-048 (Rev 1) C. M. Parker sssssasasasaaasassssssaaaasasaassaaaasssssassasassssaaassaaseasaasaasasasss Appendices - Individual Event Evaluations

.. 1A Generator Trip 1B Turbine/Generator Trip with Bypass Failure 1C Pressure Regulator Failure - Closed Turbine Trip Isolation of All Main Steam Lines Isolation of One Main Steam Line 5 Loss of Condenser Vacuum 6 Loss of Feedwater Heater 7 Shutdown Cooling Malfunction 8 Inadvertent Pump Start 9 Control Rod Withdrawal Error 10 Fuel Assembly Insertion ll Control Rod Removal Error Pressure Regulator Failure Open.

12A 12B Inadvertent Opening of All Bypass Valves 13 Inadvertent Opening of S/R Valve 14 Loss of Feedwater Flow 15 Loss of Offsite AC Power 16 Recirculation Control Failure - Decrease 17 Trip of One Recirculation Pump .

18 Trip of Both Recirculation Pumps 19 Recirculation Pump Seizure 20 Recirculation Control Failure - Increase

21. Startup of Idle Recirculation Pump 22 Loss of Shutdown Cooling 23 Feedwater Controller Failure - Maximum Dbmand 24 Control Rod Drop Accident 25A Pipe Break Inside Containment Large 25B Pipe Break Inside Containment - Intermediat 25C Pipe Bre nsx e Containment - Small 25D Pipe Bre k Inside - Containment and Radiological 26 27 Fuel Pipe Han 1%

Bre id t!g, Prtcdt,'J. ) d,, (W$ [ f:.I>..! )j 28 29 Shutdown Shutdown xth u Control Rods 30 Overpres ure Protection 31 Rotated 32A Flood 32B Low Reservoir 32C Tornado 32D Earthquake 32E Fire 33 Loss of Fuel Pool Cooling/Makeup

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Dates /N BFHWSG3-04S (Rev 1) C. M. Parker WP'PR%%%WNQWRWRRQRNNRRRRNSRR Appendices - SSA Input Evaluations

,A Input Guidelines Bl Licensing CoaNn.tment/Requirement Evaluations - CEB Items B2 Licensing Coamitment/Requirement Evaluations EEB Items B3 Licensing Conxaitment/Requirement Evaluations - MEB Items B4 Licensing Coamitment/Requirement Evaluations - HEB Items B5 Licensing Ccaanitment/Requirement Evaluations - GEC Items C Confirmation Evaluation of FSAR Questions (Volume 7)

SAPE SHVTNNE ANALYSIS Page 4 of 30 Prepared: Z Z' Date:

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BFN-OSG3"048 (Rev 1) C. M. Parker oassaassooaossaaossaaaoaaossssoaaaaooaaaaaoaaaaaaaaaaassaaoaassooaaaaseooooooaooooose 1 e 0 PURPOSE The purpose of this document is to identify and systematically review the events (i.e. abnormal ope'rational transients, accidents, and special events) that have been previously analyzed for the Browns Perry Nuclear Plant (BPNP) and are required to be analyzed as part of the design bases. This review will be primarily focused at identifying the systems or portion of systems that are used in the most recent documented analysis'an example is the BFNP Reload

'Analysis, see reference 4.1b) of the applicable event. The systems identified in this review will be based on the documented method chosen to ensure the applicable criteria (see section 6.0) for the event are met. Although other shutdown sequences may meet the criteria, the shutdown'ath identified in this analysis will follow the analyzed cases to ensure the criteria will be met.

Although this analysis was prepared with primary focus on the restart of BFNP Unit 2, the bulk of this analysis is expected to be applicable to safe shutdown of any and all of the BFHP units'.0 SCOPE AND METHODOLOGY

,2.1 Events Considered As stated in the purpose, all events that are a part of BFNP design bases and have previously been evaluated will be addressed in this analysis ~

The methodology used to identify all events that are part of BPNP design basis.was to review the BPNP PSAR (reference 4.la) and the current BPNP. Unit 2 Reload Submittal (refer . and determine all events that we e as part of the 1 censing basis. A deta' o the uestions raised~%FPW i 1 7 of the PS re Provided in section .Q~iSPQ)dtxtd'e't/Ab!

consider ~ g(revn~ ~~doq~4t d xn section 4.3 and Appendice Pl 4ugh" W. Zhe methodology for the ce of these rev ews is given in Appendix A. is t individua ly evaluated i ces lA through 33 as listed in Table 2-1. st shows the applicable category for each event (transient, accident, or special). The category of the event is based upon the most recent method of addressing the event.

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E C ~ rEck C. M. Parker Date.'lKg aaaaaaaaaaaaaaaaeaaaaaaeaaaaaaaaaaaaaaaaaaaaaeaeeaaaaeasseaeeasaaeaaessssss 2.2 Criteria Used for Anal sis The criteria used in the analysis are those used for the original design basis. The criteria were derived from the BFNP FSAR and are included in section 6.0 ~

The methodology of the analysis will be to provide a reference to the most recent analytical evaluation of the event, where applicable, which evaluates the response of the plant to the event (e.'g. see reference 4.lb) ~ This analytical evaluation will have shown that the criteria are met for the applicable category of the event when credit is taken for specific safety actions.

2.3 Safet Actions Re uired The safety actions that are used in this analysis are identified in section 7.'0 with their relationship to the criteria provided.

The safety actions for each event 'are determined by reviewing the analytical evaluation of the event and 'identifying what mitigative and/or preventive actions were utilized to ensure that the criteria will be met.

2.4 S stems Identified The systems are identified which are required to perform the

~

safety .actions for which credit is taken in the analysis for the event as identified in sections 2.2 and 2.3 above. These systems are identified by their respective system number and a brief description of the function they are perfozming,for the event.

All primary protection systems .are to be identified in this ana e r 'nder to be determined in more detailed, system a lyses to fol Kin the Baselin P o~ >

path i vd ok', jihg7< r ~t~ t5e~', g+ e e syem'.

It i ses";~'e-.pr'ovi'd'ed'ich:;demodstr'ath segg,e that.'.""shni'-"pyi~ cia.oni ho@id~ e provide a

' a are met.

nated if anal Chit, th '"% 1 cable criteria can still be met.

2.5 Extent of-Shutdown The analysis in this document is provided to identify the systems which perform limiting and/or mitigation of the events that are wi.thin the design basis't: also documents a safety-grade system path which can achieve a stable hot shutdown once the mitigation phase is complete. Safe shutdown means that (1) the reactivity of the reactor is kept to a margin below criticality consistent

SAFE SHOTDORE hEhLTSIS Page 6 of 307 Prepared: Date: ~+(f1~

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BFNWSG3-048 (Rev 1) C. H. Parker aaaaaaaaaaaaaaaaaaaaaaaa aaaaaeaaaeeaaaaeaaaaaaaaaaaaaaaaeaaaaaaaaaeeaaaeaa with the unit Technical Specifications, (2) core decay heat is being removed at a controlled rate sufficient to prevent core or reactor coolant system thermal design limits from being exceeded, and (3) components and systems necessary to maintain these conditions are operating within their design limits (reference 4.1u). Por special events, the anilysis identifies the systems by which mitigation will be performed to the same extent as .the original design basis.

Table 2-1 Safe Shutdown Anal sis Events

~Aeedix Descri tion Cate~or +

lA Generator Trip AT 1B Turbine-Generator Trip with Bypass Pailure AT 1C Pressure Regulator Failure - Closed AT 2 Turbine Trip AT 3 Isolation of All Hain Steam Lines AT 4 Isolation of One Hain Steam Line AT 5 Loss of Condenser Vacuum AT 6 Loss of Feedwater Heater AT 7' Shutdown Cooling Malfunction AT Inadvertent Pump Start AT 9 Control Rod Withdrawal Error AT 10 Puel Assembly Insertion . AT ll Control Rod Removal Error AT 12A Pressure Regulator Failure - Open AT 12B Inadvert'ent Opening of All Bypass Valves AT 13 Inadvertent Opening of S/R Valve AT 14 Loss of Feadwater Flow AT 15 Loss of Offsite AC Power AT 16 Recirculation Control Pailure - Decrease AT 17 Trip of One Recirculation Pump AT 18 Trip of Both Recirculation Pumps AT 19 Recirculation Pump Seizure ta 20 Recirculation Control Failure - Increase

.21 Startup of Idle'Recirculation Pump AT 22 of Shutdown Cooling 'oss AT 23 Feedwater Controller Failure - Maximum Demand AT

• a Indicates e categories: AT a Abnormal T ansient, A a Accid t, S e Special peep:,eeee e A P/P4 I i I )I

i SAPK 'KaTDOW hMLTSIS BFNWSG3-048 (Rev 1)

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Table 2-1 Safe Shutdown Anal sis Events Continued

~Aeedix ~Cate x~or 24 Control'Rod Drop Accident A 25A Pipe Break Inside Containment - Large A 25B Pipe Break Inside Containment - Intermediate A 25C .Pipe Break Inside Containment - Small A, 25D Pipe Break Inside - Containment and Radiological A 26 Fuel Handling Accident A 27 Pipe Break Outside Containment A 28 Shutdown From Outside Control Room S 29 . Shutdown Without Control Rods S 30 Overpressure Protection $

31 Rotated or Mislocated Bundle Error A 32A Flood S

'2B Low Reservoir S 32C Tornado S ~

32D Earthquake S 32E Fire S 33 Loss of Fuel PoolI Cooling/Makeup S d

Indicates first event of these categoriese AT ~ Abnormal Transient, A ~ Accident, S ~ Special r

SAFE SHUTDOWN ANALYSIS Page 8 of 307 Prepared: Dace:

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3. 0 ASSUMPTIONS a 3.1 This analysis addresses all reactor operating states 'for BFNP Unit 2, but'assumes (where necessary) that Units 1 and 3 are maintained shut down with fuel unloaded and the fuel pool gate in place. r 3.2 The listings of licensing Commitments/Requirements (C/Rs) used in this analysis (Appendices Bl-B5) are uncontrolled printoutsof C/R database. It is assumed that theysincerepresent a complete, completion of the the'FNP accurate compilation of ccamitments made FSAR (reference 4.la). See Appendix A for a'description of their use here.

3.3 The trip actions of the Intermediate Range Monitoring feature of the Neutron Monitoring System are assumed to remain as valid protection 'steps, although a unique failure mode has been identified in SIL-445 (reference 4.1q) ~

3.4 The high reactor water level (L8) initiation of MSIV closure described in reference 4.ls is assumed to no longer be a valid coaxnitment for the plant, and will not be included in the system safety action requirements'.5 The criteria>stated in this analysis (section 6) are assumed to be appropriate for BFNP.

4.0 f SOURCES OF DESIGN INPUT INFORMATION 4.1 References

.BFNP Final Safety Analysis Report (FSAR) through Amendment 68 (Volumes l,through 7) ~

b. BFNP Reload Licensing Report, Unit 2/Cycle 6, April 1985 C ~ Nuclear Performance Plan, Volume III (BFNP). 'VA

d. BFNP Updated FSAR (USAR)

e. of Nuclear Engineering Calculati S3-001'ivision S d es 22 860716 275) o e of Federal Regulations,'0CFR100.

ge rimary Containmeng Qqlat~gy~bpI E]JEfygn))r agrsm; e e)Pres.F 8r3bj4j;;VQEP27, Series, cd%'si n 4,

, h, i ' 'ag~s.'.1'204V'Apj(51"0 +~~Vol~" sc llaneous c b Djagfamk~ TVkt RHEA.hgs %5H614-5; revisio 12, 61 , revision 11; 45N614-13, revision 7.

0 SLY SHUXDOQS BPN-OSG3-048 (Rev 1)

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SSSS $ $ SS SSS $ $ SS SSS S S S SS $ $$ $ $$ SS $ $ $ $ $ $ SS SSSS $ $ S SSS S S S S S S S SS S S $ $ S S S $$ $ S $$$ $ S S 4.1 References (Continued) i ~ Letter, R. J. Clark to D, R. Muller (NRC) dated January 23, 1986 and attachment "Browns Perry Nuclear Station Radiological Impact of Ventilation Damper Closing Time During a Design Basis Puel Handling Accident."

j. Letter, D. R. Huller (NRC) <o S. A. White, dated March 28, 1986,

Subject:

Secondary Containment Isolation Damper Closuze Times.

k. 'BPNP, PSAR Volume 7, Question 13.11, Abnormal and Emergency Procedures and Related Training.

l. NEB RAC 1341 (B45 860626 019)5 TVA letter dated Pebruary 11, 1982, from L. M. Mills to H. R. Denton (NRC),

Report TVA"EG-047> TVA Reload Core Design and Methodology for BFNP ~

m. NEB RAC 1720 (B45 860701 046), PSAR Section 1.5.1.5, Nuclear Safety Design Criteria for Transients (Type 5-2) ~

n. NEB RAC 1719 (B45'60701 047), PSAR Section 1.5.1.5, Nuclear Safety Design Criteria for Accidents (Type 5-3).

. o. NEB RAC 1718,(B45 860701 048), PSAR Section 1.5.1.5, Nuclear Safety Design Criteria for Special Events (Types S-4/S-5).

.p. TVA letter to NRC dated January 31, 1986, Enclosure NEDE-31119, (B45 860225 826); also referenced by several other 'coamitment/requirements (e.g., GEC BTG 1003, B45 860619 891)

q. General Electric Service Information Letter, SIL"445, Septembez 10, '1986, Blown IRM fuses

r. TVA-RLR&02, Revision 1, April, 1985, Reload Licensing Report for Browne Perry, Unit 2, Cycle 6 s ~ EEB"WCW"l255 (B43 860617 939), BPN SSER, Section 7 '.1-1, ode.

ta BPNP Technical Specifications ua arinan aatianal ataddard @pl~-58/-P))1 4.2 Re

~/ a 4.2.1 A ystematic review was performed of all the questi hs and r ponses contained in Volume 7 of the'SAR (refere ce 4.la).

ocument t e incorporation (or exclusion) of each question with zespect to the SSA. These questions documented AEC/NRC concerns and TVA conmitments throughout the .initial licensing period and into early operation. (Section 4.3 addresses subsequent licensing coaxnunication and caanitments) ~ The process established identification of each item for SSA (event) or System Evaluation Report (SYSTER) application.

SAFE SHOIDOW hHhLTSIS Page 10 of 307 Prepared: Date:

E. C. rt Checked: Date: d'0 PF BFHWSG3-048 (Rev 1) C. M. Parker ssaeaesaesaaaeesaasasaaeesasasaeaaaaassssssassasssssssesaseaaaseaeaaaaaaase 4.2.2 The questions were tabulated and evaluated according to the guidelines given in Appendix A of this calculation. Each question was assigned a category from the following list:

(a) QSSA Include in the SSA..

(b) QSER - Appropriate for SYSTER consideration. ~

(c) QOP - Operational QXL - Later commitments (or Unit 1 and/or 3).

items'd)

(e) QNA - Items which are not applicable to the SSA or SYSTER effort for other reasons).

Each category is discussed in more detail in Appendix A.

4.2.3 The disposition of each item is recorded in the tables in Appendix C,, together with the justification and other information accumulated for each item during the review.

4.2.4 A significant number of these items do apply to the treatment of safe shutdown during various shutdown sequences. The appropriate event appendix number is included in the tabulated results. Some are the source of including an "event" which was not included in the initial FSAR analyses (e.g., Question Q14e5 led to the incorporation of Event Appendix 12B in the SSA) ~

Many of them are direct requirements on one (or more) systems and are designated SER for easy identification of SYSTER inputs., Each SSA event appendix which has been noted documents the use that was made of each item designated as QSSA.

Table 4.2-1 lists ' the results of the question review f'r all.

it s are listed in that are expected to have SSA impact.

the or er ated event appendix The n

Th i Qp sf~page,qf the:.tabl;e: i) ac4ujlthg cbtiipb~ p$ uplicate it 'hhy'ppe'ar kgain,',.sorted ".)id'gjjently.dn a,k sequent ed on this pa e ia'the table). Hare'ee'r,'kdptrdte'appdjadly p e to more clearly show events identified in Kh j econd and third even d). These extra columns re occasionally used when an item app to two or three events..

The remaining pages are in order of the first SSA event appendix number column (noted). Item(s) at the end are labeled SSA (instead of an event number) since they are expected to have broader application in the calculation (e.g. criteria, several events, etc.).

're parid!

E Ce Checked! Date:

BP&OSC3~8 (Rcv I) C, M. Parker TABIZ 4-2 I Rivi~ of PSAR - Volume 7 oucstions Sorted b Event/Ao endix*

SSA SSA SSA FBAR-V7 QUEST FSAR EVAL EVT EV2 EV3 QUEST DATK SECTION CAT KEYWORD 1 KEYWGRD 2 NO ~ NO NO.

aaaaeaaa RQRQR RRQRRRNR%%  %%% NQRWRNNQN% WNSSSSRCISRCRRR WAR U4 1 711206 4.4 QSSA NSSRV 'EPORT 30 Oi~ 04~

U4.2 711206 4e4 QSSA NSSRV . INFO 30 018 04 R10. 2 710522 4. 4, 4. 6 QSSA ACCUNUL INFO O1 .03 25

.QB 12 710325 8,5i 10 6 QSSA DWEl L BLWR INFO . 25 15 V6 720602 3. 6 QSSA NUCL PARAN INFO 19 18 16 QB. 4 710325 8.4. 1 QSSA DISL GENRT INDEPT PWR 15 25 QB. 7 710325 Bo4 QSSA PWR FAIL DES INFO 15 25 QB. 8 710325 QBSA DGBD FAIL DKS INFO '15 25 8.5'10325 QB. 9 8.5 QSSA DG LOAD . DES INFO 15 25

98. 10 710325 8+5 QSSA D G RELAY 'DEB INFO 15 25 i2D QB. 11 /10325 8.5 QSSA DG CONTROL INFO 15 25 DB. 13 710325 B. 5 QSBA DG LOAD DEB INFO 15 '5 QB. 15 QB. 19 710325 8 5 710325 8. 5 QSSA D GEN QSBA D. GEN OPR DES INFO INFO '5 '

1,5 25 25 R7.18 710522 7 12 QSSA SGTS PRN INFO . 26 25 27 R7 ~ 2P 7105-2 7~ 25 27 Qio 2~ 710325 1 l

<Itsy~',9%Tgg f g%4 ISO j 4 26 27 R7. 33 710522 7, ~ 4 ~ 7. QSSA CONT IBO INFO Q4. 7 710325 14 5. ~ 1 ~ 2 QSS .VKRPR RLF ADS ANAL .01B 30 Q14 ~ 1 . 710325 14 55RM26 ~ QSBA SAFE BHUTD AC/DAN/EQK 15 328 32D R10. 1 710522 10. 1 l QSBA FIRE PRO FLOOD 32A 32E 27

'duplicate items sorted to shoe events in second and third chant columns Rl

Page /2- o f '3~

Prepared:

t Date. ~~~~

C e Checked: Date: ~Hi //

BFNWSC3-048 (Rev I) C. M. Parker aaassaaaaasssssaaassasssasaaassaaasaassaaaaasaaasasaasaaaaaaaasaaaasaaaaaas TABLE 4-2.1 Review of PSAR - Volume 7 uestioas Sorted b Event/A endix BSA SSA SSA

'FSAR-V7 QUEST FSAR EVAL EVT EV2 -EV3 QUEST aaccccca DATE SECTION aaccca c aaaaacaaa CAT KEYWORD aaa aacaaaccac 1 I'EYWORD 2 aaacaacacc ~NO. NO. NO.

aaa ccc R10. 2 710522 4. 4, 4. 6 QSSA ACCUMUL INFO D1 03 25

94. 7 7]032 ~

14 ~ 5. 1 ~ 2 QSSA OVERPR RLF ADS ANAL 01B 30 25 U14. 1 711206 14.5.2t 1 QSSA FW TR FUEL INFO

93. ]0 7] 0325 3. 7,'6. 2 QSSA FUEL CLAD OPCR 10 914. 5 710 '25 14,5,4 QSSA BYP VLV ANAL

98. 1 >]0325 8 it a4~ ~ ~ a QSSA AC POWER SINGL FAIL 15

'98. 4 710325 8.4.1 QSSA DISL GENRT INDERT PWR 15 25

98. 7 7 ] 0~Ac'.4 QSSA >WR FAIL DES INFO 15 25

98. 8 710325 8.5 QSSA DGBD FAIL DES INFO 15 25 98,9 710325 8.5 QSSA DG LOAD DES INFO 15 25

98. 10 710 <<25 B. 5 QSSA D G RELAY DES INFO 15 25 32D

98. ]1 710 t2 8 5 QSSA D'G CONTROL INFO 15 25

98. 13 710 <<25 8. 5 "

QSBA DG LOAD DE 15 25 QB. 15 710 ~t25 GEN DES N 15 25

98. s.'. 15 25 Q 5)Q~+'QBStI;Cflp "'~

icl'14.

1 7] '>

D 6C/DAN/E 1. 15 32B 32D R7. 8 71 4~12 ~ '-'SSA HVAC 15 R]4. 2 7]0 22 14.5. 5. 4 98 'CIRC SEIZ INFO V6 7?06 2 3 6 QSSA NUCL PARAM INFO 19 18 16 V7 720814 3.6 QSSA COEF TRANS INFO

Prepared Page l3 of 3W e Date: Z-E C. r Checked! Datc: P7 BPNWSG3-048 (Rcv I) C. M. Parker TABLE 4>>"2 ~ 1 Review of FSAR - Volume 7 eationa Sorted b Event/A endix+

SSA SSA SSA FSAR-V7 QUEST FSAR EVAL . EVT EV2 EV3 QUEST aaaaaaaa DATE aaaaaa SECTION aaaaaaaaaaa CAT'KEYWORD aaaa aaaaaassaaa 1 KEYWORD 2 aaaaaaaaaa ~

NO. NO ~ NO.

aass aass QS. 12 710325 8. 5s 10s b QSBA DWELL BLWR INFO 5 15 Qf 0. = 7 f 0>>25 f 0. 12 QSSA CR VENTLN DES INFO 25 26 27 Q12. 18 710325 14 bo 3 ~ QSSA LOCA DISCUSS R7. 1 f 71 05.2 5. 2 QSSA VAN INFO R7. 2>> '7105 . 2 '7 4. 3. 4 QSSA CORE SPR INFO 25 R7.24 7f0522 7.4 . 'SSA LPCI INFO 25 R7.25 71052'2 7 ' QSSA LPCI ' INFO 25 R7.26 7f0522 7,,4 QSSA CONT. SPRY INFO

/

R7,27 7105 2 7 4 ~ QSSA ECCS LOGIC INFO R7. 32 7f0522 7 3.2 ~ QSSA CONT SEAL INFO R7. 33 .7105.2 7.3 ~ 4~7 QBSA CONT ISO 25 27 R7 ~ 34 710522 6. 5. 2s 5 QSSA ADS INTRLK INFO R14. 1 710522 5.2, 14sb QSSA CONT RSPNS INFO S1-4 :710730 6.0 QBBA ECCS ANAL INFO T1-3 711 SE INERTG-. 25 06.1 711 Qa. 11 710 25A Qss 7 710 5 5.3.5 QBSA PIP PENT. TEST INFO l Q5. 15 7103 QSSA O'R11 CONTN 25A Q5, 16 710325 '6.5. 1 QSSA MIBL PROTC DSGN INFO 25A

1 Page J4 of 3+7 Prepared: n>>>>>>: ~M E. C e Checked: Date: T BFNWSG3"048 (Rev I) C. M. Parker RSSRRWWRRQRRRQNNQQWSSQWWQRRWRRRRSNRQWNWNQNNWRQ%%RRRÃRQQRRWRRQKRRRRRRRWRQS'SR TABLE 4-2.l Review of PSALM - Volume 7 Ouestions Sorted b Event/A oendix SSA SSA SSA FSAR-V7 QUEST 'SAR EVAL EVT EV2 EV3 QUEST CXRDRCRRXSRR

. DATE NÃ'>>NC'RC SECTION CAT KEYWORD SCIIWWWRkSWRWN %%%CA QRRÃKNIWOSSR 1 KEYWORD 2

%5RCtClCSCCCRDCRK ~

NO. NO. NO.

RCRR RRC

94. 10 710325 7.12 QSSA LEAK DET DSGN CHANG 25D Q5. 1 710325 5.2.2 QSSA PRI CONT DSGN INFO 25D Q5. 8 710325 5.3.4.2 QSSA LEAK RATE TEST INFO 25D Q5. 10 710325 5. 2. b. 2 QSSA COMBT GAS DBGN INFO 25D Q5. 14 710325 5. 3>> 5>> 2 QSSA FILTR SENS DSGN INFO 25D Q5. 17 710325 5. 2, 2 QSSA PRIM CONTN DSCjN INFO 25D

95. 19 710""5 5~ 2~ 4 6

~ 'QSSA S ING FA I'L, DSGN INFO 25D R7. 18 710522 7.12 QBSA SGTSs PRM INFO 26 25 27 R7. 29 710522. 7. 12 QBSA RB MON INFO 26 25 27 R5 ~ i>> 710522 5 QSSA PIPE BRK ANALYZE R7. 17 7105-2 7.2,7,3 QBBA TB, RWCU INFO 27 a3. 7 710 i25 3.8.3 QSSA STDBY LIQD NC R7. 30 710522 4. 4 SA ATWS-RPT INFO Q4 ~ 710325 4. 4 U4. 1 711206 4.4 QSSA MSSr~V"" "'REPORT 30 01 04 U4. 2 711206 4.4 QSSA HSBRV 30 01B 04 Q,6 710325 2 3 ~ QSSA MAX FLOOD INFO Q2. 8 710325 2. 3. 4 QSBA MAX FLOOD WIND+WAVE 32A 710522 2 'A QSSA FLOOD DAM CALC 710522 10.11 QSBA FIRE PRO FLOOD 32A 32E 27

i rags ~5'oz'34 Prepared: Date: ~Qgp Co Checked: Date! P BPHWSG3-048 (Rev I) C. M. Parker TABLE 4.-2.l Revie~ of '.FSAR - Volume 7 estions Sorted bv Event/An endix SSA BBA SSA

'FSAR-V7 ..QUEST FSAR EVAL EVT EV2 EV3

'QUEST DATE SECTION CAT KEYWORD 1 KEYWORD 2 NO. NO ~ NO aaaaaaaa aaaaaa aaaaaaaaaaa aaaa aaaaaaaaaa aaaaaaaaaa ace aaa a>>

Q12.17 710325 12. QSSA TORNADO FEATURES 32C

04. 4 710325 APPEX. C QSSA SYS DBE OP DSGN COND 32D Q12. 2. 14, 710325 QSSA SEI S INSTR INFO 32D Q12.16 710325 12 '.2.S QSSA CRANE SEIS DES INFO 32D Q13. 1 1 710325 1'3. 'QSSA ABNORM OP EMERG PROC BSA

. Prepared:

Page 16 Date:

of 307 E C. c t Checked: ~

Date: /g S'7 BFNWSG3-048 (Rev 1) C. M. Parker aeaeeaeaaaeaesoossessaaewsseaassaeaaeaaaassasseaessesesaaaszaaaaaeaaaaaeaaeaaessasezegs

'4.3 Review of Licensin Commitments/Re uirements 4.3.1 A systematic review was performed of all the licensing coamitments and requirements as sorted from the licensing data base according to the following categories'.

o (809) -'eneral Design Topic: Design Basis Events o (813) - General Design Topic: Fire o (815) - General Design Topic: Habitability o (826) - General Design Topic: Pipe Break o (834) General Design Topic: Regulatory Compliance o (837) - General Design Topic: Safety Analysis o (843) - General Design Topic: Station Blackout o (845) - General Design Topic: System Interaction o (999) >> Applicable Systems: All Safety Systems This review was-intended to clearly document the incorporation 0 (or exclusion) of each item with respect to the SSA. These reference documented AEC/NRC concerns and TVA comnitments throughtout the years of BFNP operation. (Section 4.2 addresses initial licensing caamunication and commitments.)

The process established identification of each item for SSA (event) or System Evaluation Report (SYSTER) application.

4.3.2 The references were .tabluated and evaluated according to the guidelines given in Appendix A of this calculation. Each question was assigned a category from the following list:

t (a) SSAl - Include in the SSA.

(b) OBCR - Obsolete items.

(c) INFO - Information from the FSAR oi D&AR.

(d) OPCR - Operational items.

(e) SER - Potential SYSTER consideration.

(f) NCR - Not a commitment.

(g) XLCR - Later commitments (or Unit 1 and/or 3).

(h) CRNA - Not applicable.

Each category is discussed .in more detail in A 4.3.3 The disposi item is recorded in the tab)

Appendice Bl through B5, to g ether w t o a other info ation cc Each of th s p di 4 d g

'j~% ehEes which have bee accumula te s gned Co d engineering branch (CEB, EEB, MEB NEB) > plus group of references to Genera pany items (GEC) l Vg r." ~ i L

~ ~ 4

Page 17 of 307 Prepared: Date:

E C c Checked: Date: ~HIP BPNWSG3-04S (Rev 1) C. M. Parker S'asssaesssaooaaaooeraaaaa3ISIeaaaSaeaaaaaaaaaSSaaaSaaaaaeSaasaocSaaSSaeOaSaaaSaaae

'.3.4 A number of these items do apply to the treatment of safe shutdown during various shutdown sequences. The appropriate event appendix number is included in the tabulated results.

Some are the source of including an "event" which was not included in the 'initial PSAR analyses (e.g., TVA letter dated December 2, 1977 and others associated with it from J.

tp A. Schwencer led to the incorporation of Appendix E'illeland

. 33, Loss of Puel Pool Cooling/Makeup).

Many of them are, direct requirements on one (or more) systems and are designated SER for each identification of STSTER inputs. Each SSA event appendix which has been noted documents the use that, was made of each item designated as SSA1.

Table 4.,3-1 lists the results of this review for all items .from Appendices Bl to B5 that are expected to have SSA impact. They are listed in the order of the anticipated event appendix numbers.

The first two pages of the table are actually canposed of duplicate items (they appear again, sorted differently, on a subsequent page in the table). However, they are specially sorted to show events identified in the second and third event columns (noted). These extra columns were occasionally used when an item appeared to apply to two or three events.

The remaining pages are in the order of the first SSA event/

appendix number column (noted) . Items at. the end are labeled SSA (instead of an event/appendix number) since they are expected to have broader application in the calculation (e.g.

criteria, several events, etc.).

Page d" of Prepared: 3'atet E C. ce r Checked: Date: 7 BFHWSG3&48 (Rev 1) C. M. Parker Table 4.3-1 Review of Licensi (C/Rs) Sorted b Event/ oendix*

PRI CIR SSA SSA SSA BFH ~ ~~~ ~~ ~~ SER SORT SRT EVAL CR SOURCE SOURCE EVT EV2 EV3 Uniq C/R HUNSER sax sssmsss RINS (DATE) sass asssss HO SYS KEY PG CAT REFEREHCE SSS $ $ $$ CSSSRSSRSRSS SCSSSR RRR ~~

'DATE HO. NO. NO. KEYMORD 1 KEYMORD 2 Unit RSRXRRCRRR C~tCCRRR CR$ %

HEB RAC1593 845 S&0630 61 0 837 39 SSAI D&AR A7 0 SSA 01A 18 FUEL-DESN TRANS-'PERF 0 HEB MRB}050 B45 8&0623 644 I 837 43'SSA1 TVA LTR WRC 821015 30 01B SRV-NODS HEB SAC 1 &07 B45 860702 45 92 83? 39 SSA1 TVA LTR+EHC2 840123 ssA 0& APRN-SCRAN TH-POM-MOH 0 HEB DLN1013 B45 8&0628 725 0 837 14 SSA1 SER A-II I-A 680606 01 09 FUEL-MARSH TRANS-NODS 0 EEB MCM}255 B43 860617 9a9 &4 Sa4 sa ssA} SSER 7.2.1-1 0 }2A 12B 23 MS}V-TRIP HI-RX-LVL 0 HEB RAC}334 845 860626 12 777 809 40 SSAI TVA LTR+EHC2 750122 25 13 T-SPEC-CHG 105 PSIG 0 HEB PAC1424 845 860626 629 I 837 3& SSA} TVA LTR 560326 25C 13 ADS-L06IC JUST-NO?37 0 HEB RAC1347 B45 S&0626 25 777 S34 69 SSAl TVA LTR+EHC2 750122 25 13 LO-PR-ECCS 105 PSIG 0 HEB RAC1306 845 860624 68 3 834 68 SSA} TVA LTRtEHCL 840709 14 15 LO-LVL-ISO SETPT-Ll 0 MES MMAI}25 B44 8&0731 24 23 809 26 SSA} LTR TO HRC 801203 14 15 25D RHRSM ESF-SIGNAL 0 NEB RTM}3}9 BCC 860807 112 777 813 4& SSA} TVA L/HRC 840612 32E }C FIRE+I,S/AC ANALtsCNED 0 GEC RNR1054 B45 860618 881 200 999 2a ssA} GE-22A IOC7RO 0 15 15 25 STHSY<ITE 0 HES PAC}486 B45 860628 689 0 '09 4} SSAI SEP. 9.0 &80606 SSA 24 SAFET-AHAL SER-CHCLUS 0 HEB RAG}&01 B45 860630 '6 85 809 4C SSA1 SSERI 3.4 0 09 24 REACTV-CTR RSCS 0 EEB MCM1256 $ 43 S&0617 940 92 834 53 SSAI SSER 7.2.1"2 0 09 24 15'XAPRN-SC NOD-RANGE 0 HEB RAC1754 B45 860701 719 :23 809 48 SSAI TVA LTRfEHCL 6S0531 15 25 RHRSM-DESC STHBY-COOL' HEB RAC1080 845 860&19 40 7'77 -837 21 SSAI TVA LTRtEHCL 780523 14 25 RCIC/HPCI ISOFLO-DLY 0 GEC RNR1063 845 860618 890 74 826 13 SSAI 22A}345tBtAC 0 25A 25B 25C RHR-SYST DESH-SPEC 0 NEB LMB}018 B44 860617 15 777 834 60 SSAI TVA LTRtEHCL 840702 25 25D CONT-IHERT AI.T-HITR6H 0 HEB RAC1308 845 860624 71 85 837 33 SSA} TVA LTR ENCL 820120 SSA 27 SCPM-P IPHG DESCRIANAL 0 HEB PSN}054 B45 860701 36 777 837 42 SSAI TVA LTRIENC} 790424 25D 27 COHT-ISO ECCS-IHIT . 0 NEB RAC1751 B45 860701 722 777 815 5 SSA1 TVA LTR-AEC 680531 09 28 RX"CHTRL SYST-DESCR 0 EEB PFS1487 B43 860&19 845 26 813 9 SSAI FSAR 7.1S.2- ,0 32E 28 FIPE-PUMPS BCKUP-.CTRL 0 MEB RTM1123 B44 860818 121 30 813 25 SSA1 TVA LIHRC EH 760620 32E 28 FIRE"DNPRS DG-BKUP-CR 0 MEB RTM1188 $ 44 860818 }86 303 8}3 33'SSA} TVA I.IHRC EH 7&0620 32E 28 F }PE-DOORS CB-STRMY1C 0 NES PTMI192 B44 860818 190 303 813 33 SSA1 TVA L/HRC EN 760&20 32E 28 FIRE-DOOR BATT/BCKUP 0 MES RTMl}93 B44 860S18 191 39 S}3 34 SSA} TVA L/HRC EH 7606'20 32E 28 FIRE-EXING BYUPIBATT 0 ES RTMI}97 BCC 8608}S 195 303 13 34 SSAI TVA LI MES PTM1198 844 860818 196 777 3 34 SSA} TVA L/HRC EH 760620 32E 28 FIRE+ROT NAIHCR NEB PTM1290 BCC 8&0807 83 64 TV '3 ~ . STD HES PAC143'9 845 860626 645 1 V 7 v HES PAC}2&3 845 860624 16 99 TV T ~c

'this page sor ted to show events in second and third event columns.

Prepared<

~ Co Checked: Date!

BHfWSG3&48 (Rev I) . C. M. Parker

'able 4.3-1 Reviev of Licensi (C/Rs) Sorted b Event/ endix*

PRI CIR SSA SSA SSA BFN ~~~~0~~~ SER SORT SRT EVAL CR SONCE SONCE EVT EV2 EV3 Uniq

'IBS CIR NUBBER SR@ ERSSSXSC CEB <61011 IDATEI SRRS CSSSSS ~NO SYS SOS B41 860604 549 777

~

KEY'S 809 SOS I

CAT REFERENCE RSSSCSSSRRRR SSAI TVA LTR SRSRRR %SR 740225 32A 32B

~~

DATE NGs NO. HO. KEYMGRD I RRCRRSRSSC KEYMORD RSSRRCSSSS 2 Unit SESS FLOOD . 0 NEB RAC12IS 845 860623 656 777 809 37 SSAI DIAR A3-3.1 0 25 32C CL-H:GHPS LOADCNBS' CEB <61227 B41 860707 262 7?7 809 6 SSAI NBS 721004 32A 32D FLOOD SEISNIC 0 NEB MMA1004 844 860609 5 27 809 24 SSAI NENGS '30502 25A 32D CCM+UNPS SEISNIC-EQ 0 NEB MMA1012 B44 S60609 13 777 809 24 SSA1 NBN 721004 32A 32D FLOOD-MAVE SEISNIC 0 EEB SRR1173 843 860609 943 200 834 4 SSA1 LTRW/HERC 800806 32A 32D 15 GHSITE+OM LOVLTMYS 0 NEB DLNI047 B45 8&0&28 97 79 837 15 SSAI TVA LTRtENCI 780609 26 33 REFUEL<RH LD/HHD<IN 0 NEB RAC1 279 845 860624 34 0 809 39 SSAI MAR A4 4.4 ~0'25 SSA LOCA-10NIN NMCTION 3 HEB RAC1756 845 860701 717 74 S09 48 SSAI TVA LTR+ENCL 680531 25 SSA RHR-DESCR 0 NEB MRB1079 B45 860701 716 64 809 52 SSA1 TVA LTRtENC 680531 25D SSA CONT-DESC PRltSECOHD 0

'I

~as page sorted to show events xn second and the.rd event columns.

Pollowing pages sorted in order of first SSA event column.

1

'I

lt Prepared:

Eo C.

Checked: Date: $ '~

IIPH-OSG3-048 (Rev 1) C. M. Parker WQRRÃNSRQRQRSSRRQRWNNSSRWSSH Table 4.3<<1 Revie~ of Licensin (G/Rs> Sorted b Event/A endix PRI '/R ~

SSA SSA SSA

~ BFH ...." .. SER SORT SRT EVAL CP. SOURCE SOURCE EVT EV2 EV3 2 4

'C/R NUNBER RINS {DATE) NO SYS KEY P6 CAT REFERENCE DATE MO. HO. HO. KEYMORD 1 KEYMOPD SRR SRSCC KtR 'N?CETIC C zc OSRCSSSCCSSS

'N tCCS SCC2 Q IHC CCRC'EB Jlk 937 14 SSA} A-111-A 680606 01 09 FUEL-HARSH TRANS-NODS 0 DLN}013 845 860628 725 0 SER 809 38 SSAI RELOAD-1 2 XEB B45 860624 lb 99 TVA LTRtEHCL 780925 OIB 31 ADD-TAILPP I RAC1263'EB 837 37 SSA} 781205 0}B 30 RPLC-SRV RAC}439 B45 860626 645 1 TVA LTR 845 10 SSA} 0 02 P-REB/BYPS DESN-SPEC 0 BEC 6DC}031 B45 960619 887 47 22AI}85 R 6 809 37 SSA} 750028 09 RMEHHAL BETAB/BEXL 0 .

NEB RAC1242 $ 45 860623 690 0 TVA LTRtEXCL 809 44 SSA} 09 24 REACTV-CTR RSCS 0 XEB RAC}601 B45 860630 46 85 SSER1.3.4 0 837 38 SSA} NCPR-!.INTS IHIT-CORE 3 HEB RAC}475 845 860629 705 0 SSER 8-3 0 09

}5IAPRN-SC NOD-RAHBE 0 EEB MCM}256 843 860617 940 92 834 53 SSA} SSER 7.2.}4 0 09 24 815 5 SSAl 28 RX-CHTPL SYST-DESCR 0 XEB RAC}751 $ 45 860701 722 777 TVA LTR-AEC 680531 09 XEB RAC}671 $ 45 860701 647 777 837'0 SSA1 DLAR Ab PB-4 0 11 HUCL-EICUR RFUEL-NODE 0 837 28"SSA1 NSL-LO-PR SETPT-CHB 0 HEB RAC86 845 860623 719 64 TVA LTRtEHC3 770202 12A 934 53 SSA} }-l 0 12A 12$ 23 NSI V-TRIP Hl-RX-LVL 0 EEB MCM1255 B43 860617 939 64 SSER 7.2. S45 13 SSAI IHADV-SRV 0 BEC RRB}033 B45 860621 987 I SIL 177 760615 13 809 26 SSA} 15 ?5D RHRSM ESF-S}6NAL 0 Bhh 860731 24 23 TO NRC 801203 14 NEB MMA1125 , LTR

                                                         '837           21 SSA1                                                      25          RCIC/HPCI     ISOFLO-DLY 0 NEB    RAC}080      $ 45 860619     40 777                                           TVA LTRtENCL             780523      14 937         22 SSA}                                                                 SAFETHHAL L?-SETPT           0 NEB RAC1083          845 8606}9      43 777                                           TVA LTRtEHCL 780523 14 837         41 SSA}                                                                  AUI-HT-RNV 6EH-DESCR         0 NEB RAC}725         $ 45 860701     37 777                                            TVA LTP+EHC} 790424 14 HEB RAC1306          B45 860624      68       3
                                                      "      834         68 SSA1         .TVA LTRtEHCL 840709                  14     15          LO-LVL-ISO SETPT-Ll          0 40 SSA}                                                                  HPCI-COOL6                   0 NEB PAC1413          845 860626     94     73            809                          TVA lTRtEHCL 810000 15 999         23 SSA}                                                      15    25    STHBY-LITE                   0 6EC RHR}054         B45 S606}8 881 200                                               6E-22A104780                     0 15 834           2 SSA}                                                                 RPS-NB-SET LOSS-ACPOM 3 EEB     Ell}0}4      B43 860609 991         99                                        TVA LTR -HRC 780322 15 809         48 SSA1                                                      25          RHRSM-DESC STHBY-COOL 0 NEB RAC1754        "845 860701 719         23                                         TVA LTRtEHCL 680531 15 29 SSA}                                                                  HAT-CIRC      OPEP.-1?HPS    0 HEB RAC1213          $ 45 860623 642        68           837                           TVA LTR    8-12 771107              18 809         41 SSAI                                                                  SINBL-lOOP PNP-SEIZUR        I HEB RAC1435         845 860626 64}         68                                         TVA LTRtENCI 780928                  19 809         37 SSA1                                                                  RDA-NITBTH                   0 NEB RAC}215         $ 45 8606'23 653       95                                         SSER}    9.4                    0 24 809         37 SSA}                                                                  RDA-NOD'S                    0 NEB RAC1217         $ 45 860623 655        85                                         SSER}    9.4                     0 24 B43 860613 988         85            837         '7   SSA1 TVA LTR                 7207}S 24                       RDMRTH-NIH                   0 EEB RFS}277 EEB 6PP.}403         B43 860618 949           0
                                                         '09               12 SSAl          TVA LTR                 800903 25                      ECCS          DC FAILURE 0 809         36 SSA}                                                                  LOCA-IHT-C RHR-NODS           0 HEB RAC1203         845 860623 627         74                                         TVA LTRtEHC2 '771228 25 809         37 SSA1                                           0 ?5      32 C        CL-}<ONPS LOAD-CONBS 3 HEB RAC1218          $ 45 860623 656 777                                               D}AR A3-3.1 809         38 SSA}                                                                 LOCAWHAL APP-K-ECCS 0 XEB RAC}248         B45 860623 697            0                                        TVA LTRtEHCL 750829 25 809         38 SSA}                                                                                         EC    0 NEB RAC}268          B45 860624 21 777 09          9 SSA1                                                                 I.OCA-}ONIH NAH-ACT    OH    3 HEB RAC1279          $ 45 860624     34       0                                        DEAR Ah    4.4                   0 25      SSA UR    0 NEB RAC1284         $ 45 S60624     39        0         'S09         39 SSA}           T        +   D           8082 0

HEB RAC1334 B45 860626 12 777 Al lh A 7 E? 2 845 860626 97 0 5 1.'l T . NEB RAC}416 0 UT 0 HEB RAC1444 845 860626 650 74 MTN N 2 0 09 41 SSA} A-A-IIIB 680606 25 SAFT-FETUR FUHCT-C IT 0 845 860628 724

                                                           ~

XEB RAC1468 SER HEB RAC}476 B45 860628 703 0 '09 41 SSA} SSER 8-6 0 25 lOCA-ANAL ECCS-CR 0 PY 0 NEB RAC1636 $ 45 860701 698 74 ,$ 09 45 SSA} ME.

Prepared! E. C. Checked:

         'P&GSG3~8                 (Rev 1)                                                 C. M.       Parker Table 4      '-1          Revie~ of Licensin                    (C/Rs            rteol b     Event/A        endix PRI        C/R                                      SSA SSA SSA
  'BFX  ~ ~ oeo ~~~                     SER      SORT SRT EVAL     CR SOURCE        SOURCE EVT EV2 EV3                              U C/R XUNBER        Rlls  }DATE)   XO SYS         KEY  PB   CAT   REFERENCE 0$ $ %$ $ &$$$ t  WSSSS   ~ ~

DATE NOo XO. HO. KEYMORD 1 SOS SSRSSSNRSt KEYMORD 2 SSWSESSRSS 0 N NEB RAC}639 B45 860701 695 777 809 45 SSAI SER 6 I & 2 0'25 BEN-DESH ECCS 0 NEB RAC1755 B45 860701 71S 777 809 48 SSAI TVA LTRtEXCL 680531 25 AUTO<CCS CStLPCI 0 NEB RAC1756 B45 860701 717 74 809 4S SSA} TVA LTRtEXCL 680531 25 SSA RHR-DESCR 0

 . HEB RAC1011        B45 S&0617    39      0      999  25 SSA1 TVA     XENO  P-3 670320 25                ACRS-NTB    SIXBL+AIL 0 NEB RAC}253        845 S60624    .5 74      . 837    31 SSAI TVA LTR             801001 25              ICOHH-STBY HIH'.REDIT 0
'EB     RAC1395 NEB RAC}408 B45 860626 BC5 860626 76 89 74 74 837 837 35 SSAI TVA LTR 36 SSA} TVA LTR      ~

WC 7&0&}1 25 760611 25 LPC}-FLOMS SAFET-ANAL 0 LPCI-FLOMS SAFET~ 0

                                                                                                                              '0 NEB LMB1018        Bhh 860617     }5 777        834  60 SSA} TVA LTRtEHCL 840702 25             25D     CONT-INERT ALT-HITR6H 0 HEB RAC1347       '45   S606?&   25 777         834  69 SSAI TVA LTRtEHC2 750122 25             13      LO+R<CCS 105 PS16        0.

NEB R6X1060 Bhh 860604 80 777 809 23 SSA1'RC<ALC 710510 25A LOCA-STXLH

                                               "                                    73050?   25A 32D       CCMHUXPS    SEISNIC<Q 0 NEB MMA}004        844 860609      5   27       809  24  SSA}oNENOS NEB'RAC}084     '45     860619 44 74           809  32  SSAI TVA LTRWCL 780523           25A   ,       LOCA-LPCI   LOOP/SETPT 0 6EC BT6100}        B45 860619 889        0 837       11  SSAI LTRWEDO(R?) 850500          25A           LOCMXAL 10CFR50.46 '2 XEB RAC}020        845 860617 5}        3      837  19  SSAI, TVA LTRtEHC3 840?22        25A           LO-LVL-SET BRPI-TO<} 0 NEB RAC}335        845  8606?6    13  777     .834   68  SSA} TVA LTR MC 751201           25A           LPC}-NODS   APPEH-K      0 BEC RXR1063        B45 8&0618 890 74           826  13  SSAI 22A1345tBtAC             0  25A 258 25C RHR-SYST       DESH-SPEC   0 XEB RAC}374        B45 860626    54      I'09        40  SSA} TVA LTRtEHCL 770000         25C           LOCA-AXAL 2ADS-OOS,      0 B45 860626    98          ~ 809  40  SSA1 TVA LTR            780420   ?5C           1.0CA-ANAL   ADS-OOS     0
. XEB RAC}417                               1 XEB RAC}424       B45 860626 629        }      837  36  SSAl'VA LTR             860326   25C 13        ADS-LO6IC    JUST-H0737 0 NEB RAC}7?'9       845 860701    ?6    &4      834  73  SSAI TVA LTRtEHCL 791017         25C           ENERB PROC POST-ACCID     0 NEB RAC}212        845 860623 641       I      845  16 SSA} TVA LTRtEHCS 77}0?8 25C                    ADS-VLV      ?-OUT-SRVC 0 NEB R6N1061        844 860604    Bl 777        809  23 SSAI RC-CALC           . &S}106 25D              ENERB-VEHT RI-TOWTNO 0

'EB DLN}008 B45 860&28 733 84 809 27 SSAI TVA LTR-E3)4 791210 ?5D COHT-CAD BEH-DESM 0 HEB LJK}03C B45 8606?3 652 31 809 30 SSAI SSERl S.4 0 25D C-RN-VEHTH COHF-BDC 0 NEB LJK1073 B45 860701 40 64 809 31 SSA} TVA LTRtEXC 790424 25D COHT-VEHT BEH-DESC 0 NEB MRB1079 B45 860701 716 64 809 52 SSAI TVA LTRtEHC &S0531 25D SSA COXT-DESC PRltSECOXD 0 BEC BMFIOSI 845 860621 94C 74 837 12 SSAI SIL 15}R} 760730 25D COHT-COOLS SHBL+AILR NEB MMA}084 Bhh S60616 110 777 837 14 SSAI TVA LTR WC 750114 25D EECM/RHRSM 0 0'ECH-SPEC XEB RSN}054 B45 86070} .36 777 837 42 SSA}, TVA LTRtEHC} 790424 25D 27 CONT-ISO ECCS-}HIT 0 B44 860616 108 23 834 6} SSA} TVA LTR -HRC 801203 25D RHRSM PERF-REQTS 0 NEB MMA}OB2 XEB RSN1024 B45 860623 651 777 834 74 SSA TVA LTR-XEB RSN1057 845 860701 23 777 8 4 SSA} TVA LTRtEHCL 791017 25D CONT-ISO XEB DLN}010 B45 S60628 728 64 8 14 SSA} B45 860628 97 79 8 NEB DLN1047 NES RAC}035 B45 860616 18 777 8 6 ",'jkfPPg( j XEB RAG}309 B45 860624 '72 0 8 39:S ) @b44 LD SC NS I " 8 41 SSA} ooo 0 o oooooo~oo LOCA-XSLW NS1 VC-TINE 0 NEB RAC1487 845 8606?8 690 845 860626 23 0 8 34 SSAl TVA LTR 750214 2727'6 IHT-FLOOD BRKWUTSID 0 NEB RAC}345 NEB RAC1346 B45 860626 24 777 839 34 SSA} TVA LlR NEB MRBIOS6 B45 860814 627 777 826 20 SSAl FACIHP<IC 740628 HELB-BDC-4 EIENPTIOH 2

0 Prepared: E. C. e Page Date.'~p

                                                                                                           ~      f ~3D BYES G3-048 (Rev 1)

Checked: Date: ~ 7 C. M. Parker aaasaaaassassssassasssssssaassassssasaasaseasaassassasassssasaaasssssassass Table 4.3-1 Review of Iicensin tC/Rs> SorteR bv Event/A endir~ PRI OR SSA SSA SSA BFH ......e. SER SORT SRT EVAL CR SOURCE SDUPCE EVT EV2 EV3 U I SINS }DATE> ssss ssssss HO SYS KEY PS CA7 sss sss', ssss sss ssss REFEPEHCE sssssssaas asst ~ DATE HO. HD. HD~ KEYkORD sss sss ssssssssss I KEYMORD sss ~ss 2 s BEC BDC1009 B45 860617 907 777 809 21 SSA} BE 22A}470R6 0 28 BKUP-CXTRL DES-SPEC 0 6EC PDK1077 B45 860&23 965 31 843 2 SSAI SIL 421 850328 28 CROO&4VAC EVAL-LOSS 0 HEB RAC}348 B45 860626 26 &S 837 34 SSAI TVA LTR 74123} 29 AIMS-EVAL ADMPT 0 HES MPB}050 SC5 860623 644 837 43 SSAI TVA LTR -NRC 8210}5 30 0}B SRV-NODS 0 HEB RAC}179 B45 860623 727 0 809 35 SSAf TVA LTR 7607}4 31 FUEL-LD-ER 0 HEB DLN}108 BI5 860628 3 79 837 f7 SSA} TVA LTR>ENCL 7S0925 31 ANAL-CHB BNDL-LDERR 2 HEB RAC}410 B45 8&0626 9f 0 837 36 SSA} TVA LTR -HRC 760714 31 FUEL-LDERR TH-LIN-AHL 0 CEB -CB}011 B41 860&04'549 777 809 I SSAI TVA LTR 740225 32A 32B FLOOD 0 CEB -C61214 B41 860707 273 77? 809 5 SSA} TVA LTR 720124 32A FLOOD 0 CEB -CB}227 BII 860707 262 777 S09 6 SSAI NENO 721004 32A 32D FLOOD SEISNIC 0 NEB R6N1013 BH 8&0604 33 777 i 809 22 SSAI NENOWTTACH 720824 32A FLOOD 0 NEB R6N}014 844 860604 34 777 809 22 SSA} NENO-ATTACH 720911 32A FLOOD 0 NEB MMA1001 BII 860&09 2 777 809 24 SSA} TVA LTRS 720816 32A FLOOD NARHIHBS 0 NEB NMA}008 SCC 860702'8 777 809 24 SSA} NENO 711208 32A FLOOD MA7ER-VEL 0 NEB MMA}010 844 8&0&09 11 777 809 24 SSA} 710302 32A 'ENOS FLOOD DES-BASIS 0 NEB MNA}0}2 844 8&0&09 13 777 809 24 SSA} HEND 721004 32A 32D FLOOD-MAYE SEISNIC 0 NEB MMA1013 BCI 860609 14 23 809 25 SSA} NENO 720824 32A FLOOIH'ROT RHPSM-PNPS 0 NEB MMA}0}C BH $ 60&09 }5 777 S09 25 SSA1 NENO 721011 32A FLOOD-PROT 0 HEB RACI}98 B45 860623 700 777 809 36 SSAI TVA LTRtEHCL 740225 32A FLOOD-NODS 0 CEB 3EP1001 BII 860606 265 777 834 I SSA1 NENO TO HRC 770421 32A FIRE-DODRS NOD-DESH 0 EEB 6PR}173 B43 860609 943 200 834 4 SSA} LTR-E2/I-HRC 800806 32A 32D 15 OHSITE+OM LOVLT-RLY6 0 EEB RFS1279 B43 860613 990 0 809 20 SSA} TVA LTR 720718 32D SEIS-NOHIT 0 CEB 3DH}043 841 860818 265 0 999 10 SSAI LTR TO TVA 750917 32E HRC-COHCRH FIRE-EVALH 0 BEC BT61003 BIS 860619 891 0 999 22 SSAI LTRtXEDC RPT 0 32E 50.48eAP-R FIRE-PROT 0 EEB RFS}487 B43 860619 845 26 813 9 SSA} FSAR 7.}8.2- 0 32E 28 FIRE-PUNPS BCKUP-CTRL 0 BEC PDK}043 'B45 860623 -ARNS 0 NEB RTk}0fl 844 860619 813 12 SSA} FIRE REC I-A 760802 32E FIRE-TESTS TRA SEALS 0 NEB R7M}04& BII 860&19 -SYST 0 NEB RTN1047 BH 8&0619 I LV-SM 0 NEB RTN1048 844 860619 I jgLAjL}IR~'.~il Pet} g ~) q L-SM 0 NEB RTM1049 BH 860619 I 74 I QS'AI Aa L-XRC E}f'S&O}3} 'PfRk- -CLOS 0 NEB RTM}050 BIC 860619 &9 8}3 16 SSAI TVA L-.HRC EH S60131 32E FIRE-NODS REN PDMER 0 NEB RTM1051 BII 860619 I TVA L-HRC EH 860131 32E FIRE-NODS 3SR SEPAR 0 NEB RTM1052 BH 860619 }24 777 813 17 SSA} TVA L-HRC EN 860131 32E FIRE-NODS CBL-SHTDMH 0 NEB RTM1053 844 860619 125 200 813 17 SSAI TVA L-HPC EH 86013} 32E FUSE-DISCH TIES+REFD 0 NEB RTM1054 844 860619 126 200 813 17 SSAI TVA L-HRC EH'8&0}31 32E FIRE-NODS CONNON-PDM 0 NEB RTM}055 BIC 860619 127 200 813 f7 SSAI 7VA I.-HRC EH 8&0131 32E F IRE-NODS CONNH<HCL 0 NEB RTM}05& 844 860619 128 200 813 17 SSA} TVA L-NRC EH 860131 32E FIRE-NODS CDNNH-ENCL 0 NEB RTM}059 BIC $ 60619 131 200 813 IS SSA}q TVA L-HRC EH 860131 32E ENER-LITES NOD-SCHED 0 NEB RTN}0&0 BH 860&19 132 777 813 18 SSA} TVA L-HPC EH 860131 32E FIRE-SHTDH REQD-EQUIP 0

Prepared s E C. er . Checked: BFHWSC3-048 (Rev 1) C. M. Parker ssssssssssss ass ssssssssssssssssss sssssssssssssssssssssssssssssss Table 4.3-1 Reviev of Licenein (C/Rs> Sorted b Event/A endix PRI CIR SSA SSA SSA BFN ......" . SER SORT SRT EVAL CR SOURCE SOURCE EVT EVZ E" U C/R NUHBER RRS RSStSCSS R}NS 1 DATE) SERE SSSSRC HO SYS: SOS SOS KEY PS

                                                                  'ygmy gag CAT pygmy REFERENCE gggSSSSSSSSS      $ $$$ $ $ ~     S++    ggasgssssc    sss5$ $ RSCS NEB RTM1061              B44   8606)9 133 777'           f3 18  SSA} TVA     LMC Bl 860131         32E          FIREHREAS FIRE-BARRS 0 lfEB RTM)062             844 860&f9 )34 777            8}3 19   SSA} TVA L-HRC EN 860f31 32E                    FIRE-SHTDH EQT-AVAIL 0 lfEB RTM}075             844 860619 )47 777 .          S)3 21   SSA1 TVA LAC EH $ 60131 32E                     FIRE-SHTDH OPERWCTS         0 NEB RTM1118              844 8608}8 116           0 .813 25     SSA} TVA L/NRC EH 760620 32E                    FIRE-BRISD COVER-EQPT 0 NEB RTM}123               B44 860S}8 121        30      813  25 SSA} TVA L/NRC EN 7606ZO 32E 28                  F}RE-DNPRS DB-BKUP<R 0 NEB RTM}$42              844 860818 140 244            813  27 SSA} TVA L/NRC Bf 760620 32E                     FIRE~RH       PU)fP+OUSE    0
             )fEB RTM)}43              B44 8&08)8     )4) 39         813  27 SSA) TVA      LINC EH     7&0620 32E             FIRE~)f CBL-TUNNEl 0 NEB RTM)153              B44 860818 151 777            813  28 SSA1     TVA L/NC EH      760&20 32E             FIRE-PROT     4K-BD-Rff     0
                                                                   )

NEB RTM1155 844 860818 )53 3O3: 8}3 29 SSA) TVA L/HRC EH 7&0620 32E FIRE-DODRS CB-CORRIDR 0 NEB RTM})57 B44 860818 155 303. '..8)3 29,SSAl TVA L/NRC EH 760620 32E FIRE-DOORS )fDDS-B-LBL 0 lfEB RTM1)58 844 860818 156 39 l8}3 29 SSA} TVA L/HRC Bl 760620 32E FIRE-EITNG CB-CORRIDR 0 NEB RTMl)65 B44 860818 )63 3O3 }8}3 30 SSAl TVA L/HRC EN 760620 32E BATTERY')f FIRE-BARRS 0 ffEB RTMl}67 B44 860818 )65 303 813 30 SSA1 TVA UMRC EH '760620 32E FIRE-DOORS BATT-BHN 0 NES RTM)1&S 844 8&OS)8 166 777 813 31 SSA} TVA L/lfRC EH 760620 32E FIRE-BARRS BATT-BD-R)f 0

         'fEB        RTM}}70          B44 860818 168 303             813  3} SSAf,    TVA L/NRC EH 760620 32E                 FIRE-BARRS DC/BATT-Rff 0
             )fEB RTM)171             B44 860818 169 303             813  31 SSA)     TVA L/NRC EN 760620 32E                  FIRE-DOORS SEAL-MALlS 0
   ~

NEB RTM}176 844 8&0818 )74 303 813 31 SSA) TVA l/NRC EH 7&0620 32E CO)fPUTRW)f SEALWALLS 0 NEB RTM1177 844 8&0818 175 777 813 31 SSA1 TVA l/NRC EH 760620 32E CD)fPUTR-RH FIRE-DOORS 0 HEB RTMl)83 B44 860818 18) 3O3 8)3 32 SSA1 TVA L/HRC EH 7&0620 32E FIRE-DODRS CBLSPRD-RN 0

             )fEB RTM1184             B44 8&OS)8 182 777             813  32 SSAl     TVA L/NRC EH      760620'32E             CSR+ENETR EVAL-SEALS 0 NEB RTM1188               B44 860818     $ 8& 303        8}3 33   SSA1    TVA L/HRC    Bf 760620      32E 28       FIRE-DOORS CB-STRMY1C 0 HEB RTM1190              844 860818 188 303             813  33 SSA1     TVA L/HRC    Bf 760620      32E                        SHTDH-BDS     0   'IRE-DOORS
          'EB       RTM}f91           B44 860818 189 303             813  33 SSA1     TVA L/HRC EH 760620 32E                  FIRE-DOORS SHTDH-BDS        0
            })EB RTM)192              B44 Sb0818     )90 303         8)3  33 SSA}      TVA UHRC EH 760620 32E 28               FIRE-DDDR     BATT/BCKUP .0 NEB RTM1}93               844 860818 191. 39             813  34 SSA1     TVA L/HRC    Bl   760620 32E 28          FIRE<ITHB BKUP/BATT 0 NEB RTM1)96               844 8&OS)8 194 777             813  34 SSA1      TVA L/NRC EH 760620 32E                 FIRE-EITHS NAIH-CR           0
          . NEB RTM}}97               844 860818 195 303             813  34 SSA1      TVA L/NRC EH 760620 32E 28              FIRE-DODRS NA}H-CR          0
            )fEB RTM}198              B44 8&OS18 )96 777             8}3 34   SSA1     TVA L/HRC    Bf 760620 32E 28                               -CR      0 lfEB RTMl)99              S44 8&0818 197 303             813  34 SSA}     TVA L/NRC EH 76                          FIRE-DOORS    C  -COL  Rll   0 NEB RTM)201               B44 860818             03   '13     35 SSA1                         60620 32E             FIRE-     S     IC-BLDS' NEB RTM$ 202             844 8608)8            303                   Af    TVA L/HRC EN 760620           E                         L)/C-BLDG 0 NEB RTM)204               B44 86081                      8}3 35   SSA}     TVA       C                                      THS RLY-,, RH       0 NEB RTM1211              844 860807             30       8)3                                                           Ig  RPIS    ATION        0 NEB RTI)1220              B44 860807                                                                                 IRH'ROT     REPL -DOOR 0 NES RTI))P79             844 860807                                                                                 E)fERS-LITE S        AREA 0
       .                                                0'200 HEB RTM1280               B44 860807                     813  40 SQ)       TVA L/HRC         8}03)8 32E                       E SHTDH-SUPL 0
         , )fEB   RTM1284'EB B44 860807                      8)3  40 SSA)      HEND/ITEN 3                              FIRE-CRIT     CBL-COATHS 0 RTM1285           B44 860807                      813  41. S                  C       830105 3?E             FIRE-SHTDH CORE. ANAL 0 NEB RTM1286              B44 860807                               SSA}     TVA ).INRC    ~   830105 32E             FIRE-SHTDH TORUS-TENP 0
       . NEB RTM1287              844 860807       80      1      813  41 SSA)       TVA L/HRC        830105 32E             FIRE-SHTDH SPUR-SRV          0

~ ~

Prepared: E. ~ er Checked- Date: BPNWSG3-048 (Rev 1) C. M. Parker aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa Table 4.3-1 Revie~ of Licensin (C/Rs) Sorted b Event/An endix PRI C/R SSA SSA SSA BFH ~ ~ ~ ~ ~ ~~ ~ SER SORT SRT EVAL CR SOURCE SOURCE EVT EV2 EV3 U C/R HUNBER RINS }DATE) NO SYS ggg gg ggggg gggg gggggg ggg ggg gggg ggg gggg KEY P& CAT REFERENCE ggggggg~ggg ggg g ~ DATE HO, HG. MO. KEYNORD ggg ggg 1 KEYMORD 2 $ NEB RTM}28& 844 860807 81 777 813 41 SSA1 TVA L/HRC 830105 32E FIRE-SHTDH 7}673-}HOP 0 NEB RTM1?$ 9 Bhh 860807 82 82 813 41 SSA} TVA L/NRC 830105 32E FIRE-SHTDH DSWSE 0 NEB RTM1290 B44 860807 83 64 813 41 SSA1 TVA L/HRC 830105 32E 2& FIRE-SHTDH BKUP+LVL 0 NEB RTM}291 B44 860&O7 84 2 813 42 SSA} TVA L/HRC 830105 32E FIRE-SHIDH CST-UHAVL 0 NEB RTM1292 $ 44 860807 &5?77 813 42 SSAl TVA L/HRC 830105 3?E FIRE-SHTDM EQPT-AVA}L 0 NEB RTM}297 B44 &60&07 90 777 813 43 SSAI TVA L/HRC 751209 32E FIRE-SEALS CSR/CREEN 0 NEB RTM}29& $ 44 860807 91 303 813 43 SSA} TVA L/HRC 751209 32E FIRE-JUSTH RB-OPENING 0 NEB RTM}299 B44 860&07 92 3O 813 43 SSA} TVA L/HRC . 751209 32E CTRLSCBLS PB-VEHTLH 0 NEB RTM}300 B44 S60807 93 777 813 43 SSA}. TVA L/HRC 751209 32E FIRE-BARRS JUSTH-SEP 0 NEB RTM1301 B44 860807 94 ?77 8}3 44 SSA}$ TVA L/HRC 751?09 32E FIRE-PROT NG-AISL-CS 0 NEB RTM}303 $ 44 860807 96 77 813'4 SSA}5 TVA L/HRC '751209 32E FIRE-SUPPR RDMST-MATR 0 NEB RIM}310 B44 860807 103 303 813 45 SSA1 TVA L/HRC &411}5 32E FIRE-JUSTH RB/IB-DOOR 0 HEB RTN131} B44 860&07 104 777 813 45 SSA1 TVA L/HRC 841115 32E FIRE-JUSTH .ADJ-CIRCTS 0 NEB RTN}313 B44 860807 }06 777 813 45 SSA1 TVA L/HRC 75120& 3?E FIRE-PROT CBL-FLANSI 0 NES RTM}314 Bhh 860&07 107 ?6 813 45 SSA1 TVA }.IHRC 75}20& 3?E FIRE-SPRAY CBL-CROSS 0 NEB RTM1319 Bhh 860807 ll? 777 813 46 SSA1 TVA L/NRC 840612 32E 15 FIREtLS/AC AHALtSCHED 0 NEB RTM1322 B44 860807 }15 303 813 hb SSA} TVA L/HRC 840605 32E FIPE-JUSTH BAT-RN"SEP 0 NEB PTN1324 $ 44 860807 117 303 ~ 813 47 SSA} TVA L/HRC 840925 32E FIRE-PROT REPLC-DOOR 0 NEB PTM132& 844 860&07 121 777 813 47'SSAl TVA L/HRC 811230 32E FIRE-PROT OXSIT-TRCX 0 NEB RTM}330 $ 44 $ 60807 123 777 813 48 SSA} TVA L/HRC 830322 32E APP+CRIT PGN/OP-ACT 0 NEB RIN1331 B44 860807 124 '303 813 48 SSA1 TVA L/HRC 770421 32E F}RE-DOORS CONC-HALLS 0 NEB RTM1332 $ 44 86OSO7 125 303 8}3 48 SSAl TVA l/HRC 770421 32E FIRE-MORS NASOH-NAlL 0 NEB RTN}333 844 &60&07 126 303 813 48 SSA1 TVA L/NRC 770421 32E FIRE-DOORS NETAl-PART 0 NEB RTM}334 B44 860807 127 303 813 48 SSA} TVA L/HRC 770421 32E FIRE-DOORS CO2-RN-DNP 0 i NEB P.TM}335 844 860S07 128'03 '13 48 SSA} TVA L/HRC 770421 32E FIRE-DOORS ?-HR-HALLS 0 NEB RTM1336 $ 44 860807 129 303 813 49 SSA} TVA L/NRC 770421 32E FIRE-DOORS &DAN-BOLT 0 NEB RTN}337 B44 860807 130 303 813 49 SSAl TVA L/HRC 770421 32E FIRE-DOOPS ACCES-HDNR 0 HEB DLN1064 B45 860628 78 39 813 55 SSA1 TVA L/HRC 8}1103 32E FIRE-PROT. LL-RADMST 0 HEB DLN}066 845 860628 62 79 809 27 SSA1 TVA LTR<HCL 771116 33 FUEL"STOR& Hl-DEHS-RK 0 HEB DLN}072 $ 45 &6062& 6& 78 809 ?7 SSA1 4 33 FUEl"STORS TH-HYD.DES 0 HEB DLN}OS4 $ 4 5 &606?& TVA LTR-EMC} 771202 33 FUEL-STOR& DESH-RACKS 0 6 O628 42 78 809 28 SSAI DL HEB DLN 0 28 HEB DLH 0 B45 860628 677 78

                                                                     ~      $ 5}I}SN @0 882]!.Lb-,'t8. '.(37,'QL>8t87 )8887lh 78'EB NENO    2 I '9    4 QI22 F+OGLCGGL SAFETY-DES 0 FPOOL-NKUP EECMtHGSE TH-DESH      FUEL-POOL 0

0 NEB DLN} l8 4 $ 8l'tlb? '13 w P t73t) t37 UI3LL>78788328 8..., ~ t$ 33 E}}ERG-NKUP FUEL-POOL 0 SEC BT&l 0 j}g Pkbl'7, 088.'i79 ~S4 58'SA} LTRtEHCL 771100 33 HIDEHWACK EVAL-LDCNB 0 HEB RAC1 $ 45 86II626 19 0 809 40 SSA} TVA LTRtEHCL 82 RLMETHGD RETRAH 0

                               $ 45 860628 689                0 1'EB RAC}4 6                                                   &09 41 SS              .          6&0606 SSA 24         SAFETWHAL SER<NCLUS 0 MEB  RAC}7 9             $ 45 8 0                                          SSA} FSAR 1.5.1.6            0 SSA           ACC ID-CRIT              0
      &EC RNR}0                B45 86061& 887              64         999 23 'SSA} 6E-2?A1}3?R5               0 SSA           CONT-ISO                 0 4'

Page 26 of 307 Prepared: Date: ZQ//)/7 C Checked:

                                                        ~

Date: /f P7 BPNWSG3"048 (Rev 1) C. M. Parker 5 ~0 DOCUMENTATION OF ASSUMPTIONS POR ANALYSIS 5.1 Assumption 3.1 defines the assumed configuratio'n of the plant at the time of the restart of unit 2 ~ It represents operating limits on units 1 and 3. It is unverified. Technical Justification: This assumption has been established as the baseline program basis for Unit 2 restart evaluation. Review of this document and subsequent system evaluations are needed before progressing to a different set. of operating conditions for the three units. 5.2 Assumption 3.2 concerning the Commitment/Requirement (C/R) lists used in this analysis is unverified. Technical Justification: Use of a direct printout from the C/R data base was considered to be the most technically correct source for these references. While careful use of these references was made, a method of verified transmittal of the licensing C/Rs for this type of use . remains to be established. 5.3 Assumption 3.3 concerning inclusion of the IBM scram trip feature in spite of the identified failure mode (reference 4q) is unverified.'echnical Justification: The IRM scram function has served Browns Perry throughout its operation. A decision concerning this relatively new, unique problem has not yet been made. By keeping its protection role visible in the SSA, clearer input to the decision process is available. Resolution of this issue for BFNP is needed. 5.4 Assumption 3.4 is unverified. It neglects the apparent conmitment made early in project communication (reference 4;ls)

        <vari
         .to incorporate logic to close the MSIVs on a high water level (L8) signal when the reactor is operating in any non-RUN mode.

Technical Justification: This item was not included because no design specifications, drawings or Tech Specs have been found to substantiate its current commitment. It is apparent that this feature has not been carried into the design; however, comnunication(s) which retracted this commitment must be found to verify this assumption. 5.5 Assumption 3.5 is vera wed until the appropriate BFNP Design ter i a~r <pgy tet on is issued. T c Lo Q The criteria~stated fS dgg~jg; Q $ 1,' s e

     >~analyses for BPNP a                <~dt,ed. tg b4 ~cg          ho%;

Design Criteria doc nts;:ih'ich;&q a>so bhjhg p5eQ in he Unit 2 restart progr

SAFE SHUTDORf AEAIYSIS Page 27 of 307 Prepared: Date: C c Checked: Date: / BFN-OSG3-048 (Rev 1) C. M. Parker

   ~ tSRRRNQCC    CSCZCCRNDPRRRXISCtSQSOSQlSSRSRQSSZSSS'SRRSQRRRWRSSQRISRNC2$   5CCtQ'$0$ $ tRCCCWQOSCQDQ 6.0     SAFETY CRITERIA FOR ANALYSIS The    following list of safety criteria (unacceptable safety results) were     utilized in the Safe Shutdown Analysis of each event. They are derived from the FSAR criteria (reference 4.1a), and are listed by event category. Current analysis criteria are also considered as presented in reference 4.1-1. The numbers (e.g., 2-1) and the title
          .(e.g., Rad Release) are used in the event tables to identify each i'tern ~

6.1 Cri.teria (Unacce table Results) for Planned 0 eration+

a. (1-1) The release of radioactive material to the environs to such an extent that the limits of 10CFR20 are exceeded (Rad, Release).

b. (1-2) Fuel failure to such an extent that were the freed fission products released to the environs via the normal discharge paths for radioactive material, the limits of 10CFR20 would be exceeded (Fuel Failure).

c. (1-3) Nuclear system stress in excess of that allowed for planned operation not considered by plant safety analysis (Syst Stress).

6.2 Criteria (Unacce table Results) for Abnormal 0 erational Transients (S ecific reference 4.lm)

a. (2-1) The release of radioactive material to the environs to such an extent that the limits of 10CFR20 are exceeded (Rad Release) .

b. (2-2) Excessive fuel failure calculated as a result of the transient (Fuel Failure).

c. (2-3) Nuclear system stress in excess of th transients b c es Syst Stress)

d. (2-4) C transie t n taiament stress'eg

                                                            ~

KPplfcable 'induslg]eoges wheP jpnt'a~ent " Ls required Cont Stress). ,.!i ..

                                                                ~

qP Note: Criteria fr Planned Operation are no~~ x.s Safe Shutdown A lysis, but are included here to show the complete FSAR set. Rl ~ ~

                                               ~g

SkTE SHUXDORS kML'XSIS 'age 28 of 307 Prepared> E ~ Eck Date'ate:

                                       , Checked:

BPNWSG3-048 (Rev 1) C. M. Parker ssssssasssrssssraassrrsssssssssssssssssasssssssssssasasasassssrsssrarssasans 6.3 Crite ia (Una'cce table Results) for Accidents (S ecific Reference

4. n

                  )

Idd gg'f

a. (3-1) Radioactive material release to such an extent that the guideline values of 10CFR100 would be exceeded (Rad Release).

b. (3-2) Puel cladding temperature in excess of 2200oP or peak fuel enthalpy greater than 280 cal/gm (Puel Pailure).

c. (3-3) Nuclear system stresses in excess of that allowed for accidents by applicable industry codes (Syst Stress).

d. (3-4) Containment stresses in excess of that aI,lowed for accidents by applicable industry codes when containment is required (Cont Stress) ~

e. (3-5) Overexposure to radiation of plant operation personnel in the control room (Pers Overexp) ~

6.4 Criteria (Unacce table Results) for S ecial Events (S ecific Reference 4.1o)

a. (4-1) The inability to bring the reactor to 'the cold shutdown condition by'use of the controls and equipment which are available outside the control room (Local Shut).

b. (4-2) The inability to shut down the reactor independent of control rods (Shut No Rods).

INC ~ (4-3) Pailure of required safety-related equipment due of, special phenomena (e.g., tornado, dam to'onsequences failure,-etc.) (Equip Pail)

                        \

(4-4) Fuel cladding temperature in excess of 1500oP (Puel Pailure). (reference 4.1p) (4 r presure vessel pressure in excess of 1 75 psig (RP (4- igp or cont tate e itageII(t li1g'pntnsisd 'ods283op fC ond n). (re n%e 4.lp) eenewly italo e defined criteria related to speeifia special events.

Prepared: Checked: E C C er Page 29

                                                                                'Date: ~~of Dete: ~Y/

307 BFNWSG3-048 (Rev 1) C. M. Parker aaeaassssaINOOOSaeSsaaaSSSaeeaaaaaaSSgWasImaaaaaaSaSsssaSsaaCeSSOSaeCraOINCSSSSSeeaaaaaae 7 0 SAFETY ACTIONS AND DEFINITIONS Each reactor abnormal transient, accident, or special event,was evaluated to establish what was needed to meet the applicable safety criteria (from section 6) that was challenged by the event. The required safety actions were identified as listed below. This set of actions was derived from those defined in Appendix G of the BFNP FSAR. Also shown are some of the key relationships to compliance with the safety criteria of section 6. 7.1 Safet Actions for Abnormal 0 erational Transients Relationshi to Safet Criteria

     *17 Scram  fast insertion of                 a)     Primary action to prevent exces-control rods to achieve                       sive fuel failure.

nuclear shutdown. b) Assist in meeting industry codes for system stresses. 18 Pressure Relief a) Primary action to meet industry (Pres Relief) codes for system stresses. 19 Core Cooling a) Prevent excessive fuel failure by avoiding or minimizing any uncovering of the core. 20 Reactor Vessel Isolation a) Primary action is to provide (RPV Isol) assurance that the potential release of radioactive material is limited. b) Provides action in some events to stop or reduce uncontrolled depiessurization so that system stresses remain within applicable industry codes. 21 Restore AC Po a Provide assurance of long term (Rest AC Pow) fuel protection an con axnmen and/or fuel pool cooling shoul Q

                                            )>>~P                  glypt oa ))            e  1
                                               )) Qyactions  .    '46/ 'QP(t,+            ua are sufficient      erev this action .is. identified.           ort S
                                               ~
                                                        -4eaa      ~  ~

require standby AC power are shown under the applicable safety action (e.g. 'power for core

                                                       ~

cooling). 1

Page 30 of 307 Prepared: E C. er Date: ~/ Checked: Date: If M BPNWSG3-048 (Rev 1) C. M. Parker SSSSSSSSSSSSSSSSSSSSSSDSSSSSSSSSSSSSSSSSSSSSSS'SSSSSSSSSSSSSSSSSSSSSSSSSDSSS 7.1 Safet Actions for Abnormal 0 erational Transients (Continued) Relationshi to Safet Criteria

  ~22 Restore Plant to                     a)   Establish normal conditions via Normal Conditions                     normal procedures so plant has (Rest Normal)                         ability to prevent unacceptable safety results should. any subse-quent event occur. Normal con-trol systems are expected to be utilized by the operator. Should control actions be unavailable, safety shutdown will'e utilized (usually manual scram and heat removal).
  ~23 Limit       the Magnitude            a)   Significant analysis about or Rate of the                        equipment assumption related to Disturbance                           assurance that excessive fuel or (Lim Disturb)                         system stress conditions will be avoided.
  ++24    Power Reduction                  a)   Assist in preventing excessive (Pow Reduce)                          fuel failure.
  ++25    Water Level                      a)   Avoid excess moisture and Reduction (Level                      potential stresses in steam line Reduce)                               and turbines by terminating water supplies and tripping turbines water level is excessive.

if e 30 Containment Cooling a) Maintain containment temperature (Cont Cooling) within required limits and pressure within allowable values in applicable industry codes.

i Prepared: Checked: E C. c Page 31 Date: Date:

                                                                                          ~~

of 307 BFNWSG3>>048. (Rev 1) C. M. Parker assassasaaasasssaessssaasossasassaseaaacawassasassesasasacssssrasscsssssssssssssssssassse 7.2 Safet Actions for Accidents (In Addition to Those for Transients) Relationshi to Safet Criteria 26 Establish Primary a) Prevent excessive radioactive Containment release. (Est Pri Cont) 27 Establish Secondary a) Prevent excessive radioactive Containment release. (Est Sec Cont) 31 Provide CRD Housing a) Prevent excessive fuel failure Restraints to Stop due to loss of capability to Control Rod Ejection- provide core cooling and Passive (Stop RodEjec) flooding. 32 Limit Reactivity Insertion a) Prevent excessive cladding tem-Rate - Passive (Lim Ins perature. Rate) b) Prevent system stresses in excess of allowable values in applicable industry codes. 33 Restrict Loss of Reactor a) Prevent excessive cladding tem-Coolant - Passive perature. (Lim LossCool) 36 Control Bay Environmental a) Prevent overexposure to radiation Control (Cont Bay Env) of control room personnel.

Prepared: E~ C~ rt Page 32 Date: ~~ of 307 Checked: Date: BPNWSG3"048 (Rev 1) C. M. Parker ~~~~~W&~~WWW~ 7.3 Safet Actions for S cial Events Nos Relationshi to Safet Criteria 38 Shutdown Reactor Prom a) Prevent inability to bring Outside Control Room reactor to subcriti.cal condition ( "d Outsd) if control room is inaccessible') gl 39 Coo down Reactor Prom Prevent inability to cool reactor Outside Control Room if control room is inaccessible. (Cooldn Outsd)

  +440  Provide integrity of                         a)      Avoid  failure of enclosed safety-required structures                                  related equipment due to (Struct Integ)                                   ~   phenomena such as flood, tornado, etc.

b) Ensure boundary of structures which contain radioactive

                                                           .materials.
  ~41 Provide adequate water                         a)      Avoid   failure of safety-related drainage at the site                                 equipment by draining excess (Drainage)                                           off the site so water does not enter designated buildings, and by preventing back flow into building drains.

42 Shutdown Without Control a) Prevent inability to bring Rods (Shutdn NoRod) reactor to subcritical condition independent of control rods.

  ~43   Spent Fuel Pool Cooling                      a)      Avoid spent fuel uncovery and or  Makeup (PuelPool            Clg)                 potential release of radioactive material into the refueling zone of the secondary containment.
  ~44 Provide                     Event              a)      Avoid excess' containment conditions by Seism'afe Shutdown (Seismic Shtd                . ki ii(i"

i 's 'p .

                                                             ~@               u)utgpwn
                                                       ,ifcdcf'skC, sf ks cC 8@jg'iI spic
                                             ~ ',l     + ~ogctp)1gs ~;, j c

geese 'safety action identificatYo~amlbeas. used throughout th analysis) most are derived from the PSAR, Appendix G,

   ~Newly defined safety actions which are not highlighted                               in       the PSAR, Appendix G.
                            ~ ~

Page 33 of 307 Prepared: Date:

                                                       ..Ecer Checked:                             Date:        /g BFNWSG3"048 (Rev       l)                          C. M. Parker aaoaaaaaaaaaaaeaaaaaaaaaaaaaaaeaaawaaaeaaaeaeaaaaeaaaeaaeasaeaaeaeaaaaaaaaaa 7.4   Safet    Function   Codes The  following special function codes are defined to designate special characteristics of systems which perform the safety actions identified above.'hey      are derived from the designators utilized in Appendix               G of the   FSAR  (reference 4.la)

Code G (General) The system is required to operate in a normal manner so as to avoid the possibility of an unacceptable safety result. SF (Single The system is required so that an essential safety Failure) action will meet the single-failure criterion as stated by the nuclear safety operational criteria (reference 4.la, Appendix G.Z.2). SFS (Shared This symbol is used to indicate that the syst: em Single Failure) shares with another system the obligation to meet the single-failure criterion. R (Restricted) One or more of the system's functions must either not be acting or not be capable of acting in order to satisfy 'operational nuclear safety criteria while the reactor is in the designated operating state. L (Limit) One or more of the key process parameters must be limited to satisfy nuclear safety operational criteria while the reactor is in the designated operating state. M (Manual) Credit is taken for personnel action in Appendix G) of the corresponding system. MSF (Manual Co xn nd single failure compliance (Single Failure) is required. A (Auto) ik'ote ~~a46pjgej.c fa 4 'p j'ux This code identifies those sysYems wh tion =as auxiliaries to the front-line safet system.'

SLY, SHQiTN%% AlKiKXSIS of Prepared: F~F E C~ e Page 34 Date: M~i/g+ 307 Checkedy Date! ~ wr BFHWSG3-048 (Rev 1) C. M. arker

                                                       %%%%RRRWQRQRNNHHW%%RRWRQSRRR 7 '   Reactor       ratin 'States ratin State                            Definition Reactor vessel-head    off Reactor shutdown Atmospheric pressure Reactor vessel head on Reactor shutdown Pressure less than 850 psig Reactor vessel head. on Reactor not shutdown Pressure less than 850 psig Reactor vessel head on Reactor shutdown Pressure greatez'han 850 psig Reactor vessel head on Reactor not shutdown Pressure greater than 850 psig Notes:   -   Original evaluations (reference 4.la) included operating state B (reactor vessel head off, reactor not shutdown, atmospheric presure). This operating state (reactor critical with the vessel head off) is no longer considered to be normal and is excluded from this analysis. Should any z'elated testing be proposed in the future,       it must be evaluated completely as a special case.
             -   Shutdown,.as used    in these definitions, means that the effective multiplication factor (Qff), is sufficiently less than 1.0 such that the full withdrawal of any one control rod could not produce      criticality under    the most
               ~

restrictive potential conditions of temperature, pressure, score age> and fission product concentration. (Item 40,. section l."2 of the PSAR, zeference 4.la.) Core decay heat is being removed at a contzolled ra stable c m. zn coze and reactor coolant system thermal d ign limits ~

Page 35 of 307 Prepared: M Date: Z~/IT 87 M. Checked: Date: r BFNWSG3-048 (Rev 1) Ce M. Parker srssassssssssassssssrsssarssssssssaassrsssssssssssssaasssaasasrsaaaaaaaasaa s 7~6 STANDARD SEQUENCES The following list of Standard Sequences were utilized in the Safe Shutdown Analysis of each event. They are derived from the Baseline Program analyses based on current design drawings. The purpose of the Standard Sequences is to develop a consistent set of actions for distinct initiation signals (i.e., Scram on high reactor vessel pressure, HPCI initiation on low water level (L2), etc.). Each event covered in the appendices to this report selects the appropriate standard sequence(s) (identified in the writeup) and adds any other event~nique safety actions. 7.6.1 Cooldown/Shutdown (SSA Safet Action 22) Normal isolated, normal non-isolated, and safety shutdown standard sequences are shown in Tables .7.6-1, 7.6<<2, and 7.6-3. 7.6.1.1 The normal, isolated shutdown sequence shown in Table 7.6-1 includes system actions as associated with maintaining core cooling and containment cooling for depressurization and long-term shutdown. It is assumed that the reactor has previously been scrammed. High pressure coolant is supplied by RCIC (71) and/er'PCI (73), while low pressure supply is by RHR-LPCI (74). Since heat is being passed to the suppression pool and into the drywell during the shutdown process the systems associated with heat removal from these areas are also shown in Table 7.6-1. The primary systems involved are RHR (74)- pool cooling, plus RBCCW (70) for drywell cooling. Some of the support systems for all of these functions are also included in the table. 7.6.1.2 If the reactor is not isolated (or scrammed), but shutdown is required, normal operational systems would be used to the extent possible (e.g. turbine controls passing steam to the coolant supplied by the feedwater system). main cond However, fa and tra 'afety 'equired, isolated. shutdown becomes necessary, It the unit would th ld e 'lar actions7.6-2to 'ted be scramme shutdown in the oyf e AP.'fit~ shows thts case, includi g t et manu 2'ptefs:iaj tpit~t gable 7.6-1. 7.6.1.3 Table .6-3 shows a wiaisssmtr 6ataty+ cysts j(nl ~hII fo t assumes that reactor scted htk so) o h'e sequence. already taken place. core cooling and containment s stems shown provide.'a ~ ab e

SLY SHUZSUBf JRKLXSIS Page 36 of 307 Prepared: >Pl. Date: 2~i'~/87 Mo ~ te BFHWSG3~8 (Rev 1)

                           . Checked!
                                  .            C. M. arker n'see:     ~W 7 '.2 Scram (SSA Safet        Action 17)

Low reactor water level (L3),.high drywell pressure, high reactor pressure, high neutron flux (detected by IRMs, see discussion'which follows), high neutron flux (detected by APRMs), generator trip, turbine trip, and MSIV closure scram standard sequences are shown in Tables 7.&4, 7.6-5, 7.6<<6, 7.6-7, 7.6-8> 7 6-9> 7.6-10, and 7.6-11 respectively. Response of the'CRD System (85) is added in each event that requires scram (to avoid duplicate listings of this basic action where more than one scram path occurs). All of these scram functions are identified in the tables as actions which must satisfy single failure requirements and must be fail-safe from the viewpoint that loss of power initiates the function. Some are specifically identified in the SSA as portions of the minimum protection actions required to shutdown from an accident event. Recent discussions have addressed the trip function from the Intermediate Range Monitors (IRMs) of the Neutron Monitoring System (92). SIL-445 (reference 4.lq) points out a unique failure of fuses on the negative (chassis) side of the circuit which could go undetected, yet could disable the function if enough IRMs are effected. At this time, final resolution of this .issue for BPNP is not complete. Therefore the SSA has treated IBM functions in the following manner: ~

l. All places where IBM trip would be expected are still shown in the appropriate events..

2. It is still identified as requiring satisfaction of single fa'ilure requirements (Safety Function code SP); however,

                                                                   ~

a note is added to each entry which references this section of the SSA (7.6.2) to alert everyone of the issue.

3. Backup scram from the Average Power Range Monitors (APRMs) function of the Neutron Monitoring System (92) is included in the two primary events for which licensing analysis has regularly been submitted (reference 4.1r) Control Rod (Appendix 24).. The SIL (reference 4.1q) states 'ccident that tPe APRM scram is sufficient for these events, justifk,ng operation o an interim period) until full resolution of the issue. is complete Pinal dis oui i Sf.t'e jdsdfgg pe~ ed 1 for Browns Pe befoje; the X+ ~id&Q $ c n:..

officiall down-gr'added br de146A8 Irked tfh~~$ a e 'v ua i S ~

SALE SHOTDOW hKLLTSIS Page 37 of 307 Prepared: W, Date: ~2. R/87 Mo ~ nst Checked: Date: BFNWSG3-048 (Rev 1) C. M. Parker aaaaaaaaaaaaaawaaaaaaaaeaaeaaaea~arassaeeeeaaaeaaeaaaaeeaaaaaaaeeaaaaeassaaaa

    .7.6.3   Pressure Relief (SSA Safet          Action 18)

SRV opening on high reactor pressure, opening of main turbine bypass valves (generator trip), and opening of main turbine bypass valves (turbine trip) pressure relief standard sequences are shown in Tables 7.6-12, 7.6-13, and 7.6-14 respe'ctively. 7.6.4 Core Coolin (SSA Safet Action 19) ECCS actions at low reactor water 'level (Ll), ECCS actions on high drywell pressure, HPCI initiation on low reactor water level (L2), and RCIC initiation on low reactor water level (L2) core cooling standard sequences are shown in Tables 7.6-15, 7.6-16, 7.6.17, and 7.6-18 respectively. 7.6.5 Reactor Vessel Isolation (SSA Safet Action 20) Low reactor water level (Ll), low reactor water level (L3), high drywell pressure, HPCI steam line low pressure, RCIC steam line low pressure, main steam line low pressure, and low condenser vacuum reactor vessel isolation standard sequences are shown in Tables 7.6-19, 7.6-20, 7.6-21, 7.6-22, 7.6-23, 7.6-24, and 7.6-25 respectively. Extensive use of the BPNP Technical Specification (reference 4.1t), Table 3.7-A was made in constructing these tables. 7.6.6 Power Reduction (SSA Safet Action 24) Recirculation pump trip (on generator trip, high reactor pressure, low reactor water level (L2), and turbine trip) power reduction standard sequences are shown in Tables 7.6-26, 7.6 27, 7.6 28, and 7.6 29 zespectxvel 7.6.7 Water Reduction (SSA Safet Action 25) HPCI, RCIC, main, and feedwater turbine trip (on high reactor water level (L8)) water level reduction standard. sequences are shown in T b e 7.6-31, 7.6-32, and 7.6-33. 7.6.8 Containme (SSA Safet Action

                                         '  ~~

Manual s ja'on pboL',~ling,pgqg~~js in Table 7.6-3 . 7.6.9 Secondar Containment Punctilios'i( Sa c Systems ident' d in'Section 8.2 for secondary con aT e t are shown in Table 7.6-35.

Page 38 of 307 Prepared! Date: 2~/9~87 Checked! M.. G nate Date! W /P BFNWSG3-048 (Rev 1) C. M. Parker

  ~  TABLE ttOttBER      STANDARD SE UENCE DESCRIPTION

7. 6-1 Normal Isolated Cooldown/Shutdovn

7. 6-2 Normal (Non-Isolated) Cooldown/Shutdovn

    '7 ~ 6-3     Safety Cooldown/Shutdown 7   6-4     Scram on Low Reactor Water Level (L3) 7  '-5      Scram  on High Dryvell Pressure 7 ~ 6-6     Scram  on High Reactor Pressure 7 ~ 6-7     Scram  on High Neutron Flux (IRM)

7. 6-8 Scram on High Neutron Plux (APRM) 7 6-9 Scram on PCV Closure, Generator Trip, Power Above 30Z.

7 '-10 Scram on Stop Valve Closure, Turbine Trip, Power Above 30X 7 '-11 Scram on MSIV Closure 7.6-12 SRV Opening on High Reactor Pressure 7.6-13 Opening of Main Turbine Bypass Valves, Generator Trip 7.6-14. Opening of Turbine Bypass Valves, Turbine Trip 7.6-15 ECCS Actions at Lov Reactor Wate 1 (Ll) 7.6-16 ECCS Actions on High Drywell Pres 7.6-17 HPCI Initiation on Lov Reactor Water Level (L2) 7 '.18 . RCIC Initiation on Lov Reactor Water L'evel (L2) ~ 7.6-19 Isolation Actions on Lov Reactor Water Level (Ll) 7 '-20 Isolation Actions on Lov Reactor Water Level (L3) 7.6"21 Isolation Actions on High Dryvell Pressure 7.6-22 HPCI Isolation on Low Steam Supply Pressure 7.6-23 RCIC Isolation on Low Steam Supply Pressure 7.6-24 61'osing of MSIVs on Main Steam Line Lov Pressure 7 '-25 Closing of Turbine Stop '& Bypass Valves on Lov Condenser Vacuum 7.6-26 Recirculation Pump Trip on PCV Closure, Generator Trip, Pover Above 30X 7.6-27 Recirculation Pump Trip on High R'eactor Pressure 7.6-28 Recirculation Pump Trip on Low Reactor Water Level (L2) 7 '-29 Recirculation Pump Trip on Stop Valve Closure, Turbine Trip, Pover Above 30X 7.6-30 HPCI .Turbine Trip on High Reactor. Water Level (L8) 7 '-31 RCIC Turbine Trip on High Reactor Water Level (LS) 7 7

       '-32
       '-33 7.6-34 1% Turbine T xp Manual Su
                                        'ctor Main Turbine Trip on High Reactor Water Level (LS) ssion Pool Cooling Water Level (LS) 7.6-35      Secondary C

SAFE SH(ITDDMN ANALYSIS Page ~of 3o'7 Standard SSA Sequences Table 7.6- 1 ~ PreparedM+ Date: z/'/BIB I BFN-OSS3-04B (Rev 1) Checked:~ . Date: ~/g,y/g r UNACCEPTABLE RESULTS SAFETY ACTION TVA SYS SYSTEN

                                                       'AFE'UNC TITLE       Code   Title         NO.        XANE        CODE                CONNENTS a NDRNAL ISOLATED COOLDO)I/SHUTDOMN        ARTSD)     Y" Fuel'Failure     19     Core Cooling     I  HAIN STEAN                Provide stean for HPCI (73) turbine, .

Fuel Failure 19 Core Cooling 1 NAIN STEAN Provide stean for RCIC (71) turbine. Fuel Failure 19 Core Cooling 2,CONDENSATE Provide noraally open water supply for RCIC (71) and HPCI (73) systeas. Fuel Failure 19 Core Cooling 3 FEEDMATER SF Provide low reactor water level signal (L2) to HPCI systea (73). Fuel Failure 19 Core Cooling 3 FEEDMATER Provide path for HPCI (73) flow to the vessel through the feedwater spargers. Fuel Failure 19 Core Cooling 3 FEEDMATER SF Provide lo>> reactor water level signal (L2) to RHR systea (74) logic for RCIC systea (71) initiation. Fuel Failure 19 Core Cooling 3 FEEDMATER Provide path for RCIC (71) flow tn the vessel through the feedwater spargers, Fuel Failure 19 Core Cooling 64 PRI CONTAINNENT ~ Provide alternate water supply for HPCI systea (73) froa suppression pool. 'Fuel Failure .Provide alternat~ water supply for RCIC ~ 19 Core Cooling 64 PRI CONTAIHNENT systen (71) through Core Spray systew (75) piping froa suppression pool. Provide suppression pool level indication. Accept RCIC and HPCI turbine exhaust stean. Fuel Failure 19 Core Cooling 67 EECM N Support RHR (74) LPC! operation. Fuel Failure 19 Core Cooling 71 RCIC SF Provide ECCS ATU power to FM systea (03) low water level (L2) instruaentation for HPCI systea (73) initiation, Fuel Failure 19 Core Cooling 71 RCIC SF Provide ECCS ATU power to FM systea (03) low ~ater level (L2) instruoentation for RGIC systea initiation. Fuel Failure 19 Core'Cooling 71 RCIC RCIC initiation on low reactor ~ater level (L2) signal frow RHR systen (74). Fuel Failure 19 Core Coo reactor water level (L2) signal. Fuel Failure 19 Core Coo n 7'R t a r~t(orQou press ls 1 on RH er level" Fuel Failure 19 Core Cool ng 74 RHR Send FM systea (03) low reactor water n1 la e ~ sys ea (71). Fuel Failure 19 Core Cooling 75 CORE SPRAY Provide flow path for water to RCIC systea (71) frow suppression pool (64). F uel Failure 19 Core Cooling 200 ELECTRICAL Support shutdown core cooling functions. Fuel Failure 30 Cont Cooling I NAIN STEAN Controlled nanual depressurization of RPV using SRVs> if suppression pool

                                                                   'eaperature     approaches   allowable Iinits.

If necessary> provide nanual ADS.

SAFE SHUTDD(N ANALYSIS Page4bof 80'7 Standard SSA Sequences Table 7<<6- 1 PreparedMW ate> z./I8/87 BFtHSS3-048 (Rev 1) Chected Date> ~~~ UNACCEPTABLE TVA 'AFE RESULTS SAFETY ACTTOH SYS SYSTEN 'UXC TlTLE Code Title NO. NANE CODE COXXEHTS (%53, ~t) Cont Stress 30 Cont Cooling 10BOlLER VNTS B Provide path for aain steaa,systea (01). DRN SRVs stean blowdown (frow controlled aanual depressurixation of RPV) to

                                                                 . suppression pool                             Pf Cont Stress   30     Cont Cooling   23 RHRSi(            .)SF     Support RN systea (76) suppression pool cooling node. Support drYwel) cooliny function.                                    PI Cont Stress  30      Cont Cooling   6I PRI COHTAI)O(ENT   'rovide          suppression pool teaperature and level indication to support RHR systea (74) suppression pool cooling node,~@~.~     E Cont Stress   30      Cont Cooling   H  PRl COHTAl)IEHT            Accept SRVs steaa blowdown (froa boiler t

vents and drains systea~ 10) to suppression pool. Cont Stress

                                                                                                   .~

30 Cont Cooling 67 EECN II Support RHR systea (74) suppression pool cooling function, Support drywell cooling f<<<<ctio<</Porta <<:I<<% Cont Stress 30 Cont Cooling 70 RBCCQ N Provide drywall cooling when power and cooliny water are available. g) Cont Stress 30 Cont Cooling 74 RHR NSF Provide suppression pool cooling function. Cont Stress 30 Cont Cooliny 200 ELECTRlCAL Support shutdown containoent cooling functions.

                                                                                ~        ~

SAFE SHUTDOMX AHALYSIS of 30'7 Standard SSA Sequences Tab]e 7.6- 2 Checked'agell Prep'ared: Pt /~ate> 2//8/87 BFH-OS63-048 (Rev I) Date: +z s/t'r UHACCEPTABLE TVA SAFE RESULTS SAFETY ACT]OH SYS SYSTEN FUXC TITLE Code Title HO. HANE CODE CONNEHTS a+ HORNAL (HOH-ISOLATED) COOLDOMH/SHUTDOMH Fuel Failure 17 Scran 85 CRD SF Nanual scran (if needed) frow reactor protection systea (99) 'will activate the CRD systen to insert control rods. Scraa function only. Fuel Fai]ure 17 Scran 99 REACTOR PRDTECTH NSF Nanua]'scran (if needed) before wain steaa systen (01) isolation. Fuel Failure 19 Core Coo]ing 1 NAIH STEAN Provide stean for HPCI ('73) turbine. Fue] Failure 19 Core Cooling 1 NAIX STEAN Provide stean for RCIC (71) turbine. Fuel Failure 19 Core Cooling 2 COHDEHSATE Provide noraally open water supply for RCIC (71) and HPCI (73) systeas, Fuel Fai]ure 19 Core Cooling 3 FEEDMATER SF'rovide low reactor water level signal (L2) to HPCI systee (73). Fuel Fai]ure 19 Core Cooling 3 FEEDMATER , Provide path for HPC] (73) flow to the vessel through the feedwater spargers. 3 FEEDMATER SF Provid~ ]ow reactor water level signal (].2) to RHR systea (74) logic for RC]C systen (71) initiation, Fuel Failure 19 Core Cooling 3 FEEDMATER Provide path for RCIC (71) flow to the vesseL through the feedwater spargers. Fue] Failure 19 Core Cooling 64 PRI COHTAIHNEHT Provide alternate water supply for HPC] systea (73) frow suppression poo]. Fuel Failure 19 Core Cooling 64 PRI COHTAIHNEHT Provide alternate water or RCIC systea (71) ore Spray systea (7 ) ron suppression pool .P

                                                                    ~ appraaataa   paa) 1     ,rttrprpah Ia       Qp Rplp aat:Rpat,ltaItipaagxtaitptbtaptrp a 4 Fuel Failure 19      Core Cooling   67     CM              gn ).'.%Aport,))HR;;,(74)Qk! '()er'h4ok.

Fue] Fai]ure 19 Core Cooling 71 R C wvV,. Cp 'i', Provide XCCS AT]f.p'owhr 'to FM systen (03) g!~' 'oss wat* level (L2) instrunentati HPCI systea (73) initia Fuel Failure 19 Core Cooling 71 RCI SF 'rovide EC~CS TU wer to FM systea (03) low water ]eve] (L2) instrunentation for RC]C systen initiation. Fuel Failure 19 Core Cooling 71 RC]C SFS RC]C initiation on low reactor water level (L2) signal frow RHR systea (74). Fuel Failure 19 Core Cooling 73 HPCI SFS HPCI initiation on FM systea (03) ]ow reactor water level (L2) signal. Fuel Failure 19 Core Cooling 74 PHR NSF Mith a reactor low pressure pernissive and RCIC/HPCI loss due to vessel depressurizationp nanua))y start one RHR puap in LPCI node to aaintain RPV water level. ,Fuel Failure 19 Core Cooling 74 RHR SF Send FM systen (03). Iow reactor water level signal (L2) to initiate RCIC systea (71).

, SAFE SHUTDOMN ANALYSIS Standard SSA Sequences Table 7,6- 2 BFXWSS3<4B (Rev 1) Prepared)~. Checked) Page4lof BQ t atea ~1~+ 7 ate>~F/+ UNACCcC'TABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN "FUNC TITLE Code Title NO. RANE CME CO!SENTS Fuel Failure 19 Core holing 75 CORE SPRAY 'rovide flow path for water to RCIC systea (71) froa suppression pool (64). [ Fuel Failure 19 hre Cooling 200 ELECTRICAL Support shutdown core cooling functions. Rad Release 20 RPV Isol 1 RAIN STEAN )IF . Nanually close NSIVs and steaa line drains after aanual scree frow reactor protection systea> 99 (if needed). Fuel Failure 30 'Cont Cooling 1 NAIN STEAN NSF . Controlled aanual depressurtxation of RPV using SRUs> if suppression pool teeperature approaches allo~able Iiwits. If necessaryt provide aanual ADS. Cont Stress 30, Cont Cooling 10 BOILER VNTS L . Provide path for wain steaa systea (01)

                                             'DRN                               SRUs   steaa blowdown (froa controlled wanual depressurixation       of RPV) to suppression pool                                      gl Cont Stress    30      hnt   Cooling   23   RHRSM                )SF         'Support RHR systea (74) suppression pool cooling aode. Support drywell cooling function.

Cont Stress 30 Cont Cooling 64 PRI CI(TAINlSiT Provide suppression pool teaperature and tu level indication to support RHR systea ~~ i7li suppressibn pro! cooling Inde,d~gysg<g Cont Stress 30 Cont Cooling 64 PRI CONTAIXNENT ~ Accept SRVs stean blowdown (frow boiler vents and drains systear 10) to suppression pool, Cont Stress 30 Cont Cooling 67 EECM

N Support RHR .systea (74) suppression pool cooling 'fungtiog., Support dry>>ell cooling

                                                                          ". 'unctioni f'~~~~'~~                     E~~h Cont Stress    30      hnt   Cooling 70    RBCCM    ,     .
                                                               '.:N     'rovide            drywall cooling     when power and
               'I
                                                                           .:cooling water ari available.
                                                                  '(:. "

Cont Stress 30 Cont Cooling 74 RHR :NSF Provide suppression pool cooling function. Cont Stress 30 Cont Cooling 200 ELECTRICAL.

                                           ~       '    , "
                                                                 ,    (
                                                                            ,Support functions, shutdown
                                                                                                   ~

containaent cooling

                                                                                                                              ~

, ~

                                                                                                                    ~
                                                                                                                       ~,
                                                                                                                       ~
                                                                                                                          ~

0' J I J ~

SAFE SHUTDOMN ANALYSIS Page+Sf 3u 7 Standard SSA Sequences Table 7.6- 3 BFH-'OS83<48 (Rev I) PreparedM+ Checked! Date: ate: ~/~ 2'j/8/E)"/ UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN FUNC TITLE Code Title HO. HARE CODE CONNEHTS aa SAFETY COOLDOMH/SHUTDOMX Fuel failure 19 Core Cooling I NAIH STEAN Provide steas for HPCI (73) turbine. Fuel Failure 19 Core Cooling I NAIH STEAN NSF Depressurize using ADS valves to establish loss pressure core cooling functions. Fuel Failure 19 Core Cooling 2 CDHDEHSATE Provide noreal)y open water supply for HPCI (73) systeo. This safety action way occurs but is not required. Fuel Failure 19 Core Cooling 3 FEEDMATER SF Provide low reactor ~ater level signal (L2) to HPCI systeo (73). Fuel Failure 19 Core Cooling 3 FEEDMATER Provide path for HPCI (73) flow to tbe

                                                               ,vessel through the feedwater spargers.

Fuel Failure 19 Core Cooling 18 FUEL OIL SF Support diesel generator systea (82). Fuel Failure 19 Core Cooling 64 PRI COXTAI)IEHT Provide alternate water supply for HPCI systee (73) frow suppression pool, Provide suppression pool level indication. Accept HPCI turbine exhaust steaa. Fuel Failure 19 Core Cooling 67 EECM SF Support RHR (74) - LPCI and/or diese) generator systea (82). Fuel Failure 19 Core Cooiing 71 RCIC Provide ECCS ATU power to FM systea (03) low water level (L2) instruaentation for HPCI systew (73) initiakion. Fuel Failure 19 Core Cooling 73 HPCI SFS HPCI initiation on FM systea (03) loss reactor ~ater level (L2) signal. Fuel Failure 19 Core Cooling r 74 RHR NSF Mith a reactor low pressure peruissive and HPCI loss due to vessel depressurization~ aanually start one RHR puap in LPCI code to waintain RPV water level. Fuel Failure 19 Core Cooling 7S CORE SPRAY SF Provide backup to RHR (74) - LPCI aude'f necessary. fuel Failure 19 Core Cooling 82 DIESEL GENERATOR SF Supply eaergency power to w r core coo)ingi if required. Fuel Failure 19 Core Cool g 86 DIESEL START AIR SF Support diesel generator systeo (82).

.Fuel Failure 19      Core Cooli g 200 ELECTRICAL                   p     t butdp~            coo n(..f     th Fuel Failure 30      Cont Cooli g                                             d      al'zd    es urtyPta n o    RP l Vlh"                   i i   uppj sio Pnj)')

Itufe approaches allowable lieits. If necessaryt provide eanual ADS. Cont Stress 30 Cont Cooli 3 FEEDMATER SF Provide in o RHR (74) for aanual initiation of containaent/torus spray. Fuel Failure 30 Cont Cooling 10 BOILER VXTS.L Provide path for wain steaa systew (01) DRX, SRVs steaa bio>>down to suppression pool (64)

                                                                      'upport diesel Cont Stress   30,    Cont Cooling     18 FUEL OIL         SF                         generator systew (82).

Cont Stress 30 Cont Cooling 23 RHRSM NSF Support RHR systea (74) suppression pool cooling aude and/or drywell spray oode.

SAFE SHUTDOlN AHALYSTS Pag~f 90 't Standard SSA Sequences Table 7.6- 3 BF)H)S83MO )Rev l) PreparedM Checkedy

                                                                                        ~    Date!

llltml

                                                                                                    ~le/g
                                                                                                    +s3//T INKEPTADLE                         TVA                   SAFE RESULTS    -SAFETY ACT)OH        SYS       SYSTEM      FSC TlTLE     Code    TiCle         )tO.       QNE        CODE              CO)SENTS

.Cont Stress 30. Cont Cooling 6q PRl CONTATNKXT Provide suppression pool teaperature and level indication to support RHR systea I gl (74) suppression pool cooling soda. Provide dry>>ell pressure and teaperature indication to support RHR systea )74) containaent spray node. ~~~< P<g<)r Fuel Failure 30 Cont Cooling 6q PRl CNTA)NENT SRVs steaa bio>>do>>n lfroa boiler 'ccept vents and drains systeas 10) to suppression pool. Cont SCress 30 Cont Cooling 67 EECQ Support RHR systea 174) suppression pool cooling node and/or containaent spray

                                                               ~ ode~ and support diesel generator operation.                                             Ki Cont Stress   30     Cont Cooling    7C RN                     Provide suppression pool cooling function.

Cont SCress 30 ConC Cooling 74 R)N Provide contain>>ant and torus spray aodest if necessary. These aodes require that RN~1 is running and that 2/3 core coverage is indicated (feed>>ater systea> 03)o Cont Stress 30 Cont Caaling 82 D)ESEL SEHERATOR SF Supply eaergency po>>er to RHR systea (74) suppression pool coolingi if required. Cont Stress 30 Cont Cooling 86 DlESEL START A1R SF Suppart'iesel. generator systea (82), gl Cont Stress 30 'ant Cooling 200 ELECTRICAL .Suppart shutdo>>n containaent cooling f funcCions..

SAFE SHUTDOWN ANALYSlS Page@of QO T Standard SSA Sequences Table 7.6- 4 Pr'epared c 6di4ate: "-/lc)/py DFN-OSS3-048 tRev 1) Checkedy Dates +g,3/j7 UNACCEPTABLE TVA ~

                                                   . SAFE RESULTS    SAFETY ACT)ON      SYS      SYSTEN       FNC TlTLE     Code   Title       NO.       NANE    'ODE                 CONNENTS aa SCRAN TNlTTATION OH LOU NTER LEVEL EL3)

Fuel Failure 17 Scraa 3 FEEDNZER SF Provide low water level 4L3) trip signal to reactor protection systea l99) fai)chafe logic. Fuel Failure 17 Screw 99 REACTOR PROTECTN SF Provide'craa signal to the control rod drive NACRD) systea (BS) on feedwater systea 103) low water level IL3) trip signal ~

SAFE SHUTDDN ANALYSlS Page4 of 'BOY Standard SSA Sequences Table 7,6- 5 Prepared~ ateiZ/IB/87 BRH583WS (Rev 1) Checked! Date1 +zy/gr UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTIDN SYS SYSTEN FUNC TlTLE Code Title NU. IANE CUDE CDOENTS aa SCRAN )NITTAT)DN DN HlSH DRYNELL PRESSURE Fuel Failure 17 Scraa 64 PRl CDNTATNNEXT SF Provide high dry>>e)l pressure trip signal to reactor protection systea )RPS) )99)

                                                                . fail~afe logic.

Fuel Failure 17 Scraa 99 REACTDR PRUTECTN SF Provide scraa signal to the control rad drive (CRD) systea lS5) on priaary

                                                               ~

containaent systea 164) high dry>>ell pressure trip signal.

 ~
                                                                                 ' i  ~(CIA I   ',  Ai~t(1>> s<++o ~ >>ol + ' ~ P4Ir'S tH ~ ~

I

SAFE SHUTDONN AHALYSlS Page47of 307 Standard SSA Sequences Tab)e 7.6- 6 Prepared gpygatea 2/lf. g f."7 BFH-OS63<tB (Rev 1) Checked< Date: gg,y/~> UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN FUHC T!TLE Code Title HO. HANE CODE CONNEHTS

                                                                                'a SCRAN 1H111ATlOH OH H16H REACTOR PRESSURE Fuel Failure 17     Scraa          3 FEEDNATER         SF    Provide high reactor vessel pressure trip signal to reactor protection sYstea (99) fai loafe logic.

Fuel Failure 17 Scrao 99 REACTOR PROTECTN SF Provide scraa signal to the control rod drive $ CRD) systeo 185) on feeduater sYstea (Q3) high reactor vessel pressure trip signal.

SAFE SHUTDOUN ANALYSlS Page4of 30'7 Standard SSA Sequencep Table 7.6- 7 Prepared~ -~~ @+ate> Z/iH/~ SFN<S63WS (Rev I) 'heckedy Datet ~as/EF UNACCEPTABLE TVA . ~ SAFE RESULTS SAFETY ACTION SYS SYSTEN 'UNC TITLE Code Title NO. NANE CODE SCRAH INITIATION,DN HISH IRN NEUTRON FLUI Fuel Failure 17 'craa 92 XEllPON NNITOR SF Provide IRN neutron flux trip signal to reactor protection systea {99) Isee section 7ebe2) ~ Fuel Failure 17 Scraa 99 REACTOR PROTECTN SF Provide scree signal to tbe control rod

                            ~
                                      .                       drive (CRD) systea IBS) on neutron monitoring systea I92).IRN neutron flux trip signal,

SAFE SMUTDOMN ANALYSIS Page'Hof 907 Standard SSA Sequences Table 7eh- 8 Prepared. Date: 2/ld/8'7 BFH-OSS3-048 IRev I) ~ Checked: Date! Qo/t'l UHACCEPTABLE TVA SAFE RESULTS SAFETY ACTIOH SYS SYSTEH FUHC TITLE Code Ti tie . NO. NAHE 'ODE CONiEHTS a% SCRAH IHIT. ON H!BH APRH NEUTRON FLUE 92 NEUTROH HONITOR SF Provide APRH neutron flux trip signal to reactor protection systea I99) fail-safe logic. Fuel Failure l7 Scrao. 99 REACTOR PROTECTH SF Provide scraa signal to the control rod

                                                                'drive (CRD) systea [85) on neutron aonitoring systea (92) APRH neutron flux trip signal.

/ SAFE SHUTDDQI ANALYSIS Page50of 3o7 Standard SSA Sequences Table 7.6- 9 Prepared Dater 2//8/g 7 IF)H)S63WB IRev I) Checkedr Date> Yc3/pr UNACCEPTASLE Tgi SAFE RESULTS SAFETY ACTIN SYS SYSTEN 'UNC TITLE . Code Title ND. NANE Cm)E aa SCRAN IHIT DN FCV CLDSUREt 6EN TRIPt (PNR>303) Fuel Failure 17 Scrau 1 NAIN STEAN'F Provide >303 turbine first stage pressure inter)oct signal to reactor protection systes (99) fail-safe logic. ,Fuel Failure Rl 17 Scrao I7 TURSINE CDNTRDL SF Provide nain turbine control valve fast closure trip signal to reactor protection systeo 199) fai)chafe logic. Fuel Failure 17 Scraa 99 REACTDR PROTECTN SF Provide scraa signal to tbe control rod drive (CN) systeu (85) on aain turbine control val've fast closure trip signal froo turbine control systea (C7) and sain

                                                               'teaa systes 101) signal indicating turbine first stage p'reassure > 30K.

e SAFE SHUT0$ 0) ANALYSIS Standard SSA Sequences Table 7.A- 9 BFlHS63WB (Rev 1) Prepared Checked! Page50of 0ate> Oaten

                                                                                                    ~p 2/s8/g 7 Yia/pg  .

WACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN FUNC TITLE . Code Title N0. NANE mE ee SCRAN IHlT ON FCV CLOSURE'EN TRIPi <PNR>30K) Fuel Failure 17 Scran I RAIN STEAN'F 'rovide >301 turbine first stage pressure interlock signil to reactor protection systen (99) fail-safe logic Q

 , Fuel Failure 17      Scree        47 TURSINE CNNOL SF         Provide nain turbine control valve fast closure trip signal to reactor protection systeo 199) failwafe logic.

Fuel Failure 17 Scras 99 REACT0R PROTECTN SF Provide scrao signal to the control rod drive (CR0) systen ISS) on nain turbine control val've fast closure trip signal fron turbine control systea 147) and aain

                                                               'teaa systen (OI) signal indicating turbine first stage prj.ssure > 30$ .

SAFE SHUTDDHH ANALYSIS Page 5) of Za7 Standard SSA Sequences Table 7.b"IO PreparedM a- c rate: 2//e)/H7 BFH-OSB3<48 IRev 1) Checkeda Sate: ~ay/F7 UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN 'FUNC TITLE Code Title NO, HANE CODE CONNEHTS 5

                                                                           \

aa SCRAN IHIT. ON STOP VLV CLOSURE'NB TRIP)PHR>30X Fuel Failure 17 Scraa I NAIH STEAN SF Provide ) 30K turbine first stage pressure interlock signal. to reactor protection systea 199) fai loafe logic. Fuel Failure 17 Scraa 1 HAIN STEAN SF Provide nain turbine stop valve < 90K open trip signal to reartor protection systea f99). Fuel Failure 17 Scraa 99 RPS . SF Provide'scran signal to the control rod drive (CRD) systea (85) on signals froa aain stean systea (OI) indicating aain turbine stop valves < 90K opin and turbine first stage pressure > 30K.

0 SAFE SHUTOONN ANALYS1S Standard SSA Seguences Table 7.6-)1 SFN-OS83WS IRev 1) Prepared Cbeckedi Page Q,of Bi>7

                                                                                        ++Rate>

Oate! Pj)t'rP'7

                                                                                                  +>>/Ev UNACCEPTABLE                     TVA                   SAFE RESULTS    SAFETY ACT1ON       SYS      SYSTEN       FUNC TlTLE     Code   Title       'NO,       NANE        CSE aa SCRAN 1NITlATION ON NS)V CLOSURE Fuel Failure )7     Scree,l           NAlN STEAN       SF     Provide NSlV < 90X open trip signal to reactor protection systes I99) fail-sift
                                                   'uel logics Failure 17   . Scrao          99 REACTOR PROTECTN SF   'rovide scree signal to tbe control rod drive ICl5) systeo I85) on cain stean systee <01) NS)Vs < 90X open trip signal Iif in RUN soda),
                                                                                                            ~ ~

SAFE SHUTDOUH ANALYSlS Page G of 367 Standard SSA Sequences-Table 7eb-)2 Prepared~.&~~5 ~ ate: 2/I~I8' SFH-OS63-0iB (Rev 1) Checked Date: ~e+t UNACCEPTABLE . TVA SAFE RESULTS SAFETY ACTIOH SYS SYSTEH 'UNC

   'TITLE     Code    Title         HO.      HAHE        CODE             COHHEHTS ee SRV OPEHIHB   ON HI6H REACTOR PRESSURE Syst'tress    18      Pres Relief     1 HAIN STEAN       SF    SRVs open on  high reactor pressure, Syst Stress   18      Pres Relief    10 BOlLER VNTS 1           Provide path for aain steaa systea 101)

DRN SRVs steaa bloudoun to suppression pool fbi'), Syst Stress 1$ Pres Relief Sh PR1 CONTAlHHEHT 'ccept SRVs steaa hlowdoun (froe boi]er vents anb brains systea~ 10) to suppression pool.

SAFE S)ITOONN ANALYSlS Paqe54of ~'T Standard SSA Sequences Table 7.6-)3 Prepared~ Gates 2/I&/H7 SFMSSHAS lRev 1) Chactada Batt> ~/&3/~ UNACCEPTABLE RESULTS TfTLE SAFETY ACTlOH Code Title TVA SYS EE. SYSTEN EEET

                                                          'IC SAFE CtIE aa OPEN TURBlNE BYPASS VALVES GH SEHERATOR TRlP Fuel Failure 18       Pres Relief    1 NATN STEAN              llatn turbine bypass'alves open on turbine control systca (h7) generator trip signal.

Fuel Fai)urt 18 Pres Relief 47 TURBINE CONTROL signal to optn and control cain steaa

                                                                                                              'end systta 101) bypass valvts on gcncrator trip.

o E ~ ,0 od

SAFE SHUTDD)I AHALYSlS PageS of 90'1 Standard SSA Sequences Table 7.b-)4 Prepared'< v(/~Date> 2/IB/g'7 BFH<663<48 (Rev .1) Checked: Datei gal/Sr~ UXACCEPTABLE TVA SAFE = SAFETY ACTTOH

               'ESULTS SYS     SYSTEN        FUXC TlTLE     Code       Title         NO.      HADE         CDDE             CONNEHTS ~

aa OPEH TURBlliE BYPASS VA!VES OX TURBIHE TRlP Fuel Failure 18 . Pres Relief 1 NAIH STEAN Nain turbine bypass valves open on turbine control systea 147) turbine trip signal. Fuel Failure 18 Pres Relief 47 TURBIHE COHTROL Send signal to open and control oain stean. systems (01) bypass valves on turbine trip'.

SAFE SHUTDOHN ANALYSIS, PageNof 3C 7 Standard SSA Sequences

                                     'Table 7.6-15                           Prepared.P~             . gcc9atea          2//8i~

BFN-OSS3WB (Rev I), Checked< Date: ~o/gg {WACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN 'LIC TITLE Code Title NO. - Nw{E CODE CONENTS tt ECCS ACTIONS AT LOi) REACTOR {(ATER LEVEL (L1) Fuel Failure 19 Core Cooling 3 FEEDNATER SF Provide low.reactor water level (L1 1 L3 peraissive) signals to iain stean systea (OI) for ADS. Provide Low reactor

                                                                            'ressure signals to core spray systea (75) and RHR systea {74).

Fuel Failure 19 Core Cooling. 18 FUEL OLL SF Provide diesel &el oil to diesel generitor systea {82) starting on low Fuel Failure 19 Co're Cooling

                                          'I 64 PRI COHTAI)O{ENT
                                                                      ~
                                                                          ~
                                                                            . reactor water level {Ll) signa)s.

Support core spray systea {75)+by

                                                                                                                          ~ ~g~
                                                                           ~

providing roon cooling with service water froa EECQ (67) in support nf ECCS actions at low reactor water Level (Ll).

~

Fuel Failure 19 Core Cooling 64 PRI CONTAINNENT Provide torus water supply for RHR,(74) and Core Spray l75) systeas in support nf ECCS.actions at lo>> reactor water level Fuel Failure 19 Core Cooling 67'ECQ 'F (Ll). Provide cooling wate'r.to dieseL generator systea (B2) starting on low reactor water . level (LL) signals. Fuel Failure 19 Core Cooling 67 EECQ SF Provide cooling water to'RHR-LPCI {74),and/)(k assi core spray rona coolers (64) in 'support of ~> N~'Q ECCS actions at, low reactor ~ater level {Ll). Fuel Failure 19 Core Conling 68 RECLACULATION SFS Provide low.reactor pressure peraissive signals to core spray systea l75) and RHR' systea l a o>>. reactor water level lLL). Fuel Failure 19 Close R Cnre Cooli recirculation pi~chape.va(Vhs 68 RECIRCULATION SF

                                                            ~

e.". o pep i jatio'ii)p flphLs free~lkpIsy ~

                                                                        'Q   signals low reactor pressure peraissive
                                                                                                                              ~
                                                                'F
                                                                                       ~

Fuel Fiilure 19 Core Cooling 71 RCIC QgtidWCCS-AT~r o eedwater systea (O3) low water level {I.I) and FH systea and recirculation systea (68) low 'reactor. pressure instruaentation. Fuel Failure 19 Core Cooling 74 RHR SF Start puaps on low reactor water level {LI) signals froa core spray systea (75). Provide signals to close recirculation valves to recirculation systea (68).'F {)pen injection valves on low reactor pressure signals froa core spray systea {75) and RHR systea l74) LPCI initiation signals due to low water level {L1).

                                                         '.Ir,
                                                          ~       q             ~

0, SAFE SHUTDD((k ANALYSIS Page57of 9O I Standard SSA Sequences Tab)e 7.b-15 Prepared.P+ <. 4c<ate: ~llHIH'~ BFN-GS63-048 (Rev I) Checked:~ Date: +g,o//7 UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN 'UNC TITLE Coda Title NO. NRNE CODE CO)SENTS Failure 19 Core Cooling 75 CORE SPRAY SF Send feedwater systea (03) loss reactor water Ievek signals (LI) and loss reactor pressure signals (fron feedwater systea (03) and recirculation systea (b8)) to RHR systea (74). Fuel Failure 19 Core Cooling 75

                                           'uel CORE SPRAY                  Start puaps on feedwater systea (03) low C

reactor water level (Ll) signals. Fuel Failure 19 Core Cooling 75 CORE SPRAY injection valves on feedwater systea

                                                                                                       'pen (03) or recirculation systea (bB) )o>>

reactor pressure signals in conjunction with low water level (Ll) signal froa feedwater systea (03). Fuel Failure 19 Core Cooling 75 CORE SPRAY SF Provide standby diesel generator (82) start signals on feedwater systea (03) low reartor water level signals (Ll). Diesel generators'start en low reactor water level signals (Ll) froa core sp'ray systea (75). Fuel Failure 19 Core Cooling Bb DS(. BEH START SF Provide diesel starting air to diesel AIP. generator systea starting on low reactor water level (Ll) signals. 41

SAFE SHUTDOHH ANALYSIS Page93of BC 1 Standard SSA Sequences Table 7.6-16 Prepared@/ fsd6c&atel df'n /P>7 BFH-OSS3-048 IRev I) Checked! Date: +is/g 7 UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTIN SYS SYSTEN FUNC TITLE Code Title HO. NANE CODE CONNEHTS aa ECCS ACTIONS AT HIGH DRYMEU. PRESSURE Fuel Failure 19 Core Cooling I NAIH STEAN Pravide stean for MPCI (73) turbine in support of HPCI initiation on high drywe)) pressure. Fuel Failure 19 Core Cooling 1 HAIN STEAN Provide reactor vessel auto depressurrixation systea lADS) on priaary containaent systea ]64) high DH pressure signal coincident >>ith FH systea ]3) LML

                                                                       ~

(L]LL3) for nore than 120 seconds 6 either two core spray (75) puaps or'one RHR ]74) puap funp]ngt Fue] Failure 19 Core Cooling 2 CONDENSATE Provide noraally open water supply for HPCI systea ]73) initiation on high drywall pressure signals. Fuel Failure 19 Core Cooling SFS Provide low reactor pressure perwissive signals to core spray systea (75) and RHR systea (74) for ECCS actions on high drywall pressure signals, Fuel Failure 19 Core Cooling . 3 FEEDHATER P<ovide path for HPCI ]73) flow ta the I Yesse] through the feedwater spargers for HPC] initiation on high drywell pressure. fuel Failure 19 Core Cooling 10 BOILER VNTS 6 Pravide path for wain stean systea IO]) DRH reactor vessel autonatic depressurixation systea lADS) stean blowdown to suppression . paol (64). Fuel Failure 19 Core Cooling 18 DIESEL FUEL 0]L SF Provide diesel fue]'il to diesel generator systea ]82) starting on high , drywall pressure signals, Fuel Failure 19 Core Coaling 64 oRl COHTAIHNEHT dr well pressure signals to

                                                                          'core spra s y stew         l               (74)t ain stean systea <0]) for ADS.

Fuel Failure 19 Care Coa]ing '64 PR] COHTA]HNEHT P vie u a r su ly for II syl Iliny sxgna]st steaot oad Egg t4 Fue] Failure 19 Core Cooling 64 PR] COHTAIHNEHT s stea ]75)~by g~/g7 provadlng roon coo]lug wg frow EECQ ]67) for ECCS actions on high drywall pressure signal. Fuel Failure ]9 Core Cooling 64 PP] COHTA]HNEHT Accept reactor vessel autoaatic. depressurixation systea (ADS) steaa blowdown ]fran bailer vents and drains systeal 10) to suppression poo], fuel fai]ure 19 Core Cooling 64 PRI COHTA]N NT Provide torus water supply for RHR ]74) and Core Spray $ 75) systea for ECCS actions on high drywall pressure.

                            \

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                                                             ~ ~'

I t

SAFE SHUTDOHH AHALYSIS Page5Rof gO'7 Standard SSA Sequences Table 7.6-1& Prepareda k~ate: 2/Ieu 7 BFH-OS63-04S (Rev I) Checked: UNACCEPTABLE TVA SAFE Oaten'~y/d7'uel RESULTS SAFETY ACTIOH SYS SYSTEN FUNC TITLE Code Title HD. XANE CODE CONNEHTS Failure 19,Core Cooling 67 EECH SF Provide cooling water to diesel generator systea f82) starting on high drywall pressure signa)s. Fuel Failure 19 Core Cooling 67 EEL povidg cooling water to RHR-LPCI f74), and to<core spray rooa coolers f&4) (or ECCS ytctions on high drywell pressure Fuel Failure 19 Core Cooling &8 RECIRCULATIOH SFS Provide low reactor pressure peraissive signals to core spray systea f75) and RHR systea (74) for ECCS actions at high dry>>ell pressure. Fuel Failure 19 Core Cooling 68 RECIRCULATIOH SF Close recirculation discharge valves for LPCI injection on signals froa RHR systea (74) LPCI initiation signal due to hjgh reaEtnr>peraassive signals. Ncaa,/~~ Core Cooling .71 RCIC Provide ECCS ATU po~er to priaary containaent systea f&4) high drywall

p. assure instruaentation and feedwater systea .";=) and recirculation systea 16S) low reartor prepsure '.r~iruaentation.

Fuel -Failure 19 Core Cooling '73 HPCI SFS HPCI initiation on core spray systea f75) high drywall pressure signals. Fuel Failure 19 Core Cooling 74 RMR SF Start puaps on priaary containaent systea (64) high drywall pressure signals with low reactor pressure signals froa core spray systea (75). Provide signals to close. recirculation valves to r Fuel Failure 19 Core Cooling Open injection valve on low reactor

                                                                                                     ~

74 RHR SF pressure salina s+om, c r pytajac f~P It l qS))tf)dlfIII It I iptlkt'ibj " $ pnIp}>i

                                                                                    "     pgp IPflj      'pl.n ,'rn",'

Q Fuel Failure 19 Core Cooling 74 RMR I nvjpn'nr'pnnl4thnXtllH jjnlp~~n running) to aain steaa systea f01) for the autoaatic depressurization systea fADS). Fuel Failure 19 Core Cooling 75 CORE SPRAY feedwater systea (03) and recirculation systea f&B)) to RHR systea t74). Send priaary containaent systea f64) high drywall pressure signals to HPCI systea f73). 0 Fuel Failure 19 Core Cooling 75 CORE SPRAY SF Start puaps on priaary conf'ainaent systea (64) high drywall pressure signals with feedwater systea 103) or recirculation systea (68) lo>> reactor pres'sure signals.

SAFE SHUTDON ARALYSlS Page&Oof 3C / Standard SSA Sequences Table 7.6-16 Prepared~,PÃ .c. laAate: Z~I~I87 DF&OSS3WB (Rev 1) 'hected> Date) +z3/~ NACCEP TABLE TN SAFE RESULTS SAFETY ACT10k SYS SYSTE)1 , FlNC TlTLE Code Title N. NlNE CODE . CONKkTS Fuel Failure 19 Core Cooling 75 CORE SPRAY 'F Open injection va)ve on feeduater systea (03) or recirculation systea l6S) low reactor pressure signals in conjunction uith high dryaell pressure signals froa priaary containaent systea (64).

 . Fuel Failure 19     Core Cooling   75 CORE SPRAY         SF    Provide standby diesel generator 182)
                           ~  ~

start signals on priaary containaent systea f64) high dryuel) pressure signals. Fuel Failure 19 Core Cooling 75 CORE SPRAY SF " 'rovide signals lthat core spray puaps are-running) to aain steaa systea 1Q1) for the

                                                           ~      autoaatic depressurization systea 1ADS).

Fuel Failure 19 Core Cooling S2 DlESEL SEHERATOR SF Diesel generators start on priaary containaent systea 164) high dryaell pressure signals. Fuel Failure 19 Core Cooling S6 DlESEL START AIR SF Provide diesel" starting air to diesel generator systea (S2) starting on high dryuell pressure signals. 0 1 ~ ~

SAFE SHUTDOMN ANALYSIS Page&I of 901 Standard SSA Sequences Table 7.6-17 Prepared (uAate: Z/'/fy/pe BFN-OS63<48 IRev 1) Checked: Date! ~w.s/g1 UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTIOH SYS SYSTEH FUNC TITLE Code Title NO. NAHE CODE CO)SENTS a> HPCI IHITIATIOH OH LOM MATER LEVEL IL2) Fuel Failure 19 Core Cooling 1 HAIN STEAN Provide stean for HPCI (73) turbine in 1 support of HPCI initiation on low water level (L2). Fuel Failure 19 Core Cooling 2 Provide noraally open water supply for HPCI systea <73) initiation on lo>> water level IL2). CONDEHSATE-'uel Fuel Failure 19 Core Cooling 3 FEEDMATER SF Provide lo>> reictor water level signal IL2) to HPCI systea (73). Fuel Failure 19 Core Cooling 3 FEEDMATER . Provide path for HPCI {73) flow to the vessel through the feedwater spargers for HPCI initiation on low water level <L2). Failure 19 Core Cooling 64 PRl CONTAIOEHT .Provide alternate source water supply for HPCI systea (73) frow suppression pool for HPCI initiation on low water level (L2) signa)i Accept HPCI turbine exhaust stean Fuel Failure 19 Core Cooling ~ 71 RCIC SF Provide ECCS ATU power to FM systea l03) low ~ater level (L2)'instruaentation for HPCI systea (73) initiation. Fuel Failure 19 Core Cooling 73 HPCI SFS HPCI initiation on FM systea <03) low reactor water level IL2) signal.'

SAFE SHUTDOVH ANALYSIS Pagct Lof ~"I Standard SSA Sequences Table 7.6-1B Prepared -(Matc: ZXIgg8'7 SRHS83WB (Rev I) Checked! Date: p~s/(l 7 UHACCEPTABLE TVA RESULTS SAFETY ACTION 'SYS

  'TITLE     Code       Title         NO.

aa RCIC IHITIATION ON UI HATER'LEVEL (L2) Fuel Failure 19 Core Cooling" 1 NIH STEAN Provide stcaa for'CIC (71) turbine. Fuel Failure 19 Core Cooling 2 CONDENSATE Prov'ide noraally open water supply for RCIC systea (71), Fuel Failure 19 Core Cooling 3 FEEDUATER SF Provide low reactor water level signal'L2) to RHR systea (74) Ingle for RCIC systea (71) initiation. Fuel Failure 19 - Core Cooling 3 FEHIATER . Provide path for RCIC (71) flow to the vessel through the feed>>ater spargers. Fuel Failure 19 Core Cooling 64 PRI CONTAIN(ENT Provide alternate source water supply for

                                                          ~

RCIC systea (71) through Core Spray systea (7S) piping frow suppression pool. Provide suppression pool level indication. Accept RCIC turbine exhaust. stean. Fuel Failure 19 Core Cooling 71 RCIC SF Provide ECCS ATU power to FN systea (03) low water level '(L2) instrunentation for RCIC systea initiation, Fuel Failure 19 Core Cooling 71 RCIC SFS RCIC initiation on low reartor water level (L2) signal froa RHR systea (74). fueL Failure 19 Core Cooling 'N. RHR SF Send feedwater systea (03) lo>> reactor water level signal (L2) to initiate RCIC systea (71). Fuel Failure 19 . Core Cooling 7$ CORE SPRAY Provide flow path for alternate source of water to RCIC systea (71) frow suppression pool .(64). C

                    ~ ~
                         ~  ~

SAFE SHUTDDMX AXALYSIS Page(kof 30"I Standard SSA Sequences Table 7e6-)9 Prepared'&. ates  ? r') Sr) BFH-OSS3-04D (Rev I) Chectedt Date: +i.p/g7 II UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTIOH . SYS SYSTEN FUXC TITLE Code Title XO. . HANE .CODE CONNEHTS aa ISOLATIOX ACTIOXS OH LOX HATER LEVEL ILI) Rad Release 26 Est Pri Cont I NAIH STEAN SF Close NSIVs and aain steaa drain lines on low reactor water level (LI) signal froa priaary containaedt systea 164). Rad Release 26 Est Pri Cont 3 FEEDXATER Provide low reactor water level (Ll) signal to priaary containaent systea l65) fail-safe logic. Rad Release 26 Est Pri Cont 43 SNPL 6 MTR DUAL SF Close saapling and water quality systea iso)ation valves on low reactor water level ILI) signal froa priaary containaent systea f64). Rad Release 26 Est Pri Cont 64 PRI COXTAIHNEXT SF Send feedwater systea'[03)'ow reactor water level ILI) signa) to initiate aain steaa line iso')ation> systea (01) and saaplihg I water quality systea (43) line isolation;

                                                                       ~ ~ ~     e', ~ ~ 'I (g. s        4(  '   a }

1 Qi ~ ~

SAFE SHUTDOUN ANALYSIS PageWf 3O"7 Standard SSA Sequences Table ?s 6<0 PreparedW I//+ate> ZQ(8/87 SF)H)SOTS (Rev I) Checked! 'aten'zy /gal NACCEP TAB}E TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN 'U)N: TITLE Code Title N. .RANE CODE CONNENTS aa ISOLATION ACTIONS ON L(N HATER LEVEL (L3) Rad Release 20 RPV Isol 69 RQCU Close RRCU isolation valves on lo>> water level (L3) signal froa priaary containaent Rad Release 20 RPV lsol '4 RHR systea ( RHR isolation signal tripped on low water level (L3) signal

                                                                             . froa priaary rontainaent systea (64).

o

 'Rad Release    2&      Est Pri Cont     3 FEEDUATER                            Provide lo>> water level (L3) signal to reactor protection systea (99) for, initiation of L3 isolations.

Rad Release 2& Est Pri Cont 32 CONTROL AIR SF Perfora isolation action(s) upon receiving low water level (L3) isolation signal froa the Priaary Containaent systea (&4). Rad Release 2& Est Pri Cont &4 PRI COHTAIKNENT SF Upon lo>> water level (L3) signal froa reactor protection systea (99)i'nitiate 0 L3 isolation actions and send priaarylsecondary containaent isolation signals to systeas 32'5~ 69~ 74'5~ 76~ 77$ 84~ 90~ ind 94. Rad Release 26 Est Pri ConC 75 CORE SPRAY Perfora isolation action(s) upon receiving low water level (L3) isolation signal froa the Priaary Containaent systea (64). Rad Release 26 Est Pri Cont 76 CONTAIHNEXT Perfora isolation action(s) upon receiving low wat'er level (L3).i'solation signal froa INERI'ad the Priaary Containaent systea (64). Release 26 Est Pri Cont 77 Perfora isolation action(s) upon receiving ow wa e the Priaary Containaent systea (&4), Rad Release 26 Est Pri'ont 84 P$ j&firw4qlgiye4aMep{s),upon cec~jvj 9

                                                                                                                                  )f
                                                                          '.')ow wakey'AgpL3) imli(pp'n sagnaL'fp I                 tt'te Priiary (.'ontki/a&4 +step 44)';-"  '; g'.

Rad Release 26 ,Est Pri Cont 90 DIAT)OH Sr trsrfors isoi ~ ttos oriioois'i otrsstskitrtly 1 HITOR, Iow water level (L3) isolation signal froa he Priaar Containaent sy (64))~ Rad'Release 2& Est Pri Cont 94 TIP SF ~ Initiate TIP wi rawa>on o (L3) signal froa priaary containaegt systaa (64), W (W IS>~~+"~) <<+7 Rad Release 26 Est Pri ConC 99 REACTOR PRDTECTN SF Provide low water level (L3)'signal froa feedwater systea (03) to priaary containaent systea (&4) for 'i'nitiation of l.3 isolat'ions. Rad Release 2? Est Sec Cont 64 PRI CONTAINNB(T SF Perfosa isolation action(s) upon receiving low water level (L31 i'so)ation signal froa the Reactor Protection Systea (99). (See Safety Act<on 26)

                                                                                                                              ~ ~

Vr o

SAFE SHUTDOMN AHALYSlS Page&of '3O'7

                                    'Standard SSA Sequences Table 7.6-20                   Preparedx~. P~

BFN-OSS3-0kB (Rev 1) Checked x Datex +i>/gP UNACCEPTABLE TVA SAFE RESULTS 'AFETY ACTION. SYS SYSTEN '<IC TlTLE Code Title NO.. NI<E CODE CONNENTS

 'ad    Release  27   'st  Sec Cont    65 SST                        initiate   SGT plant start on low water level (L3) signal froa prieary containaent systea <6%). <See Systea 64~ Safety Action 26)

Pars Overexp 36 Cont Bay Env 3 FEED<<ATER SF Provide low water level <L3) signal to RPS systea (99) for initiation of control bay isolation. Pars Overexp 36 Cont Bay Env 31 AIR COHDlTlOMINS SF Air conditioner (AC) supp)y ducts aust isolate and Ewergency Pressurization Systea (Control Roon Eaerg. Vent. Systea) aust supply pressurized filtered air to cain control roon on low water level (L3) t

                                                                  . signal froa Priaary Containoent Systea (64).

Pers Overexp 36 Cont Bay Env 64 PRl CONTAlHNENT SF Upon 'lo>> water level <L3) signal froa RPS(99)x send isolation signal to air conditioning systea(31). Pars Overexp 36 Cont Bay Env 99 REACTOR PROTECTN SF Provide low ~ater level<L3) signal froa feedwater systea (03) to priaary containaent systea <64) for initiation of control bay isolation. Rf

                             ~~

i SAFE SHUTDON AXALYSlS Standard SBA Sequences Pagebhf ~ Ta)p)c 7.6-2l. Prepared ~ate: 2/l8li' BFN<SB3-048 (Rcv  !) Checked: Oaten'fgtcp y v o llACCEPTABLE TVA ~ SAFE RESULTS SAFETY ACT10N 'YS SYSTEN FUXC Tll1.E Code Title . NO.. NANE CODE COINENTS aa ISOLATlDH ACT1ONS ON H16H DRYUELL PRESSURE Rad Release 26 Est Pri Cont 32 CONROL AlR Perfora isolation actionls) upon receiving high dry>>ell pressure isolation signal froa thc Priaary Containaent systea l64). Rad Release 26 Est Pri Cont 64 PRl CONA)NHENT SF Provide high dry>>cll pressure signal to reactor protection ystca (99). for initiation of high dry>>ell. pressure isolations. Rad Release 26 Est Pri Cont 64 PRl COHTAlNNENT SF high dry>>ell pressure signal froa

                                                                                                 'pon reactor protection systea (99)p initiate high dry>>ell pressure isolation actions and send priaaryhccondary contain>>ant isolation signals to systeas 32p 65p 7hp 75p 7&i 77             hp 90p and   94.

Rad Release 26 Est Pri Cont 74 RHR RHR solation signal tripped on high dry>>c)) pressure signal froa thc Priaary Containacnt systea (64). Rad Release 26 Est Pri Cont 75 CORE SPRAY Perfora iso)ation actionls) upon receiving high dry>>e)l pressure isolation signal froa tne Priaary Containaent systca (64). Rad Pelease 26 Est Pri Cont 76 CONA)I(EHT SF Perfora isolation actionls) rec iying 1NERT high dry>>ell pressu atio'n signa

                                                                             ~

froa the Pr Containacnt sist 6 ). Rad Release 26 Est Pri Cont 77 RADUASTE P iso)ation aetio $ r iv ng o igh dry>>ell ppgsy i 1 '0 'gn (ron tly~lba(YCoh in n ) (64). Rad Release 2& Est Pri Cont 84 aires+~'stha$ roniastrpntsp si on rsrsiviat tt >>a(res>>ltoptsasnra isolation sian ('jIIau PFiaary Containaent ea l64). Rad Release 25 Est Pri Co er ora isolation ac' upon receiving high dry>>ell ure isqlation signal froa riaary ContainacgjpyEtg~a( Q. Rad Release 26 Est Pri Cont 4 Tlp 5 nitia!e Tip >>ithdra>>agon Yiigh Pry>>efl pressure signal froa priaary cdntainccnt systea (64).'(~>>~>' f<a'4>> Rad Release 26 Est Pri Cont 99 REACTOR PROTECTN SF Provide high dry>>ell pressure signal froa priaary containacnt systea lb4) to priaary containaent systea (64) for initiat'son of

                                                                               ))igh dry>>ell pressure isolations.

Rad Pe)case 27 Est Sec Cont 64 PRl COHTAl)IENT SF Perfora isolation actionls) upon receiving high dry>>ell pressure isolation signal froa the reactor protection systea (99). Rad Release 27 Est.Scc Cont 65 KBT initiate SBT plant start on high dry>>el) pressure signal froa priaary containacnt systca (&4). I ~ v o

                              ~

4

SAFE SHUTDON ANALYSIS Page67nf 8> t Standard SSA Sequences Tab)e 7.b-2) Prepared:hf~ Bi~Date: >~IV.~R ) BFN-OSS3-048 tRev') Checlred: Date: g~y/I 7 UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN FUHC TITLE Code Title ND, NANE CODE CONNENTS Pars Overexp 3b Pars Overexp 3l AIR CONDITIOHINS SF Air conditioner IAC) supply ducts aust isolate and Eaergency Pressurization Systea tControl Roon Energ. Vent. Systea) aust supply pressurized filtered air to

                                                             . nein control roon on high dryaell pressure signal froa prieary containaent systea tb4).

Pers Oyerexp 3b Pers Overex'p b4 PRI COHTAINNEHT SF Send high dryaell pressure isolation signal tn air conditioning systea t31).

                                                                                             ~1 SAFE SHUTDNN ANALYSlS                                   PageSof Standard SSA Sequences Tab]e 7.h-22                   Prepared                               ~Fr~/Y7 3'ates NI~~          (Rev 1)    . Checkedy                        Date:         +g.y/pg UNACCEPTABLE                     TVA                    SAFE RESUi.TS   SAFETY ACTION       SYS      SYSTEll T1TLE     Code   Tit le       ND.       NANE   'UD   FUNC E              CDNNENTS aa HPCl 1SOLATlDN DN LOQ REACTDR PRESSURE Rad Release  20     RPV  Isol     73 HPCl,                      Lou stean supply pressure    signal closes steaa supply line isolation valves.,                I gk

~g

  \

mV

                                                                  ~

l~~ g3 $ I'!fl,\ t

                                                        ~
                                                                                                     ~ ~

P SAFE SNUTDONN ANALYSIS Standard SSA Sequences Page&of 50'P Table 7.b-23 Prepared Nc/~Datet ~-~r8/87 DFMSB3-048 Nev 1) Checked: Datea Qi.y/~ UNACCEPTABLE

                                'VA                      SAFE RESULTS  'AFETY ACTION          SYS     SYSTEN        FUNC.

TITLE Code Title NO. NANE CODE ee RCIC ISOLATION ON LOM REACTOR PRESSURE Rad Release 20 RPV Isol 71 RCIC Low stean supply pressure signal closes stean supply line isolation valves.

SAFE SHUTDUHX AHALYSlS Page70of 3O 7 Standard 59l Sequences Table 7.b-24 Prepared@ 4 ates 2j'l8/&7 SRHS~S tRav 1) Checheda Sate! +~y//'7 UHACCEPTASLE TVA SAFE RESULTS SAFETY AC11DH SYS SYSTEN. FUXC T)TLE Code Title NU. NANE C(U)E 0 ea CLOSE NS1V OH NAlk STEAN LlkE L()H PRESSURE ('RUH') lsol ' Rad'Release 20 RPV NA1N STEAN SF Provide lou pressure signal (in aain steaa line at turbine) to priaary containaent

                                                                 'ystea (Q) failwafe logic. (RUH node)

Rad Release 20 RPV lsol .1 NAlN STEAN SF Close NSlVs and aain stean drain lines on low pressure signal (in aain stean line at turbine) froa priaary containaent systea

                                                                  '(H).     (RUN aode)

Rad Release 20 . RPV lsol b( PR1 CUHTAlHNENT SF Send nein steaa systea'(01) )ow pressure signal (in aain stean line at turbine) to initiate NSIV (01) closure. (RUH node) Rad Release 20 RPV lsol 99 RPS Provide RUN Node signal to systea At for lou steaa line pressure isolation

                                                                'nterloct.

i SAFE SHUTDOMH AHALYS1S Page7) of 90'7 Standard SSA Sequences Table 7.b-25 Prepared'. ~446)ate) Z/ro/S ) BFH-OSS3<hB (Rev 1) Cbeckedr Date> ~>y/Pp UNACCEPTABLE TVA SAFE'lIC RESULTS SAFETY ACTJOH SYS SYSTEN T1TLE Code Title HO. HANE CODE CONNEHTS ta CLOSE TURB STOP ! BYPASS VLVS OH COHO, LOU VACUUN Rad Release 20 RPV iso) 1 NAIH STEAN Close nain turbine stop valves on turbine control systen lh7) condenser )ox vacuue signal. Rad Release 20 RPV Isol 1 NAlH STEAN Close nain turbine bypass valves on turbine control systea th7) condenser )ou vacuus signal. Rad Release 20 RPV Tsol 47 TURBIHE COHTROL Provide condenser lou vacuun signal to close aain stean systea (01) turbine stop valves'rovide Rad Release 20 PPV Iso) h7 TURB)HE COHTROL condenser lou vacuus signal to

                                                                    ~

close aain stean systea (0)) turbine bypass valves. , ~

SAFE S)R)TMNN ANALYSIS Page7Zof '3O'7 Sta'ndard SSA Sequences Table 7,6-26 PreparedM~ W~E~ate) >z'~~/ 8 ) BF)H563WB (Rev I) Checked) Date: +g.p/pp UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN 'UNC TITLE Code Title NO. NANE CmjE CNNENTS ae RECIRC PNP'RIP ON FCV CLOSURE, SEN TRIP, PIR>30X , Fuel Failure 24 Pow Reduce I RAIN STEAN SF Provide >30X turbine first stage pressure interlock signal to reactor protection systen (99) faiIwafe logic. Fuel Failure 2l . Po>> Reduce 47 TURMNE CONTROL SF Provide nein turbine control valve fast closuri signal to reactor protection systen tgg). Fuel Failure 2A Po>> Reduce 6B RECIRCULATION L Open recirculation puap aotor'reakers on RPS (99) signal due to nein turbine control valve fast closure and > 30X turbine first stage pressure. Coastdo>>n aust be faster than assuaed an reload analysis to liait severity of the event. Fuel Failure 24 Po>> Reduce 99 REACTOR PAOTECTN SF Send signal to open recirculation puap no'tor breakers Isystea 6B) on nein turbine control valve fast closure signal froa turbine control systea (47) and oain stean systeo l01) signal indicating > 30X turbine first stage. pressure.

SAFE SHUTDOMN ANALYSIS Pageof 90> Standard SSA Sequences Table 7.6-27 ~ ~Mate: 2gle/87 (Rev I) Prepared'FH-OS63-OiB Checked) Bate> gee/i'p UNACCEPTABLE TVA , SAFE RESULTS SAFETY ACTION SYS SYSTEH FUNC TITI,E Code Title 'HO. HANE COBE aa RECIRC PUlP 'TRIP ON HIGH REACTOR PRESSURE Fuel Failure 24, Po>> Reduce 3 FEEDHATER Provide high reactor pressure signal to open recirculation H/6 drive sotor breaters (200'lectrical systea) for trip of recirculation. puaps. Fuel Failure 24 Po>> Reduc~ bB RECIRCULATION . Recirculation H/8'rive aotor 'breaters

                                                                           )200'lectrical systes) tripped on. high reactor pressure signal.

Fuel Failure 24 Po>> Redure gg REACTOR PROTECTH Provide RPS ATU po>>er to feed>>ater systea (03) high reactor pressure instruaentation for trip of recirculation puaps. Fuel Failure 24'o>> Reduce 200 ELECTRICAL Op'n recirculation H/B drive aotor breaters <200> electrical systea) on feed>>ster systea I03) high reactor pressure signal.

SAFE S)ITDNN A)UILYS1S Page74of Standard SSA Sequences Table 7.~ Prepared> ~/<Bi8 7 3'ate> IF)H)SS~S (Rev 1) . Checkedy Dater ~i/pp UNACCEPMLE TVA SAFE RESULTS SAFETY ACTlON SYS SYSTEN F)NC TlTLE Code Title N. WNE CODE aa REC1RC PUlP TR)P N LO'N NTER LEVEL tL2) Fuel Failure 24 Po>> Reduce 3 FEENATER Provide low water )eve) (L2) signal to open recirculation II/O drive ootor breakers lsysteo 200). ~ Fuel Failure 24 Po>> Reduce b8 REClRCUL'AT1% Recirculation ))/6 drive aotor breaters (200'lectrical systeo) tripped on low water level (L2). Fuel Failure 24 Po>> Reduce 99.REACTOR PROTECT Provide RPS ATU power to feedwater systea 103) low water level tL2) instruaentation for trip of ricirculation puaps. Fuel Failure 2t Pow Reduce 200 ELECTRICAL Open recirculation N/6 drive aotor breaters 1200) on feedwater systea <03) low water level {L2) signal.

                                                                                                           ~    ~

SAFE SBUTDOMM ANALYSIS Page75of QO'P Standard SSA Sequences Table 9.6-29 Preparedfgf +4 Date: > /re/6'1 BRHSS3-04S IRev I) Checked: Date: ~N/eE URACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN 'UNC TITLE Code Title NO. RANE CODE CONNEHTS aa RECIRC PNP TRIP, STOP VLV CLOSE,TRB TRIP,PMR>30X Fuel Failure 24 Pou Reduce 1 NAIN STEAN SF Provide sain turbine stop valves < 90X 1 open trip signal to reactor protection g4 systea l99), ~J Fuel Failure 24 Po>> Reduce 1 NAIN STEAN'F Provide '> 30X turbine first staoe pressure

                                ~                                 interlock signa) to reactor protection systea f99) fail-safe logic.                    8)

Fuel Failure 24 Pou Reduce 6B RECIRCULATION L Open recirculation puap aotor breakers on RPS (99) signal due.to aain turbine stop valves < 90X open and >30X turbine first stage pressure. Coastdoun aust be faster

                                                                . than assuaed in reload analysis to lisit severity of the event, Fuel Fai)ure ?4      Pox Reduce       99 REACTOR PROTECTH SF      Send signal td open recirculation puap
                                                                  ~ otor breakers Isystes 68) on signals fros aain stean systea IOI) indicating aain turbine stop valves < 90X open and >30X
                                                             ~

turbine first stage pressure. )p)

SAFE SI)TDOMN IILYSlS Page7bof 9O 7

                               'tandard    SSA Sequences Tab)e '7.6-30                Prepared   &               <<~Batea u/'I>Ild'7 BFJHS63WB (Rev      l)       Checkedy                       Datei ~y~

INCCEP TABLE TVA SAFE RESULTS SAFETY ACTlON SYS SYS TEN FNC TlTLE Code Title NO. RANE CODE '. aa HPCl STEAN SUPPLY VALVE CLOSURE OH HML (LB) Syst Stress 25 Level Reduce 3 FEEDQATER Provide high reactor >>ster level signal MLS) to HPCI systea t'l3); Syst Stress 25 Level Reduce 7l RClC Provide ECCS ATU po>>er to FM systea (03) hi,gh >>ater level [LB) HPCl trip instruoentation. Syst Stress 25 Level Reduce 73 HPCl Shutoff HPCl (if operating) on FM systea (03) high reactor >>ater level )LB) signal.

e SAFE SHUTDO)OI AHALYSIS Standard SSA Sequences Page77of $ 07

                 'VA                 Table 7.6-31                Prepared>$ 1.         $~ . r ~Date:  - r'~>IE' DFH-OS63-(AS IRev  l)       Checked>                        Date: ~~/Xr UHACCEPTABLE                                              SAFE RESULTS     SAFETY ACTIOH        SYS      SYSTEN       FUXC TITLE    'ode     Title         XO.       HANE        CODE              CONNEHTS aa RCIC S'IEAN SUPPLY VALVE CLOSURE.OH INL <LD)

Syst Stress '25 Level Reduce 3 FEEDHATER Provide high reactor water level signal (LB) to RCIC systea I7)). Syst Stress 25 Level Reduce 71 RCIC Provide ECCS ATU power to FQ systea (03) high water level (LS) RCIC trip instruaentation, Syst Stress 25 Level Reduce 7l RCIC Shutoff RCIC )if operating) on FH systea (03) high reactor water level ILB) signal.

SAFE SHUTDDN ANALYSlS Page?8of Standard SSA Sequences 2/Cl/&g 30'Motel Table 7.6-32 PreparedMA

                                   . SF)H)S$ 34$ D IRev 1)                             'heckedy bate! Q~y/~

UHACCEPTADLE TVA SAFE RESULTS SAFETY ACTION 'YS SYSTEN RNC T1TLE Code Title . ND. NANE CODE CDIOIE)ITS aa lIAlk TURBTME TRlP ON )D)L ILS) Syst Stress 25 Level Reduce 1 HAlN STEA)I Close naia turbine stop valves on high

                                 \                                       water level ILS) signal froa aain turbine
                                                                      . control'systeo (4?).

Syst Stress 25 Level Reduce ~ 'rovide high reactor water 'level signal ILB) to initiate oain turbine trip through feedwater control systea 146). Syst Stress 25 Level Reduce a!6 FEEDMATER Send high water level (LB) signal froo .

                                           . CONTROL                     feedwater systea 103) to aain turbine control systeo 147).

Syst Stress 25 Level Reduce 47 TURBlNE CONTROL On high water level (LS) signal froe FU control systea 146)> energize oain turbine trip solenoids which close the wain stean systeo (01) turbine inlet valves.

SAFE SHUTDOMH ANALYSIS Page77of 3G"7 Standard SSA Sequences Table 7.b-33 Prepared@ g ahu~Date: ~/fb'l&'7 BFN<S63-0kB .(Rev 1) Checked: Date: +~/f'7 UHACCEPTABLE ~ TVA SAFE RESULTS SAFETY ACTION SYS SYSTEH FUHC TITLE Code Title NO. NAHE CODE CO)O(EHTS ea FEEDMATER PUNP TRIP OH HIGH MATER LEVEL (LS) Syst Stress 25 Level Reduce 1 HAIN STEAN Close FM turbine 'inlet valves on high water level (LS) signal fros FM control systee (46).

                  "                                                Provide high reactor water level signal Syst Stress 25      l.evel Reduce     3 FEEDMATER (LS) to initiate RFPT trip through feedwater control systee (%b),

Syst Stress 25 Level Reduce Eb FEEDMATER On high water level (L8) signal froa FM , CONTROL systes (03)> energiae RFPT soienoids which r .

                                                            ~

close the wain stean systea (01) feedwater turbine inlet valves.

                                                    ~g     (A Ir"Q p,fj         g p ., !   ~r.  ~
                                                              ~      ~

SAFE SHUTDOO ANALYSIS PageSbf QC'7. Standard SSA Sequences Table 7.6-34 Prepared~. LL~644atet +/ re'~&'7 BF)HSB3-Ol8 (Rev I) Checkeds Dat ~:Qzp/pp UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN FUNC TITLE Code Title NO. NAKE CODE COtNENTS ae NANUAL SUPPRESSION POOL COOLINS ON NISH TENP. Cont Stress 30 Cont Cooling '?3 RHRSN ~ NSF Support RHR systea IN) suppression pool cooling aode. Cont Stress 30 Cont Cooling 64 PRI CONTAINNENT Provide suppression pool teaperature and

                                                              )eve) indication to support RHR systea I74) suppression pool cooling aodev      '      C, .

Cont Stress 30 Cont Cooling 67 EECM Support RHR systea <74) suppression poo) 4qy(Fg cooling aode, Cont Stress 30 Cont Cooling 74 RHR Provide suppression pool cooling function.

                                                                                               ~ ) '

SAFE SHUTDOMH AHALYSIS Standard SSA Sequences Page el of ~7 Table 7. 6- 35 Prepar ed rM. c~ Date: ~/f/~/cia BFH OSG3 04S (Rev ll Checked a ->~<~'pS/t1 UNACCEPTABLE TVA SAFE RESULTS SAFETY ACTION SYS SYSTEN FUNC TITLE Code Title HO. MANE CODE ~ CONNEHTS aa SUPPORT SECOHDAR'Y COHTAIHNEHT Rad*Release Rad Release 27 27 'stEst Sec Cont Sec Cont I 2 NAIH STEAN COHDEHSATE Support Support secondary secondary containaent containaent function. function.

                                                                                                                          ~function.

27'stEst containaent Rad Release 27 Sec Cont ~ 3.FEEDMATER Support secondary Rad Release Sec Cont 12 AUX BOILER Support secondary containaent function.

 'ad     Release  27     Esf Sec Cont     23   RHRSW                      Support   secondary     containaent              function.

Rad Release 27. Est Sec ConC 24 RAM COOLING SupporC secondary cont'ainaent function, MATER Rad Release 27 Est Sec Cont 25 RAM SERVICE Support secondary containient function. MATER Rad Release 27 Est Sec Cont 26 HI PR FIRE Support secondary containaent function. PROTECT Rad Release 27 Est Sec Cont 29 PDTABLE MATER Support secondary containaent function. Rad Release 27 Est Sec Cont 31 AIR COHDITIOHIHB Support secondary containaent function, Rad Release 27 Est Sec Cont 32 CONTROL AIR Support secondary containaent function. Rad Release 27 Est Sec. Cont 33 SERVICE AIR Support secondary containaent function, Rad Release 27 Est Sec Cont 34 VACUUN PRINIHG Support secondary conCainaent function. Rad Release 27 Est Sec Cont 37 GLAND SEAL MATER Support secondary containaent function, Rad Release 27 Est Sec Cont 40 STATION DRAINAGE Support secondary containaent function. Rad Release 27 Est Sec Cont 44'BLDG HEATIHG Support'econdary containaent function. Rad Release 27 Est Sec Cont 53 DENIH BACKMASH Support secondary containaent function, AIR Rad Release 27 Est Sec Cont 64 PRI COHTAIHNEHT Support secondary containaent function. Rad Release 27 Est Sec.Cont 65 SiTS Support seconda'ry containaent function, Rad Release 27 Est Sec Cont 67 EECM Support "secondary containaent funrtion Rad Release 27 Est Sec Cont 69 RMCU Support secondary containaent function. Rad Release 27 Est Sec Cont 70 RBCCM Support. secondary containaent'upport

                                                                                                                                      'uncti'on.

Rad Release 27 EsC Sec Cont ~ 71 RCIC secondary containaent function. Rad Release 27 Est Sec Cont 73 HPCI Support secondary containaent function.. Rad Release 27 Est Sec Cont 74 RMR Support secondary containnent function.. Rad Release 27 Est Sec Cont 75 CORE SPRAY Support secondary containaent function, Rad Release 27 Est Sec Cont 76 COHT IHERTIHB Support secondary containaent function Rad Release 27 EsC Sec Cont 77 RADMASTE Support secondary containaent funcCion. Rad Release 27 Est Sec Cont 78 FPC I CU Support secondary containaent function, Rad Release 27 Est Sec Cont 79 FUEL HAHDL L Support secondary containaent function. STOR Rad Release 27 Est'ec Cont B4 CAD Support secondar containaent fun t gg ~c Rad Release 27

                 >I Est Sec Cont
                        ~5h:     Ce  0                                   S~ uppor secondary containaent funchon.
                                                                                   ><~~ ~~~~4~~~o 6 ltr P)ph,< f
                                                                                               '~l p',[:I f~><f, I j', i)'<'~> !t i!,I .ii i"; <>>
                                                                                                                                 'i v 0

Page 82 of 307 Prepared: Date. ~QQ Checked: Date: / d'r BFNWSG3-048 (Rev 1) C. M; Parker RRRWSWW%%%%%RRQRNRNRWRR%%WWRRWN%%%%%%%%QRQHRRW%%R~'RWRZf+%%<%<QQW~WWRRWCNW 8~0 ANALYSIS

  ~ '.1       Summa     Results of'Anal sis The   result of this event analysis is'an explicit list of safety functions that are to be used as pert of the BFNP Baseline effort to reconfirm that the systems required for safe shutdown are indeed in place and adequately checked so that'nit 2 startup 'can proceed   in  accordance   with the   TVA  Nuclear Performance Plan,.

Volume 3 - BFNP (reference 4.lc). In many systems,'nly a small portion of the system plays any safe shutdown role, while in others, virtually 'all of the system is considered as a safety functi'on. This report .is to provide identification of the "primary".system functions those which sense. parameters and initiate actions.

         . A   key step   still  remains beyond this analysis..Xt involves area of "support" functions needed by a system to          the'mportant perform the identified safety action(s). A few of these items are obvious .(e.g., c'ooling water for a heat exchanger) and have been included whenever possible, especially           if'he   suppor't is fram a less likely source. .However, miny are, much'ore subtle (e.g., dc power for h control function', or ventilation
           'equirements) and this report must not be'nterpreted as complete-for 'these..support functions. This important,ext-level effort demands intimate'ndividual-system knowledge.'t must be performed by each responsible system engineer during the remainder of the Baseline effort. The information in this
        ..document about each "primary" 'system safety action provides' key input for this effort.'he complete list of shutdown, system
            'safety modes/actions required for all. of the systdas involved in safe shutdown, including .those which primarily perform support roles, will be documented in the System Requirements Calculations, subsequent to this. event-oriented document.

This document p esents its resu "d'rections"t events and syst a endix for each individua presents in de t a jf

            'functions requ r t                                                       i.'n t e pq@,                 ggnPP+j a e

cceau.tments. t)yfPen s

                                                                    'e
             'together to .pr vide several prot c automatic; s         are identified to be manual. A Pew p?4) preventive or l~xtx ~le-.W~QE disturbance. Section 7.6 includes docum
                                                    'the     course of the "standard" sequences which are utilised repe'atedly. for sa shutdown in many events.         In general, most transient events are

i

                                                                                                                                                             'Page 83                of       307 Prepared:                                                                       Date:
                                                                                                       'I c     rt Checked!                                                                       Date:

BFNWSG3-048 (Rev 1) C. M. Parker RR $ $ m $$ $$ $$ $$ $$ $$ $$ $ $ $$ $$ $$ $$ $$ $$ $ $ $$ $ $ $$ $$ %% $$ $$ $$ %a $$ $$ $$ $$ $$ $$ $$ $$ $$ $$ $$ $ $ $$ N% $$ $$ $$ $$ $$ %% $ $ ÃR$$ $ $ WQ $$ $$ $$ $$ $$ $$ $$ $$ $$ $$ $$ $$ $ $ $$ $ $ $ $ j taken to -stable, hot shutdown conditions with the option to restore normal power conditions or continue shutting down to cold conditions. This is co'nsistent with the extent of analysis provided in the TSAR (reference 4.la) and subsequent licensing submittals. The only event carried to depressurized (cold) shutdown is Event 28, shutdown from outside the control room,.

                         .plus those events (e.g., LOCA) that inherently depressurize the reactor.

A fully safety-equipment-on1y path to shutdown is also provided to show features in this option that must be checked in the Baseline effort event. Unique, shutdown functions are included, if needed, in each event appendix. Table 8.1-1 gives a summary of key trip functions expected for each event (by operating state).. The table is collated primarily from the event appendices to piovide a more comprehensive tabulation of the situations which utilize various protection actions. It shows several key setpoints (e.g. low water levels)

                       ~   that are used to initiate various reactor isolation functions.

It includes some Level 3 actions that may not have been highlighted in the appendices, but which are expected because of scram (and subsequent level drop) or the fact that a lower level

                      . setpoint is expected to be reached.

In section 8.4, the re'suits h'ave been sorted and presented from the other "direction," the system perspective. The actions. performed by each system are tabulated together showing the events in which that system performs a safety-related function, and in many ca'ses, the different safety functions that may be performed by the same system. By utilizing the safety function code (which defines if, a function must meet the single failure

                         ~

criteria, utilizes the function (whether if it is manual; etc.) plus knowing each event that it is an abnormal transient or a Design Basis Accident)t the system evaluation can proceed with comprehensive knowledge'f 'the ways the system is to be used in event mitigation and shutdown.

                                                                                  ~ ~ 'I

SAPK SHUXDOQR h%bXZSZS Prepared:

                                                         ~
                                                              ~

ZX 6

                                                                     ~             Page 84 of 307
                                                                                 'ote:    '~Z E    C. c er Checkeds                                    ~

Dates BFNWSG3-048 "(Rev 1) C M. Parker RRORRRW%NQSRQRNRRRRRRNRaa Table 8.1.1 Suasna of Ke Tri Actions e (Entry shoes applicable reactor operating state) Low Low , Low Hi,gh High High . Est Est Event... Lvl L3 Lvl L2 Lvl

                                '8   Lvl
                                             .'res RPV'l
                                                                 ~

Dry 'oat Pres Sec Stnby Parer Scrsin 01A, P 01B P P 01C D)F D)F 02 F P. ~ 03 CDEF CDEP 'DEF 'D;F '5 04 F CDEF CDEF P D)F 06 P P 07; D 08 P 'D, P, ~ F 09 D D 10 11 12A CDEF CDEP CDEF D,F 12B CDEF CDEF ~ D)F 13 CDEF CDEF CIEF D,P 14 CDEP CDEP D,P 15 C DER CDEP CDEF, CDEP CDEP D)P 16 a 17 18

 '19                                                                                           ')P
                                            '.'        ~
                                                            .(pp Q
                                                                              /

e e

                                  ~   ~

i Page 85 of 307 Prepared: Qc E C~ c Date: ~/ Q7 Checked: Date: BFNWSG3-048 '(Rev I) C. M. Parker aaameaea>>eeer>>aasreeaaaaaa>>aaaaaaaarra>>oaaaaaaaeaere>>>>EE>>reeerrre+ereeree->>EE>>>>eE P Table 8.1-1 Summa of Ke Tri Actions (Continued) (Entry shows applicable reactor operating state) e 4 Low Low Low High High High Est Est Lvl .Lvl Lvl Lvl RPV 'rywPres'Cont Sec. Stnby Scram Eeeee L3 L2 Ll L8 Pres Power 20 P P 21 D,F D,F 22 PJHr D 23 CDEF CDEF 'DEP CDEF D,F 24 D~F DpP ~ D,P .DtP D,P D~F 'D,F D,F 25A 25B CDEF CDEP CDEF CDEF CDEF CDEF CDEP CDE~ CDEF CDEP CDEF CDEF CDEF CDEF ',F 25C CDEF CDEF CDEF CDEF CDE~ CDEP CDEP CDEF D,F 25D CDEP CDEF CDEP CDEP~ CDEF CDEF CDEF D,F 26, A .A 27 CDEF CDEF CDEP CDEP 'DEP CDEF P,F 28 ALL, CDEP , CDEF 'DEF ALL D,F 29 D Pk 30 '. D,F 31 32A 32B e CDEP . CDEP D,F 32C I ~ 32D 32E CDEF CDEF CDEP ALL D,F 33

    *-   Shutdown with standby liquid            control.
       - SRVs used in depressuriration             role..

e fT,(f r kgb ~ Q aPE'e ~. e e

0 SllK SKIZDQK kÃ&ERSXS Prepared: V. B ac et

                                                                                   ~

Jr. Page Datet SP of 2-5-.87 30$ &IJ/0'1 Checked: Date: 5 fF BPNWSG3&48 (Rev 1) C. M. Parker QRWRNUWRÃtnN'SRRWSRQ1$ Qnenene 8,2 Seconda Containment and Safe Shutdown 8;2 I ~zpoee

                    . This section describes the secondary contaiament fimction during safe shutdown of the pla~..            It is provided ae 'a special             '

section of the report since'its function. is'omplex azd is - actually performed by a large number of systems. I 8 2.2 Descri tion of Seconda Contaimnent The total secondary contaiament coniiste of the entire reactor

                  ~
                      .building.      The secondary contaimnent ie subdivided 'into. four cones; the Unit 1 Reactor Zone (U1RZ), the'Unit 2 Reactor Zone (U2RZ), the Unit 3 Reactor Zone (U3RZ), axd the refueling:

floor walls, ceiling axd floors, piping and electrical electrical penetrations, airlock door seals, hatch seals, . ventilation system isolation logic; axd process axd ventilation

                ~        components in various pla~ systems which act together to limit air inf low into the secondary contaiament (references 4'la asd

4. le) .

The Secon'dary Contaiament 'System acts to assure that the . Standby Gas Treatment System gSGZS, 65) can maintain isolated portions'f secondary contaiament at a negative aressure with respect to the outside environment. 'he intent is that any leakage be directed iato the contaiament where the atmosphere can be 'filtered before being released to thc enviroanent. The automatic component e which:accomplish this" Smction for r'eactor building a% refueliag f loor ventilation are located in System 64 (Primary Containment System) ~ In reactor operating states C,'D> E> axe P.(ae defined in-section 7.5), the Secondary Coataiament system provides secondary containment when the primerv contaiament 'is intact.. The secondary contaiament system provides primary contaiament when the primary contaiament is open> such as during 'rcfueliag aad maiatenaace operations'n reactor operating state A. 'n . the event of rcleasc of radioactivity to;the reactor build atmosphere, the .secondary containment s es to contain the ra dioac t re controlled f teredil '"

                         'elevated'release o the eeoo5eiety oontei5ment item atmosphere (refcren f~     'I radiation dose rate f'53 o

i ~ 5 e

                                               ~ I 5
~ I.'

7g 848K SHOd%a hMLMIS Prepared: V, 0 ~ B Page Jr. Date'. ~G ~ 8Ll of 30K'<</~ Checked: Date BFNWSG3&48 (Rev 1) C. M. Parker RSRDJRNSSRRW%R%%R2SDRR%RQ%RRR%aaaeaaRaaSES%Q%QNRNII%QW%%RzCQRStQCSRRCSCs%CEQtRQRRQztRz R%CR Secondary containment is maintained at a negative pressure relative to the building exterior during zarmal operation and isolated. Under normal conditions, this negative pressure 'hen, is maintained'by controlling appropriate reactor building supply and exhaust ventilation system components.

8. 2.3 Seconda Containment for Safe Shutdown 8.2.3. 1 Secondary Containment Isolation. Ventilation Systems The reactor .building ventilation system consists of high flow volume supply axd exhaust fans aA large diameter duct work (18-inch diameter is typical). During zarmal operation, the exhaust fans remove air fran within the reactor building aR f loor ard discharge it to the outaide enviroanent.

If

                                                                                                  'efueling events occur within the plant such that radioactive material is released into 'the reac'tor building or refueling floor atmosphere, this high volume, unfiltered flow path to the outside environment is rapidly shut off upon event detection. To accomplish this, the ventilation supplv axd exhaust fans are tripped off ard isolation damoers on the ventilation lines are closed. The Standby Gas Treatment
                 'System (SOS, System 65) is also started in order to maintain the necessary reactor building negative pressure. Under isolated conditions with the SGl.'S running, reactor building secondary containment atmosphere is 'filtered by SGTS and released with dilution air flow through the plant stack.

Isolation of the ventilation systems is necessary to prevent outside air fran entering secondary contiinment at a rate where the SGZS would be unable to maintain the required building negative pressure. A building internal negative pressure assures that fission products in the reactor building atmosphere will undergo a controlled, elevated> filtered release via the SGt,'8 (65) flow path. The following trip signals isolate portions of the ventilation systems (reference 4.1g): Trip A: Reactor Low Mater Level (level 3) Trip B: Dr i ent) Hi h Pressure Trip C: Res tor Buildi'ng (unit specific enti a ion Ple diation U scale or Downscale Trip D: Ref e i A$ For safe shu refueling f or as well as the reactor biil the trip occ rred> aced trip D on any unit's portion of the

ShlK SKULD%% hl%KMXS Page 88of 30!PE<,g~ Prepared: nate: DWWr Checked: V.. 1 nchet e Date: BFN-OSG3-048 (Rev 1) C. M. Parker aaaaamscsa3IaaesnaeeaeaaeeaaINaaaeeeaasaeaaaaaecsaeaaaaaeSaaaeaaaescsaaIIeasSaaaaaaaaa refueli'ng floor isolates the entire refueling floor (reference 4.lh). Any of the trips A through D above will start the SGTS (reference 4.1g). Secondary containment isolation of the ventilation system is capable of being initiated manually or by loss of electrical power. System 64 dampers provide secondary containment isolation within ten seconds of the isolation initiation signal in a given xone (references 4.1i and 4.1i). The secondarv containment barrier limits infiltration of air such that the SGZS can maintain a negative pressure of 0.25 inch of water within the containment to support safe shutdown (reference 4.la). 8.2.3.2 Seconda Containment Isolation: Other Systems Process lines carrying liquids or gasses penetrate the secondary containment boundaries as part of their system function..These lines are not open to the secondary containment atmosphere, so that rapid isolation of these lines is not necessary upon any of the trips listed in

                ~

8.2.3.1. Therefore, manual isolation valves, check valves, and loop seals are used to limit air entering secondary containment through these lines in the event they are broken. Table 8.2-1 provides a list of systems which contain lines which penetrate secondary containment. Other system components may also be found in subsequent detailed evaluation of all secondary containment actions during the remainder of the Baseline Program. These other system components will be addressed as necessary in the system evaluation repor t.

SAFE %ITD%% MhLYSIS Page 89 of 307 Prepared: Date: g~P. Secondary Contaiamaxt V~ G B c t r. Checked: Date: ~V/5 t'> BFNWSG3-048 (Rev 1) C. M. Parker saasaatcaasaasaaccaaccssactaccaccsassassaasaasaaaaaccaaaaaaasaaaaaaaaaaaacraaaactaaca TABLE 8.2-1 S stems That Su ort Seconda Containment 1 Main Steam 2 Condensate and Demineralization 3 Feedwater 12 Auxiliary Boiler 23 RHR Service Water 24 Raw Cooling Water 25 Raw Service Water 26 High Pressure Fire Protection 29 Potable Water 31 Air Conditioning 32 Control Air 33 Service Air 34 Vacuum Priming 37 Gland Seal Water 40- Station Drainage 44 Building Heating 53 Demineralizer Backwash Air 64 Primary Containment 65 Standby Gas Treatment 67 EECW 69 Reactor Water Cleanup 70 RBCCW+ 71 RCXC 73 HPCI 74 RHR 75 Core Spray 76 Containment Inerting 77 Radwaste 78 Fuel Pool j",peal,ing and 79 Fuel Hanghggnd Stor ge 84 Containment Atmospher Dilution 85 Control Rod Drive

    ~;q;,MD,AOr~

p.Qw 9'1

SAFE SHOTDORR hMLYSIS Page 90 of 307 Prepared: Date: C. c r Checked: Date: BFN-OSG3-048 (Rev 1) C. M. Parker

                                                                   ~Mrl1l'ssoaaassaamscaasssaaaaaaasssaaasassaaassessaseaassaaassaaasssaassssaassssassssssasaasessas 8.3 Reactor   Stabilit Evalution During each reload evaluation (Section VII of the current submittal, reference 4.1.b) the thermal hydraulic stability margin of the core design is rechecked. This process primarily involves analytical evaluation and reestablishment of the approved operating range for the unit during the upcoming cycle.

The most limiting power/core flow condition is examined and stability margin is calculated. All appropriate fuel types are considered. No protective safety actions are currently required (other than operating within the prescribed constraints) since acceptable stability margins are shown (reference 4.1.b). If ever needed, protective features are associated with power monitoring (Neutron Monitor System 92), manual insertion of control rods if needed (CRD System 85), increase of recirculation flow (Recirculation Control and Pumps, Systems 96 and 68), and, if necessary, manual scram (using the RPS System 99) or automatic scram from high neutron flux (Systems 92, 99, and 85). These safety action paths are not included for stability protection in the SSA since the current core analysis demonstrates that they are only backup features, not required for any of the three units from their initial startup to this time.

SAFE SHUTIX%5 kRhLYSIS Page 91 of 307 Prepared: E Co Date: ~/9~ Checked: Date: gg BFHWSG3-048 (Rev 1) C. M. Parker aacasaasaasaaaaaaaaasssesssssasagsaaasaaaaaassaaeaaaesaaaasasaaasssssscaaa=aaaasaa 8.4 Re uired S stem Safet Actions The appendices of the report describe each transient, accident, and special event postulated in the FSAR, questions and answers to the original FSAR, and/or in the current reload licensing submittal for Unit 2, cycle 6. Section 2.2 listed all these events. The appendices list all safety functions, including sensors, logic, and actuator functions (whether automatic or manual) for which credit is taken in some form in the safey analysis documentation. Also included are any plant features which are related to analytical asumptions that limit the magnitude or rate of the disturbance caused by the event. In some transient events, actions needed to restore the plant to normal conditions are also listed. In the section, these safety-related entries are sorted and listed by system number in order to provide a more comprehensive list of requirements for use during the Safety Evaluation Report (SYSTER) preparation for each system. The SYSTER evaluation must consider the different event conditions

            'associated with each required function. A documented path of information is provided to search out the event description more fully as   needed.

The following list provided shows all systems which have been identified in some way in the Safe Shutdown Analysis. They are listed in numerical order of the system number.

Page 92 of 307 Prepared: Er. Date'. 2~1 7 Checked: E.. ke Date: r BPNWSG3"048 (Rev 1) C. M. Parker S stems Identified in the Safe Shutdown Anal sis Main Steam (01) Recirculation (68) Condensate (02) Reactor Water Cleanup (69) Peedwster (03) Reactor Building Closed Cooling Extraction Steam (05)* IRi Water (70)

• Boiler Drains and Vents (10) Reactor Core Isolation Cooling (71)

Auxiliary Boiler (12)* High Pressure Coolant Fuel Oil System (18) P.l Injection (73) RHR Service Water (23) Residual Heat Removal (74) Raw Cooling Water (24)* Core Spray (75) Raw Service Water (25)* Containment Inerting (76) High Pressure Pire Protection (26)* Q Radwaste (77) Conde'nser Circulation Water (27)* Fuel Pool Cooling (78)* Portable Water (29)* le Fuel Handling and Storage (79)* Ventilation (30)* Diesel Generator (82) Air Conditioning (31) Containment Atmospheric Control Air (32) Dilution (84) Service Air (33)* Control Rod Drive (85) Vacuum Priming (34)* Diesel Starting Air (86) Gland Seal Water (37)* Radiation Monitoring (90) Station Drainage '(40)* Neutron Monitoring (92) Sampling end Water Quality (43) Traversing Incore Probe (94) Building Heating (44)* Recirculati'on Plow Control (96)* Feedwater Control {46)* Turbine Control (47) I Reactor Protection (99) Cranes (Ill)* Deminerelizer Backwash Air (53)* )g~ Electrical (200) Standby Liquid Control (63)* Ccmmunication {244)* Primary Containment (64) Structure (303) Standby Gas Treatment (65) Off-Gas (66)* Emergency Equipment Cooling Water (67)

*System actions are not extensive.

gy Q~ fp y~fygy all e,MreA $$ $ ~~

SAFE SHJTOON) ANALYS)S Page 93 of 2M' Prepared! tel NlN STEAN SYSTE)f Checkedl E k L Parker iI2g ~+2P ~ BF)H)SS3~ (Rev 1) C. EUENT REACT iNAtXEPTABLE SAFE CR CPER RERLTS SAFETY ACTTN RK SEQ. EUENT TlTLE STATE Code Title Code Title CBBE 01A Seneratcr Trip F 2-2 Fuel Failure 17 Scran SF Provide >301 turbine first stage pressure interlcck signal to reacts protecticn systen N) fai]~fe logice 01B T/B Trip& Fails F, 2-2 Fuel Failure 17 Scraa SF Provide >301 turbine first stage pressure interlock signal to react+ protectim systen 199) fail~e logice 02,. Turbine Trip F 2-2 Fuel Failure 17 Scran SF Provide > 301 turbine first stage pressure interlock signal to reactcr protection systea N) failmfe logics 02 Turbine Trip F 2-2 Fuel Failm 17 Scrae SF Provide nain turbine stop valve < 901 open trip signal to reactcr protectim systen N). 03 lhin Stn lsolatim F 2-2 Fuel Failure 17 Scran SF Provide )SIU < 901 open trip signal to reactcr protectim systen N) fai&afe logic. 05 Loss of Vacuue F 2-2 Fuel Failure 17 Scran SF Provide > 301 turbine first stage pressure interlock signal to reactcr protectim systea N) failmfe logic. F 2-2 Fuel Failure 17 Scran SF Provide nain t~bine stap valve < 901 open trip signal to reactcr protectim systen N). 12A Pres Reg Fail~ BF 2-2 Fuel Failure 0 Scran SF Provide IS1U < 901 open trip signal to reactcr protectim systea N) failmfe logic. 12B lnadv Bpen-Bypass F 2-2 Fuel Failure 17 Scran SF Provide )61U < 901 open trip signal otectim systea N) failmfe logic. 15 Loss Bff<ite AC F 2-2 Fuel Failure Pee$ 5j~eP fiptg t 'pf j$ 4ij!@.Qgnpjfr logic.

QFE BUlDW A)ILES]S Page 9l oG85

                                                                                                        ~'repared:

Bate: NA) X STEM) SWE)l Checked) Bate ' NHM~ Nev 1) C. L Parber EVENT RBCT NNXGVARE SAFE N SER RESI.TS SAFHY ACTlSl F1K SH). aENT TTTLE STATE Code Title Code Title NK Rf Ctrl Fail~ F 2-2 Fuel Failure 17 Scraa , Provide > RC turbine first stage pressure interloct signal to reactor protectim systea 199) fail~e logics 23 Rf Ctrl Fail~ F 2-2 Fuel Failure 17 Scran Provide calo turbine step valve < %If open trip signal to reactor protectim systen 199). 27 Pipe Break Outside F 3-2 Fuel Failure 17 Scraa Provide )llV < 90l open trip signal to reactcr protectim systen {99) fail~e logic. 328 bw Reservoir F 2-2 Fuel Failure 17 Scraa Provide > 3R turbine first stage prese interlock signal to reactcr protectim systes 89) fail~e logic. This safety actim nay ocar, but is required mly if a lou vanxa

                                                                                   . turbine   trip  occurs.

328 W Reservoir F 2-2 Fuel Failure 17 Scraa Provide aain turbine stop valve < 90X open trip signal to reactor protectim systen 199). ibis safety actim nay occur, but is required mly if a lm vacuun turbine trip occursr 01A Seneratcr Trip F 2-3 Syst Stress 18 Pres Relief SF SNs open m high reactcr pressure. 01A Seneratcr Trip F. 2-2 Fuel Failure 18 Pres Relief Nain turbine bypass valves open m tgbine cmtrol systen 1l7) gewrator trip signal. This safety actim nay. occur, but is not required by this

                                                                                   -   events 018     TIS TripW Fails    .F     2-3 Syst Stress  18                         S)Ns open   m high reactor pressure.

01C PresReg Fail&use CBEF 2-3 Syst Stress 18 a 02 Turbine Trip F 2-3 Syst Stress 18 Pres Relief SF SNs open cn hI"Q "reaiti pri'siuFjr'2 Twbine Trip 2-2 Fuel Failure 18 es iaaf / turbine ccntrol systea N7) turbine

                                                                                     'trip signal. This safety actim nay occur, but is not required by this P) event.
 ~
   ~ ~

SAFE Mil AN.y51S Page. 95 of3$ 7 ppppgrg.' C- pptn ~au E. Eck Neckedly BF)HKS3~ (Rev 1) C. N. Parker 6M' REACT QNXEPTABLE SAFE HER REM;lS SAFETY ACT19I RK 5EL EItBff TlltE STA)K Code Title Code Title IE GNSAS 03 Nain Stn lsolatim COEF 2-3 Syst Stress 18 Pres Relief SF SNs open cn high reactor pressure. 05 Loss of Acuun F 2-3 Syst Stress 18 Pres Relief SF %Vs cpen m high reactm pressure. 05 Loss of Vacuun F 2-2 Fuel Failure 18 Nain turbine bypass valves open m turbine cmtrol systen 147) turbine trip signal. Ibis safety actim nay occur, but is not required by this evmt. 12A Pres Reg Fail~ COEF 2-3 Syst Stress 18 Pres ReliH 5F QNsopenmhighreactor pressure. 128 lnadv Open~ COEF 2-3 +t Stress 18 Pres Relief SF QNs open cn high reactcr pressure. 15 Loss Offdite AC COEF 2-3 Syst Stress 18 Pres Relief SF SRVs open m high reacts pressure. F 2-3 Syst Stress 18 Pres Relief SF SNs ope m high reacts pressure. 23 Rf Ctrl Fail~ F 2-2 Fuel Failure 18 Pres Relief Nain turbine bypass valves open m turbine cmtrol systen N7) turbine trip signal. 2l Rod Orop Accident OF 3-3 Syst Stress 18 Pres ReliH SF SNs open cn high reactor pressure. ZC Saall Break inside COEF 3-3 Syst Stress 18 Pres ReliH SF SRVs open cn high reactcr presare. 27 Pipe Break Outside COEF 3-3 Syst Stress 18 Pres Relief SF 51Ns open m high reactor pressure. 30 Overpres Protect COEF 2-3 Syst Stress 18 Pres ReliH 5F SNs open m high reactor pressure. 328 Lap Reservoir COEF 2-3 Syst Stress 18 Pres Relief 5F SfNs open m high reactor pressre. 328 Qw Reservoir COEF 2-3 Syst Stress 18 Pres ReliH Naan turbine bypass valves open cn turbine cmtrol systen f47) turbine trip signal. This safety actim nay occur, but is not require) by this event. 3Z Fire COEF I-5 Rpg Overpres 18 Oesign of safety features est -0

S4FE SHJHOH 484LYSIS Page 96o(Xb'7 7 NIk STE4N STSTEk Prepared'hecked! E O tei ~~~~ Oats: NHKQ~ (Rev 1) C. L Parks 88(T RE4CT Ll(4CCEPT4BLE S4FE N tRR RESKTS S4fE(T 4CTIOk flK SH}. BM TITLE ST4TE hde Title CHe Title NK (6 Nain Sts Isolatim CKF 2-2 Fuel Failure 19 Core Cooling Provide stean for RCIC (71) turbine. (6 Nain Sts lsolatim (XEF 2-2 Fuel Fail~ 19 hre Cooling Provide stean fN (KI 03) turbine in support of )Kl initiaticn m low water level (L2). (XEF 2-2 Fuel Failure 19 hre holing Provide stean for )KI (73) turbine in support of (Kl initiatim m Im water level (L2), 05 loss of Vacuus %F 2-2 Fuel Failure 19 Core Cooling Provide stean fi RCIC (71) turbine. 124 Pres Reg'Fail~ COEF 2-2 Fuel Failme 19 Care Cooling Provide stean for RCIC (71) turbine. 124 Pres Reg Fail+en (XEF 2-2 Fuel Failure 19 Ccrc Cooling Provide stean fa )KI (H) turbine in support of )KI initiatim cn Ios water level (L2). 12B lnadv Open-Bypass (XEF 2-2 Fuel Failure 19 Core Cooling Provide stean fm RCIC (71) turbine. 12B lnadv Open~ass IXEF 2-2 Fuel Failure 19 Care Ccoling Provide stean fcr )Kl (75) turbine in support of )KI initiatim cn Icw water level (L2). D lnadvert Open S(N COEF 2-2 Fuel Failure 19 Care Cooling Provide stean fm'CIC (71) turbine. D Inadvert Open S(N (XEF 2-2 Fuel Failure 19 hre holing Provide stean fcr (Kl (H) turbine in support of (KI initiation m Im water level (L2). H Loss of F)f Row COEF 2-2 Fuel Failure 19 Core holing Provide stean hr RCIC (71) turbine, 14 toss of FN Row KF 2-2 Fuel Failure 19 Core Cooling Provide stean hr (KI fTP turbine

                                                                                         'n support    of IKI initiatim m Im water level  (L3.

15 Loss Of flite 4C (XEF 2-2 Fuel Failure, 19 hre Cooling Provide stean for RCIC (71) turbine. 15 Loss Off<ite 4C (XEF 2-2 Fuel Failure 19 e holing Pran in support of. (Kl initiatim m low

                                                                                                ~(p.

-0 t Kl

SAFE QUTIM QNLYSIS f)t epared: Dates za NAIN STEAN SYSTEN Checked s Dates T BF)HKS3~ IRev I) C. N. Parker BJENT REACT MXEPTAELE SAFE IN IfER REM.TS SAFEIY ACTIQl RK SH). BJE>IT TITLE STATE Cade Title Code ~ Title mE 22 Loss Shutloe hol CD 2-2 Fuel Failure 19 hre Caoling %F Naintain pressure law enough fsr RIR systen if necessary, with Relief Yalvel s). 23 FN Ctrl Fail~ CDEF 2-2 Fuel Failure 19 Care holing Provide stean far RCIC I71) turbine. 23 FN Ctrl FaiHhx CDEF 2-2 Fuel Failure 19 Care Cooling Provide stean fsr IfCI 173> turbine in support of IfCI initiatim m law water level IL2). 24 Rod Drop Accident 3-2 Fuel Failure 19 Care holing Provide stean for IfCI 173> turbine In suppsrt of IfCI Inltiatim m law water level IL2). 258 intern bk Inside CDEF 3-2 Fuel Failure 19 Care Ccoling Provide stean fsr IfCI I73) tsrbine in suppcrt of IfCI initiatim m law

                                                                                 <<atsr level IL2).

258 Intern Brk Inside CDEF 3-'2 Fuel Failure 19 Core Cooling Provide stean fcr IfCI f73) turbine in suppcrt of IfCI initiaticn cn high dryweII pressure. I 258 Intern Brk Inside CDEF 3-2 Fuel Failsre 19 Ccrc Cooling SFS Provide reactcr vessel auto depressurriaatim systen IADS) an prieary cmtaineent systen {H> high DN presssre signal coincident with FN systen 13) UL ILILL3) far are the 120 secmds b either tso csre spray 175) puaps cr cne M I7I) puep running. 25C Snail Break Inside CDEF 3-2 Fuel Failure 19 Care Cooling Provide stean fcr IfCI 03) turbine in suppart of IfCI initiaticn m law

                                                                               , water level IL2>.

25C Saall Break Inside CDEF 3-2 Fuel Failure 19 Care Cooling Provide stean far PCI 173) turbine in suppsrt of IfCI initiation m high pressure. Snail Break Inside 3-2 Fuel j 25C CDEF

                                                                                 )ctc~rtcctt           gt~':IWtt tea, j) iccctccicttc'cc'ccct'tttt is ~litt'chi)ts DN  presssre ssgnal cosncsdent wsth Rl systen 0) Ul IL11L3) fsr esre than t

05> puaps sr me SR I7I) puap ruming ~ 4

QFE SWIIW ANALYSIS SIN SIEAII SYSTE)I P<<qw<

> ~ ~ IO s' ~ s s t ' I s I s s I ~ ~ I ~ I ~ ~ ~ ~ ~ I' t s I ' ~ ~ ~ I s ~ s 0 SfE Mmow A)NLYSIS Page 100 $ 07 Prepared: Date: NI)l STEA)l SYSTE)l .E. C. Eck lheckab IRH)SS3~ 1Rev 1) I EUM REACT )NNXPTABLE QfE CR IRR REINS SfETY ACTlnl RlC SE9. EUENT TITLE STATE Code Title Code Title CE 25A Large Break Inside Pf 3-1 Rad Release 20 RPU lsol Provide Ion pres@re signal lin aain stean line at turbine) to priaary aatainaent systea N) faille logic NN aude). This safety action aay occur, but is not required. 25A Large Break Inside 3-1 Rad Release 20 RPU Isol Close )SIUs and aain stean drain lines on Iw pressure signal lin aain'tean line at turbine) froa priaary containaent systea N) (R)I aade). This safety action aay occur, but is not required. 25C Saall Break Inside 3-1 Rad Release 20 RPU lsol Provide Iw pressure signal lin aain steaa line at turbine) to priaary contaimnt systea N) failmfe logic lNM aude). This safety acticn aay occur, but is not required. 25C Saall Break Inside Rad Release 20 RPU lsol Close )SIUs and'aain steaa drain lines on Ion pressure signai lin aain stean line at turbine) froa priaary containaent systea N) (QN aude). This safety acticn aay occN', but is Rot requffedr 25D Brklnside-Cont@ad 3-1 Rad Release 20 RPU Isol Provide Ion pressure signal lin aain 'stean line at turbine) to priaary aatainaent systea N) failmfe logic il91 node). This safety action aay occur, but is not required. 25D Brklnside-CcntLRad JKF 3-1 Rad Release lose )SIUs and aain steaa drain <<q.(n lines a si lin aain ctccc i>cc.ct ter>ice> tres >r> i /c't>cc>> ,ttt> ic>t> ccc>c>. 27 Pipe Break Outside GXF 3-1 Rad Release Sensas aeittring fN high steaa ~< ature and differential 0 pressure britche fina in the )S lines aust send closure signals to systeaN) to initiate closure of the )SIU. hi X7 SAFE BUHOtI QNLYSIS )NI)I STEA)I SYSTHI Prepared'heckedy P e101 E. C. Ec Ote: ~ ate ~%PI BF)H)SS3~ IRw 1) C. IL Parker EVDIT REACT lNCCEPTlRE SAFE N !M IKM.TS SAFHY ACTIN RK SB. EVE)IT TITLE STATE Code Title Code TNe CKE 27 Pipe Break Outside GKF 3-1 Rad Release 20 RPV lsol Close )61Vs and nain stean drain lines m high stean flw cr high stean tunnel teeperature signal froe systen Q. 27 PipeBreak Outside <C,REF 3-1 RadRelease 20 RPVlsol Provide lou pressure signal Iin aain stean line at turbine) to primy cmtaimnt systee N) fail~fe logic NSI node). This safety actim aay occur, but is not required. 27 Pipe Break Outside S f 3-1 Rad Release 20 RPV Isol Close )61Vs and aain stean drain lines m lar pressure signal Iin aain stean line at turbine) froe prinary containnent systen IM) I@M node). This safety actim aay occur, but is not required. '32B Uw Reservoir F 2-1 Rad Release 20 RPV lsol Close sain turbine stop valves cn turbine cmtrol systen I57) condenser lou vacuua signal. This safety acticn say occur, but is not required by this event. gt 32B Lcm Reservoir COEF 2-1 Rad Release 20 RPV lsol Close aain turbine bypass valves m turbine control systee Il7) cmdenser lou vacuua signal. This safety actim say occur, but is not required by this went. %0 Haraal Shutdoe COEF 2-1 Rad Release 20 RPV lsol )Ianually close )61Vs and stean line drains after nanual scraa froe reactor protectim systee, 99 Iif . seeded). 01A Seneratcr Trip F 2-2 Fuel Failure pftQrjj i've)hii/if avoid scranch th) This safety Rtlm eay occur, but is not required by this ~ ~ l SAFE S)UIESN AHALYSIS Page 102 OQRr 0 Prepared: NAIH STEAN SYSTEN E.C k r tbecked: BF)H)663~ llhv 1) C. IL Parker BIEHT REACT QACCEPTARE SAFE CR fPER RES)lTS SAFETY ACTTN RK SEL BIEHT TllLE STATE Code Title Code, Title GEE 02 Turbine Trip F 2-2 Fuel Failure 22 Ihst Homal Signal fron Turbine Con'tfol l47) 'to bypass valves provides tuv'na) control after trip fron low initial poer, avoids scran. This safety action nay occur, but is not required by this event. 13 lnadvert 0pen S)N IXEF 2-3 Syst Stress 22 Rest Rrnai N Attenpt to reclose stuck open 9P. 0)A 6eneratcr Trip F 2-2 Fuel Failure 24 PouReduce SF Provide >301 turbine first stage pressure interlock signal to reactcr protectim systen 199) fall~e logics 0)B T/6 Trip-BP Fails F 2-2 Fuel Failure24 Poulhduce SF Provide >301 turbine first stage pressure interlock signal to reactcr protectim systen 199) fail~fe log)co 02 Turbine Trip F 2-2 Fuel Failure 24 Pou Reduce SF Provide aain turbine stop valves < 90X qxs) trip signal to reactN protectim systea 199). 02 Turbine Trip F 2-2 Fuel Failure 24 Pou lhduce SF Provide > 301 turbine first stage pressure interlock signal to reactcr protection systen (99) failmfe logic ~ 5 Loss of Yacuun F 2-2 Fuel Failure 24 Pox Reduce SF Provide nain turbine stop valves < 901 open trip signal to reactor protection systen 199). 05 Loss of Yacutm F 2-2 Fuel Failure 24 Pou Reduce SF Provide > RZ turbine first stage ~ pressure interlock signal to reactcr protecticn systen 199) failmfe logics 23 FN Ctrl FaiHhx F 2-2 Fuel Failur e SF Provide aain turbine t enI Qp 23 FM Ctrl Fail~ 2 F Q'p Q 'y aJ 'i @rNlirL Rfii~binr first interlock signal t rig'ressure 'rotecti ailmfe ogico SAFE SHHMN AN.YSIS Page 103 of2()(r Prep ved! Date: ~>v~ lQIH STEAI( S6TE)( E. C Eck Oecked: Datet 5 If)H)SS3~ (Rev 1) i'VM REACT NACCEPTABLE SAFE N 5%R RES(lTS SAFETT ACTI% FIK SE(). EVENT TIRE STATE Code Title Code Title mE 328 Lw Reservoir F 2-2 Fuel Failure 8 Pw Reduce SF Provide nain turbine stop valves < 9$ open trip signal to reactor protection systen (gg). This safety acticn nay occur, but is required only if a lw vacuun turbine trip 32B Lw Reservoir F 2-2 Fuel Failure 24 Pw Reduce SF Provide > RX turbine first stage pressure interlock signal to reactor protection systen (gg) failmfe logic. This safety action nay occur, but is required only if a lw vacuua turbine trip occurs. '3 F)I Ctrl Fail~ CDEF 2-3 Syst Stress 25 Level Reduce Close nain turbine stop valves on high water level (LB) signal froa ~ aln turbine control systen N7). 23 F)l Ctrl Fail~ CDEF 2-3 Syst Stress 25 Level Reduce Close F)I hrbine inlet valves on high water level (LB) signal fron F)l control systen N). 24 Rod Drop Accident DF 3-1 Rad Release 26 Est Pri Cont SF Close lSIVs and aain stean drain lines on lw reactm uater level (L1) signal fron priaary cmtainaent systen (M). 2i Rod Dry Accident lf 3-1 Rad Release 26 Est Pri Cont SF I(ain stean line drains isolate upon <I" hip>radiaticn trip signal fron Prinary Containeent systea (6(). 25A Large Break Inside (XXF 3-1 Rad Release 26 Est Pri Cont SF Close l(SIVs and nain stean drain lines on lw reactcr uater level (Ll) .signai fron priaary containaent systen (N. 25A Large Beak Inside CDEF 3-1 Rad systen (6I) signal fv priaary a Ir>> s +p~lplv are 25B Intern Brk Inside CDEF 3-1 Rad R lease Pr'i'Cont SF 'oM fBk 4 ah st(( 'dr a~ ( lines cn lw reactcr water level (Ll) si I froa i sary ccntainaent systea (6(). e ~ >> SAFE )8)TNN NQLYSIS Page 1N oQ!lr3~7 Prepared( (late: 4 > N) k STEAt( SySTE)( E. Caked: hte: EF)Hj553~ (Rev 1) C. L Parker EUM REACT MXEPTASLE SAFE N (RR RES(lTS SAFETy ACT IN F(K 5E(l. EUM TlTlE STATE Code Title Cede Title (IE Snail Sreak inside C(XF 3-1 Rad Release 26 Est Pri Cmt 5F Close lSIUs and aain stean drain lines m lm reactcr water level (L1) signal frca priaary ccntainaent systea (6().

25) Srklnside-CmtNad IXEF 3-1 RadRelease 26 Est Pri Cont SF C(ose lSIUs and aain stean drain lines m low reactor water level (Ll) signa) froa priaary cmtainaent systea (M).

ZD Brklnside-CantNad, (XEF 3-1 Rad Release 26 Est Prl Cont SF High steaaline flow and high stean tmnel teaperature signals ve provided to priaary cmtainaent systea N), 250 itrklnside-ContNad (XEF 3-1 Rad Release 26 Est Pri Cmt SF )(ain Stean lsolatim valves and drain and vent va)ves close upm receipt of isolaticn signal froa the priaary cmtainaent systea N). 27 Pipe Sreak Outside COEF 3-1 Rad Release 26 Est Pri Cont SF Close t$ 1Us and aain stean drain lines on low reactor water level (Ll) signal fn>> priaary cmtainant systea N).PQ a~~+f'~ ~fret. WQPff1 15 Loss Off<ite AC ALL 2-1 Rad Release 27 Est Sec Cent 5upport secmdary cmtainaent functim. 250 Brk inside-CmtNad COEF 3-1 Rad Release 27 Est Sec Cont Support secmdary cmtainaent functicn. 26 Fuel Handl Ace(dot A 3-1 Rad Release 27 Est Sec Cont Support secmdary cmta(naent functim. 33 Loss Fuel Pool Cl Nl 2-1 Rad 1 t Cmt Support secondary containaent

.I,:.g.(, ps p,.;,

150 15O 5hutdown COEF 2-2 Fuel F 1 t II/ k; adj RPY jjlgi using SRUs, 'jjriz(upi if suppressim lcod iR. teaperature approaches allowable SzE Are CktMtlf9g CJ t~ VCSlpl J4 Shed'g ben(udge ~~y7iw flkkrO lr lit R ~~ c>>st4~+i~Si g~ +~alp ~ ~ ~ i >~ ','C SEE SHJEM NQLYSIS Page 105 of Preparedi Date: NAIK STEAN STSIBI E. C. Ec Oieckedi Date: 8F)H)SS3~ (Rev 1) C. )L Parker EUEKT REACT MXEPTASLE SAFE CR CPER RES%IS SAFEIT ACTIN FlK SE{L BM TITLE STATE Code Title Code Title GEE )8) Horaal Shutdown CDEF 2-2 Fuel Failure 3O Cont Cooling l6F Controlled aanual depressurizatim of RPU using SRUs, if suppression pool teaperature approaches allowable liaits. If necessary, provide aanual ADK SSD Safety Shutdown CDEF 3-2 Fuel Failure 3l) Cmt Cooling l6F Cmtrolled aanual depressuriaatim of RPU using SRUs, if suppressim pool .teaerature approaches allowable Ri liaits. If necessary, provide aanual ADS. ~ 25A Large beak Inside IEF 3-2 Fuel Failure 33 Lia LossCml L Nain steaa line flow restrictm s liait loss of coolant mtil l61U closure occurs (passive). 27 Pipe 8reak Outside CDEF 3-2 Fuel Failure 33 Lia LossCool L Flow restrictm s installed in the. Nain Steaa lines aust liait the aass flow rate of coolant being ejected frm the steaa lines following a break. This is a passive functim. 28 Shtdn Froa Outside DF 4-1 Local Shut 38 Shutdn Outsd N SEKBKE Cs STEP 3: All I61Us are aanually closed to isolate the ractw45sru~ alee ia sbcyi) 28 Shtdn Froa Dutside EF 4-1 Local Shut 38 Shutdn Outsd SF SE!SKE A cr 8, SIP 4: Close all lSIUs cn signal free systea 6l. 28 Shtdn Froa Outside CDEF 4-1 Local Shut 39 Cooldn Out essurize,the reactor pressure vessel. 2S Shtdn Froa Outside CDEF 4-1 Local Shut ],:,(.. pa 2S Shtdn Free Outside CDEF 4-1 Local Shut 3g Cooldn Outsd N ~dk ki and )6IU test circuits. 329 Earthquake ALL 4-3 Equip Fail 44 Seisaic Shtd Design of safety ea capability to provide shutdiaei following a design basis seisaic event. SAFE SHJTMN ANALYSIS 106 47 Prepared'age ense: ~~ GEE)GATE NS E. C Eck O8IMNLIIH)WITER SYSTE)I Ooked: Oaten' f7 fAH86~ IRev I) C. L Parker EVENT REACT NCCEPTABLE SAFE N 6%R RESILTS SAFETY ACTIN FIK KL EVENT TlllE STATE Code Title Code Title ORE 05 loss of Uacuua OF 2-2 Fuel Failure 17 Scran SF Provide low condenser vacuun signal to RPSI99) for scran Ietra,anticipatory signal). This safety actim is epectsd to occur, hut is not a requireeent of this event. ZB Low Reservoir If 2-2 Fuel Failure 17 Scraa SF Provide low cmdenser vacuua signa) tetra backup signal) to reactl protecticn systen I99) fail~e logic. Thjs safety actim nay occur, hut is not required by this event. 8 IIain Stn lsolaticn COEF 2-2 Fuel Failure 19 Are Cooling Provide ncraally open water supply fl RCIC systen I71). N )Iain Stn Isolation COEF 2-2 Fuel Failure I9 Core Cooling Provide ernally open water supply fcr SCI systen I73) initiatim cn low water level (L2). 05 Loss of Vacml COEF '-2 Fuel Failure 19 Core Cooling Provide na'nally open water supply fcr IfCI systen I75) initiatim m low water level (L2). 05 Loss of Vacuun COEF 2-2 Fuel Failure 19 Care Ccoling Provide smally open water supp) y for RCIC systen I71). 12A Pres Reg Fail+en COEF 2-2 Fuel Failure 19 Core Cooling Provide ncrnally open water supply for RCIC systen I71). 12A Pres Reg Fai &pen COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide nlnaily open water supply fcr )fCI systen 173) initiaticn on low water level IL2). 12B Inadv Open~ass GKF 2-2 Fuel Failure 19 Core Cooling Provide noraally open water supply for RCIC systen I71). 128 lnadv Open-Bypass COEF 2-2 Fuel Failure 19 Provide ncraally open water supply 'atim m low water level (L2). 0 P,f 1 Il/]~) ] 'g0."~J ( I'ij> SAFE 9JJIHN ANALYSIS Page.'107 of Prepared> CMENBATE AND E. C. Eck DENMfNLIIED NTER STSTS Checked> KH)BS3~ lRev 1) EVE)fT REACT MXEPTARE SAFE 0R 5%R RESKTS QFHY ACTIBI RK SE0. EVENT TITLE STATE Code Title Code Title C00E 13 Inadvert 0pen SRV C0EF 2-2 Fuel Failure 19 Core Cooling Provide ncraally open water supply fcr RCIC systea l71). 13 lnadvert 0pen SRV CDEF 2-2 Fuel Failure 19 Core holing Provide na'aally open water supply fcr ffCI systea f73) initiaticn cn low water level lL2). If Loss of F)f Flow CDEF 2-2 Fuel Failure 19 Core Coolirg Provide ntraally open water supply fcr RCIC systea l71). Loss of F)f Flow CDEF 2-2 Fuel Failure 19 Ccrc Cooling Provide Mraally open water supply fm )KI systea f73) initiaticn cn low water level lL2). 15 Loss 0f finite AC CDEF 2-2 Fuel Failure 19 Ccrc Cooling Provide nm aally open water supply fcr RCIC systea 01). 15 Loss Qff-Bite AC CDEF 2-2 Fuel Failure 19 Ccrc Cooling Provide ncraaliy open water supply for )PCI systea l73) Initiation cn 'low water level lL2). 23 F)f Ctrl Fail~ CDEF 2-2 'Fuel Failure 19 Core Cooling Provide ncraally open water supply fcr RCIC systea l71). 23 F)f Ctrl Fail~ CDEF 2-2 Fuel Failure 19 Ccrc Cooling Provide ncraally open water supply fcr )PCI systea f73) initiaticn on Iow water level IL2). 2f Rod Droo Accident DF 3-2 Fuel Failure 19 Core Cooling Provide tonally open water supply fcr IPCI systea (73) initiaticn on low water level lL2). ()kraal path, hut not required) 3-2 Fuel Failure 19 Core 'de ncraally open wa 258 lntera Erk Inside C0EF 258 lntera Brk Inside CDEF 3-2 Fuel Failure 1 ~ ~ j ~ ME S)ITHN AN.YSIS Pane IOB of g 7 Prspvsl: K latm> ~Z I66ATE NO E. c Kl(IIKSUIED)(ATER SYSTEN Checked: Date! z-ORH)SS3<(8 (Rev 1) C. )L Parker BEHT N REACT IRR (NA(XEPTARE - RES(LTS SAFE)Y 'AFE ACTIN RK SEL Et}EHT TITLE STAlE Code Title Cade Title CODE 25C Seal) Oreak Inside COEF 3-2 Fuel Failure 19 Care Cooling Provide ncrnally open water supply for %CI systea (73) initiation on low water level (L2), Noraal path, (}ut not required) 25C Snail Break Inside GKF 3-2 Fuel Failure 19 Ccrc Cooling Provide }uraally open water supply fcr )f}CI systen 03) initiation cn high drywell pressure signals. Nornal supply, but not required) 27 Pipe Oreak Outside COEF 3-2 Fuel Failure 19 Care Cooling Provide nmnally apen water supply fcr )fCI systea (73) initiaticn on low water level (L2). Nrnal supply, but not required) 32S Law Reservoir COEF 2-2 Fuel Failure 19 Care Cooling Provide ncrnally open water supply fcr RCIC systen (71). 328 Low Reservoir COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide ncraally apen water supply fcr )fCI systen (73) initiation on Iaw water level (L2). 1% ISO Shutd}we COEF 2-2 Fuel Failure 19 Care Cooling Provide }wraally open water supply fcr RCIC 01) and IFCI (73) systeas. RSO Homal Shutd!wo COEF 2-2 Fuel Failure 19 Care Cooling Provide ncrnally ap0} water supply far RCIC 01) and )FCI (73) systens. Sb Safety Shutdo}e COEF 3-2 Fuel Failure 19 Care Cooling Provide carnally open water supply f~ m 03) sy t~. Th(s safety action 'nay occur, but is not required. 06 Loss F)( Heater F 2-2 Fuel Failure 23 as+uilt design neets 100 degree F assuaptian for naxinun feedwater tenperature drop that can 'l 'g II}9% 1(i:6t(('1}ri }}, ~ t}i s}. Jp+O3)" Qksl r}! jI ac $t ( <<-'I Systens. r I ~ ~ ~ ~ Pw QFE 9UT9N ANLYSIS Page Igg Prepared: 7o'. see: ~~a~ C. Eck DENllSNLIIED NTER SSIEN Checked: Date: ZP M SF)H)SS3~ IRev 1) BlW REACT MXEPTASLE SAFE N VER RERLTS SAFETY ACTIN FIR SEQ. 99lT TITLE STATE Code Title Code Title C00E 15 Loss Off<ite AC ALL 2-1 Rad Release 27 Est Sec Cont Support secmdary containnent fmctim. ZO Srklnsid~tNad CDEF 3-1 Rad Release 27 Est Sec Cont Suppcrt secmdary cmtainnent functim. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cmt Support secondary containeent functim. Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont SuppN t secmdary cmtainnent functicn. 20 Shtdn Froe 0utside CDEF H Local Shut 3g Cooldn 0utsd Provide nN nally open uater supply fa RCIC systea OI). Noraal supply, but is not required) 3Z Ear thquake ALL 4-3 Equip Fail qq Seisaic Shtd Design of safety features oust have capability to provide shutdae follouing a design basis seisnic event. n , ~ 5~ ~ ~ SAFE 5HJ1NN ANALTSIS Page 110 of 7o7 Prepared: Data ~au> E. E Reeked: Bate: BF)H)553~ (Rev I) C. L Parker EVENT REACT MXEPTABLE SAFE OR IPER RES(LTS SAFEIY ACTION RK SEL EVENT TITLE STATE Code Title Code Title GEE 01A Seneratcr Trip F 2-2 Fuel Failure 17 Scras SF Provide high reactcr vessel pressure trip signal to reactor protecticn systea N) failmfe logic. OIB T/5 Tri~ Fails F 2-2 Fuel Failure 17 Scraa SF Provide high reactcr vessel pressure trip signal to reactor protection systea (99) failmfe logic. 01C PresReg Fai l<)ose F 2-2 Fuel Failure 0 Scraa 5F Provide high reacta vessel pressure trip signal to'reactcr protection systea (99) fail~fe logic. 02 Turbine Trip F 2-2 Fuel Failure 17 Scraa SF Provide high reactcr vessel pressure trip signal to reactcr protection systea (99) fail~fe logic. 05 Loss of Vacuua If 2-2 Fuel Failure 17 Scras 5F Provide high reactcr vessel pressure trip signal to reactor protection systee (99) failmfe logic. 13 Inadvert Open SRV OF 2-2 Fuel Failure 17 Scraa SF Provide low water level (Q) trip signal to reactor protection systea N) failmfe logic. Il Loss of R( Flow EF 2-2 Fuel Failure 17 Scras SF Provide low water level (L3) trip signal to reactcr protection systea N) failmfe logic. 15 Loss Of flite AC If 2-2 Fuel Failure 17 Scraa SF Provide low water level (Q) trip signal to reactor protmticn systei (99) falimfe logic. 23 R( Ctrl Fail~ EF 2-2 Fuel Failure 17 -Scram SF Provide high reactcr vessel pressure -trip s(goal to reactcr protecticn ea N) failmfe logic. 23 F)( Ctrl Fail-)Iax EF 2-2 Fuel Failure SF Provide Iow water level IQ) trip t fa l Large Break Inside EF 3-2 Fuel Failure Provide low water leve signal to reactor protecticn system i SAFE SmmjtN ANALYSIS Page 111 oft% FaSvsl: E. C. ' Oats ~w/ Checked: Oats ~SO 1 SRHEQ~ (Rev I) C. L Parker BM RMT MXEPTADLE SAFE OR IRR REM.TS SAFEIY ACTIN FlK SEL EVENI TITLE STATE Code Title Code Title IXXE Z8 intern Brk Inside DF 3-2 Fuel Failure 17 Scraa SF Provide low water level IQ) trip signal to reactcr protecticn systen f99) failmfe logic. Snail Dreak Inside DF 3-2 Fuel Failure 0 Scraa SF Provide low water level IQ) trip signai to reactcr protection systen 199) failmfe logic. 27 Pipe Dreak Outside DF 3-2 Fuel Failure17 Scran SF Provide low water level IQ) trip signal to reactcr protection systen f99) failmfe logic. R Overpres Protect DF 2-5 Syst Stress 17 Scran SF Provide high reactcr vessel pressure trip signal to reactor protecticn systen t99) failmfe logic. 03 Nain Stn lsolaticn CDEF 2-2 Fuel Failure 19 Core Cooling SF Provide low reactcr water level signaI IL2) to M systen (74) logic fl'CIC systen f71) initiation. 03 Nain Sta Isolation CDEF 2-2 Fuel Failure 19 Core Ccoling Provide path fl'CIC 171) flow to the vessel through the feedwater spargersa 05 Nain Sta Isolaticn IXEF 2-2 Fuel Failure 19 Core Cooling SF Provide Iow reactor water level signal IL2) to )PCI systen IH). 03 Nain Sta Isolaticn CDEF 2-2 Fuel Failure 19 Core Cooling Provide path fa )fCI <2) flow to the vessel through the feedwater spargers fa )PCI initiation on low water level (L2). CDEF 2-2 Fuel Failure 19 Core Cooling SF Provide low reactcr water level IL2) to )PCI systen IH). 05 Loss of Vacuua CDEF 2-2 Fuel Failure 19 Core ovide path fcr )fCI IH) flow to e eedwater spv n 05 Loss of Vacuua IXEF 2-2 Fuel Failure 19 Ccr Provide low reactcr water level s ten (74) logic or RCIC systen 171) inc SAFE SN/HM ANALYSIS Page 112 ofgÃP 7 Prepared: Sate: E. C Eck flecked: Sate: BRH)SN~ lRev 1) C. N. Parker EVENI REACT MXEPTABLE SAFE CR OPER REM.TS SAFEIY ACTIN RK SEO. EVM TIILE STATE Code Title Cade Title CKE 05 Loss of Vacuun IXEF 2-2 Fuel Failure 19 Ccrc Cooling Provide path fcr RCIC 171) flow to the vessel through the feedwater spargerse 12A Pres Reg Fail+en CSEF 2-2 Fuel Failure 19 Care Cooling SF Provide law reactcr water level signal tt2) to RN systen 174) logic PI ) fm RCIC systen l71) initiatim. 12A Pres Reg Fail+en CSEF 2-2 Fuel Failure 19 Core Cooling Provide path far RCIC 171) flow to the vessel through the feedwater spargerso 12A Pres Reg Fail+en 'OEF 2-2 Fuel Failure 19 Care Cooling SF Provide low reactor water level signal lL2) to Hfl systen IH). 2-2 Fuel Failure 19 Core Cooling Provide path fa )fCI IH) flow to the vessel through the feedwater spargers fm PCI initiation m low water level lL2). '2B lnadv Open-Bypass CSEF 2-2 Fuel Failure 19 Ccrc Caoling SF Provide low reactcr water level signal lL2) to R)N systen 17i) logic for RCIC systen t71) initiaticn. 15 lnadv Open-Bypass COEF 2-2 Fuel Failure 19 Care Cooling Provide path far RCIC l71) flow to the vessel through the feedwater spar gerse 12B Inadv Open-Bypass CSEF 2-2 Fuel Failure 19 Care Cooling SF Provide low reactcr water level signal (L2) to IPCI systen IH). 12B lnadv Open-Bypass COEF 2-2 Fuel Failure 19 Care Cooling Provide path fm IPCI IH) flow to the vessel through the feedwater water level IL2). IS lnadvert Open SRV IXEF 2-2 Fuel Failure 19 Care i id P g o en )i i 15 lnadvert Open SRV COEF 2-2 Fuel Failure 19 Care th far RCIC l71) flow to

~ the vessel through the f wa er spar gers 4 ~,

~ ~ 'I SAFE SHH(XN N(ALTSIS REACT% FEED(ATER SySTE)( NH)SH~ IRev 1) Prep ved: Checked: Page E. 1D of Ec Bta Sate: ~ MiT REACT MXEPIAOLE SAFE OR SB RESN.TS SAFETY ACTlN F(K SE(). SEH( Tl)LE STATE Code Title Code Title GXE V lnadvert Open $ N COEF 2-2 Fuel Failure 19 Care Cooling SF Provide lw reacta water level signal (L2) to ffCl systee (75). 15 lnadvert Open SN COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide path fcr )fCl (TS) flw to the vessel through the feeduater spargers fN fPCI initiaticn m lw eater level (L2). D. lnadvert Open 9N COEF 2-2 Fuel Failure 19 Care Cooling SF Provide lw reactN uater level signals (Li). Provide lw reactcr pressure signals to me spray systee (75) and M systen (7I). Provide lw reactN water level (L1 0 U) pernissive signals to cain stean systen (01) fcr AOS. 1I 'oss of F)( Floe COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide lw reacts water level signal (L2) to RH(systea (7I) logic fa RClC systee (71) initiatim. ll Loss of F)( F(w COEF 2-2 Fuel Failure 19 Core Cooling Provide path fcr RClC (71) flw to the vessel through the feeduater spargers )I Loss of FN F(w COEF 2-2 Fuel Failure 19 Core Cooling SF Provide lw reactcr water level signal (12) to )fC1 systee (75). 1I Loss of F)( Flw COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide path fa ffC1 03) flw to the vessel through the feeduater spargers fcr (Kl initiatim m lw water level (L2). 15 Loss O(Hite AC COEF 2-2 Fuel Failure 19 Core Cooling SF Provide lw reacta water level 'signal (L2) to RN systee (74) logic fN RClC systen (71) initiatim. 15 Loss Dff-Site AC COEF 2-2 Fuel Failure 19 lin Provide path fcr RCIC (71) flw to 15 Loss Off-Site AC COEF 2-2 Fuel Fa}ure 19 SAFE SNllUN NCLYSIS Page 114 of 7 Prepared: Date: zo Reeked: Date: OF)H)S63~ IRev I) C. L Parker 4 BBIT REACT NNXEPTASLE SAFE N OPER RESN.TS . SAFETY ACTIN RK SED. BENT TITLE STATE Code Title Code Title GEE 15 Loss Dffdite AC COEF 2-2 Fuel Failure 19 Care Cooling Provide path fcr IPCI 173) flw to the vessel through the feeduater spargers fcr PCI initiation m lw uater level IL2), 15 Loss Dff<ite AC CDEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide lw reacta uater Level ILI 0 L3 peraissive) signals to aain stean systea 10)) fN AOS. Provide lw reartcr pressure signals to core spray systen f75) and R)R systen (74). 22 Loss Shutdoe Cool ACD 2-2 Fuel Failure 19 Core Cooling Indicate React@ water level and pressure in the cmtrol roon. 23 FN Ctrl Fail~ COEF 2-2 Fuel Failure 19 Core Cooling SF Provide lw reacta uater level signal IL2) to M systen l74) logic fN RCIC systee f71) initiation. 23 FM Ctrl Fail~ COEF 2-2 Fuel Failure 19 Core Cooling Provide path fcr RCIC I71) flw to the vessel through the feeduater spargerso 23 FN Ctrl FaiHhx COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide Iw reacta water level signal IL2) to )PCI systea 173). 23 FM Ctrl Fail-Nax COEF 2-2 Fuel Failure 19 Core Cooling Provide path fcr tfCI l73) flw to the vessel through the feeduater spargers fN'KI initiation m lw uater level IL2). 24 Rod Drop Accident OF 3-2 Fuel Failure 19 Core Cooling SF Provide lw reactor water level IL1 I L3 pernissive) signals to nain stean systea l01) fcr AOS. Provide lw actcr essure signals to ccrc spray systen f75) (74). 24 Rod Drop Accident IF 3-2 Fuel Fail e) 24 Rod Drop Accident OF 3-2 Fuel Fail e 19 Core Cooling Provide path fcr )PCI (73) flw to I spargers for )PCI initiatim m Iw eater level lt2), ~ VVE SWIHN ANALTSIS Page g) of @CD 7 Prepared: Date: ~eaa at IIEACTI FEEONTER STSTEN E. er Ooked Date: BRH)SS3~ tlhv 1) C. L Parker e BJENT IIEACT MXEPTARE SAFE R CPER IIEM.TS QFHY ACTISI FIK SEO. BJENI TITLE STATE Code Title Code Title CODE 25A Large Beak Inside CDEF 3-2 Fuel Failure 19 Core Cooling SFS Provide lw reactor pressure per>>issive signals to core spray syste>> 175) and M syste>> 175) fcr HXS acticns m high dryuell pressure signals. 25A Large Break Inside GKF 3-2 Fuel Failure 19 Core Cooling SF Provide Iw reacts water level signals ILI). Provide lw reactcr pressure signals to core spray syste>> 175) and M syste>> (71). 25B Inter>> Brk Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide lw reactcr water level ILI I L3 peraissive) signals to>>ain stea>> syste>> l01) fcr AOS. Provide lw reMa pressure signals to core spray syste>> 05) and M syste>> l7$ ). 25B Inter>> Brk Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide lw reacts uater level signal IL2) to )fCI syste>> 173). 258 Inter>> Irk Inside COEF 3-2 Fuel Failure 19 Core Cooling Provide path fcr )PCI 173) flw to the vessel through the feeduater spargers fl )fCI initiaticn m lw eater level (L2). 2% Inter>> Brk Inside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SFS Provide Iw reactl pressure per>>issive signals to are spray syste>> (75) and M syste>> (74) fcr HXS actims on high dry>>ell pressure signalse ZB Inter>> Brk Inside CDEF 3-2 Fuel Failure 19 Core Cooling Provide path for )PCI 03) flw to 'the vessel through the fee4ater spargers for )Kl initiatim cn high lLprassur 2X S>>all Break Inside COEF 3-2 Fuel Fa 'de lw reactor uater level (LI ID -0 spray syste>> (75) and (7f). M syste>> SAFE SWAN AHALYSIS Prepared: Page 116 ~ REACT% FEHNATER SySTE)I Gxcksl: Date: BF)H)SS3~ IRev 1) C. II. Parker MN REACT MXEPTl8E'AFE ~ N SER REM.TS SAFHY ACTIEII RK SB. BM TITLE STATE Code Title Code Title ImE 25C Snail Break Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide lw reactor water level signal IL2) to )PCI systen I73). 25C Snail Erea41nside CDEF 3-2 Fuel Failure 19 Ccrc Cooling Provide path for )PCI I73) flw to the vessel through the feeduater. spargers for IPCI initiaticn on lw water level It2). 2R Snail Ik'eak inside COEF 3-2 Fuel Failure 19 Core Cooling SFS Provide lw reactcr pressure pernisslve signals to core spray systen I75) and RHI systes I7$ ) fcr GXS actions on high dryuell presare signals. 25C Saall Break Inside CDEF 3-2 Fuel Failure 19 Core Cooling Provide path fcr IKl 03) flw to the vessel through the feeduater spargers fa IfCI initiation cn high dr)nn.ll pressure. 3-2 Fuel Failure 19 Core Cooling SF Provide lw reactor water level (Ll $ asd 27 Pipe Break Outside COEF reactor pressure signals to ccrc spray systee 05) and M systen 174). 27 Pipe Break Outside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide lw reacta water level signal ILi9 to )FCI systea I73). 27 Pipe Break Outside IXEF 3-2 Fuel Failure 19 Core Cooling Provide path fcr IFCI I73) flw to the vessel through the'eeduater spargers fcr )Kl inltlaticn on lw uater level IL2). 32B Lw Reservoir COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide lw reactcr uater level si 62) to RIR systen I7I) logic fe'ClC en 32B Lw Reservoir IXEF 2-2 Fuel Failure SAFE 58)NN A(ALYSIS Page 117 of Prepared)' REACTI FEEONATER StSIEH E. Dmk ed) RHKQ~ (Rev 1) C. L Parker BM REACT IACCEPTAELE SAFE (I( 5%R RES(LTS SAFEIY ACTI(I FlK SED. E(IEHI TITLE STATE Cede Title Code Title NK 320 Low Reservoir CKF 2-2 Fuel Failure 19 Core Cooling SF Provide lw reacta water level signal (L2) to ffCI systea (73). COEF 2-2 Fuel Failwe 19 Ccrc Cooling Provide path fm )fCI (73) flw to the vessel through the feedwatef spar gers fof )fCI initiation ce low water level (L2). IS0 ISO Shutdoe COEF 2-2 Fuel Failure 19 Core Cooling SF Provide lw reactor water level signai (L2) to (PCI systea (73). ISD ISO Shutdoe COEF 2-2 Fuel Failure 19 Care Cooling Provide path for IKI (73) flw to the vessel through the feedwater spafgefs C0EF 2-2 Fuel Failure 19 Core Cooling SF Provide lw reactor water level signaI (L2) to RH( systea (7$ ) logic fa RCIC systea 01) initiatiw. ISO 150 Shutdown NEF 2-2 Fuel Failure 19 Qre Cooling Provide path fcr RCIC (71) flw to the vessel through the feelhiater spar gef so t6D kraal Shutdtae (XEF 2-2 Fuel Failure 19 Core Cooling SF Provide lw reactcr water level signal (L2) to IPCI systea (73). HSD Horaal Shutdwn COEF 2-2 Fuel Failure 19 Core Ccoling Provide path fcr )PCI (73) flw to the vessel through the feedwater spar gers HSD Hcraal Shutdae NEF 2-2 Fuel Failure 19 Core Cooling SF Provide lw reactl >>ater level signa) (L2) to R)R systea (7$ ) logic fl RCIC systea 01) initiaticn. HSO kraal Shutdae NEF 2-2 Fuel Failure 19 Care Cooling Provide path fcr RCIC (71) flow to the vessel through the feedwater spar 9ef so SSD Safety %utdsa COEF 3-2 Fuel Failure 19 e Cooling SF Provide lw reacts water level (73). SSD Safety Sutdwn COEF 3-2 Fuel Failure 19 ea if( I vesse tlfr~4 spar gers' i SAFE MT(01( N(ALySIS P '118 of Prepared) C- I) tex ~i~/ E. Ec theckedx 0ate> IBHSQ~ (Rev I) C. L Pvker SW' REACT tNKEPTAKE SAFE PER RES(LTS SAFE(Y ACTIS( RK SEL BE)(T TITLE STATE Code Title Code Title CEE 05 Loss of Vacwa F 2-1 Rad Release 20 RPY Isol Feedwater puap trip signal (to systea

01) upon cmdenser lw vacuua signal.

32() Lw Reservoir CKF 2-1 RadRelease 20 RPY Isol Provide condenser lw vacuua signal to aain steaa systea (01) to trip feedwater turbine. This safety actim aay occur, but is not required by this event. ly 06 Loss FN Heater F 2-2 Fuel Failure 23 Lia Msturb L asWilt design aeets 100 degree F assuaptim fcr aaxias feedwater tiaperature drop that can cccur fl any single action a failure of FN heater ts). See also Condensate t02) and Extractim Steaa (05) systeas. 23 FN Ctrl. Fail~ (XEF 2-2 Fuel Failure 23 Lia t)isturb L The upper liait cn feedwater flow ast restrict flw to about 130 percent [Reload assuaption]. )l Loss of Rt Flw IF 2-2 Fuel Failure" 2f Pw Reduce Povide lw water level (t2) signal to oxen recirculatim H/6 drive anti breakers (200, electrical systea) fcr trip of recirculatim puaps. This safety acticn is expected to ocor, but is not a requireaent fm'his event. IEF 2-3 Syst Stress 24 Pwhduce Provide high reachr pressure signal to qadi recirculatim t(IS drive aota breakers (200, electrical systea) for trip of recirculatim puxps. OB lnadv Puap Start IIF 2-3 Syst Stress 25 Level de htgh reactcr water level signal (LB) o 06 Inadv Puap Start IF 2-3 Syst Stress 25 Level yt 1 1 if IRx 23 Rt Ctrl Fail~ GXF 2-3 Syst Stress 25 Provide high r t leva

~ signai (LS) to initiate san turb]ne control systea I

'119 Prepared> Oaten '+M+~~ Chechdt Oate: SS IHHjSS3~ IRev 1) Par)fer SE)IT REACT lNXEPTAKE SAFE OR SEL 'ct TTTLE SER RESILTS STATE Code Title SAFETY ACTTSI Code Title FIR IE 23 FN Ctrl Fail~ GKF 2-3 Syst Stress 25 Level Reduce Provide high reactcr wats'evel signal ILB) to initiate RFPT trip through feehater cmtrol systee W). 2I Rod Orth'ccident OF 3-3 Syst Stress 25 Level Reduce Provida high reacts water level signa) ILS) to IffCI systen 173). O3 )lain Stn lsolatim IXEF 2-1 Rad Release 2b Est Pri Cant SF Provide lw water level IQ) signal to reactm protecticn systee I99) fcr 1nitiatim of Q isolatims. This safety action is expected to cccN', but is not a requireeent this since there is no fuel failure. fN'vent GKF 2-1 Rad Release 26 Est Pri Cont SF Provide lw uater level IQ) signal to reactor protectim systen I99) fcr initiatim of Q isolatims. This safety acticn is epected to occur, but is not a reguirenent fcr this event since there is no fuel failure. 13 Inadvert Open SRV COEF 2-1 Rad Release 2b Est Pri Cent SF Provide lw eater level IQ) signal to reactor protectim systen I99) fN initiaticn of Q isolatims. This safety action is evpected to occurs but is not a requireeent fcr this event since there is no fuel failure. 1$ Loss of Rf Rw CKF 2-1 Rad Release 26 Est Pri Cent SF Provide lw water level IQ) signal to reactor protectim systen I99) fcr initiatim of Q isolaticns. This safety actlm is expected to DccN', hilt is not a requireeent fcr this event since there is no fuel failure. 15 Loss Off-Site AC NL 2-1 Rad Release 2b Est Pri Cont SF Provide lw uater level IQ) signal to reacti protectim systen (99) fcr initiatim so a lms. ls safety actim is espected to occur, ,0 II f@ffaW+i lcpmp fIfpfllvmfIf. SAFE QUMN AHALTSIS Page 120 of Prepared: lhta ~~4 REACT% FEHNTER STSTE)t IF)HSS3~ tRev 1) Giecked: E C. Ec L Parker ate: ~W GENT REACT NQPTARE SAFE N IPER RESN.TS SAFETY ACTIN RK SEt). BBIT TITLE STATE Cade Title Code Title GEE 23 Fit Ctrl Fail~ GXF 2-1 hd Release 26 Esk Pri Cant SF Provide law water level tQ) signal to reactor protectim systen (99) fcr initiation of Q isolatians. lhis, safety actim is expected to cccN, but is nat a requirenent far this event since there is no fuel failure. 2h Rad Brap Accident BF 3-1 Rad Release 26 Est Pri Cant SF Provide law water level tQ) signal to reactcr protectim systen tgg) far initiatim of Q isolatims. 2l Rod Erap Accident rF 3-I Rad Release 26 Est Pri C t SF Provide low reactm water level tU) signal to priaary cmtainant systen N) failmfe logic. IXEF 3-1 Rad Release 26 Est Pri Cont SF Provide law reactm water level tU) signal to pri nary cmtainant systen N) failmfe logic. 25A Large Break Inside tXEF 3-1 Radhlease 26 Est Pri Cmt SF Provide law water level tQ) signal to reactm protectim systen tgg) fcr initiatim of Q isolaticns. 25B Intern Brk Inside tXEF 3-1 Rad Release 26 Est Pri Cmt SF Provide law reacta water level tL1) signal to priaary cmtaiiment systea N) failmfe logic. 258 Intern Brk Inside tXEF 3-1 Rad Release 26 Est Pri Cant SF Provide law water level tQ) signal to reactar protecticn systen tgg) far initiation of Q isolatians. 25C Snail Break Inside tmEF 3-1 Rad Release 26 Est Pri Cent SF Provide low reactcr water level tLI) signai to prinary cantainaet systen N) failmfe logic. 25C Snail Break Inside GXF 3-1 Rad Rel Provide law water level tQ) signal to reactor pr initiation of Q isolatims. SAFE 9UHXN NNLTSlS P '121 of y Prepared> E. Eck Checkedy IF)Hap3~ Ilhv I) C. L Parker 99lI REACT MXEPTARE SAFE N SER RESETS SAFETy ACTlII FIK SEL BENT TlTLE STATE Code Title Code Title GIE 25> Brklnside+atNad COEF 3-1 Radlhlease 2b Est Pri Cmt SF Provide lw <<ater level IQ) signal to reectcr frotectiro epetee ttt> fcr I Rl initiatim of Q isolatims. 25D Brklnsi~tSad GEF 3-1 RadRelease 2b Est Pri Cent SF Provide lo<<<<ater level signals to priaary cmtainaet systen IM) fcr initiatim of cmtaimnt isolatim. eepfof~fpf. Rf e.f~p>> I 27 Pipe Break Outside COEF 3-1 RadRelease 2b Est Pri Cant SF Provide lo<<reactm <<ater level IL1) signal to prinary cmtainaent systen N) failmfe logic.,~ ~<p 24 Rlv xsol oM roper eeeeree~. ~< 3-1 Ihlease 4k-fsh4w4ant Provide Io<<<<ater level IQ) signal ~q 27 Pipe Break Outside COEF Rad SF to reactcr systea Igg) fcr proton initiatim of tS'isolatl s. 4~~> 15 Loss Offdlte AC ALL 2-1 Rad Release 27 Est Sec Cmt Support secondary cmtainlent fmctim. ZiD Brkinsid~tLRad COEF 3-1 Rad Release 27 Eg Sec Cmt Support secmdary cmtai~t fmctim. 2b Fuel Handl Accidnt A 3-1 Ihd Release 27 Est Sec Cmt Support secmdary ccntainnent fmction. 33 Loss Fuel Pool CI ALL 2-1 Rad Release 27 Est Sec Cont Support secmdary cmtainaent functim. IS Loss Off<ite AL'OEF & Cont Stress 30 Cmt Cooling SF Provide <<ater level indicatim and pernissive signal that 213 core coverage is aaintaimd Ifl M, 7l, spray node). Brkinsi~Nbd 3-4 Cont Stress 3D Cont Cooling SF Provide indicatim aand peraissive ZiD COEF signal Ito ~ 74) that 213 ccrc containant and tcrus spray nodes). SSO Safety Shutdoto IXEF & Cont Stress 3O tp Ibftppjffii&cIirc p eIceti~f!;, ~ . I sprayo Rl SAFE 90)NN AHALYSIS Page.)22 of 3 7 C c hler:~M E. Ec Oatex TF)HBS3~ IRev 1) C. L Parks'E)IT REACT MXEPTiKE SAFE OR 5'ESIITS SAFHY ACT151 RK SEO. BE)IT TlTLE STATE Code Title Code, Title CXE O3 IIain Stn lsolatim COEF 34 Pers Overexp 36 Cmt Bay Env SF Provide lcxx,uater level IQ) signal to RPS systen N) fcr initiatim of cmtrol bay isolatim. This safety . acticn is expected to occur, but is not a requireaent fcr this event since there is no fuel failure. 05 Loss of Vanan IXEF 3-5 Pers Overexp 36 Cant Say Gw SF Provide low water level IQ) signal to RPS systen N) for lnl'tlatim of ccntrol bay isolatim. This safety acticn is expected to occur, but is not a requirenent fa. this event since there is no fuel failure. 13 lnadvert Open SRV DKF 3-5 Pers Overexp 36 Cmt Say Env SF Provide lw uater level IQ) signal . ~ to RPS systen N) fcr initiatim of cmtrol bay isolatim. This safety actim is expected to ocnr, but is not a requireaent fcr this event since there is no fuel failure. 14 Loss of FM Floe IXEF 3-5 Pers Overexp 36 Cmt Say Env SF Provide its uater level IL3) signal to RPS systen N) fcr initiaticn of ccntrol bay isolaticn. This safety actim is expected to occur, but is not a requirenent for this event since there is no fuel failure. 15 Loss OfHite AC ALL 3-5 Pers Overexp 36 Cent Say Env SF Provide lou rater level IQ) signal to RPS systen N) for initiatim of ccntrol bay isolatim. This safety actim is expected to occur, but is not a requireaent fcr this event since there is no fuel failure. 23 F)I Ctrl Fail~ COEF 3-5 Pers Overexp 36 Provide low uater level IQ) sx t fPQ~t I ) for initiatim of Hefb i / ljIimji e jest/ (~Pg-'jj ( not a rYqulreI&k since there is no fuel failure. hjs.md> I,, S)UMI ANALYSIS bye SAFE Prepved) E C 123 lIitn ~z/ Checked: 0t~ NH)SS3~ (Rev I) C. L Parker GPENT REACT NGXHABLE SAFE N CPER RESILTS SAFETY ACTIII FIK SEL 8M TITLE STATE Code Title Code Title IIE 24 Rad Orap Accident IF 3-5 Pers Overep 36 Cent Bay Env SF Provide law water level IQ) signal to RPS systen 199) for initiatim of control bay isolatim. 25A Larae Break inside EKF 3d Pers Overesp 3f Cant hy Env SF Provide low water level IQ) signal to RPS systen Igg) for initiatim of cmtrol bay isolatim. GKF 3-5 Pars Overezp 36 Cmt hy Env low water level IQ) signal Intern Brk inside 'oProvide systen 258 SF RPS Igg) fm initiatim of cmtrol bay isolatim. 25C Snail Break inside IXEF 3-5 Pers Overep 36 Cant hy Env SF Provide low water level IQ) signai to RPS systen Igg) for Inibatim of cmtrol bay isolatim. 250 Brklnside4mt4Rad CDEF 3-5 Pers Overexp % Cant hy Env 9: Provide low water level IQ) signal 'to RPS systen Igg) fcr Inltiatim of cmtrol bay isolation. 27 Pipe Break Outside CBEF 3-5 Pers Overexp 3/ Cont hy Env SF Provide law water level IQ) signal to RPS systen N) fcr initiation of cmtrol bay isolaticn. 28 Shtdn Fran Outside ALL 4-1 Local Shut 3g Cmldn 0utsd Reactcr vessel water level indicatlm is provided at the backup control center ~ 28 Shtdn Frow Outside 'OEF 4-1 Local Out 39 Caoldn Outsd Reactcr vessel pressure indicatim is provided at the backup cmtrol centero rsll 4AM c&C'+/ 4 fST g~gWd.~g core. cwcr>> ~y,to ~ a Shtdn Fros Outside COEF 4-1 Local Shut 3g Cmldn Outsd 4~ de/caw)4~~f Provide path far RC1C I71) flow sel through the fesdwater / spargers 29 Shutdwn N/0 Rods OF 4-2 Shut No Rads 42 g 'a+%rngy I I gt)~I ilylj@i~!Ig i '.; jq 29 Shutde N/0 Rods BF 4-2 Shut )Io Rods 42 Provide high reactor pressure signai 0 drive not cr breakers I2lO, electrical systee) I trip of recirculatim pusps. BUEQN NAYS)S Page 12i of 3 7 SAFE Dl)NCT)N SIM SSIEN hspwsd: E.C Ec ate: ~M Checked: NHSQ~ IRev 1) C. )i. Parker BM REACT MXEPTlKE SAFE IR IPB 8ESN.TS SAFETY ACTTN RK SEg. BM TlTLE STATE Code Title Code Title imE 06 Loss RI Heater F 2-2 Fuel Failura 2E Lie Msturb L c~ f~44~design c4eee4~as~lt eeets 100 degree F assuaption for aaxiam feeduater teeperatN e drop that can occur fcr any single action N failure of F)i heater Is). See also F)IIIG) and CondensateI02) Systee. R~ SAFE 9//TOQMI AMIS Page 125 Prep ved: Bate: BOILER 0RAB6, VE)IIS Ii Checkedl Parks'o7 9ate: BRH)S63~ IRev 1) C. )I. SSTE)I 10 EVE)IT REACT MXEHARE SAFE Im IPER RESILTS SAFETY ACTICH FlK SE0. EVE)IT TTTLE STATE Code Title Code Title C00E 01A 6eneratcr Trip F 2-3 Syst Stress 18 Pres Relief Provide path fa aain stean systen I01) SNs stean blodoe to suppressim pool I6l). 018 T/6 Tri~ Fails F 2-3 Syst Stress 18 Pres Relief Provide path for nain stean systea I01) SNs stean blcadon to suppressim pool I6l). 01C PresReg Fail&ose CDEF 2-3 Syst Stress 18 Pres Relief Provide path fcr aain stean systen . I01) SNs stean blodaa to suppressim pool 16t). 02 Turbine Trip F 2-3 Syst Stress 18 Pres Relief Provide path fcr nato stean systen I01) SRVs stean blondan to suppressim pml IN. 03 IIain Stn isolation IXEF 2-3 Syst Stress 18 Pres Relief Provide path fcr aain stean systen I01) SRVs stean bloudon to suppression pool IN. 05 Loss of Vaaaa F 2-3 Syst Stress 18 Pres Relief Provide path fcr nain stean systen I01) SRVs stean blodao to suppressim pool N). 12A Pres Reg Fail+en CBEF 2-3 Syst Stress 18 Pres Relief Provide path fcr nain stean systen I01) SRVs stean blodoun to suppression pml IN. 128 lnadv 0pen~ass CCEF 2-3 Syst Stress 18 Pres Relief Provide path fcr nato stean systen I01) SNs stean bloudcun to suppressicn pool IN, 13 Inadvert 0pen SN CKF 2-3 Syst Stress 18 Pres Relief Provide path fcr nain stean systen IOl) stean blcwdoun fran stuck open SN to suppressim pool IN. 15 Loss Offdite AC Rel Provide path fcr nain stean 'I/~P rQ$ $ iQ/ri, a,. FM Ctrl Fail~ F 2-3 Syst iL Pres llalisf Ptovidl)a5'l'ah rjfsu rifI'" - ~ I01) SNs stean blcaaho icn pool IN; to SAFE SHJIHN NOESIS Page'126 of Prepared: Bate: > e BOILER ORAIHS, VERBIS Ir %HE~ (Rev I) STSIE)( 10 Seckedc E C. L Parker Oate: 9'VD(T REACT MXEPTARE SAFE CR OPER REM.TS SAFETy ACTIN RK SEL EVE)(I TITLE STATE Code Title Cade Title GEE 24 Rod Orop Accident OF . 3-3 Syst Stress 18 Pres Relief Provide path for aain steaa systea (01) SRVs steaa bloadoe to suppressim pool (Q), 25C Saal 1 Break Inside (XEF 3-3 Syst Stress 18 Pres Relief Provide path fcr aain steaa systea l01) SRVs stean bloadcae to suppresrim pool N). 27 Pipe Break Outside COEF 3-3 Syst Stress 18 Pres Relief Provide path for aain steaa systea (01) SRVs stean blodtae to suppression pool N). lAlso suppcrt AOS discharge in Safety Action SI9, if needed). COEF 2-3 Syst Stress 18 Pres Relief Provide path for aain steaa systea (01) SRVs steaa blcwdoe to suppressim pool (M). 328 Los Reservoir COEF 2-3 Syst Stress 18 Pres Relief Provide path fcr aain steaa systea (01) SRVs steaa blaue to suppressim pool N). 22 Loss Shutdown Cool CO 2-2 Fuel Failure 19 Ccrc Cooling Provide path for aain steaa systea(01) SRVs steaa blcaa)oan to suppressim pools ZB Intera Brk Inside 'OEF 3-2 Fuel Failure 19 Ccrc Cooling Provide path for aain steaa systea (Ol) reactor vessel autoaatic depressuriaatim systea (AOS) steaa blondie to suppressicn pool (bl). ZC Saal1 Break Inside COEF 3-2 Fuel Failure 19 Core Cooling . Provide path for aain stean systea (01) reactor vessel autoaatic depressuriaatim systea (AOS) stean bloadoan to suppressi ). ISO ISO Shutdoe Provide path for aain steaa t e l' sur at c tljj r 160 Noraal Shutdso s 50 t rssrrisrr Provide path fcr aain steaa syst /' (01) oa cmtrolled aanual depressurczatcm of RPV) to suppressim pool %E BUHOM QNLYSIS Page 127 of Prepared: SOILER NQll8, %HIS b B.MN SYSTEII Chested t Date! SRHKH~ IRev I) SYSIBI 10 BE)IT REACT MXEPTAKE SAFE OR IPER RESILTS SAFEIY ACTTN FIK SED. EVERT TlTtK STATE Code Title Code Title CME SSD Safety Shutdoe IXEF 3-2 Fuel Failure 30 Cont Coolioy Provide path for nain stean systea I01) SINs stean bloudao to suppression pool N). ~ 28 Shtdn Froa Outside COEF 4-1 Local Shut 9 Cooldn Outsd provide path for nain stean systen I01) SRVs stean blanche to suppression pool IM). SAFE SHIIT% ANALYSIS PaLe 12S of Prep ved: Date: AUIILIARYSDILER SYSTEN F C.Ec Checked: Date! iRHM~ (Rev I) C. N. Parker SYSTEN 12 BM ISED. BJEIIT TITLE REACT IHAIXEPTASLE lRR RESULTS STATE Cade Title . SAFEIY ACTION Code Title SAFE RK CME 25D Srklnside+mtSad CDEF 5-I Rad Release 26 Est Pri Cont Supper t prleary cmtainMlt functim 15 Loss Dffdite AC ALL 2-1 Rad Release 27 Est Sec Cant Support secmdvy cmtalneent functim. 25D Srklnside4atSad CDEF 5-I Rad Release 27 Est Sec Cmt Support secmdvy containaent functim. 26 Fuel Nandl Accidnt A 5-I Rad Release 27 Est Sec Cant Support seccndary containment functicn. Loss Fuel Pool CI ALL 2-1 Rad Release 27 Est Sec Cant Suppcrt secondvy cmtaineent fmcticn. SAFE QUINN N(ALTSIS 129 flKL OIL SySIE)I Prepared: E. C Ec Itete: ~m7 Checked> NHM~ (Rev 1) C. L Parker 'EACT SYSIE)I 18 EYB(T IMXEPTARE SAFE N OPER RESILTS SAFEPl ACTIQI FIK SEQ. BM TITLE STATE Code Title Code Title CNE 13, Inadvert Open SRV COEF 2-2 Fuel Failure 19 Core Cooling SF Provide diesel fuel oil to diesel generate systN (8?) starting m lw reactcr water level (Ll) signals. This safety actim nay occur, but is not required by this event since noma) ac is available. 15 Loss Offdite AC C0EF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide diesel fuel oil to diesel generahr systN (8?) starting m Iow reactor water level ILI) signals. 24 . Rod Ory Accident OF 3-2 Fuel Failure 19 Core Cooling SF Provide diesel fuel oil to diesel generatcr systN (8?) starting m Ical reactcr water level IL1) signa)se 25A Large Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide diesel fuel oil to diesel generatcr systN IB?) start)+ cn high drywell pressure signal. 25A Large Break Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide diesel fuel oil to diesel generatcr systN (8?) starting m Ia reactcr water level (Ll) signal. 258 Intern Erk Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide diesel fuel oil to diesel gemratcf systN (8?) starting m Iw reactcr water level (Ll) signals. h 258 Intern Brk Inside C0EF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide diesel fuel oil to diesel generatcr systN (82) s'tartlng m high drywell pressure signals. 25C Seal I Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide diesel fuel oil to diesel generatcr systN (8?) starting cn low reactcr water level (LI) signals. 25C Snail Break Inside 3-2 Fuel Failure 19 Core Ccoling SF Provide diesel fuel oil to diesel "~Pl.,J)h~VQ "',<<',"~If 27 Pipe Break Outside 3-2 Feet 2ei teeH teWteolie(32 'ete(tltedree(t ftteteett ieetteeet g8lerats'ystN (8?) starting m col reactcr water level ILl) signals. SAFE SWTNN ANALySIS of 3~7 Page 130 'repared: -C Date: RKL 01L STSTE)( F C Ec Checked! Date: SF)HIS83W8 (Rev I) C. N.'Parker SYSTE)( 18 BM I SE(). BBlT TITLE REACT [PE MXEPTAILE STATE Code RESETS Title SAFETT Code ACTIN Title SAFE RK CODE '2E Fire Nl fH Fuel Failure 19 Ccrc Cooling Design of safety features est include Appendix R fire consideraticns. SSD Safety Shutdoe CDEF 3-2 Fuel Failure 19 Ccrc Cooling SF Support diesel generatm systea (82). 15 Loss Dff-Site AC CDEF M Cent Stress 30 Cent Cooling Provide suppcrt to diesel generatcr systen (82). 250 8rklnsidHmtM(ad CDEF 3H Cont Stress 30 Cont Cceling SF Provide fuel oil to diesel genaratcr systen (82). 32E Fire CDEF f-6 Cent Condtns 30 Cont Cooling Design of safety features est include Appendix R fire ccnsiderations. SSD Safety Shutdoe CDEF 3H Cent Stress 30 Cant Cooling SF Supp(rt diesel generator systee (82). 28 Shtdn Fron Outside kL H Local Shut 39 Cooldn Outsd)( Provide diesel fuel oil to diesel fc'. generator systee (82) if ncraal ac poler is lost. 320 Ear thqua'ke ALL 1-3 Equip Fail 4 Seisaic Shtd Design of safety features aust have capability to provide shutd(xe fotlcwing a design basis seiseic event. SAFE SHJlHN AHALYS]S P e 131 Qo7 'I Preparedl ( Ilats 'E~J>~/ M SE)wlCE NTER STSTEN E. Checked: Oate: BF)H)663~ (Rev 1) C. N. Parker 6E)fI REACT MXEPTABLE SAFE OR IPER REStlTS SAFETT ACT lN RK SD). MIT T)TLE STATE Code Title Code Title COOE Loss Shutdoo Cool ACO 2-2 Fuel Failure 22 Rest Horaa) )) Suppcrt operation of M174) Reactcr Shut~mling. 25D Brklnsid~tMhd COEF 3-1 Rad Release 2b Est Pri Cont Support priaary ccntainaent functim. 15 Loss Offdite AC ALL 2-1 Rad Release 27 Est Sec Cmt Support seccndvy cmtainaent function. 2% Brk)nside+xtSad COEF 3-1 Rad Release 27 Est Sec Cont Support seccndary containaent functim. 2f Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont Support seccndary cmtainaent functicn. Supper t secmdary cmtalneent function. 0)B T/6 Trip-BP Fails F 2-4 Cont Stress 30 Cent Cooling ))SF Suppcrt M systee f74) suppression pcol cooling axle. 01C PresReg Fail<lose . CDEF 2-4 Cont Stress 30 Cont Ccoling ))SF Suppcrt M systea <74) suppression pcol cooling node. 03 Nain Sta Isolation COEF 2H Cant Stress 30 Cent Cooling lSF Suppcrt M systee f74) suppression pool cooling node. 05 Loss of I/acuua COEF 2H Cmt Stress 30 Cent Ccoling lSF Support M systee f74) suppressicn node. 12A Pres Reg Fail+en COEF 2-4 t fygnp,wf 5upsat IN systse I') suppressim gk ~ I 128 lhadv Open Bypass COEF 2-4 t Stress pcol cooling node. 13 lnadvert Open SRI/ COEF 2H Cmt Stress 30 Cmt Cool ang p pool cooling node. 15 Loss Off Site AC CDEF 2H Cont Stress 30 Cont Cooling NSF Support M systee l74) suppressicn pcol cooling aode. SAFE BUESN A)4)LYSIS Page i32 37 0 Preparedl Oate: M %NICE )(AIER SYSTEN E. Ec Checked: Oate: BF)H)SS3~ (Rev I) C. L Parker BM REACT (NCCEPTAKE SAFE OR OPER RES(LTS SAFEIY ACTIN FlK SEO. EVENT TITLE STATE Code Title Cade Title C(XE I5 Loss Of flite AC COEF 2H Cont Stress 30 Cont Cooling NSF Rqpcrt M systen (74) dryuell spray ~ ode of dryuelI cooling (if necessary) ~ 22 Loss Shutdoe Cool CO & Cmt Stress 30 Cmt Cooling N Support operatim of M(74) pool cooling'F 24 Rod Orcp Accident 3-4 Cent Stress 30 Cmt Cooling NSF Suppcrt M systea (74) suppression .pml cooling ecde. 25A Large Break Inside COEF 3H Cent Stress 30 Cent Cooling NSF Support M systee (74) suppressicn pool cooling code. 258 Intern Brk Inside COEF 3-4 Cmt Stress 30 Cont Cooling NSF Support M systen (74) suppressicn pool ccoling node. ZC Seal( R eak Inside COEF 3H Cmt Stress 30 Cmt Cmling NSF Support M systen 174) suppressicn pool cooling node. 25O BrklnsidHmtUbd COEF 3-4 Cent Stress 30 Cent Cooling NSF Support M systee (74) suppressim pool cooling node. 27 Pipe Break Outside CDEF 3-4 Cont Stress 30 Cont Cooling NSF Suppcrt M systee (74) suppressicn pool cooling node. 328 Lcw Reservoir COEF 2-4 Cooling NSF Support M systea (74) suppressim e. 32E Fire a ISO ISO Shutdoo ress X Cent Cralin9 l6F Su@art BR systems (7l) suppr48n/ I cooling node. Suppcrt dryueil cooling un 1 NSO lorna) Shutdoe COEF 2-4 Cont Stress 30 Cont Cooling NSF Support M systen (74) suppressim pool cooling node. Support dryuell cooling functim. SSO Safety Shutdown CKF 3H Cmt Stress 30 Cmt Cooling le Suppcrt M systen (74) suppressim pcol cooling node and/cr dryueil .I spray lodh SAFE 9UTHN OQLYSIS Page 133 /of Prepared: lhta ~~X+ 7 M %NICE NTER SYSTEII E: Pecked: Date: SS SRHSQOQ (Rev I) .C. IL Parks DIEHT REACT MXEFTARE SAFE 0R I PER RERLTS SAFEIY ACTIN FNC SEQ. BERT TITLE STATE Code Title Code Title CEE 28, Shtdn Froa 0utside Nl H Local Shut 3g Cooldn 0utsd If RIB Service l4ter Systee required fe operatim of M in all nodes. ALL q-5 Equip Fail 0 Struct lnteg Provide wheeler Lale level indi catim/al are at elevatim 55B feet and rising, as all as at elevatim 56l feet and rising. ALL 3-I Rad Release qI Drainage Provide iheeler Lake level indication/alvn at elevaticn 558 feet and rising, as nell as at elevatim 56l feet and rising. 20 Ear thquake Nl l-5 Equip Fail l4 Seisaic Shtd Design of safety features ast have capability to provide shutdoun following a design basis seisaic event. 3c$ 7 SAFE SIJTHN AN.YSlS Pa9e VA Prepared: C bate: Nt Cm.1% NTER SySTBl F Checked: bateau IRHSS3~ IRev 1) C. L Parker BW REACT MXPTAKE -SAFE N CPER RESlLTS ~ QFEIY ACTlN FllC SER. EIBIT TlTLE STATE Code Title Code Title IE 15 Loss bffdite AC tU. 2-1 Rad Release 27 Est Sec Cont Support secondary containeent function. 250 Brklnside+mtSad CCEF 3-1 Rad Release 27 Est Sec Cont Suppcrt seccndvy ccntalneent functicn. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont Support seccndvy containeent functicn. 33 Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont Support seccndary containeent fwxticn. SAFE SNIIHN LYSIS P 135.of )67 Prepared: Oate: . NHE63~ IIhv 1) Checked: E. C. L Pvker Oate: 69/i REACT INKEPTARE SAFE IR IM REStLTS SAFETY ACTTN FlK SEL 88lT TlTlE STATE Code Title Code Title COOE 15 toss Offdite AC ALL 2-1 Rad Release 27 Est Sec Cont Suppart seem) vy containeent functicn. 25D Orklnsi~tlflad COEF 3-1 Rad Release 27 Est Sec Cant Suppart secondary containeent function. 26 Fuel Randl Accidnt A 3-1 Rad Release 27 Est Sec Cant Suppart seccndvy cantaineent function. 33 Loss Fuel Pool CI AU. 2-1 Rad Release 27 Est Sec Cant Support secandilry cantaineent function. QUIN ANN.TS1S D6 nf po7 SAFE Prepired: Page 0ate: ~~ S HIQH%ESRIE FIRE E. N0TECTIN STSTB Checkedy Sate: M 7 SRHM~ fRev 1) C. L Parker BiW REACT MXEPTARE SAFE N 8%R REStLTS SAfEIY ACTIN FBC SHL BM TlTLE STATE Code Title Code Title GEE 1S, toss Offdite AC Nl 2-1 Rad Release 27 Est Sec Cent Suppof't secondary Gxltalneent functicn. 250 lwrklnsi~tNad C0EF 3-1 Rad Release 27 Est Sec Cont Support secondary ccntainnent functicn. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont Support secondary ccntalneent functicn. 33 loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont Support seccndary coltaibwlt fmtim. SAFE BITKN ANALYSIS P eD7of $ Q7 Prepared: 0ate: CBl096B CIRQLAT)IS E. NTER SYSTEN Checked: 0ate: IRHE63~ (Rev l) C. N. Parker I BQT SB). BENT TlTLE REACT 6%R MXEPTARE REStLTS STATE Code Title SAFETy Cade ACT)N Title SAFE RK C00E ZS Lcw Reservoir 0F 2-l Rad Release 17 Scran Provide indication of lm fcrebay level in the control roon for paar reduct10l I') lf necessary scraL ZS trw Reservoir C0EF 2H Cont Stress R Cant Cooling N Nanually break vacuue to prevent backflw into the flebay. 32B Ltw Reser voir ALI 24 Cant Stress 30 Cant Cooling SF Trip cooling tour lift puep if discharge teeperature is excessive. SAFE SHlllNl A%LYSIS DB Qo7 Page Prepared: C bate; POMLE lTREATEb) NTER E. er 0WRINlllN SYSTE)1 Checked: bate: NHEM~ IRev 1) C. )L Parker SYSTE)l 29 BM REACT MXEPTAKE SAFE OR CPER REStLTS SAFHY ACTlQI RK SHl. BDIT TITLE STATE Code Title Code Title IE 15 Loss Offdite AC Nl 2-1 Rad Release 27 Est Sec Cont Support seccndary ccntaineent function. 25D Brklnsid~tNad COEF 5-1 hd Release 27 Est Sec Cont Supper't secondary contalM%t functim. 26 Fuel Hand) Accidnt A >] Rad Release 27 Est Sec Cont Support seccndary containeent functicn. 9 Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont Support seccndary ccntainaent function. i SAFE SHmXN NNLYSIS p Prepared: Oate: + . UENTILATIN SYSTBI ERHKS3~ (Rev I) Checked: E. C. er IL Parker Sate: BE)IT 8EACT MXEPTAKE SAFE N OPER RESEITS SAFEIY ACTIN RK SEQ, BBlT TITLE STATE Code Title Code Title COOE 2S, Shtdn Froa Outside ALL 4-I Local Bit 3g Ceo)do Outsd Support systee required. 3X Tornado ALL 4-3 Equip Fail 0 Struct Integ L Suiiding eahaust fans and ducting cn the roof ast ulthstand full tornado design loads fpassive). 320 Earthquake NI l-3 Equip Fail 4 Seisaic Shtd Oesign of safety features aust have capability to provide shutdoe following a design basis seisaic event. . ~ pp) Y J .. W~w I $ 4gtl qr ~ 'I 4 ~ ~ g1 f ~ ~ loof Qc7 SAFE QUEM NIALTSIS AIR CNDITICNI% Prepared: P E.. e C hter ~~r ICIXLINrKAT1%) SSTBI s'hecked; Date: TF)H)SS3~ IRev 1) C. L Parker STSTE)I 31 EVE)IT REACT LNXEPTl8E SAFE CR IM RESI1.TS SAFEPf ACTIN RK SEO. EVE)IT TITLE STATE Code Title Code Title CKE 15 Loss Off<ite AC ALL 2-1 Rad Release 27 Est Sec Cont Support secondary containaent function. ZD Srklnside4mtNad CDEF 3-1 Rad Release 27 Est Sec Cont Support secondary containnent functicn. 21 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont Support seccndary ccntainnent functicn, 33 Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cent Support seccndary containment functicn. D3 Nain Stn isolation CDEF 3-5 Pers Overexp 36 Cont Day Env SF Air ccnd. IAC) supply ducts isolate I berg. Pres. Systen IControl Rcon berg. Vent. Systen) supplies pres. ,0 filtered air to KR cn LM. IQ) signal fron systen 6i. This safety action is expected to occm, but is not a requirenent for this event. 05 Loss of Vacuua CDEF 3-5 Pers Overexp 2 Cont hy Env SF Air cond. IAC) supply ducts isolate I berg. Pres, Systen (Control Roon berg. Vent. Systen) supplies pres. filtered air to KR cn N. IQ) signal froa systen bl. This safety action is expected to occur, but is not a requirenent for this event. 13 lnadvert Open 9N CDEF 3-5 Pers Overexp 3I Cont Day Env SF Air cond. IAC) supply ducts isolate h es. ys en ro berg, Vent. Systen) supplies pres. I jfcngph q. j ii] jHi!)!js:.!aspic)g to~84!$ not a requirenent fcr this event. lf Loss of F)I Flaw 2-Cent Day Env SF--4r"cond~HAD)-supply-berg, Pres, Systen Kcntrol Roon berg. Vent. Systen) supplies pres. filtered air to KR cn L)L IQ) signal fron systen M. This safety acticn is expected to occur, but is not a requirenent fcr this event. Sf'HllMNANALYSIS p Il) of Prepared; Date: AIR GNDITIN1% E. c ert ICm.l &%N1%) SYSTE)I Checked: D te: BF)H)SS3~ IRev I) C. IL Parker EVENT REACT MXEPTABLE SAFE IS CPER RERLTS SAFEIY ACTICII RK SED. EVENT TITLE STAlE Code Title Code Title GXE 15, Loss Dff<ite AC ALL 3-5 Pers Dverexp 36 Cont Bay Env SF Air ccnd. NC) supply ducts isolate 6 Eeerg. Pres. Systee IControl Rcoa Eaerg. Vent. Systee) supplies pres. filtered air to IKR m LN. (Q) signal fron systeo 64. This safety acticn is expected to occur, but is not a requireeent fcr this event. 23 F)f Ctrl Fail~ CDEF 3-5 Pers Overexp 36 Cont Bay Env SF Air ccnd. NC) supply ducts isolate 6 Energ. Pres. Systen ICmtrol Rona Eeerg. Vent. Systen) supplies pres. filtered air to )KR on LM. (Q) signal froa systea 64. This safety action is expected to occur, but is not a requirenent for this event. 24 Rod Drop Accident EF 3-5 Pers Dverexp 36 Cont Bay Env SF Air conditioner IAC) supply ducts ast isolate and Eaergency Pressurization Systen ICmtrol Roon Eaerg. Vent. Systea) aust supply pressurized filtered air to nain control roca on low water level (Q) signal froa Prinary Ccntainaent Systen 164). 2l Rod Drop Accident EF 3-S ~Dz+nv, free R)6190), AC supply air ducts /ST kl gtr 'Q ccntrol roon, 2M Large Break Inside CDEF 3-5 Pers Overexp 36 Cent Bay Env SF Air conditimer NC) supply ducts ~ ust isolate and Eaergency Pressurization Systea (Cmtrol Roon Eaerg. Vent. Systen) aust supply pressurized filtered air to nain control roon on low water level IQ) signal froe Prinary Containaent Systen f64). i SAFE SNf0M A)(ALYSIS P ll2 of Prepared: lkte: ~Blot/ AIR CNOITIN((% E. Ec ((m.l &HEAT)%) STSTE)( Checked: Bate: 7 BF)H)SS3~ (Rev l) C. )L Parker STSTB)( 3l EVE((T REACT Q(A(XEPTAKE SAFE OR 5%R KklTS QFHY ACTINI RK SE1 EVD(T TITLE STATE Code Title Code Title CNE '25A Large Break Inside COEF 3-5 Pers Ovirexp 36 Cent Bay Env SF Air conditioner (AC) supply ducts <<st isolate and bergency Pressurization Systea (Control Rx>> Energ. Vent. Systea) aust supply pressurized filtered air to,aain ccntrol roon on high dryweII pressre signal fron prinary containnent systen (BS). 258 Intern Brk Inside CDEF 3-5 Pers Overexp 3( Cont Bay Env SF Air ccnditioner (AC) supply ducts <<xst isolate and bergency Pressurizaticn Systen Kcntrol Roon berg. Vent. Systen) aust supply pressurized filtered air to aain control roon on Ix>> water level (L3) signal frco Priaary Ccntainaent Systen N). 258 Intern Brk Inside (XEF 3-5 Pers Overexp 3f Coot Bay Env SF Air ccnditioner (AC) supply ducts ~ ust isolate and bergency Pressurization Systen (Ccntrol Roon berg. Vent. Systen) est supply pressurized filtered air to nain control roon on high dr pell presare signal froa prinary ccntainnent systen (M). ZC Snail Break Inside (XEF 3-5 P Air ccnditicner (AC) supply ducts aust isolate st en f@ly iaaf)! ed r to'a5n control roc>> on low water level (L3) signaI fro>> Prinary Contai 2X Sna(l Break Inside COEF 3-5 Pers Overexp 36 Cont Bay Env SF Air conditicner (AC) supply ducts ast isolate and bergency Pressurizaticn Systen Kontrol Roon berg. Vent. Systen) aust supply pressurized filtered air to aain ccntrol roc>> on high drywell prasare signal fron priaary cmtainant systen N). CI7 SAFE BILDNN ANWILYSIS Page IQ of Prepared: AIR CNDITINIHS ICILINRKATI%)SYSLE)I Checked: Date: M P7 BF)H)SS3~ (Rev I) C. )f. Parker EVENT REACT MXEPTABtE, SAFE OR CPER RESILTS SAFELY ACTIN RK SEL EVE)IL TIL1E STATE Code Title Code Litle IE '25D Brklnsid~tNad CDEF 3-5 Pers Overexp 36 Cent Bay Env SF Air ccnditioner IAC) supply ducts ~ ust isolate and Energency Pressurization Systea ICcntrol Roon berg. Vent. Systen) est supply pressurized filtered air to nain ccntrol roon on IosE ssater level IL3) signal froa Priaary Containaent Systen N). 25D Brklnsid~tlRad CDEF 3-5 Pers Dverexp 36 Cont Bay Env SF Air conditicner (AC) supply ducts oust isolate and Eaergency Pressurizaticn Systea IControl Roon Energ. Vent. Systen) <<Est supply pressurized filtered air to nain -0 ccntrol roon on high drysseII pres@re signal fron prinary ccntainnent systen 161). ZD Brklnsid~tNad CDEF 3-5 Pers Overexp 36 Cont Bay Env SF Air ccnd. <AC) supply ducts aust isolate k Energ. Pressurization Systen ICcntrol Rcon Energ. Vent. Systen) an't supply pressurized, filtered air to KR cn high radiation in reactor building exhaust Isysteas b41 90) cr control bay inlet 190 26 Fuel Handl Accidnt A 3 5 s Dverevp SS Sent hy Env SP Eeerten y 'ty Sypttnvt tt tv t S tent 'QI 'yp IES ssPtntet Sestet tn pressurize KR to prevent inleakage of radioactive aa p y is being provided. 26 Fuel Handl Accidnt A 3-5 Pers Overexp 36 Cont Bay Env SF Air conditioner IAC) supply ducts isolate and the AC systen supplies filtered air to the nain ccntrol roon shen high radiaticn initiation signai is received froa systea M (reactor . building) cr systea g0 KR air inlet ducts. SAFE SHJIHN ANALYSIS Page I44 $ 47 Date:~ ~ e Prepared'. AIR CMITIWI% (GXLIMrKAT1%) SYSTE)( Checked: Date: BRHKS3~ (Rev I) C. L Parker EVBIT REACT (N(XEPTABLE SAFE OR KR RESETS SAFEIY ACTI(N RK SEQ. EVENT TITLE STATE Code Title Code Title CNE 27 Pipe Break Outside C0EF 3-5 Pars Overexp 36 Cent Bay Env SF Air ccnditicner (AC) supply ducts ~ ust isolate and Energency Pressurization Systea'Control Roon Energ. Vent. Systen) Nxst supply pressurized filtered air to aain control roon cn Isr water level (L3) signal froa Prinary Containnent Systea (64). 27 Pipe Break Outside 'DEF 3-S Pers Overexp 36 Cent Bay Env SF Air Conditioning (AC) supply ducts aust isolate L Energency Pressurization Systea (subsysten of AC systen) Nst supply pressurized filtered air to aain ccntrol roon nhen radiation levels above noraal ,0 are sensed by systen ducts. g() in AC inlet 33 Lo:s Fuel Pool Cl AtL 3-5 Pers Overexp 36 Cent Bay Env SF The Control Building energency pressurization function provides clean air to the control roon after receiving signal fron priaary containnent system (64). 28 Shtdn Froa Outside ALL 4-1 Local Shut 3g Cooldn Outsd Support systen required. 32D Earthquake ALL 4-3 Equip Fail 44 Seisaic Shtd Design of safety features <<st have capability to provide shutdao foll@sing a design basis seisnic event. SAFE S(ITIN N(ALYSIS CIM. AlR SYSTE)( SF)HKS3~ (Rev 1) Prepared: Checked: ~ P E.C C. e145 C Ec L Parker gal Date: Date: ~ ~W BtENT REACT MXEPTAKE SAFE 0R (PER RESKTS SAFETY ACTli F(N; SEL BM T1TLE STATE Code Title Code Title GXE 03 'lain Stn lsolatim C0EF 2-1 Rad Release 26 Est Pri Cent SF Perfcrn isolatim actim(s) upon receiving lw uater level (Q) isolation signal fron the Priaary Cmtaimnt systen (M). This safety acticn is expected to occur, but is not a requirenent fcr this event since there. is no fuel failure. 05 Loss of Vacuua C0EF 2-1 Rad Release 26 Est Pri Cont SF Perfora isolation acticn(s) upon receiving lw water level (Q) isolatim signal fron the Primy Cmtainaet systen (6(). This safety actim is expected to occur, but is not a require+et fcr this event since there is no fuel failure. 13 Inadvert Open'SRV C0EF 2-( Rad Release 26 Est Pri Cent SF Perforn isolatim action(s) upon receiving lw water level (Q) W. ~ e isolaticn signal fron the Prinary Cmtainnent systen IN. This safety ac'tllxl ls evpected 'to occurD but 1$ not a requirenent fcr this event since there is no fuel failure. ll Loss of Rl Flw C0EF 2-1 Rad Release 26 Est Pri Cmt 5 Perfcra isolaticn acticn(s) upm receiving lw uater level IQ) isolatim signal froa the Priaary Cmta(naent systen (6() ~ This safety actim is expected to occur, but is aot a requireeent fcr this event since there is no fuel failure. 15 Loss Off-Site AC Nl 2-1 R ad Release 26 Est Pri Cmt SF Perfmn isolat') upon w uater level (L3) isolatim $ (goal ff hggi~g)en 6h >,. sa y eeet' )tttpaai eapeana tt ((DD t y)p k >',ai~'.$ ngt (requjre(en%for"this event ', s '7, ',ginsecthtre is no fuel failure. att F)I Ctrl Fail~ hale 3b Est Pri Cont SF Perforn 'HorCactim(s) upm e>ving lw uater level (Q) e isolation signal fron the Prinary Ccntainnent systen N). This safety action is eapected to occur, but >s sot a require+et fcr this event since there is no fuel failure. ' e ~ ~ ~ ~ ~ (IT(M AN.YSIS Ilbof 3 07 ~ SAFE P e Prepared: CN((HL AIR SYSTE)( Checked: Sate: BF)H)SS3~ (Rev 1) C. IL Parker BE((T REACT MXEPTARE SAFE (tR KR = REM.TS - SAFETY ACTIN FlK SEL 6E((T TITLE STATE Code Title Code Title IE 2l, Rod Brop Accident BF 3-1 Rad Release 26 Est Pri Cant SF Perfcre isolaticn action(s) upon receiving lw water level (Q) isolaticn signai fron the Prinary Ccntainaot systee (6I). 25A Large beak Inside IEF 3-1 Rad Release 26 Est Pri Cmt SF Perforn isolaticn actim(s) upon receiving lw ggater level (Q) isolatim signai fron the Prieary Containeent systee (6I). 25A Large Break Inside NEF 3-1 Rad Release 26 Est Pri Cont SF Perfcra isolaticn acticn(s) upon receiving high dryggell pressge e isolatim signal froa the Priaary Contain+et systen (6I). 258 Intern Brk Inside (XEF 3-1 Rad Release 26 Est Pri Cmt SF Perfcre isolatim acticn(s) upon receiving lw ggater level (Q) isolaticn signal froa the Priaary Cmtalneent systen (6I) 258 Intern Brk Inside CKF 3-1 Rad Release 26 Est Pri Cont SF Perfore isolaticn actim(s) upm receiving high dryuell pressure isolaticn signai fio~ the Priaary Cmtaineent systea (6I). 25C Saall 8'elk Inside COEF 3-1 Rad Release 26 Est Pri Cont SF Perfcrn isolatim actim(s) upm \ receiving lw uatsr level (Q) isolation signal fron the Priaary Containeent systen (6I). r 25C Seal) Break Inside (XEF 3-1 RadRelease 26 Est Pri Cant cn actim(s) upm rereirieg high grteett greeeere i i igVS(ggeth(gd tape)tr) gc 25() Brkinside+xtLRad CBEF 3-1 9 Pri Cmt Vc ":Wfcrnadolfti(n ictim(s) upon receiving Iw eater level (Q) isolaticn s')aLfc ~ .. Ea~naineent systen (6I). 25() Brklnside-CantNad RadRelease 26 Est Pri Cont SF Perfcrnisolatim acticn(s) upm receiving high dryggell propre isolaticn signal froe the Priaary Cmtaineent systen (6I). 'er, ~ ~ ~ ~ SAFE SRITNN A%LYSIS P 147 Prepared: Oate: GNlRL AIR SYSTEN Checked( KHSM~ (Rev 1) BID(T REACT INGEPTARE SAFE OR SG REM.TS SFETY ACT) 9( F(K SEQ. BIENT TlTLE STATE Code Title Code Title CXE 27, Pipe Sreak Outside COEF 3-1 RadReiease 26 Est Pri Cmt SF Perfore isolatim acticn(s) upon receiving lou eater level (L3) isolatim siqnal fron the Prinary Cmtainant systee (M). 15 Loss Of finite AC kL 2-1 Rad Release 27 Est Sec Cont fmctim. ZD Brklnsid~tSad COEF 3-) Rad Release 27 Est Sec Cont Suppa't secmdary containaet fmctim. 2b Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont Support secmdary ccntaineent functim. 33 Loss Fuel Pool Cl Nl 2-1 Rad Release 27 Est Sec Cont Support secmdary cmtaineent functicn. 32O Earthquake Nl 4-3 Equip Fail 44 Seisaic Shtd Oesign of safety features aust have capability to provide shutdown follorinq a desiqn basis seiseic event, D7 SAFE SIIITONI AN.TSIS SERVICE AIR SYSTBI I)'repired: thected: IRHS&0h8 Ihv BM REACT INNXHARE SAFE N CPER REM.TS %EIy ACTISI FIR SB. SW TllLE STATE Code Title Cade Title CEE 259 Srtlasi~tthd IXEF 5-1 hdRelease 26 EstPri Cant Suppart priavy cmtainaet fmctim. t 15 Loss OIHite AC ALl 2-1 hd Release 27 Est Sec Cant Suppart secmdery cmtainant fanctim. 250 Srtlnsi5e+mtSal 'EF 3-1 hd Release 27 Est Sec Cent Sappcrt secandvy cmteinant fmctim. Fuel Hendl Accidat A 5-1 hd Release 27 Est Sec Cant Sappart secandery cmtelnmt functim. Q Loss Fuel Pool CI Nl 2-1 hd Release 27 Est Sec Cent Support secmdery ccntainnent fmctim. 0 SAFE MIIIMNRLYSIS Page 1I9 y7 Prepared: Date: , NMN PRIIIIIS SYSTBI NKKN~ (Rev I) heckedt E. C. IL Parker Date> BM IGCT MXEPTASLE SAFE N 9%R RESN.TS SAFETY ACTISI flw,'ED. E%NT TITLE STATE Code Title Code Title CIXE 1$ , Loss Nfdite AC Nl 2-1 Rad Release 27 Est Sec Cent Support secandary aataineent fmticn. 25D h klnside+mtQbd GKF 3-1 Rad Release 27 Est Sec Cont Support secondary containant fmticn. 2b Fuel Hmll Acci dot A 3-1 Rad Release 27 Est Sec Cont hppart secondary untainaet functim. Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont Support secondary umtainant functim. SAFE QNMNI NILYSIS 150 of Prepared' hte: MO SEAL INTER SYSTHI Chectedt hte> SFIHK&OQ IRev II C. IL Parter BBIT REACT lNKEPTARE SAFE IN SER RESILTS SAFEIY ACTTSI RK SB. BM TTTLE STATE Code Title Code Title GXE 15 Loss Offdite AC Ail 2-1 Rad Release 27 Est Sec Cant Support seca&ry ccntaiaant fmtim. 259 Brklosi~tifIad CCEF 5-1 Rad Release 27 Est Sec Cont Support seccndary ccntaiaaat function. 2f Fuel Handl Accidnt A 5-1 Rad Release 27 Est Sec Cent Support seccndary untai~t functicn. 2 Loss Fuel Pool CI 'U. 2-1 Rad Release 27 Est Sec Coot Support secondary ccntai~t fmticn. QFE QlAIQM AN.YSIS Page 151 Preparedh STATIC ONllNK S6lP Qwckedt NHKQOl8 (Rev 1) 69T REACT NCXPIARE SR HER RESI.1S SAFETy ACTTII RK Kl. BM TlllE STA1E Cade Title fade Title GK 15, Loss Of&ite AC Ni 2-1 Rad Release 27 Est Sec Cont hpport secondary aatainmt functim. 25D lk'klnsi~URad, EKF 3-1 Rad Release 27 Est Sec Bat hppaA secondary cataiamt hactim. 26 Fuel Handl Accidnt A &1 Rad Release 27 Est Sec Cont Support sectary ccntaineeat functim. 33 Loss Fuel Pool Cl Nl 2-1 Rad Release 27 Est Sec Cont hpport secceiary aataiaaat fictile. ALL 3-1 Rad Release ll Crainage Il Provide valve cioare cr piping geoeetry to prevent >>atlr Ihlcb -0 flooded the base of the stack, standby gas treateent, and off gas building, froe floang into tbe radeaste building. '0 ~ 1 f, 1 ~ ~ SAFE SHIIIOM AN.YSIS 'Pap 1 Xr>4~ Prepared: Datr. SQPUIS 0 NTER IN.ITY E. Ec SYSIE)I Oocked> htn ~~ RHKQ~ IRev 1) C. )L Parker BE)IT REACT MXP?ARE SAFE

6) CPHI REKTS SfEIY ACHSI FIMPS SEL B81T TITtE STATE Code Title "

Code Title NK 2$ Rod Drop Accident lf 3-) hd Release 2b Est Pri Cmt SF CIae saepling and water quality systee isolatim valves m low reactor water level ILI) signal frm prieary cmtaimnt systew Ibl). 2i Rod Drop Accident TF >I hdReiease 2b Estpri C tSF Reactor water saeple lines isolate upm big calo stean line radlatim signal frm Priwary Containeent systee N). 25A Large beak Inside CKF 5-I hd Release 2b Est Pri Cont SF C)ose saepling and water quality systen isnlatim valves cn low reactcr water level ILI) signai freya prieafy cmtalneent systen IQ) ~ 258 intern Srk inside CDEF 5-1 hd Release 2b Est Pri Cent SF Close saepling and water quality systee isolatim valves m low reactcr water level IL1) signai frm prieary cmtainent systen IM). 25C Saall Sreakdnside GXF H hd Release 2b Est Pri Cont SF Close saepling and water quality systee isolation valves m low reactcr water level ILI) signal fra prieary ccntaineent systee IM). 25D Brklnsi~tSad CDEF 3-I hd Release 2b Est Pri Cent SF Close saapling and water quality systea isolaticn valves m low reactcr wats level ILI) signal frca prieary cmtaimnt systee IM). 25D Srklnsid~tSad Release 2b Est Pri Cmt Support prieary cmtaineent functlm. 27 Pipe Sreak Dutside IXEF 3-I FnAf'wj4 g(iqu$ 'y($ Qeg",jI priaary cmtaineent systea N). ~ ~MtVs)ke Isa 4.1~~~> R ~ SAFE SHJINN NNLTSlS QD7 Prepared< irate: ~ Zb M1LMS KATIIS STSTBl lbccted: san NH%3~ llhv 1) EIM 88CT QNXEPTARE SfE N IPER RERLTS SAFETY ACTIN RK KL 69IT TITLE STATE Code Title Code Title GEE 15 Loss Of&ite AC 2-1 hd Release 27 Est Sec Cmt Support secondary ccntaineent fmctim. 2$ hklosihe4ntiad IXEF 3-1 hdRelease 27 Est Sec Cmt Support secaxlary cmtainaat fmctim. 26 Fuel Handl Accidnt A 3-1 hd Release 27 Est Sec Cont Support secmdary cmtainemt flscti04 33 Loss Fuel Pool Cl Nl 2-1 hd Release 2l Est Sec Cont Support secmdary cmtainemt functim. SAFE SR)INN M.YSIS 15f of Prepared> I)ate: FEB%TER GNRL SYS181 E. Oecked) fRHS&OQ IIIev 1) C. L Pvhr SEAL lMXPTAfLE SAFE SB IKRLTS 'SAFETY ACTTIN FlK STATE Code Title Cade Title GK 5 Ri Ctrl Fail~ Send high uater level (LS) signai froe feeduater systea IN) to cain brbine aatrol systee I f7). 25 RI Ctrl Fail~ CKF 2-5 ~ Stress 25 Level Reduce Ih high eater level IL8) signal froa Ri systee I01), energize 8FPT solenoids Sich close the cain stem systee I01) feeduater turbine Met valves. 'C ~. '1) J 0 y C ~ ~ SAFE SWMN NNLYSlS Prepared: Sate: T)RNFDGNTN GKRn. SYSTBl bshed: St~ NHM~ iRev 1) BM REACT MXPTA8LE SAFE SR HER REStlTS SAFE)Y KflSI RK KL BLENT TlTlE STATE Code Title Qde Title GEE 'I 01A Senerata Trip F 2-2 Fuel Faili e i7 Scraa Provide nain brbine cmtrol valve fast clmare trip signal to reactcr protectim systen l'tg) fail~e 1ogi co 016 T/6 Trip% Fails F 2-2 Fuel Failure 0 Scraa Provide nain turbine cmtrol valve fast clcgre trip signal to reacbr protectim systee fgg) fail~ logice 1$ Loss Off+ite C F 2-2 Fuel Failure 0 Scrae Provide aain t&ine ccntrol valve fast clare trip signal to reactcr protectim systen N) faiHde logics '1A Senerata Trip F 2-2 Fuel Failure 18 Pres Relief Send signal to open and cmtrol aain stean systee f01) bypass valves m generatcr trip. This safety actim ~ ay occur, but is not required by this event. 02 Turbine Trip F 2-2 Fuel Failure 18 Pres Relief Send signal to open and ccntrol cain stean systen (01) bypass valves m turbine trip. This safety actim nay occi, but is not required by this event. r 05 Loss of Uacuue F 2-2 Fuel Failure 18 Pres Relief Send signal to open and ccntrol nain stean systen l0]) bypass valves m turbine trip. This safety actim nay occe, but is not r 5 F)f M Fail~ F 2-2 F Frllurr 1$ Pres Rrlirl l Frnl FF~ VrFFrlrn )NtP4 r ';, dg4rr AFeFs)0iplbjiiskrrrlSF err ZS lac Reservoir Mes%lief Send signal to open and ccntrol nai stean systee ves m turbine trip. This safety actim nay occur, but is not required by this event. ga7 SAFE MITMN NLTSIS Page 156 of Prepared! Date: TIIKSE)GQTN GNR1 E SySEI Chected> Date: If)HISS~ (Rev 1) C. L Pr hr SyST1)I I7 BBIT REACT MXEPTKE QFE N SER IIESI.TS SAFETY ACT15l RK KL BJE)IT TlllE STATE Cade Title Cade Title CIXE 05 Loss of Vacuua F 2-1 Rad Release 20 RPV lsol Sand 1DI Rois signal ta )lain Stl& Systea 101) fcr aain turbine stap valve claare. 05 loss of Vagus CDEF 2-1 RadRelease 20 RPV lsol Send signal to aain stean systea 101) to clase turbine b)pass valve yea very lm candenser vanae signai. 32S Law Reservoir F 2-1 Rad Release 20 RPV Isol Provide condenser lm vacuue signal to close aain stean systea I01) turbine stop valves. This safety acticn aay occur, but is aat required by this event. XS Lou Reservoir CD6'-1 hd Ihlease 20 RPV lsol Provide ccndenser lw nasa signal to close aain stean systee 101) turbine bypass valves. Kis safety action aay acnr, but is not required by this event. 01A 6eneratm Trip F 2-2 Fuel Failure 22 Rest kraal After 6eneratcr Trip, signal bypass valves to aaiatain presare ccntrol. ihis sdaty acti+ my ocmr, hut is aot required by this event. I g~ 02 Turbine Trip F 2-2 Fuel Failure 22 Rest kraal After Turbine Trip, signal bypass valves to naiatain Tressure control. iiis saaty acum my cmr, Xi is ant required by this event. 06 Loss RI Heater F 2-2 Fuel Failure 22 Rest Maraal II Load deeand setpoint is u'sed to obtain reducticn ken in autcaatic ~ ade of recirculation floe control. 12A Pres Reg Fail+en CCEF 2-5 Syst Stress 25 Ua h LOS Stean Raw through tgbi ne f~f~ plus bypass val gh's t ILf >v gaydlt&fa .  ! ,.* o; SAfE SIJIDNI AN.YSIS Page 157 oI )of Prepared: Date: TIIHE%RATII CIIIKL E. C SySIBI Reeked) hti: ~Y>> IRHSSRN IRev I) . C. L Parter I EIM SEC. BSIT TlllE REACT CPER INCCEPTARE RESILTS STATE Code Title SAFEIY ACTIQI Code Title SAFE FIlC IE NlSGS 01A Sewratcr Trip F 2-2 Fuel Failure 24 Pottt hduce SF Proride cain turbine cmtrol valve fast claare signal to reactcr protectim systen Igg). 0$ T/6 Tri~ Fails F 2-2 Fuel Failure 24 Pox Reduce SF Provide cain turbine cmtrol valve fast clare signal to reactm protectim systen Igg). 23 F)I Ctrl Fail~ CCEF 2-5 Syst Stress 25 Level Reduce Cu high uater level ILS) signal froa Rl tmtrtt tyttes litt, ansyiza.itin ( + turbine trip solenoids Mch close the nain stean systen 101) turbine inlet valves. '$ 07 SAFE SHJMII AN.YSlS bge 158 Prepared> htes MllKNLllERNORM E AlR SYSTBI Owcked: Sate SS NHM~ llhv 1) C. 1L Perker BM REACT MXEHAKE . SAFE N SER REM.TS SAFETY ACTISl RK KL SENT TITLE STATE Cade Title Code Title CNE 1$ Loss Dff&te AC Nl 2-1 hd Release 27 Est Sec Cent Support secandery ccntsineent fuacticn. 2$ Srklnsi~tNal GKF 5-1 hd Release 27 Est Sec Cant Suppart seccndery anteinaat functicn. 2b Fuel Hsndl Accidnt A 5-1 hd Release 27 Est Sec Cant Support secandery cmtainaet fmticn. Loss Fuel Pool Cl ALL 2-1 hd Relesse 27 Est Sec Cont Support secexbry canteinmt functim, i '3 o7 SAFE SNJIIQN AN.YSlS Prepared: Date: .0 STN)8Y LllSID GNHQ. SYSIEN FlHK&OQ IRev l) Oechdi C. L Parker Date: t$ BE)IT 89K lNNXEHARE SAFE N IPER RESILTS SAPEIY ACTlll RK SEL BM TlT1E STAlE Code Title Cade Title CNE 25D Erklcsid~tSal IXEF 3-l Rad Release 20 RW lsol Suppcrt RPY isolation functim. 25D Srklnside+atLRad IXEF 3-) Rad Release 2I Est Pri Cant Support prinary ccntainant functim. 29 Shuts )I/O Rods IF 4-2 Shut )Io Rods 42 Shutdn NoRod Ii initiate standhy liquid ccnbol systea injecticn 9iven indicaticn of incoeplete shubhe uith rodsISysten 85 l 92). I)co)ter injecticn of the soluticn in the reacta. Send si9nal to R)G)$9) to isolate. '5 o7 ~ QFE QSTMN AN.TSIS P )60 of Prepared: Sate: PRINN GNAINBIT E STSTEN Checked: hte: KH)S&0l8 (Rev 1) C. L Parker EUM REACT SACCEPTAKE SAFE N 5ER REStLTS SAFETT ACTIN RK SHl. MIT TITLE STATE Code Title Code Title CCKE 13 Inadvert Open SRU IF 2-2 Fuel Failure 17 Scrae Tenperature senars in the suppressim pool supply inhrnatim for the operator 'to nake a 'eeded decisicn to nanually scran. 25A Large Break Inside EF 3-2 Fuel Failure 17 Scrae SF Provide high dryuell presan trip signal to reactcr protecticn systen fRPS) Hg) failmfe logic. 258 Intern Ik k Inside OF 3-2 Fuel Failure 17 Scraa SF Provide high dryveil presare trip signal to react+ protectim systen IRPS) (gg) failmfe logic. 25C Snail Break Inside OF 3-2 Fuel Failure 17 Scraa SF Prwide high drynell pressre trip signaI to reactcr protsctim systen IRPS) fgg) failmfe logic. OIA Seneratcr Trip F 2-3 Syst Stress 18 Pres Relief Accept SRUs stean blodae I fry boiler vents and drains systen, 10) to suppressim pool. i 018 T/6 Trip& Fails F 2-3 Syst Stress 18 Pres Relief Accept SRUs stean blodoe ffra boiler vents and drains systen, 10) to suppressim pool. 01C PresReg Fail<lose COEF 2-3 Syst Stress 18 Pres Relief Accept SRUs stean bloabe lfron boiler vents and drains systen, 10) to suppressicn pool. 02 Turbine Trip F 2-3 Syst Stress 18 Pres Relief Accept 95s stean blodcae (fry boiler vents and drains systen, 10) to suppression pool. 03 Nain Sta Isolaticn COEF 2-3 Syst Str )l 1~@AD@< '"p~"A' ~~ 05 Loss of Uacuua F 2-3 Syst Stres 8 Qiefs c < 'cct'55k stean blodae ffrcn boiler vents and drains systen, 10) to supprqssim $ o7 SAFE SHlMN A)IILYSIS hge 161 Prepared: Sate: PRNRf GNAINBIT STSTE)l Checked: Bate: lF)H)S83~ (Rev 1) C. IL Parker REACT lNKEPTAIE SfE SER RERI.TS SAFETy ACTIN fiK STATE Code Title Code Title GXE 12A, Pres Reg Fail+en IXEF 2-3 Syst Stress 18 Pres Relief Accept 9Ns stean blon)oe Hra bailer vents and drains systen, 10) ta suppressim pool. 128 Inadv Open~s GKF 2-3 Syit Stress 18 Pres Relief Accept 9Ns stean bio!doe Hrm boiler vents and drains systen, 10) to suppressim pool. 13 Inadvert Open 9N CBEF 2-3 Syst Stress 18 Pres Relief Accept stean blon)oe I fran boiler vents and drains systen, 10) frcn stuck apen 9N to sqyressim paul N). 15 Loss Off<ite AC IXEF 2-3 Syst Stress 18 Pres Relief Accept 9Ns stean blon)oe ffrm 0 23 F)f Ctrl Fail~ F 2-3 Syst Stress 18 Pres Relief boiler vents and drains systen, 10) to suppressicn pool. Accept 9Ns stean hlodoe Hron bailer vents and drains systen, 10) to suppressicn pool. 2f Rad Bry Accident BF 3-3 Syst Stress 18 Pres Relief Accept 9Ns stean hlodoe Ifrm bailer vents and drains systen) 10) to suppressim pool. 25C Snail Break Inside COEF 3-3 Syst Stress 18 Pres Relief Accept 9Ns stean blodoe vents and drains systen, 10) if'oiler to suppressicn pool. 27 Pipe Break Outside COEF 3-3 Syst Stress 18 Pres Relief Accept 9Ns stean bloudoe Hrm VE l to suppressicn pool. GEF 2-3 hg!Iil~ !;.. gvt j4 sf+ )~!(fieg'., I j tb'suppressim poo . XE Lau Reservoir GEF 2-3 Accep boiler vents and drains systea, 10) to suppressicn pool. 03 )fain Stn Isolatim GEF 2-2 Fuel Failure lg Ccrc Cooling Provide alternate uurce uater apply I fcr RCIC systen l71) through Core Spray systen f7') piping frca suppfessim pool ~ Pfpi1de suppresslm pool level indicaticn. Accept RCIC turb>ne ezhaust stean ~ I' SAFE BSlNN ANALYSIS P IQ Prepared: hte: ~~+f'7 PRIlARY GNAINM SWE)l Decked: hte: BF)H663~ IRev I) BSIT REACT MXEPTAKE . SAFE OR OPER RERLTS SAFEIY ACTI% RlC SEQ. BENT TIRE STATE Code Title Code Title NK hT, Naia Sta lsolatim COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide alternate sarce water supply fcr )KI systea le) frm suppressim pool fl )KI initiatim m lcw water level 02) signal. Accept )KI turbine exhaust steaa. 05 toss of Vacuua COEF 2-2 Fuel Failure 19 Ccrc holing Provide alternate source water supply fcr )Kl systea G3) frm suppressim pool fm IKI initiaticn m Im water level 02) signal. Accept )Kl turbine exhaust stean. 2-2 Fuel Failure 19 Ccrc Cooling Provide alternate source water supply fcr RCIC systea 171) through Ccrc .0 Fail~'EF , Spray systea (75) piping frm suppressicn pool. Provide suppressicn pool level indicatim. turbine exhaust steaa. Accept RCIC 12A Pres Reg COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide alternate source water supply fcr RCIC systea l71) through Ccrc Spray systea 175) piping froa suppressicn pool. Provide suppressim pml level Indicaticn. Accept RCIC turbine exhaust stem 12A Pres Reg Fail+en COEF 2-2 Fuel Failwe 19 Ccrc Cooling Provide alternate aarce ly fcr roa suppress pool for )Kl initiaticn m Iqw t .P5pf }K3 ~ l6$ @g 12B Inadv Open-Bypass Pf luis;gf, +> g"J source water apply fcr RCIC systea 171) through Ccrc Spray systea l75) pool. Prande suppremcn pool level indication. Accept RCIC turbine exhaust stean. 12B Inadv Open-Bypass COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide alternate smrce water supply fcr )KI systea I75) froa suppressim pool fcr )KI initiaticn m low water level IL2) signai. Accept )KI turbine exhaust. stean. i SAFE BSTNN QNLYSIS IQ $ 07 Prepared: hta ~~vJ PRIIARY (XNAIMSIT E. SYSTE)( Chechd: 0ate: 7 SRHM~ (Rev 1) C. L Pate BENT REACT NQPlARE SAFE N (PER RESETS SAFETY ACTIN RK SEL PENT TITLE STATE hde Title hde Title C(IE D, Inadvert 0pen SRV (XEF 2-2 Fuel Failure 19 Care Cooling Provide alternate smrce uater supply fcr RCIC systea 01) through Core Spray systea (75) piping froa suppressim pcol. Provide suppressim pool level indicaticn. Accept RCIC turbine exhaust steaa. D Inadvert 0pen S)N (XEF 2-2 Fuel Failure 19 hre Cooling Provide alternate source water supply . fcr )fCI systea (73) free suppressim pool fl %CI initiatim m lw uater level (i2) signai. Accept IfCI turbine exhaust steaa. g f(g(74) ~d 15 Inadvert 0pen SN (XEF 2-2 Fuel Failure 19 Core Cooling Support>core spray systea 05) by 0 providing row cooling uith service uater froa EEQ ((7) in support of HXS actims at lw reactor uater level (Ll). 9 Inadvert 0pen 9N (XEF 2-2 Fuel Failure 19 Core holing Provide tcrus water supply fl M (74) and hre Spray 05) systeas in suppa t of HXS acticns at lw reactor water level (Ll). Il Loss of R( Flow C0EF 2-2 Fuel Failure 19 Core Cooling Provide alternate source>>ater supply for RCIC systea (71) through Ccrc Spray systea (75) piping froa suppressim pool. Provide suppressim pool level indication. Accept RCIC turbine exhaust stem I . Loss of FN Flw 2 2 Fuel Failure 19 Core Cooling Provide alternate source uater ly turbine exhaust steaa. 15 Loss 0ffdite AC fcr RCIC systea (71) through Core Spray systea (75) piping froa

0 suppression pool. Provide suppresslcn pool level indicaticn.

turbine exhaust stean. Accept RCIC 1 SAFE QITOOW ANALYSIS Page IM 7~7 Prepared: bate: PRINN GNTAIQBiT STSTE)( Checked: batex ERH)SH~ (Rev I) C. IL Parka BBIT REACT QIA(XEPTAOLE SAFE N Hat RES(LTS S(IFETY ACTIlN FDC SEO. SENT TITLE STATE Code Title Code Title IE 15, Loss Offdite AC COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide alternate sutrce water supply fa IKI systen (75) fran suppressim togae pool fcr IKI initiaticn m low water level (L2) signal. Accept IKT turbine exhaust steaL p;>If gMaCg4)~< 15 Loss Off<ite AC COEF 2-2 Fuel Failure 19 Core Cooling 4 spray systen (75) by providing roon cooling with service water free EEN (57) in support of HXS actims at Icw reactor water level (Li). 15 Loss Off-Site AC COEF 2-2 Fuel Failure 19 Core Cooling Provide taus water supply fcr R% (71) and Core Spray (75) system in suppat of HXS acticns at low reactcr water level (LI). 22 Loss Shutdoo Cool CO 2-2 Fuel Failure 19 Core Cooling Accept 9Ns stean blowdown (fro>> boiler drains, vents and blowdown systen, IO) to suppressicn pool. 25 f)( Ctrl Fail~ COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide alternate sarce water supply fa RCIC systee (71) through Core Spray systen (75) piping frow suppressicn pcol. Provide suppressim pool level indication. Accept RClC turbine exhaust stean, F)I Ctrl Fail~ ( XEF 2-2 Fuel Failure 19 Ccrc Cooling Provide alternate smrce water supply fcr IKJ systen (75) frow suppressicn pool fa'KI initiaticn on low water level (l2) signal. Accept IKI turbine exhaust stean, qc $7 @g 21 Rod brop Accident 3-2 Fuel Failure 19 Core ing-- <core spray systen 05) y providing roon coolin with 'ce tp IPE '. <Qmtjcgthsg lg,'t tcr t >'1 .,I > <~ 4 21 Rod Orop Accident uh Failure 19 Ccrc Cooling Provide tcrus water supply spray'ys suppcrt of HXS acttms at low reacta water level (Ll). fa'71Lan(LCcwe ens in SAFE MIT(QN ANALTSIS P 165 37 Prepared: C, babe: ~su PRINN (XI(TAINT E. C. Eck SSTE)( Checkedy Date: ZB SV BF)H)SB3~ (Rev 1) C. L Parker BBIT REACT lNKPTARE SAFE N IPER RESTS SAFETy ACTI% FIK SEL GENT TITLE STATE Code Title Code Title CKE

24. Rod Drop Accident OF 3-2 Fuel Failure 19 Ccrc Cooling Provide alternate scajrce water supply fcr IPCl systen (73) fry suppressicn pool for IfCl initiaticn cn lcIF water level I(2) signai. Accept )PCI turbine echaust stean.

25A Large Break inside C0EF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide high drywel1 pressN e signal to ccrc spray systen (75) and IR systen (7$ ). (C Ig Qe(14)a~d 25A Large Break inside COEF 3-2 Fuel Failure 19 Care Cooling Jere spray systen 05) by providing roon ccoling with service water fry EEDI (67) for HXS acticns on high drywell pressure signal. 25A 'arge Break inside COEF 3-2 Fuel Failure 19 Core'Cooling Provide tms water supply fa Nf( (7l) and Core Spray (75) systen fcr ECS acticns cn high drywell pres@re. (c, lN /Ac(~w)~ 8 25A Large Break inside COEF 3-2 Fuel Failure 19 Cere Cooling J~tI(cure spray systen (75) by providing roon cooling with service water fron EECM (67) in support of HXS acticns at low reactm water level (Ll). 25A Large Break inside COEF 3-2 Fuel Failure 19 Ccrc Cooling Provide tcrus water supply for RN (74) and Ccrc Spray 05) system in support of HXS actions at low reactcr water level IL1). c yys yydsy'.&a)M 258 Intern Brk inside C0EF 3-2 Fuel Failure 19 Core Cooling Psppwb~ssre spray syslee lyby by ovidi roon cooling with service water fron EEDI (6 ) in supp HXS actions at low reactor water drb i p.spau su)) Aa 25B intern Brk inside iydy 2-2 Fuel Failure lp yare ysu i Ir 'd Is)I(ualppts'by~ltd (74) and Ccrc Spray (75) system >n support of HXS acticns at low s s, ~ 0 SAFE SHJIIM NNLYSIS Page 166 7 Prepared Date: ~Ill/ PRINN C(NTAIQ%N E.C Ec SYSIE)I Checked: Bate: If)HS)3~ (Rev I) C. L Parker BM REACT MXEPTABLE SAFE N SER RES(LTS 'AFEIY ACTIN RK SEL BENT TITLE STATE Code Title Code Title C(IE 258 lntera Brk Inside CBEF 3-2 Fuel Failure 19 Core Cooling Provide alternate srarce water supply fcr fPCI systea l73) frca suppressim pool fcr IfCI initiatim m la water level (L2) signal, Accept ffCI turbine exhaust steaa. 258 lntera Brk Inside (XEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide high drywell pressure signals to ccrc spray systea (75), fN systea (7$ ), and aain stean systea l01) fcr A(6. 258 Intera Brk Inside CBEF 3-2 Fuel Failure 19 Ccrc Cooling Provide alternate source water supply fcr IfCI systea (73) froa suppressicn pool fcr IPCI initiatim m high drywell pressure signals. Accept Wl turbine exhaust steaa. 258 Intera Brk Inside CDEF 3-2 Fuel Failure 19 Core Cooling Q(p1 gfggrQ ~ Support~ccrc spray systea l75) by providing rooa cooling with service water froa EEO( l67) fN ECCS acticns on high dryweil pressure signal, 258 lntera Brk Inside CBEF 3-2 Fuel Failure 19 Core Cooling Accept reactcr vessel autoaatic depres5N'iaatim systea (ABS) steaa blowdoo lfroa boiler vents and drains systea, 10) to suppressicn pools 258 lntera Brk Inside (XEF 3-2 Fuel Failure 19 Core Cooling Provide tcrus water supply fm RN (74) and Core Spray (75) systea fcr HXS actims m high drywell pressure. k~~pg /HE~<)~~ 25C Saall Break Inside COEF 3-2 ilure 19 Ccrc Coolin ~ Supor~tccre seas stea 05) by providing rona cooling eth sennce '~ water froa EED( l67) in support of actiafs<a('IceT,reactcr='wdta,' f 'di '<;"-; 25C Saall Bre'ak Inside (XEF 3-2 uel Failm e 19 Ccrc Cooling Provide trna'water supply fa II( . (7i) and Core Spray (75) systeas in lapp>t ol reactcr water level (Ll). SAFE QUMN N)ALYSIS P !bi Prepared: Date: tu

0 PRWRY SRHKM~

GNAINSll'WHl (Rev 1) Checked: E. C. L Parker Date: EVBIT REACT MXPTABLE SAFE KR IRR REBLTS SAFEIY ACTItN flK SHL 881T TIRE STATE Code Title Code Title ONE 25C Snail Break Inside CKEF 3-2 Fuel Failure 19 Ccrc Cooling Provide alternate smrce water supply fcr )fCI systee (73) frow suppressim pool fcr IPCI initiatim m low water level IL2) signal. Accept SCI turbine exhaust stean. 25C Saall Break Inside CKEF 3-2 Fuel FailNe 19 Core Cooling SF Provide high dryweII presscre signals to ccrc spray systee f75), R)R systea 174), and aain stean systee 101) fcr ADS. 25C Snail Break Inside CKEF 3-2 Fuel Failure 19 Ccrc Ccoling Provide alternate source water sqqly fN )PCI systea 173) froa suppressicn pool fcr )fCI initiation cn'high dryweII pressure signals. Accept IfCI turbine exhaust stean. g~P ~g)Pn J 25C hall Break Inside CKEF 3-2 Fuel Failure 19 Core Cooling Suppcrfpccre spray systee f75) by providing roon cooling with service water froe EECM (67) fcr HXS actims on high drywell pressN e signal. 25C Snail Break Inside CDEF 3-2 Fuel Failure 19 Core Cooling Accept reactor vessel autoeatic depressNizaticn systen IADS) stean blowdown ifroa boils vents and drains systea, 10) to suppressicn pool. 25C Snail Break Inside CKEF 3-2 Fuel Failure 19 Ccrc Cooling Provide torus water supply for R)gt (7I) and Ccrc Spray 175) systen fcr ECCS actions cn high dryweII qc<gr; @yg~<~~k 2-Fuel Failure 4y:s 27 Pipe Break Outside CKEF 19 r Ccrc Cooling Sugp~crp spa~ sysPae d75) f V). level (LI). v 27 Pipe Break Outside CDEF 3 onde tcrus wat'er supply fcr R% 1') and Ccrc Spray 175) systens in support of HXS actims at low reactN water level (LI). ~ ' ~ \> SAFE SEEM ANALTSIS Page 1 $ 07 Prepared: Dots ~~( PRI)gRT GNAlttBIT SSTE)I Checked: Date: EF)H)SS3~ IRev 1) BENT REACT MXEPTAKE SAFE OR SB KBLTS SAFETT ACTIN RK SEL aENT TITLE STATE Code Title Code Title GEE 27, Pipe Sreak Outside CDEF 3-2 Fuel Failure 19 Ccrc Cooling Provide alternate source water supply fcr IfCI systea (73) froa suppressicn pool fcr PCI initiation m low water level ll2) signal. Accept IfCI turbine exhaust steaa. 32D Low Reservoir CDEF 2-2 Fuel Failure 19 Ccrc Cooling Provide alternate source water supply fcr RCIC systea l71) through Core Spray systea 175) piping froa suppressim pool. Provide suppressicn pool level indication. Accept RCIC tm bine exhaust steaa. 328 Low Reservoir CDEF 2-2 Fuel Failure 19 Core Cooling Provide alternate source water supply fcx IFCI systea l73) froa suppressim ,0 pool for IPCI initiation on low water level lL2) signal. Accept IPCI turbine exhaust stean. ISD ISO Shutdown CDEF 2-2 Fuel Failure 19 Core Cooling Provide alternate water supply fcr IXI systea l73) froa suppressim pool. ISD ISO Shutdcao CDEF 2-2 Fuel Failure 19 Core Cooling Provide alternate water supply for RCIC systea 01) through Core Spray systea l75) piping froa suppressicn pool. Provide suppressim pool level indication. Accept RCIC and IfCI turbine exhaust steaa. NSD Noraal Shutdc<<o CDEF 2-2 'lure 19 Core Cooling Provide alternate water supply for )FCI systea (73) frca suppression gp<<f 3 f <s expKted to occur,but is not a requireoent fN this event. 15 Inadvert Open SRV COEF 2-1 Rad Release 26 Est Pri Cmt SF Upcn LM. (U) signal frca RPS (99), nitittHMmtatIYa) s I send prinary/seccndvy cmtain t , c,l . ~u,>; Pgso,, Iaticn(n)9@, ) o $Q 2, 65, and N. cur,but ~ " '-'4s"not r+ikeaenYfor t a s event. Il Loss of FN FIxxx (XEF 2-1 Rad Release 26 Est Pri Cont SF cn (99), initiate L5 isolation acticns I send prinary/secmdary containnent isolaticn signals to systens 32) 65) 69, 75)75} 76, 77, N, 90, and 9f. This action is expected to occur,but is not a requireaent for this event. SAFE 9UHXN ANALYSIS Page 171 os 7 Prepared: Oate: Zo 0 PR])(ARY SYSIE)( CNTAI)ST BF)HM~ (Rev 1) Checked! ]jate: BM REACT MXEPTABLE SFE (R (PER RES]LTS SAFETY ACTIN FllC SE0. BIENT TIT]E STATE Code Title Code Title IE 1S Loss Off<ite AC ALL 2-1 Rad Release 26 Est Pri Cont SF UPm LIL (Q) signal frow RPS (99)f initiate Q isolation acticns t send prinary/secmdary cmta]nnent isolaticn signals to systens 32>> 65, 69, 74, 7S, 76, 77, 84f 90, and 94. This actim is expected to occm'>>but is not a requirenent fm this event. 15 Loss 0f f-Site AC Cmt Stress 26 Est Pri Cont SF Initiate ccntainnent isolatim signal/action m loss of ncrna] ac puffer (va]ves and systens as~ fai]safe ccnditim). 23 R( Ctrl Fai]~ CDEF 2-1 Rad Release 26 Est Pri Cont SF Upm lÃ. (Q) s]gna] frow RPS N), initiate Q isolatim actims (f send pr]narylsecmdary cmta]AKOt isolatim signa]s to systens 32f 6S, 69, 7lf 7Sf 76f 77f Slf 90f and 94. This acticn is expected tn occur,but is not a requirenent fN this event. 24 Rod Brop Accident ]IF 3-] Rad Release 26 Est Pri Cmt SF Upon low water level (Q) signal fron reactcr protectim systen N), initiate Q isola cns and 'naary/seccndary c t ainnent isolaticn signals to sy t 32f 6S, a.,e .,w,69>> 74f 7S) 2l)f and 94. 24 Rod Erop Accident Rad Ref]ease 26: '+i+ '~(SF, 6))k))J)fff)j ) )m reactcr water level (LI) s> to initiate na]n stean line i laticn, sng 4 water quality systen (43) line isolatim. 24 Rod Brcp Accident Rad Release 26 Est Pri Cmt SF Upon nain stean line high radiatim trip signal (froa reactor protecticn systenf 99)f send isolaticn signals to wain stean (01) line drains and reactor water saaple lines (43). 25A Large Break Inside (XEF 3-1 Rad Re]ease 26 Est Pri Cmt SF Send feedffater systen (03) low . reactcr water level (Ll) signa] to initiate na]n stean systen (0]) and sanp]ing line isolaticn, 1 water quality systen (43) line isolatim. Scs1 ~ SAFE QUINN ANALYSIS Page 172 Prepared: Date: PRINN GRAINS E SYSTE)I Oseckedc Sate If)H)SS3~ IRev I) C. L Parker EVE)IT REACT IMXEPTABLE SAFE DR IPER RESILTS SAFEIY ACTISI Ftw: SG). EVENT TITLE STAlE Code Title Cade Title INE 25A Large Break Inside NEF 3-1 Rad Release 26 Est Pri Cmt SF @m law water level IL3) signal frm reactor protecticn systen I99) s initiate L3 isolaticn acticns and send prinarylsecondary cmtainaent isalatim signals 'to systens 32s 65s 69s 7ls 75s 76s 77s Bfs 90s and 'N. 25A Large Break Inside NEF 3-1 Rad Release 26 Est Pri Cant SF Provide high drywell pressure signal to reactor protectim systen I99) far initiatim of high dryssell presscre isolatims. ZA Large Break Inside NEF 3-1 Rad Release 26 Est Pri Cmt SF Qm high drywell pressure signal free reactor protection systen I99)s initiate high drywel1 pressure isolaticn acticns and send prinarylsecandary-ccntainaent isolatim signals to systens 32, 65, 7is 75s 76, 77, Bfs 90, and 9I. 25A Large Break Inside CDEF 3-1 Rad Release 26 Est Pri Cant SF, High stean line differential presscre senscrs initiate sianal to )lain Stean Systen I01) fcr isolatim. 258 Intern Brk Inside NEF 3-1 Rad Release 26 Est Pri Cmt SF Send feechsater systea 06) low reactor water level ILI) signal to initiate naia stean line isalaticn, systen I01) and saepling Is water 1 t 258 Intern Brk Inside GEF 3-1 Rad Release 26 Egtpr) Cpt~.,~ Qwatpcl&e(,]L3)),$ fran d t isolaticn signals to systees 69 94. 258 intern Brk Inside C0EF 3-1 Rad Release 26 Est Pri Cont SF Provide high drywell pressure signal to reactcr protecticn systen I99) fcr initiation of high drywell pressure isolatims. SAFE SHJTNN AN.YSIS Page )73 of Prepved: .~~ PRI!ARY SYSTE)( BF)H)SS3~ C(NAINfM (Rev 1) Checked: E.C C. )L Ec Pvker REACT MXPTAKE SAFE EPER RES%IS SARY ACTIN RK KL ElJENT TITLE SINE Cade Title Cade Title (IE 258 lntera Brk Inside lXEF 3-1 Rad Release 26 Est Pri Cant SF Upcn high drywell pressure signal fraa reactcr protecticn systea (99) f initiate high drywell pressure isa)atlm actims and send priaary/secmdvy ccntainaeft isolatim signals to systeas 32f 65f 7<f 75, 76f 77, 8(f 90, and >. 258 Intera Brk Inside CBEF 3-1 Rad Release 26 Est Pri Cant Establishing priaary ccntainafxft is covered in Appendix ZD of this analysis. 25C Saall Break Inside (XEF 3-1 Rad Release 26 Est Pri Cmt SF Send feedwater systea (N) law reactor water level (LI) signal to initiate aain stean line isolaticn, systea l01) and saapling k water quality systea N3) line isolaticn. 25C Seal I Break Inside (XEF 3-1 Rad Release 26 Est Pri Cant SF Upm loi fh))ec gnal fraa reactcr protection systea l ), Initfate 0 igata any and t f 65r ',s 94. 5o 25C Saall Break Inside 4 %adQIease 26 Est Pri Cant SF Provide high drywell ignal o ection systea (99) fcr initiatim of high dryweII pressure Isolatims 25C Saall Break Inside COEF 3-1 Rad Release 26 Est Pri Cant SF Upan high drywell pressure signal fraa reactcr protecticn systea (99) f ~ initiate high drywell pressure isolaticn actims and send priaarylsecmdvy cmtainaent isolation signals to systeas 32, 65, 75f 75f 76, 77, Blf 90, and 94 Zb Brklnsid~tkRad (XEF 3-1 Rad Release 26 Est Pri Cmt SF Seel feedffater systea (N) low reactm water level (L1) signal to initiate aain stean line isalatim, g systea (01) and saapling h water quality systea (Q) line isolatim. SAFE SHJ(MN Q(ALYSIS Page 17 f07 Prepved: bate: PRIIIARY GNAIQM SYSIEN Checked: bate: NHM~ (Rev I) REACT MXEPTABLE SAFE CPER RES%IS SAFETY ACTIN RK STATE Code Title Code Title (mE '250 Brklnside+mtSad C(KF 3-1 Rad Release 26 Est Pri Cont SF Opon low water level (Q) signai frow reactor protectim systen (99), initiate L3 isolatim actims and send pri nvylsecmdvy cmtainnent isolatim signals to systl% 32, 65,

69) 7l~ 75~ 76'7s Bls 90'nd 9l.

250 Brklnsid~tSad IXEF 3-1 Rad Release 26 Est Pri Cont SF Provide high drywell pressure signal to reactor drywall protecticn systen (99) fcr initiatim of high drywell pressure isolatims. 25D BrLlnsid~tSad COEF 3-1 Rad Release 26 Est Pri Cmt SF I/pen high pressure signal froa reactor,.protectim systen (99), initiate high *ywell pressure isolaticn acticns and send prinarylsecmdvy containaet isolaticn signa)s to systens 32$ 65 f 7l$ 75$ 76) 77, Bl) 90, and 9l. 25D Brklnsid~tSad CDEF 3-1 Rad Release 26 Est Pri Cont L Passive: Rust naintain required leaLage rates to Secondary Ccn'tal nnent 27 Pipe Break Outside IXEF 3-1 Rad Ihlease 26 Est Pri Cent SF Send feedwater systen (03) low reactcr>>ater level (Ll) signal to initiate wain stean line isolaticn, systen (01) and sapling I water quality systen (l3) ine isolaticn: 46-M~nl 27 Pipe Break Outside Rad Release k4+1. 'j+5 SF 'i I/pm Iow water level f+tef~~plsys+ Jjn(tie ( peal ) i juijcn 'jictim'q'y ~s l 69 7l~ 75~ 76'7s Ns 90~ and 9l.. rC 27 Pipe Break Outside (XEF 3- t Pri Cont SF ann steu systen ( 1) signal (high stean tunnel teeperature cr high steaa flow) to initiate nain stean systen (01) line isolatim and sapling 6 water quality systen (l3) line isolaticn. SAFE )SINN AtNLYSIS Page 175 Prepared: PRI)NY GNAIQEHT E. SYSIB) Oechd: If)H)M~ Ilhv 1) C. L ParLer 4.. ~ 9%XI REACT INCEPTASLE SAFE N $ %R RERLTS, SAFEIY ACTIN RK KL BM TIRE STATE Code Title Code Title CNE 'N Hain Sta Isolation CDEF 2-1 Rad Release 27 Est Sec Cmt SF Perfla isolaticn actimls) upm receiving low water level IQ) isolatim signai frow the Reactl Protectim Systen Igg). This safety actim is Npected to occur) hut is sot a requlH%flt fcr thIS event since there is no fuel failura. 05 hss of Yacuua CDEF 2-1 Rad Release 27 Est Sec Cent SF Perforn isolatim actimfs) upon receiving low water level IQ) isolaticn signal frm the hactcr Protecticn Systen VH). This safety actim is expected to occur, but is aot a requireeent fcr this event since there is no fuel failure. r D Inadvert Dpen S)N CDEF 2-1 bd Release 27 Est Sec Cmt SF Perfora isolaticn acticnts) upon receiving low water level IQ) isolatim signal froa the Reactor Protectim Systea N). This safety acticn is expected to occur, hut is aot a requireeent fcr this event since there is no fuel failure. ll Loss of F)f Flow IXEF 2-1 Rad Release 27 Est Sec Cent SF Perfaa isolation acticnis) upm receiving lm water level IQ) isolatim signal froa the Reactm ~ Is safety action is expected to occur, but is ~epjr~t f+lg + Loss Dff-Site AC hdJReliase'7'Est Sec Cmt SF'Perfhrd" isola\1m actionis) upon 15 ALL Q receiving low water level IQ) iso Protection Systee Igg). This safety act<m is eapected to occN', but is aot a requireeent for this event since there is no fuel failure. ALL 2-1 Rad Release 27 Est Sec Cont Support secmdary ccntaineent functim. - SAFE SUIQN lNLYSIS Page 176 97 Preparedt Bate: PRINN EQ. 9 GNAIQBll'YSTEN Checked: Bate: BF)H)SS3~ (Rev I) C. L Parker EIIEHT REACT MXEPTABLE SAFE N NB RESILTS SAFETY ACTIQI RK SEL BENT Tlllf STATE Code Title Code Title GXE '15 Loss Offdite AC ALL 2-1 Rad Release 27 Est Sec Cent SF Send start signal to SSTS lb5) m loss of meal ac power. 23 FN Ctrl Fail~ GKF 2-1 Rad Release 27 Est Sec Cont SF Perhra isolation acticnls) upon receiving low water level lQ) isolaticn signal froa the Reactor Protectim Systen f99). This safety acticn is expected to ocnr, but is sot a requireNnt for this event since there is no fuel failure. 2f Rod Brop Accident IF 3-1 Rad Release 27 Est Sec Cont SF Perfcrn isolaticn actimls) upm receiving Iow water level lQ) lsolaticn signal frow the Reactar ~ Protecticn Systee f99). lSee Safety Action 26) 25A Large Break Inside G)EF 3-1 Rad Release 27 Est Sec Cont SF Perfcrn isolatim acticnls) qxn  %, ~ receiving Iow water level lQ) isolation signal fran the Reactcr Protection Systea l99). lSee Safety Acticn 26) 25A Large Break Inside CBEF 3-1 Rad Release 27 Est Sec Cent SF Perfcrn isolation acticnls) upm receiving high drywall pressure Isolatim signaI froe the reac proEect>cn systen 199). irjeMng>'lowest& 258'ntern Brk Inside ~~Jyipqp, pfs~w".,"givipt~gsticn jtti Ie'veI, IL3) m(ll+'",.l'<; Systee 199). lSee Safet 'rotecticn Actim 26) 258 Intern Brk Inside GKF 3-1 hd Release 27 Est Sec Cmt SF Perftrn isolaticn acticnls) upm receiving high dryweII pressN'e isolatim signaI fry the reactcr protectim systen l99). 25C Saal I Break Inside GEF 3-1 Rad Release 27 Est Sec Cmt SF Perforiisolatim actimls) qm receiving low water level IQ) isolatim signal frow the Reactcr Protectim Systen 199). ISee Safety Actim 26) ~ ~ I 1 ~ SAFE 9UlNN AIILTSIS Prepared: ate: PRINN GRAINS STSTEH Checked) bate: AF JC7 EF)H)S(b4I8 (Rev 1) C. )L Parker BSIT REACT MXEPTAKK SAFE N 5%R RESETS SAFETY ACTIN RK SE(}. BBlT TITLE STATE Code Title Code Title CNE '25C Snail Sreak Inside CSEF 3-1 Rad Release 27 Est Sec Cmt SF Perfa a isolation acticn(s) upon receiving high drywell prcsure isolaticn signaI frow the reactcr protection systee PB). 25() Srkinsid~tNad (XEF 3-1 Rad Release 27 Est Sec Cmt SF Perfore isolaticn actim(s) upm receiving Iow water level (L3) isolatim signal frow the Reactar Protectim ~en ('5). (See Safety Acticn 26) 25() Srkinsi~t(f(ad CHIEF 3-1 Rad Release Per fera isolatim acticnls) upon receiving high drywall presare isolatim signal froa the reactcr protectim systee (99). 25() Brklnsid~tNad (XEF 3-1 Rad Release 27 Est Sec Cont Support secmdary containeent functim. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont Support secmdary ccntaimnt functicn. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont SF Upm receipt of high Radiatim Nonitcring systee ('5) signal) send a signal to activate the Standby Sas Treatment Systen (65)~&~he 1'Sot 4!! t A .. W<pl 27 Pipe Break ()utside  !!al Rel<<<<21 Bt Eec!'<<t.V<; fer!!rI,!!!tlitg(dgn! ) upm (L3) Safety ","; '< a ~ Acticn 26). This safet im is ury but is not requiredo Rl 33 Loss Fuel Pool CI ALL 2-1 Rad Release 27 Est Sec Cmt Suqxrt secmdary containant fmcticn. 33 Loss Fuel Pool Cl NL 3-1 Rad Release 27 Est Sec Cont SF Upon receipt of high bdiatim Honitcring Systen (go) signal, provide signal to activate standhy gas treateent systea (65)~~he mls tA ~ r le~ a>r. gccr(<<(r! R! QFE SHIIION AN.YSIS Page 178 of 'Prepared: PRlSN CITlONEN Quacked: E. Ec SYSTBI IRH)S63~ IRev 1) C. L Parter S6Rl Q IKL BJE)IT EUE)IT T1TLE REACT IPER. MXPTARE REKTS STATE Code Title SAFEIY ACTTOI Code Title SAFE FIR IIE '01S T/S Tri~ Fails F M Cmt Stress 30 hnt Cooling Provide Rppressim pool teeperature and level iedicatim to squirt M systea I70 sqyressim pool ~ odeum'lro yoo oooli~q. cooling fC (l7 01C PresRegFaiH:lose IXEF M Cent Stress R Cont Cooling Provide suppresaim pool teaperabre axl level indicatim to support M systen I74) sqqressicn pool aoling ~ ode'/Sa room cealikg, W~gg IG IIain Sta lsolaticn CKF 2H Cont Stress 30 Cmt Ccoling Provide suppressicn pool teeperabre and level indicatim to support M systee 17') suppressim pool aoliog code ~ls ~~ cool~ .0 05 Loss of Uacus IXEF 2H Cmt Stress 30 Cmt holing Provide suppressicn pool teeperabre and level indicatim to srgert M code) n[so ~ systee I7I) aqpressim pool cooling coollHy i /~CD I> 12A Pres Rag Fail~ C0EF M hnt Stress R Cmt holing, Provide suppressim pool teeperabre and level Indicaticn to support M systee I7$ ) suppressicn pool cooling ao/SP 12S lnadv Cpen~ass -Pp&l ature M ailing ~~rolP7 D lnadvert 0pen SRU and level indicatim to support M systee I7$ ) suppressicn pool aeling ' ode] g$ o ~m coolie.g+ JM~ 15 Loss Of finite C IXEF 2H hnt Stress 30 Cent holing Provide suppression pool teeperatm and level indicaticn to support M systen 17') suppressicn pool cooling ggy> who .yoo~ caaifagf+C 22 Loss Shutdoo Cool A 2H hnt Stress R Cont holing Reeove heat free above open vessel. SAFE SIHOM NN.YSIS Page 17g of o7 Preparedl PRIINY CIITAIIBIT E. Eck SYSIE)l Dleckedl If)H)S63~ Ilhv I) C. L Parker EtM IIEACT IIAXEPTARE SfE IR 5%R IIESI.TS SfHY ACTISI FQt SEL GE)IT TITLE STATE hde Title hde Title Cm)E 24 Rod Orop Accident IF M hnt Stress 30 Cont Cooling Provide Qpllessim pool teeperature and level lndlcatlm to suppcft NR systee 17$ ) suppressim pool aoling ~ ode y))o voo~ ~lr<<g. gC ~s IJ'1 25A Large Ik eak Inside IXEF M hnt Stress 30 Cont holing Provide suppressim pool teeperater>> and level iodicaticn to support NR systee (70 sqlpressicn pool cooling node 7 a)s~ s ~ ~l~J~/P7 Z8 Intere Srk Inside CHIEF 3H Cmt Stress 30 hnt Cooling Provide appression pool teeperatcre and level indicatim to @@part NR systee I7f) sqqressim pool cooling ~)SQ ~OrI<< ~I>Hgt+ ywE7 25C Snail Break Inside CCEF 3H Ccn't Stress 30 Cent Cooling Providesuppressicn pool teeperature and level indicatim to support NR systee (7I) suppressim pool cooling ~ ode p)sc ~~ ~)i'<y,~~//f P< 250 Erklnside+mtNad COEF 3H Cont Stress 30 Cmt Cooling Provide suppressim pool teeperatore and level indicatim to support NR ~ode>n)A ~ systee I74) suppressicn pool cooling cooli~y. /gal. ~~/f1 25D Brklnsld~tSad CtEF 3H hnt Stress 30 Cmt Cooling SF Uacuue breaks systee prevents dryeell cr Tcrus negative preaare froe daeaging ccntainaent structure. 27 Pipe Sreak Outside IXEF 3H hnt ess 30 hnt Cooling Providesuppression pool teeperabee 32$ Lou hservoir COEF & hnt ess 30 hnt holing Provide suppressim pool teeperature and systee I7l) suppressicn pool cooling ISO ISO Setdown COEF M hnt Stress 30 Cmt holing ~ odeum abo leo ~ neliy. g~]~ Provide suppressicn pool teeperature f1 and level indicatim to support NR systee I7$ ) appressim pool cooling node nb 4 re<~ 4iMI~K)r f I SAFE S(ITHN NOLTSIS Page )80 of )c) l Preps ed: Date: Z PRI)NT CI(TAINSIT E. k STSTE)( Oecked: Date: IRHM~ (Rev 1) C. L Parker REACT MXEPTAME SAFE 5%R RENTS ' SAFE(V ACTISI F(K SEL GM TITLE STATE Code Title Code Title GEE 'ISD ISD Shutdofo CKF M Cmt Stress 30 Cant Cooling Accept S(Ns stean bloffdofaf (fry boiler vents and drains systen, 10) to suppressim pool; (XEF 2H Cont Stress 30 Cmt Cooling Provide suppressim pool teeperatea and level lndlcatim 'to suppcÃt R% systea (7l) suppressicn pool ccoling ~ ode> AtS'o paar 4eoliffg. E~~ ~la KD Hcraal Shutdofe . (XEF M Cont Stress 30 Cmt Cooling Accept S(Ns stean bloffdaef (frcn boiler vents and drains systaf 10) to suppressim pool. 'SD Safety Shutdofe (XEF 3H Cmt Stress 30 Cmt Cooling Provide suppressim pool teaperature and level indicaticn to apert (N systen (7f) suppressicn pool cooling ~ ode. Provide dr)ffell prcsare and teeperature indicaticn to suppfrt M s sten (70 cmtainaent spray aude. rov:da RH<474)mao jul'"b SSD Safety Shutdae (XEF 3-2 Fuel Failure 30 Cent Cooling Accept %Vs steaa bloffdae (frcn boiler vents and drains systenf 10) to suppression pool. 03 ((ain Sta Isolaticn (XEF 3-5 Pers Overexp 36 Cmt Day Env SF ~ RPS(gg) Iaf ffater level (Q) f send signal fron isolaticn signal to air ( safety actim is expected to cxxurf but is not a r renent fcr, this (( t f"vo(psfj, ip rgb'j%$ ur...!L, 05 Loss of llacuua (XEF 3-5 Pers Dver f Ilay Enfc 3 IIpcn lou &e'r."Ie've) (L3);.signal frost k RPS(gg), send isolaticn signaI to air conditicning systee (31). This "~yaHcim is expecWefTb R5a'f but is not a requireeent fa this event since there is no fuel failwe. 13 lnadvert Dpen S(N CDEF 3-5 Pers Overexp 3( Cmt hy Env SF Upon Iaf ffater level (Q) signal fron RPS(gg) f send isolation signaI to air cmditicning systen (31). This safety acticn is expected to occd'f but is not a requireeent fcr this event since there is no fuel failtn. ~ ~ ') 07 SAFE SHIINN NCLYSIS Page 181 Prepared: Sate: PRIIIARY GNTA1IM E. C Eck SYSTE)i theckedx sat: BF)H)SS3~ IRev 1) C. L Pxker BM- REACT MXEPlABLE SVE IRR RESILTS SAFEIY ACTIIN RK SEL BE)IT TlllE STATE Code Title Code Title GEE ll Loss of F)i Flow IXEF 3-5 Pars Overexp 36 Cmt hy Env SF Open low water level IQ) signal free RPSIlI9) x send lsolatlcn signai to air cmditiming systea Dl). This safety actim is expected to mLr, but is not a requireaent fcr this event since there is no fuel failtre. 15 Loss Offdite AC NL M Pers Overexp 36 Cmt hy Env SF Opm lcw water level IQ) signal froa RPSN), send isolation signal to air cmditiming systea Dl). lhis safety actim is expected to cccxr, but is not a requireaent for this event since there is no fuel failur. 23 F)i Ctrl Fail~ GXF 3-5 Pars Overexp 36 Cent hy Env SF Opnn low water level IQ) signal froa RPSIgg), send isolatim signal to air cmditiming systea I31). This safety actim is expected to occLr, but is not a requireaent fm this event since there is no fuel failLre. 24 Rod Srcp Accident SF 3-5 Pers Overexp 36 Cmt hy Env SF ~ low water level IL3) signai froa RPSN), send isolaticn signal to air cmditicning systeaI31). 25A Large Break inside CSEF RPSN), send isolatim signal to ai p pondLM~ysteaPDQ f, j~ lila ~l x x 25A Large Break Inside 5 I a):a'tm(. +xIviF pPgpj a cn x 's)gnal to air cmdit>cmng sys ea Dl). 258 lntera Brk Inside IXEF 3-5 Pars Overexp 36 Cmt hy Env SF Opan low water level (Q) signal froa RPSIgg), send isolaticn signal to air \ cmdltiming systeai31). 258 Intera Brk Inside CSEF 3-5 Pers Overexp 36 Cant hy Env SF Send high drywall pressure isola'tim signal to air cmditiming systea Dl). 25C Saall Ik eak inside CKF 3-5 Pers Overexp 36 Cent Bay Env SF Open low water level IQ) signal free I, RPSIgg), send isolatim signal to air canditiming systeaDI). 3o7 ~ SAFE S)$ )NN ANALYS1S P 18? Prepared: Oate: PRTMY QNA)lt8lT E. SYSTE)i Oiecked: ltate: TF)H)S&Ol8 IRev 1) C. L Parker EVENT REACT QNKEPTASLE SAFE N

• 5%R RESETS SAFETY ACT1% RK SEL EVENT TlTLE STATE Code Title Code Title CKE 25C Saall beak Inside IXEF 3-5 Pars Overexp 36 Cont Say Env SF Send high dryuell pressure isolatim signal to air cmditicning systee l31).

250 Srklnsi5e+mtlM ~ COEF 3-5 Pers Overexp 36 Cmt hy Env SF tjpm lou water level lQ) signal froe RPSlgg), send isolatim signal to air cmditiming systee13)). 250 SrklnsidHmtQhd COEF 3-5 Pers Overexp 36 Cmt i)ay Env SF Send high dryuel1 pres@re isolattm signal to air cmditicning systen 01). 250 Hrklnsid~tLRad COEF 3-5 Pers Overexp 36 Cmt hy Env SF Logic provided fcr high reactcr building radiaticn signa)s frm systen g0 and sant to Systen 31. 26 Fuel Rand) Accidnt A 3-5 Pers Overexp 36 Cent hy Env SF Provide signal to initiate isolatim of cmtrol rom air supply h initiate eeergency pressurizaticn functim of Cmtrol Sui)ding HVAC ISystee 31) after receipt of reactcr building exhaust high radiatim signal frm systee 90. 27 Pipe Break Outside CTKF 3-5 Pers Overexp 36 Cent hy Env SF Open lou uater level lQ) signal froe RPSlgg), send isolaticn signal to air cmditiming systeel31). 33 Loss Fuel Pool 1 NL 3-5 Pers exp y v te the eeergency pressurixati functicn of i it < F." p >'t ) q ii m iN. v, pe sstrolibpl(gi 01). 2S Shtdn Froe Outs Vg qI'jj~'~! par'e )Qr.""' si ",'its, 'to nhn t 3: Send tee l01) to close all SIVs aft0 rip of the IPS RPS isolaticn logic. 2S Shtdn Froe Outside IXEF H Local Bet 3g Cmldn Outsd Pool teeperature and level indication is provided at the Rackup Control ~ Center. ProVi& R)4f-'oc~ b4) Cooling, W<C +Mp7. ') o7 SAFE SHJMN NQLTSIS Page ll6 of Prepared) Oate: PRINART GNAIQM E. k SSIEN ONcked> Pate: %HEN~ IRev 1) C. N. Parker BENT RFXT MXP7ARE IS- IVER RESILTS SAFETY ACTIQl RK SEL 6ENI TIRE STATE Code Title Code Title CNE GERMS XL 2S ~Q ~QQg(g Shtdn Froa Outside C&F+I CeCal Slabs IXEF 5-I Local Shut yg S C (a.~ Cooldn Outsd R ~ 44L aAW~f;~4~W( ptp45t~ ~ a '4 sly ~~ ~C Provide alternate smrce water supply I mw. for RCIC systee I71) through M ill) systee piping froe suppressim pool. 2S Shtdn Froe Outside IXEF 4-) tocaI Sit 39 Cooldn Outsd Accept QNs steae blowdae Ifrrs boiler vents and drains systee, 10) to suppression pool. Nl 4-3 Eguip Fail 0 Seisaic Shtd Oesign of safety features ast have capability to provide shutdown folloring a design basis seisac event. SAFE SHJEM AN.TSIS P 'repared: STOaSr SAS TREAT)(T E. Oec'ked: Oate; .IF)H)SQ~ (Rev I) C. 1 Parker EUM REACT MXEPTAKE SAFE OR CPER RES%IS SAFETY ACTISI RK SEO. EUM TITLE STATE Code Title Code Title GEE (0 Nain Stn iso] atim (XEF 2-1 Rad Release 27 Est Sec Cent SF Initiate SST plant start cn ]or water level (Q) sic@ frow pr]nary cmtainnent systen (M). This safety actim is apected to occur, but is not a requireant fcr this event since there is no fuel failcre. 05 Loss of Uacuun (XEF '2-1 Rad Release 27 Est Sec Cont SF initiate SST plant start cn ]m water level (Q) s]qna] froa prissy cmtainaent systen (M). Ibis safety action is expected to ocnr, but is not a requireaet for this event since there is no fuel failure. ]3 Inadvert'pen SRU CKEF 2-1 Rad Re]ease 27 Est Sec Cent SF Initiate SST plant start cn ]a water level (Q) signa] frow priaary cmtaineent systen (M). This safety actim is expected to occur, but is not a requirenent for this event since there is no fuel failure. ]4 Loss of F)( Fiow CDEF 2-] Rad Release 27 Est Sec Cent SF Initiate SST plant start cn ]os water level (Q) sicko] frow pr]nary ccntainnent systen (M). This safety aHTR not a requlrmnt fcr this event 15 Loss Offdite AC 1 e fbi,fst@gt<k r i ik level ( k<plptit)i a t s)qna] fr'4 priary cmtainnent systen (N. This safet not a requirenent fcr this evmt since there is no fuel fa]lure. 15 Loss Off-boite AC AU. 2-1 Rad Release 27 Est Sec Cont Suppcrt secmdary ccntainnent fmct(m. 15 Loss Offdite AC Nl 2-1 Rad Release 27 Est Sec Cont F lorna] cmtainaent ventilatim is lost. The Stindhy Sas Treatnent Systen is started m loss of ncrna] ac poor siqna] (ron primy ccnta]nnent systen (6l) to naintain secmdary cmtainaent ]ntegr]ty. rod ~ SAFE SI(T(OI( A)NLYSIS Page 185 of Preparedt ()ate: STNfaiSY SAS TREAT)(T ChKked: Date: BF)H)563~ (Rev 1) BENT REACT (NA(XEPTARE SAFE (I Mt RE5(LTS SAFEIY ACTI(I( RK SEL E(JE)(T TITLE STATE Cade Title Code Title CO()E 23 R( Ctrl Fail~ C(IEF 2-1 Rad Release 27 Est Sec Cmt SF Initiate SST plant stvt on low water level tQ) signal fran priaary cmtainaent systen tbf). This safety actim is espected to occur, but is not a requirenent fl this event since there is no fuel failure. 24 Rod ()rop Accident lf 3-1 Rad Release 27 Est 5ec Cant SF Initiate %T plant start cn law water level tQ) signal frow prinary cmtainnent systee N). lSee Systen bl, Safety Actim 2b) 25A Urge Break Inside CBEF 3-1 Rad Release 27 Est 5ec Cont SF Initiate SST plant start m Iaw>>ater level tQ) signal frm priaary cmtainaent systee N). tSee Systen M, Safety Actim 2b) 25A targe Break Inside Rad Release 27 ~ Est Sec Cant SF Initiate SST plant start m high drywel I presme signal fran priaary cantainaent systen tbh). 258 Intern Brk Inside CEF 3-1 Rad Release 27 Est Sec Cant SF Initiate %T plant start m law water level (Q) signal fran prinary cmtainaent systen N). binary l5ee Systen M, y 25B Intern Brk Inside q high free prieary 25C Snail Break Inside Est Sec Cont SF Initiate SST plant stvt m Iaaf water containaent systea N). (See Systen M, Safety Action 2b) 25C Snail Break Inside (XEF 3-1 bd Release 27 Est Sec Cont SF Initiate SST plant start m high drywell presare signal free prinary cmtainnent systen (Q). 25(j Brklnside4xtÃad C(IEF 3-1 Rad Release 27 Est Sec Cont SF Initiate %T plant start m low water ~ level (Q) signai fran priwary cmtunnent systen (M). l5ee +ten bl, Safety Acticn 2b) 7 SAFE SWMM INLySIS STANON $8 TREAT)IT Prepared '. Page )86 Oate: Checked: Bate> BF)HSS3~ IRev 1) C. L Parker BE)IT REACT QIACCEPTAKE SAFE OR SG RESETS .SAFETY ACTI% FIK SEQ. BENT TITLE STATE Code Title Cade Title CNE '250 Brklnsi~tNad IXEF 3-1 Rad Release 27 Est Sec Cmt SF Initiate SST plant start m high dryuell pressure signal free prinary cmtaineent systen IQ). 25D Brklnside-CantSad COEF 3-1 Rad Release 27 Est Sec Cent Support secandary ccntaineent fmctim. 2h Fuel Hand) Accidnt A 3-1 Rad Release 27 Est Sec Cmt Support secmdary cmtalllKAt fmcticn. 2b Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont SF Standby Sas Treatnent Systee ~aintains negative vacuue in the seccelary ccntainaent. Activated fran a signal fran priaary cmtainmt systen I64). 27 Pipe Break Outside COEF 3-1 Rad Release 27 Est Sec Cmt SF Initiate SST plant start cn Ical uater level IL3) signal froe prieary ccntainaent systee N) ISee Systee Q, Safety Actim 2b). This safety action is expected to occur, but is not required. 33 Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cant Support seccndary cmtaineent functicn. 33 Loss Fuel Pool Cl Nl 3-1 Rad Release 27 Est Sec Cant SF )laintain the reacta building at a negative vacuua upon excessive radiation signai fran priaary containeent systee Ibl). SAFE SflIOQN QQLYSIS Page 167 of Dj Prepared: 57~ STSTEN Dwchdt ate: RHKH~ fRev 1) BM REACT LMNXPTAKE SAFE N fPER RESETS SAFETY ACT1Cff RK SEO. BM TTTLE STATE Cade Title Code Title GEE '31 Eundle Error lf 2-1 Rad Release 22 Rest Ncreal N Control plant offgas process to assist ccntrol rod/paar changes as operation is adjusted to stay uitbin aoraal operating liaits, 25D Srklnsid~tLRad CKF 3-1 Rad Release 27 Est Sec Cent L Passive: Nain stack necessary to provide analyaed distributicn of gases exhausted by SSTS (6$ . ~ 0 ~ SAFE QUINN ANALYSIS E)ARSE)D m.1% ERIIPIM )(ATER SYSTEN NHM~ (Rev I) Prepared: Checked: Page 188 E~ Ec C. L Parker gua7 Date: Oat~ ~ 9 fP'M SYSIEN 67 REACT (MXEPTAB(E SAFE N (EB RES(LTS SAFETY ACTIBi RK SH). GENT TITLE STATE Code Title Cafe Title (mE 13 lnadvert Open 9N COEF 2-2 Fuel Failure 19 Core Cooling SF Provide cooling water to diesel generatcr systen (82) starting on Iow reactcr water level (Ll) signals. Ibis safety action nay occur, but is not required by this event since ncrnal ac is available. 13 Inadvert Open SRV (XEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide cooling ~Qq tgR)R~~~py (74) and.ccrc spray>rue coolers (64) in support of HXS actions at low reactcr water level (LI). 15 Loss Off-Site AC CKF 2-2 Fuel Failure 19 Core Cooling SF provide cooling water to diesel generatcr systee (8?) starting on low reactor water level (LI) signals. 15 Loss Off<ite AC COEF 2-2 Fuel Failure 19 Core Cooling SF Provide cooling what tgP~I~<gp and core spray>roon coolers (64) '74) in support of HXS act(ms at Iow reactcr water level (Ll). 22 Loss Shutdoo Cool ACO e 19 Core Cooling N Support operation of M(74) IJCI 24 Rod Crop Accident 3- Qusf F/jR 1$ lpga!CoolkpSf> If poj~ligy water to diesel I VI( r. 24 Rod Orap Accident Fuel Failure 19 Ccrc Cooling SF Provide cooling watggt 5%~I (74) and ccrc spray roon cooTers (64) an sup reactcr water level (LI). ZA Large Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide cooling water to diesel generatcr systen (8?) starting on high drywell pressure signal. Large Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Prtmde cooling water to I ~@~ (74) and to core spray>r oolers (64) )or HXS actions on high drywell pressure SAFE '(HJEM AN.YS]S fKREE)Cf Ellip)ENT Prepared: Page E. I Ec $ 4)7 gate: ~7 M.I% NTER SYSTE)( DIecked: Bate: BF)H)SB3~ Nev 1) C. L Parker SYSTE)( 67 I EVENT SH). EVENT TIILE REACT PER IA(XEPTAELE RESI.TS STATE Code Title SAFETY Code ACTIN( Title SAFE RK (IE 25A Large Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide cooling water to diesel generator systen (B2) starting on Iow reactm water level (Ll) signal. yC~ 25A Large Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Frmide cmtiog mtarQo ~t (7l) and cm e spray r oo ers N) in support of HXS actions at low reactm water level (L1). 25B Intern Brk Inside COEF 3-2 Fuel Failure 19 Care Cooling SF Provide cooling'water to diesel generatcr systen (82) starting on low reactor water level (Ll) signals. 25B Intern Brk inside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide cooling water tg N~~l f~~ (7$ ) and cN e spray>rona ccoTers N) in suppcrt of HXS acticns at low reactor water level (LI). 25B Intern Brk Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide cooling water to diesel generatcr systen (B2) s'taf ting on ll essure signals. 258 Intern Brk Inside COEF 3-2 F(OP 'l l9 Cleat(tgtt'rovide cmt tag eate~rt ~t(4 25C Seal) Break Inside COEF 3- 9 Ccrc Cooling SF Provide cooling water to diesel 9 reactor water level (Ll) signals. 25C Snail Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF ~ Provide cooling wat~er to ~I '~Py c<~ (7i) and core spraygooa cco ers N) in support of HXS actions at low 'reactor water level (L1). 25C Snail Break Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide cooling water to diesel generator systen (B2) s'tartlng ol high drywell pressure signals. ZC Snail Break Inside CCCF g-9 Fmi Failure l9 Ccrc Ccoliog CF Provide cooliog eater tgII~F ~4 (7I) and to ccrc spray>roon coolers (bl) fcr HXS acticns on high dryweli pressure SAFE SHJTIOII ANALYSIS Page 190 $ d) 7 Prepared: / E)6 6EKY GDIP)BIT I.I% NTER SYSTEN Checked: Date: BF)H)663~ IRev I) C. N. Parker SYSTEN 67 I BE)IT SEO. BJE)IT TITLE REACT IVER MXEPTABLE STATE Code RESILTS Title ~ SAFEIY ACTIQl Code Title SAFE FlK CODE '7 Pipe Break Outside CDEF 3-2 Fuel Failure 19 Core Cooling SF Provide cooling water to diesel generatcr systee IB?) starting cn Iow reactcr water level ILI) signals. 27 Pipe Break Outside COEF 3-2 Fuel Failure 19 Care Cooling SF Provide cooling wag>tgfi-LPCI f~4 ( I74) and ccrc spray~roan caolers 164) in support of HXS acticns at law reactcr>>ater level ILI). 3?E Fire ALL 4H Fuel Failure 19 Ccrc Cooling Design of safety features an't include Appendix R fire canst der atl ense ISD ISO Shutdoo COEF 2-2 Fuel Failure 19 Care Cooling N Support M I74) lKI operaticn. NSD Horaal Shutdcee COEF 2-2 Fuel Failure 19 Core Ccoling N Support M I74) LKI operaticxl, SSD Safety Shutdcxo COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Support M I74) - lKI andhr diesel generatcr systea IB?). 2? Loss Shutdown Cool ACD 2-2 Fuel Failure 22 Rest Horaal N Support cperaticn of MI7\) hactcf Shut~coling. IS Loss Off<ite AC 2-1 Rad Release 27 Est Sec Cont Support seccn y functicn, KO Brklnsid~tNad 26 Fuel Rand) Accidnt Rad Release 27 Est Sec Cont Support secondary ccntainaent Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cant Suppart seccndary ccntainaent functicn. OIB T/6 Trip-BP Fails F 2H Cant. Stress 30 Cont Cooling NSF Suppmt M systee I74) suppressicn pool cooling aade. 01C PresReg Fai)<lose CDEF 2-4 Cant Stress 30 Cent Ccoling NSF Support M systei I74) suppressicn pcol cooling sade. M systee 'DEF 03 Nain Sta Isolation 2-4 Cont Stress 30 Cant Cooling NSF Suppcrt I74) suppressicn pool cooling sade. 9UEM ANALYSIS SAFE Prepilred) oC P 191 Date; -0 E)GKE)D GUIP)8fT mLI% )IATER STSTEN IF)H)SS3~ IRev 1) STSTEN 67 Oecked: E. Ec C. N. Parker Date: EUE)IT REACT MXEPTARE SAFE CR CPER RESILTS SFETY ACTICII FBC SED. EUE)IT TITLE STATE Code Title Code Title mE '05 Loss of Uecuun IXEF M Cant Stress 30 Cant Cooling NSF Suppcrt M systen 174) suppression pool cooling node. 12A Pres Reg Fail+en CCEF 2H Cont Stress 30 Cant Cooling NSF Suppcrt M systee f74) suppression pool cooling node. 12B IMdv Open+piss CCEF 2H Cent Stress 30 Cant Cooling NSF Suppart M systea 174) suppression pool cooling node. 13 Inadvert Open SRU IXEF 2-4 Cont Stress 30 Cant Cooling NSF Suppcrt M systee 174) suporessian pool cooling cade. 15 Loss Offdite AC COEF 2H Cont Stress 30 Cont Cooling NSF Support M systee f74) suppressicn pool cooling node. 15 Loss Off-Site AC CDEF 2-4 Cant Stress 30 Cant Cooling N Support drpeII cooling function by M systen 15 Loss Dff<ite AC CDEF 2H Cont Stress 30 Cont Cooling NSF RBCCM (~~)~'- systen f70<)cr Provide suppcrt to di f74) I generator ) systea IB2). 22 Loss Shutdoe Cool CD 2H Cent Stress 30 Cant Cooling N Support operation of MI74) pool cooling. 24 Rad Drop Accident DF 3-4 Cent Stress 30 Cant Ccoling NSF Support Msysten (74) suppressicn pool cooling aode. 25A Large Ik eak Inside Cooling NSF Suppart M systee I7 ) qsi ZB Intern Brk Inside 'i;.' 'pcol cool'ing node. 25C Snail Break Inside IXEF 3H t Stress 30 Cont Cooling NSF pool cooling node. ZD Brklnsid~t4Rid CKEF 3H Cant Stress 30 Cent Caoling NSF Suppcrt M systee <74) suppressicn pool cooling node. 27 Pipe Break Outside 'KF 3H Cant Stress 30 Cont Cooling NSF Support M systea 174) suppression pool cooling cade. SAFE SHHHN AN.YSlS P e)92 >1 Prepved: Date: Z4' BERKS GN)IPISfT . E. NLllS NTER SYSTEN Checked: Date: NHEH~ lRev 1) C. N. Parker SYSTEN 67 DENT REACT MXEPTASLE SAFE OR 5%R REQLTS ' SAFETY ACTIN RK SEO. BM TlTLE STATE Code Title Code Title C00E Low Reservoir CDEF 2H Cent Stress 30 Cont Cooling NSF Support M systen (74) suppression pool cooling node. 32E Fire CDEF 44 Cont Ccndtns 30 Cont, Cooling Design of safety features <<ist include Appendix R fire Support consideratims. ~ 'SD 1SO Shutdoe CKF 2-4 Cont Stress 30 Cont Cooling N Support RHt systen 174) suppressia pool cooling functicn. ~. gCC'~+ drywe),l cooling Ae) functicnSf NSD Nornal Shutdiwe CDEF 2-4 Cont Stress 30 Cent Cooling N Support IR systen 04) suppressicn pool cooling function. Suppcrt drywell cooling functicnA~~ ~f/nor-. +E~>~/~ SSD Safety ShutdoIi C0EF 3-4 Cont Stress 30 Cent Cooling NSF Support M systen 174) suppressicn peal cooling node and/or ccntainaent spray node, and support diesel generatcr operatl coo 28 Shtdn Fron Outside ALL. 4-1 Local Shut 3g Caoldn Outsd N Support systea required fcr M 174) and Diesel Seneratcr (02). ZX Tcrnado Nl 4-3 Equip Fail 40 Struct Integ L Circulating water puaps <<St aaintain capability for shutdoe heat reaoval under tanado design conditions fpassive). 1$ Loss Off-Site AC ALL 2-1 Rad ease. *43 FuelPoolCool N Attach a fire hose to the EEC)f to ain the water level in the fuel pool o 33 Loss Fuel Pool Cl ALL 3-1 Rad e g RAa7ih,  ;"jn4i()g @ grFplNe )tat@1 'f @e.EEC)fko jn tJe3$ .'t ~> vh, 320 Earthquake ALL 4-3 Shtd Design of safety features <<ist have cap tdae following a design basis seisaic event. SAFE SH)TNN ANALYSIS Page Prepared: Date: V~ss . REACT% I(ATER RECIIQLATICN SYSTE)I )F)H)SS3~ IRev I) Checked: C. L .E t Parka'1 Date: BIBIT REACT IMXPTASLE SAFE OR IFEI( RESILTS SAFEIY ACTIOI( RK SEO. EIJENT TITLE STAlE Code Title Code Title IE '3 Inadvert Open SW CCEF 2-2 Fuel Failure 19 Ccrc Cooling SFS Provide Iow reacts pressure peraissive signals to ccrc spray systen (75) and R)f( systen (74) fm BXS actions at Iow reactcr water level ILI). 13 Inadvert Open SRV COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Close recirculation discharge valves for lKI injection cn signals frcN RH( systen (74) indicating lKI initiation signal due to low water level (Ll) in conjuncticn with Iw reactcr pressure pernissive signal. 15 Loss Off-Site AC COEF 2-2 Fuel Failure 19 Core Cooling SFS Provide Iow reactor pressure peraissive signals to core spray systen (75) and M svstee 04) f(r ECCS actlcns at Iow feactcr water Ielvl (Ll). 15 Loss Off-Site I COEF 2-2 Fuel Failure 19 Core Cooling SF Close recirculation discharge valves fm IJCI injection cn yignals froe If( systen (74) indicating LKI initiation eve 'on signal due to Iow water reactor pressure peraissive sign <<ith low 24 Rod Orcp Accident OF 3-2 sysEss'I'ysl'sss'48) sys)sy,)ys')'Jsy, HXS actions at Iow reactor water I). 24 Rod Drop A:cident OF 3-2 Fuel Failure 19 Core Cooling SF Close recirculation discharge valves fcr lKI injection on signals fron R% systen (74) indicating lKI level (LI)a reactor pressur 'ow initiation sign I due to low water eraissive signal. ZA Laroe frea'k Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SFS Provide Iow reactor pressure peraissive signals to ccrc spray systen (75) and RH( systen (74) (..., acticns at high dryweII fa'CCS pressure. SAFE )UEM ANALYSIS Page 19 of Prep ved: Date: REACTI NTER E. RECIRC(UITIN SYSTE)( Checked: Date: BF)H)SS3~ (Rev I) C. L Parker E(JENT REACT IACCEPTABLE SAFE N EPER RES(LTS 'AFETY ACTIN FlK SEL EVENT TITLE STATE Me Title Code Title CDDE 25A Large Break Inside CDEF 3-2 Fuel Failure 19 Core Cooling SF<<endue Close recirculation discharge valves fcr LKI injection on signal fry M systei (74) LKI initiaticn signal to high dryueII pressN'e,4n ~4K IPaajunaiewsith lm reactm peru( sslve sl9,lal ~ ~4~ ZA Large Break Inside COEF 3-2 Fuel Failure 19 Care Cooling SFS Provide Icw reactcr pressure peraissive signals to core spray (75) and R)fl systea (74) fcr 'ystei ECCS acticns at Iou reactor water level (LI). ZA Large freak inside CDEF 3-2 Fuel Failure 19 Core Cooling SF Close recirculaticn discharge valves fcr LKI injection cn signal fra M systea (74) indicating IKI initiation signal'gue to Iu uater level (Ll), ~l reactor pressure peraissive signd. 25B Intera Brk Inside CDEF 3-2 Fuel Failure 19 Core Cooling SFS Provide Iae reactcr pres erN slve 19 al toe ray s 74) fcr re t uater 'Z 1 "I .' ii' ~ '" IdeVeI'IILI)l / ~ 25B Intern Brk Inside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SF Close recirculation disch e valves cr ln]ec ion cn sl s froa I'l~, RHR systea (74) indicating lKI initiation signal due to Iw ua~t reactor pressure pereissive signal. ZB lntera Brk lr,side CDEF 3-2 Fuel Failure 19 Core Cooling SFS Provide Iou reactm pressure peraissive signals to core spray systea (75) and R)fl systea (74) fa'CCS acticns at high dr)nell pressure. ZB Intera Brk Inside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SF Close recirculation dis;harge valves fcr LKI injection on signals fraa M systea (74) lPCI initiaticn ~CC. signal due to high dg pell pressure~ ~l lou reactN'p'/@+M~ pef Qsslve signals. g.+ivy ~ ~u 3W ~ SAFE QUINN ANALYSIS e tt5 Prep ved: e Date: REACTN NTER E.. E SYSTB( 'EIRC(1ATIN Checked: ttete: BFN~~ (Rev 1) C. L Parker DM REACT (N(XEPTABLE SAFE OR (fG RES(LTS SAFElY ACTIN FlK SEQ. EitEHT TITLE STATE Code Title Code Title (IE 2X Snail Break Inside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SFS Provide Iw reactor pressure pernissive signals to ccrc spray systen (75) and Nf( systen (74) for E(XS actions at lw reactor uater level (LI). 25C Saall Break Inside CBEF 3-2 Fuel Failure 19 Core Cooling SF Close recirculation discharge valves fcr LKI injection on signals fron R% systen (74) indicating LPCI initiation signal due to lw sat level (LI),i &/> reactm pressure pernissive signal. 25C Snail Break Inside CDEF 3-2 Fuel Failure 19 Core Cooling SFS Provide Iw reactor pressure peraissive signals to core spray systen (75) and IR systen (74) fcr E(XS acticns at high drytte)I pressure. ZC Snal( Break Inside CDEF 3-2 Fuel Failure 19 Core Cooling SF Close recirculaticn discharge valves for LKI injection on signals froa RN systen (74) LKI initiaticn s'gnal due to hioh drvueli pressure,c(nd w reactcrpr~" ~ W Paissive signals. Mgmlt1 27 Pipe Break Outside s'e,'ence@ c pernissivegiggaIT go gory ray

74) fcr ter C

27 Pipe Break Outside CDEF 3-2 Fuel Failure 19 Core Cooling SF Close recirculaticn dischv valves an ) ec >on on si 9 nals fran RH( systea (74) indicating LKI initiaticn signal due to lw t PC level (LI)>' w0" reactor pressure pernissive signal. 25D Brklnsid&cntNad CDEF 3-1 Rad Release 20 W lsol Suppcrt W isolation function. 17 Trip I RKirc Punp CDEF 2-3 Syst Stress 22 Rest Ncraal )( Restart puap and restore the systen fN n(baal operation via na'nai procedures. SAFE S)TDNN ANALYSIS Pa e 19 3%7 'Prepared: S ~e: ~iZi fSCTOR NTER E.C Ec RECIRQ1ATIOH STSTEN Checked: Date: ~ IAH86~ (Rev I) C. N. Parker GENT REACT MXEPTABLE SAFE OR IPER RES%IS 'AFETy ACTIN FlK SED. BSH TITlE STATE Code Title Code Title imE 18 Trip Both Recirc. CDEF 2-2 Fuel Failure 22 Rest Homal N Restart each puap per ncrnal puap star t procedures. 21 Startup Idl Recirc CDEF 2-2 Fuel Failure 22 Rest Hiraal N Establish initial pour level properly, and balance both loops after puap start )7 Trip I Recirc Puap F 2-2 Fuel Failure 23 Lia Disturb L Recircul~ticn puap notor inertia coastdao/cmsuned in the FSAR,w I l 18 Trip Both Recirc. F 2-2 Fuel Failure 23 Lia Disturb L Recirculaticn pusp aotor inertia coastdoo assuaed in the FSAR is iapcrtant to the results of the trip event. 20 Recirc Fail-incr F 2-2 Fuel Failure 23 lia Disturb L F98 analysis assunes a W of full speed/sec rate of change fcr lR setspeed ccntroller failure. 20 Recirc Fail-Ircr F 2-2 Fuel @ate 2~iYiib ion flow controls shall be naintained nical 21 Startup Idl Recirc CDEF 2-3 Syst Stress 23 Lin%s urb i4 Ff t/ xAhtk Rb)I))9y Loop befcre puap start. 01A Generator Trip F 2-2" Fuel Failure 24 Pox Reduce L Open recirculation puap notor breakers on RPS.(99) sional due to ~ ain turbine ccntrol valve fast closure and > 30'. turbine first stage pressure. Coastdoe aust be faster than assuned in reload analysis to Iinit severity of the event. 01B I/O Trip-BP Fails F 2-2 Fuel Failure 2f Pou Reduce L Open recirculation puap enter breakers cn RPS 199) sional due to ~ ain turbine ccntrol valve fast closure and > 307 turbine first stage pressure. Coastdoe aust be faster than assuaed in reload analysis to liait severity of the event. ~ " ~ ~ I ~ SAFE QUTIXN ANALYSIS Page )97 e - REACTOR NATER RECIRCKATISI SYSTE)I ERHKH~ (Rev I) Prepared; Checked: C. )I. Parker I GE)IT REACT NNXEPTADLE SAFE OR OPER RESETS 'AFEIY ACTIN FIK SEO. MIT TIILE STATE Code Title Code Title CODE '2 Turbine Trip F 2-2 Fuel Failure 24 Pow Reduce L Open recirculaticn puap outcr breakers m RPS I99) signal due to ~ ain turbine stop valves < 9R open and >XI turbine first stage pressure. Coastdoe aust be faster than assuaed in reload analysis to linit severity of the event. 05 Loss of Vacuun F 2-2 Fuel Failure 24 Pow Reduce L Open recirculation ouxo notcv; breakers cn RPS I99) signal due to ~ ain turbine stop valves < 9'pen and >302 turbine first stage .pressure. Coastdon aust be faster than assuaed in reload analysis to liait severity of the event. I4 Loss of F)I Flow 2-2 Fuel Failure 24 Pow Reduce Recirculation IJ6 drive notor breakers f200, electrical systen) tripped on low water level IL2) signal. This safety action is expected to occur, but is not a requireaent for this 2S F)i Ctrl Fail~ ure 4 Pow Reduce L Open recircula'tion g+ ptp

brmkpr<crCPS Igg)kiipal'Idu '

xl~n fbi'pop 'val/y'4, 'ressure. Coastdoe aust be faster than assuaed in reload anal si e event. 39 Overpres Protect CDEF 2-5 Syst Stress 24 Pow Reduce Recirculation $ 6 drive notor breakers f200, electrical systen) tripped cn high reactor pressure signal. ZB Low Reservoir F 2-1 Rad Release 24 Pow Reduce Reduce core flow in respmse to signal frow recirculation flow control systeo l96). ,0 328 Low Reservoir F 2-2 Fuel Failure 24 Pow Reduce SF Open recirculation puao breakers m RPS I99) s>goal. notcr Coastdoo est be faster than assuned in reload anlysis to Iinit severity of event. This safety actim nay occur, but is required mly if Im vacuua turbine trip occurs. SAFE BUTD0W ANALYSIS Page lgS of $ 0y Prepared: ttnte: ~nl REACTOR )IATER F t IECIRDLATIN SYSTE)I Checked: Date: 8B SRH663~ (Rev 1) C. L Parker BE)IT REACT MXEPTASLE SAFE OR OPER REStlTS 'AFEIY ACTIN RK SEO. BENT T]TLE STATE Code Title Code Title IIE 'ZD Srklnsid~t Mad COEF 3-l Rad Release 26 Est Pri Cont Snppcct ycimcy cmteinnent fmctim. 28 Shtdn Froa Outside ALL 4-I Loca) Out 3g Ccoldn Outsd )I Position Recirculation valves to perait LPCI node of M. 29 Shutde )I/O Rods lf 4-2 Shut No Rods 42 Shutdn )IoRod Recirculaticn N6 drive notor breakers f200, electrical systen) tripped on its uater level (Q). 29 Shutdtyn )I/O Rods EF 4-2 Shut No Rods 42 Shutdn HcRcd Recirculaticn N6 drive enter breakers <200, electrical s)ysten) tripped cn high reactor pressure signal. y SAFE SHJTD(N ANALYSIS Prepared: P cd& 8' REACTOR NTER CLENP E. SYSTEN Checked: Date: 8RHSQ<48 (Rev 1) C. N. Parker SYSTEN 69 EVM REACT IN(XEPTAOLE SAFE N (RR RES(LTS SAFE)Y ACTIO(( RK SEO. 8 cHl I'ITLE STATE Code Title Code Title CODE '03 Nain Stn Isolation COEF 2-1 Rad Release 20 RPV Isol SF Close R)G) isolation valves m low water level (L3) sinnal frow prinary cmtainaent systea (6$ ). This safety action is expected to occur, but is not a requireeent fcr this event since there is no fuel failure. 05 Loss of Vacuua (XEF 2-1 Rad Release 20 RPV Isol SF Close RKU isolatim valves m low water level (Q) signal frow priaary ccntainaent systen (64). This safety actim is expected to occur, but is not a requirenent fcx this event since there is no fuel failure. 0 13 Inadvert Open SRV COEF 2-1 Rad Release 20 RFV Isol SF Close R)G) isolatim valves cn low water level (Q) sinnal fron prinary ccntainaent systea (6(). This safety action is expected to occur, but is not a requireeent fcr this event since there is no fuel failure. Il Loss of F)( Flow COEF 2-1 Rad Release 20 RPV Isol SF Close MU isolation valves on low water level (Q) sinnal fron priaary ccntainnent systen (M). This safety ec o cccur, but is not a requirenent for this event > jppprrvqo8un ~Up ~~ 15 Loss Off-Site AC ALL 2-1 Rad Release w'ahr I&ef (U) signal fron priaary containnent systen N). This safety a t not a requirenent for this event since tta'e is no fuel failure. FR Ctrl Fai)~ COEF 2-1 Rad Release 20 RPV Isol SF Close RXU isolation valves m low water level (L3) signal froa priaary ccntainaent systen (6(). This safety action is expected to occur, but is not a requireeent fcr this event since there is no fuel failure. ~ ~ ~  % ~ SAFE BUlMN ANALYSIS Page $ 0/ Prepar8)l ate: ~x.t E REACT% NTER CLEAIIs E. E ert SYSIEN Checked; Sate: ~48' RHM~ IRev 1) C. L Parker SSTE)1 69 EVENT REACT MXEPTAHE SAFE OR IRR RERlTS ~ SAFETY ACTIN RK SEO. EVENT TITLE STATE Code Title Code Title CNE '$ Rod Dry Accident OF 3-1 Rad Release 20 RPV Isol SF Close MJ lsolatim valves Nl low water level IQ) signal fra priaary cmtalnaent systN 164). 25A Large Break Inside CKF 3-1 Rad Release 20 RPV Isol SF Close M) isolatim valves m low water level (Q) sional froi priaary cmtaina8lt systN 1'), 25B Intern Brk Inside COEF 3-1 Rad Release 20 RPV lsol SF Close IG) isolaticn valves m level (Q) signal fra priaary 18'ater coltalna8l't systN lbt) ~ 25C Soall Brea'k Inside CDEF 3-1 Rad Release 20 RPV Isol SF Close M) isolaticn valves m Iow water level IQ) signal fra pri'aary cmtalna8lt systN Ibi) ~ 250 Brklnside+mtNad COEF 3-1 Rad Release 20 RPV lsol SF Close M) isolatim valves m low water level iQ) signal fra priaary containaent systN 16)) .. 27 Pipe Break Outside CDEF 3-1 Rad Release 20 RPV lsol ,SF Close RKU isolaticn valves low 27 Pipe Break Outside 3-1 <I i I t Rad Release f t RPV tt l"0 iso) J! D,'t :I s a'Y'i,jtttsftlivetfot)sftld <<I "! v' SF t s~tIttkstssge1 '. vP s Senscrs aonitcring <tt. l< ~ riaary tNperatur in the ataosphere and in drains in t eM) high teaperature signal to priiary containaent systei I6I). 27 Pipe Break Outside COEF 3-1 Rad Release 20 RPV Isol SF On signal fra systN 64, close isolatim valves thigh IG) tNperature). 25D Brklnside-Cent tRad CDEF 3-1 Rad Release 26 Est Pri Cont Support priiary ccntainaent fmctim. 15 Loss Off~ite AC Nl 2-1 Rad Release 27 Est Sec Cont Support seccndvy cmtailaent functicn. 25D Brklnsid~tSad COEF 3-1 Rad Release 27 Est Sec Cent Support seccndarycontainaent functicn. SAFE BUEM NQLYSlS Page 201 Prepared: e REACTIR NTER SYSTEN lRH6H~ CLENP llhv 1) Checked: BM I SEL BM TITLE REACT IM IMXEPTASLE RESETS STATE Code Title SAFETY Code ACTlSI Title SAFE FlK lmE 2b Fuel Nandl Accidnt A 3-1 Rad Release 27 Est Sec Cont Support secmdvy ccntainaent functim. 33 Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont Support secmdary cmtainnent functicn. 28 Shtdn Froa Outside ALL 4-1 Local Shut 39 Cooldn Outsd N Nanual isolatim capability outside ~ain control rona. h 29 Shutde )t/0 Rods 4-2 Shut No Rods 42 Shutdn NoRod Reactor l4ter Cleanup systen ast isolate upon initiation of the Standby Liquid Control systeeN). SAFE BUIUN NQLTSIS Page 202 of ) c)7 Prepared: REACTI HIILD1% CLOSED E. GKM NTER SYSTEN checked: Date: ERH66~ (Rev 1) C. 1L Parker SYSTEN 70 BIBIT REACT llfACCEPTAfLE SAFE fN CPER RESLLTS SAFETT ACTlN RK SEIL EIM T1TLE STATE Code Title Code Title GXE ZD Srklnsibe+mtNad CKF 5-1 Rad Release 2f Est Pri Cmt Support prieary cmtaineent fmctim. 15 Loss Of flite AC ALL 2-1 Rad Release 27 Est Sec Cmt Support seccndary cmtainant function. 250 Brklnsid~tLRad CDEF 5-1 Rad Release 27 Est Sec Cont Supper't secmdary cmtalneen t fmctim. 2b Fuel Nandl Accidnt A 5-1 Rad Release 27 Est Sec Cont Support secmdary cmtainant fmcticn. Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cent Support secmdary ccntaineent funct 1 me 15 Loss Off<ite AC REF 2H Cont Stress 30 Cont Cooling N Provide drywell cooling when power and cooling water are available. 1SD 1SO Shutdoe CDEF 2H Cont Stress R Cont Cooling N Provide drywell ccoling when poar and cooling water are available. NSD kraal Shutdown GXF 2H Cant Stress 50 Cmt Cooling N Provide dryweLL ccoling when power and ccoling water are available. 2S Shtdn Froe Outside CDEF f-1 Local Shut 5g Cooldn Outsd N Provide drywell cooling ~ power and ccoling water are available. Qo7 SAFE SUTDSH AN.YSIS Page 2Q Prepared: Date: Z REACTOR GK MATIN NUlS SYSTHI [%ected: Date: NHM~ IRev I) SYSIEN 71 )j: EVM REACT MXEPTAKE SAFE OR IPER REQLTS SAfEIY ACTIN RK SEQ. EVENT TITLE STATE Code Title Code Title CXE NfSGS Q Nain Stn lsolatim CKF 2-2 Fuel Failure 19 Core Cooling SF Provide ECCS ATU pea to F)f systen (Q) lw water level IL2) Instruaentatim fcr RCIC systen Initiatlme Q Rain Stn Isolaticn 'COEF 2-2 Fuel Failure 19 Care Cooling SFS RCIC initiaticn m lw reactor uater level IL2) signal frw R)ft systen Ill). Q Nain Stn Isolation COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide HXS AIU pwer to FU system fQ) lw uter level IL2) instruaentatim fe'PCI systea tH) initiaticn. 05 Loss of Vacuua COEF 2-2 Fuel Failure 19 Care Cooling SF Provide HXS ATU paar to P)f systea IQ) lw uater level IL2) instruaentatim for fPCI systen fyI) initiatim 05 Loss of Vamp'XEF 2-2 Fuel Failure 19 Ccrc Cooling SF. Provide HXS AIU pwer to Rl systen IQ) lw uater level IL2) instruaentatim for RCIC systen initiatim. 05 Loss of Vacuua COEF 2-2 Fuel Failure 19 Ccrc Cooling SFS RCIC initiaticn cn lw reacts uater level (L2) signaI fran M systen 12A Pres Reg Fail+en FaPyq %e~Qoij@SF~grip HXS AIUgoer to 8 fj I,ig), iw;4fer )evelyn('@ $ J's r t tihhati'cn< 12A Pres Reg Fail+en li SFS RCIC initiatim m lw reactor 04). 12A Pres Reg Fail+en COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide ECCS ATU pwer to FU systen fQ) lw uater level IL2) Instruaentatim fa ffCI systen (8), Initiatim. (0 0 SAFE 5WlMN ANLySIS REACTIR ONE GKlM STSTBI WHSE~ 15RATIQl tRev 1) Prepared: Ooked: 31 atm Oate: ~ ~ 7 SSTE)I 71 EVENT REACT MXEPTAELE SAFE OR SB RESILTS SAFETY ACTION FQC SH) ~ BBIT TITLE STATE Code Title hde Title CNE '128 Inadv Open~ass COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide HXS ATU peer to F)f systee (N) low water level IL2) iostrueentatim fe RCIC systen initiatim. 12S lnadv Open~ass IXEF 2-2 Fuel Failure 19 Core Cooling SFS RCIC initiatim m low reactor water level IL2) signai frow M systen f74). 12S Inadv Open-Sypass COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide HXS AIl) power to 8 systa ILT) low water level (L2) instrueentatim for kPCI systen Iyl) ini tiatim.. D lnadvert Open SRV COEF 2-2 Fuel Failure 19 Core Cooling SF Provide HXS AIU power to 8 systen NQ Iw water level IL2) instrumtatim fe'CIC systen initiatim, D Inadvert Open SN COEF 2-2 Fuel Failure 19 Core Cooling SFS RCIC initiaticn cn 1<x reactor water level (12) signal free R)R systen 170. D Inadvert Open SRV COEF 2-2 Fuel Failure 19 Core Cooling SF Provide HXS AIU power to FM systen

06) low water level IL2)

Instruaentatim fcr lfCI systen f73) initiatim. D lnadvert Open 9N COEF 2-2 Fuel Failure 19 Core Cooling SF Provide HXS AIU power to feedwater systee I05) low water level ILI) and FN systen Qbll fKlrcu~41m 6 (b8) low reactcr pressre a4,)l p Loss of.FN Flow CKF 2-2 Fuel Failure 19 e 8glljy < frvqdsQS,At/~@ 8 (Q) low water level (L2f tlkJ ( e. 1 At@ l instrmntatim for RCIC systen Il Loss of Rf Flew COEF 2-2 Fuel Failure 19 Ccrc Cooling SFS RCIC initiatim m low reactor water level IQ) signal frow RN systen l75). $07 SAFE Sf)INN AN.YSIS REACT(l( (XRE IL1% SYSTB( 15RATI(I prep ed) "Page 2% of E. C5 Dlts ~7 SB 0)ected> If)H)SSRNS (Rev 1) C. L Paher ~ ! SYSID( 71 SENT REACT MXEFIARE SAFE N 5%R RES(LTS SAFETY ACTIIN RlC SEO. aENT TITLE STATE hde Title hde Title GK 'll Loss of FN Flow- CKF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide HXS ATU power to FN systea (N) Im water level (L2) Instruaentatim fcr )fCI systea (H) initiatim IS Loss Off<ite AC (XEF 2-2 Fuel Failure 19,Core Cooling 5F Provide HXS AT() peer to FN systea NI) Im water level (L2) lnstrueentaticn fcr RCIC systea initiation. 15 Loss Off<ite AC (XEF 2-2 Fuel Failure 19 Ccrc Cooling SFS RCIC initiaticn m low reactcr water level (L2) signal frm M systea (75). '$ Loss Offdite AC (XEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide HXS ATU paar to FN systea .0 ((6) Icw water level (L2) instruaentaticn initiatim. fe'KI systea (H) 15 Loss Off<ite AC (XEF 2-2 Fuel Failure 19 hre Cooling SF Provide HXS ATU power to feedwate systea (03) low water level (Ll) and FN systea and rec(rculatim systea (N) Im reactcr presme instruaentatim. FN Ctrl Fail~ (XEF 2-2 Fuel Failure 19 hre holing 5F Pr oa toFN ystea ((6) l ow water level (L2) 23 FN Ctrl Fail~ @. 2, i 'Fiiijirjrt Gjeiho1isj+', PCS'initiatili cii ia rmis ts level (l2) signal frm R)w syst (7I). FN Ctrl Fail~ p~uel Failurf'l9 Ccrc Cooling SF Provide HXS ATU paar to FN systea (S Im water level (L2) instruaentatim fa IPCI systea (9) initiatim Rod Orop Accident IF 3-2 Fuel Failure 19 Care holing 5F Provide HXS ATU poor to feedwater systea N) Im water leeR (LI) and FN systea and recirculaticn systea (68) Im reactcr pres@re instruaentatim. '307 SAFE SIIIIIONI AN.YSIS Patt tt6 of Prepared: e IKACT(R CNK DXU)6 SYSTE)I BFIH)SSl~ (Ihv IS(UT19( 1) SYSTHI 71 Checkedl E L Parker Date: P7 EtM IKACT QNQXPTARE SAFE N CPER IKRLTS SAFEIY ACTI(l FQC SEL HJENT TlllE STATE Cade Title Cade Title (mE 2I Rad Brap Accident IF 3-2 Fuel Failure 19 Care holing SF Provide HXS AIU power to RI systen (N) law water level (i2) instrueentatim far IFCI system (H) Initiatimo 25A L<<ge Break Inside C(EF 3-2 Fuel Failcre 19 Ccrc Cooling SF Provide HXS ATU power to prlnary ccntainant systen N) high drywell preslaf e lnstrueentatim anil feedwater systee (N) and recifculatim spiN (N) law reactcf presslre instrmentatim. 25A Large Break Inside IXEF 3-2 Fuel Failure 19 Care Cooling SF Provide HXS ATU power to feedwater systen (N) low water level (Ll) and ~ RI systen and recirculatim systen LN) law reactor pressl e Instrueent at lm 258 Intern Erk Inside (XEF 5-2 Fuel Failure 19 Ccrc Cooling SF Provide HXS AlU power to feedwatsr systee (N) law water level (LI) and F)1 systee and recirculaticn systee IN) law reactor presme instrueentatlm 258 Intern Brk Inside COEF &2 Fuel Failure 19 Care Cooling SF Provide HXS AlU power to F)I systee (N) law water level IL2) Instrueentatim far IKI systee le) initiatim. 258 Intern Brk Inside CBEF 3-2 Fuel F 'cntainant systee I p ~ po ~ISA'pe lnstfoaen+I,ol and Ibl) high dryweII f .".',.:, )h fair i+en.(N),'and". a ~ 25C Snail Break Inside IXEF 3-2 Fuel F ttttm tMt ta attr tent tUt tttl F)1 systen and recirculatim systea (68) law reactcr presme Instr Lw%ta'tlm (:O. SAFE S)NTON)f m3% NH86~ NQLTSIS REACTIR ONE STSEI ISCLATIN Nev I) Prepared: Checked: Page 207 E. C. L Parker of er Data Ote: ~ SPfl SySIE)) 71 BM REACT MXEPTAELE SAFE OR HER RESKTS SAFETy ACTIN RIC SEL GENT TITLE STATE Code Title Code Title C00E '25C Snail Break Inside CEEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide ECCS AlU poor to FM systen IOS) Im water level IL2) instruaentaticn fm )fCI systee (73) Oil'tia'tlcn, 2X Snail Break Inside CKF 3-2 Fuel Failure 19 Core Cooling SF Provide HXS AIU power to prinary ccntaincent systen IQ) high drywell pressure instruaentaticn and feedwater systen IOS) and recirculaticn systen (N) Icw reactcr pressure instruaentaticn. 27 Pipe Break Outside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide ECS AlU power to feedwater systen IOS) Im water level ILI) and Rf systen and recirculatim systen IN) Im reactor pressme instruaentatim. 27 Pipe Break Outside COEF 3-2 Fuel Failure 19 Care Cooling SF Provide HXS AlU power to F)l systen {M) low water level IL2) Instruaentatim for IPCI systen (73) Initiatlm. 32B Low Reservoir C0EF 2-2 Fuel<<Failure 19 Core Cooling SF Provide ECS All) power to F)f systea IOS) Iu water level (L2) Yen initiation. 32B Low Reservoir C0EF 2-2 Fue Reservoir CEEF 2-2 Fue Failure)9 Ccr 'n 328 Low N) Iow water level IQ) instruaentatim fcr )FCI systee 173) initiatlmo IS0 ISO Shutdoe COEF 2-2 Fuel Failure 19. Ccrc Cooling SF Provide HXS AlU poor to Rf systen (M) Iow water level IL2) instruaentatim for )fCI systea 173) initiatim. SAFE 5)STlNN QNLYSIS Prepared: Page 20() of bate: ~ REACTI GIE 15(U)TII E GKM STS18( Ouckah Sate> ~~~ El .SRH)SS3-0(S Iev I) C. L Parker SYSTE)( 71 BSIT KKT MXEPTARE SAFE (a CPER RES(LTS SAFETY ACTllf( RlC SEL BM TITLE STATE Code Title Code Title GEE '15b 150 Shutdoe GKF 2-2 Fuel Failure 19 Core Cooling SF Provide ECS NU power to FM systea (I) low water level (L2) instruaentatim for RCIC systen ini'tiat i Nv ISb ISO Shutdoo CSEF 2-2 Fuel Failure 19 Ccrc Cooling SFS RCIC initiatim cn Iw reactor water level (l2) signal froa R)f( systen (74). GEF 2-2 Fuel Failure 19 Core Cooling SF Provide HXS NU pea. to R( systa (03) low wats'evel (L2) instruaentaticn fl )fCI systen (73) initlatim. Ncrnal Shutdoo 2-2 Fuel Failure 19 Core Cooling Provide HXS ATU poor to R( systen ~ R50 CSEF SF (03) low water level (L2) ,- instruaentatim for RCIC systee initiaticn. N50 kraal Shutdoe (XEF 2-2 Fuel Failure 19 Core Cooling SFS RCIC initiatim cn low reactcr>>ater level (L2) signal fron R)a systen (7I). SSb Safety %utdae CSEF 3-2 Fuel Failure 19 Core Cooling SF Provide HXS ATU Daki-eysteN wats'evel (L2) instruantatim fl HCt ~pa $ 73t < + vvjqllvlve!vtv ~ $ ('~ 13 lnadvert bpen SIN CCEF 2-1 '-' closes'stea ~ supply line isolaticn valves. 15 Loss bff<ite AC CSEF 2-1 Rad lsol Low stean supply pressure signal closes stean supply line isolation valves. 25D Srklnsi~t!Rad CSEF 3-1 Rad Release 20 RPV lsol Low stean supply pressure signal closes stean supply line i la 'ow stean supply pressure signal closes stean supply line isolatim valves. l/vl'vvvvrad v~C p~~ ~~p ~ ~ SAFE SHIM'NLTSIS Page 20g of )a7 Prepared: Oate: REACTOR NK ISRATI(N E. CmUM STSTEN Chected: O te: BRH)SS3~ (Rev I) C. L Parker PSIEN 71 ELW REACT SIZABLE SFE N IRR RESKTS SAFETT ACTISI FllC SE9. BHIT TITLE STATE Code Title Code Title mE NKNIS 27 Pipe heat Outside CKF 3-1 RadRelease 20 8VIsol SF High steaalinebifferential presare signal cr high stean line space teeperature signal closes stean supply line isolation valves. 08 lnadv Puap Start F 2-2 Fuel Failure 22 Rest Naraal N Provide systen start indicatim so qxntor <<n stop systen and retm to Mraai. 08 lnadv Puap Start EF 2-3 Syst Stress 25 Level Reduce Provide ECCS Alll pour to RI systen (N) high uater level (LS) IPCI trip Instruamtatim. 08 lnadv Puap Start lf 2-3 Syst Stress 25 Level Reduce Provide ECS ATU p~ to Rf systen (N) high uater level (LB) RCIC trip Instrueentatim. 08 lnadv Puap Start OF 2-3 Syst Stress 25 Level Reduce Shutoff RCIC (if operating) m Rf systea (03) high reactcr uater level (LS) signal. 24 Rod Orap Accident lF 3-3 Syst Stress 25 Level Reduce Provide HXS ATU paar to Ri systea (N) high uater level (LS) IPCI trip instrulenta'ti cn. 25O Brtlnside+mt(Bad COEF 3-1 Rad Release 26 Est Pri Cant Support priaary cmtainaent functim. 15 Loss Off<ite AC Nl 2"I Rad Release 5iippart secondary ccntainst 25O Brkinside+ntNad (XEF 3-1 Rad Relea c ) NU 2b Fuel Handl Accidnt A 3-1 Rad Rel 27 Est Sec Cent Support secondar con '3 Loss Fuel Pool CI ALL 2-1 Rad Release 27 Est Sec Cont Support secmdary ccntainaent fwctim. 28 Shtdn Froe Outside COEF H local Shut 3g Cooldn Outsd N Naintains ncraal >>ater inventory in the reactcr vessel shih presare above )00 psig. Nanual control fry lactup panel. SAFE SHJIMN QN.YSIS Page 210 oI go7 Prepared: C 9 l . ~M(l IIISH PRESQRE GXUNT E. IQECTIQi STSTE)I 03ecked: Sate: ~S M EF)HM~ Iev 1) C. L Park ElIENT REACT lNQXPTRE SAFE N 5%R RESILTS SAFETY ACTlTSI FllC SEL 6E)IT T1TLE STATE Code Title Code Title GXE N IIain Stn lsolatim COEF 2-2 Fuel Failure 19 Ccrc Cooling SFS IfCI initiatim m RI systen IN) low reactm water level IL2) signal. 05 Loss of Vacta3n COEF 2-2 Fuel Failure 19 Core Cooling SFS IfCl initiaticn on RI systen (N) low reactor water level IL2) signal. 12A Pres Reg Fail+en COEF 2-2 Fuel Failure 19 Care Cooling SFS IfCI initiatim m Ri systen IN) low reactor water level (Q) signal. 12B lnadv Open~ass COEF 2-2 Fuel Failure 19 Core Cooling SFS IfCI initiaticn m Ri systen LN) low reactor water level IQ) signal. a 13 Inadvert Open %V CCEF 2-2 Fuel Failure 19 Ccrc Cooling SFS IfC1 initiatim m RI systen IN) Im reactor water level IL2) signal. COEF 2-2 Fuel Failure 19 Ccrc Cooling SFS IfCl initiatim cn RI systen IN) low reactcr water level IQ) signal. 15 Loss OI flite AC COEF 2-2 Fuel Failure 19 Ccrc Cooling SFS IfCl initiatim m Rl systen (N) low reactcr water level IL2) signal.. 23 RI Ctrl Fail~ COEF 2-2 Fuel Failure 19 Core Cooling SFS IfCl initiation m RI systen IN) lm reactcr water level IQ) signal. 24 M Orop Accident OF 3-2 Fuel Failure 19 Core Cooling PS IfCl initiatim on RI systen (N) low reactcr water level IQ) signal. 258 intern Brk inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SFS IfCI initiaticn m Rt systen IN) low reactcr water level IQ) signal. 25B intern bk inside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SFS IfCI initiatim m cae spray systen 25C Snail Break inside CKEF 3-2 Fuel ai gQ ~e spoil~~ If51 jnipakiggp, Risysten IN) lm 25C Snail Break inside TXEF 3-2 Fuel u ureW 'are C'ooling'FS'- If'neitiati'm'h c'creamy systen '.".:;" 175) high drywell pressure signals. 27 Pipe Break Outside IZFF 3-2 Feel Failree le Dre Crarliee aFF IKI ieitiarrre mm9 aeaFas FIFT5i reactcr water level IQ) signal. SAFE QITIOM QNLTSIS Page 211 Preparedl IGEH PRESRRE IU)IT E. c T)I)ECTIN SYSTE)I Reeked: Bates BAH)SS3~ IRev 1) C. IL Parker 69iI IEACT SNXEPTARE SAFE OR SER RESIITS SAFETT ACTIN flK SB. 'IE)IT TITLE STATE Code Title Cade Title IXIE 32B Law Reservoir IXEF 2-2 Fuel Failure 19 Care Cooling SFS IPCI initiatim on Rf systen (03) law reactor water level IL2) signal. ISO 1SO Shutdlne CTEF 2-2 Fuel Failure 19 hre Cooling SFS SKI initiatim cn F)f systen I03) Ital reactcr water level (L2) signal. COEF 2-2 Fuel Failure 19 Ccrc Cooling SFS IfCI initiaticn an Rf systen I03) la reactor water level (L2) signal. SSO Safety Shutdoe " IXEF 3-2 Fuel Failure 19 Care Caoling SFS IfC1 initiaticn cn Ri systen f03) lcw reactcr water level IL2) signal, 13 lnadvert Open SIN IXEF 2-1 Rad Release 20 SV lsol Law stean supply pressure signal closes stean supply line isolatim valves. 15 Loss Offdite AC GKF 2-1 Rad Release 20 RPY lsol Law stean supply pressure signal closes stean supply line isolatim valveso 25A Large Break inside COEF 3-) RadRelease 20 Rpglsol Low stean supply pressure signal closes stean supply line isolaticn valves. P 25D Brklnside+mtÃad COEF 3-1 Rad Release 20 RPVlsol Law stean supply pressure signai I atim vai~. (un'~. 4~ Qs<l$ 7 27 Pipe Break Outside COEF 3-1 <,3) I6 1sIll, q" .-.Law;Steh SupjI}y'pre'ssure ffg'nal: '"," ';) 'l 'a j Clos'er'MeILn Iy,line IWlatj'an;: QC /!wC g4ggj fQt~~tjy 3-1 ical SF 'Vn. High stean line differential presslre 2T Pipe Break Outside CKEF signai cr > ~sean1anFspacF-teeperature signal closes stean supply line isolatim valves. 08 lnadv Puap Start F 2-2 Fuel Failure 2? Rest )kraal II Provide systen start indicatim so aperatcr can step systen and return to ncrial. EF 2-3 Syst Stress 25 Level Reduce Shutoff )PCI Iif operating) m Rf systen 103) high reactcr water level ILB) signal. SAFE BUIIOM A)NLYSIS Prepare Page 21 $ 01 Data ~/f Checked: Date: BIW REACT INGXHAfLE SAFE CR KR REM.TS SAFETY ACT)9i FIK, SED. GAS TlTLE STATE Code Title Code Tit)e NK 24 Rod Drop Accident 'IF 3-3 Syst Stress 25 Leve) Reduce Shutoff )fCI fif operating) cn F)) systee (03) high reactcr water level iLS) signal. ZD Srklnside-CmtNad CDEF 3-1 Rad Release 26 Est Pri Cant Support prieary cmtaineent functim. I 15 Loss Off+ite AC NL 2-1 Rad Release 27 Est Sec Cant Suppart secmdary ccntainmt functim. 25D Srklnsi~tSad CDEF 3-1 Rad Release 27 Est Sec Cmt Support secmdary cmtainaent functicn. 2b Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cmt Support secmdary ccntainaat fmcticn. 33 Loss Fuel Pool C) NL 2-1 Rad Re)ease 27 Est Sec Cmt Support secmdary cmtainesnt functim. 27 Pipe Break Outside CDEF 3-2 Fuel Failure 33 Li~ LossCool L Flw restrictors instal)ed in the stean supply lines to the )PC) ast , liait the sass f)ou rate of coolant being ejected froe the stean line fo))oring a break. This is a passive functim. 28 Shtdn Froe Outside COEF 4-1 Loca) Qgt 39 Coo)dn Outsd )I ))anual)y tripped off by overriding suitches at the back up cmtrol 'enter to prevent reactcr vessel overfill. 320 Earthquake ALL 4-3 Equip Fail 44 Seisaic Shtd Desi n of ave capability to provide shutdcae) g U ~ k SAFE SHJT(M AN.YSlS Page 2D nQN'S>7 Prepared Oater e RES)(INL )EAT RBI)VAL STSTE)( fF)HISS3~ (Rev 1) STS(BI 7i 02ected: E. C. L Par)rer Ote: >>iF EVENT REACT LNXEFlAME SAFE SER REM.TS SAFETy ACT)5( FlM: SEL EVE)(T TTTLE STATE Code Title Code Title CKE N )(ain Stn lsolaticn (XEF 2-2 Fuel Failure 19 Care Cooling SF Send feedwater systee N) low reactcr water level signal (L2) to initiate K1C systee 01). 05 Loss of Vacwe, COEF 2-2 Fuel Failure 19 Core Cooling SF Send feedwater systen (0$ low reactcr water level signal (l2) to initiate RClC systen (71). 12A Pres Reg Fail~ C(EF 2-2 Fuel Failure 19 Core Cooling SF Send feedwata systee (03) low reacta water level signal (L2) to initiate RC1C systea (71). 12S lnadv Open~ass COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Send feedwater systen N) low reactcr water level signa) (L2) to .initiate KlC systee (71). D 1nadvert Open SRV COEF 2-2 Fuel Failure 19 Core Cooling SF Seod feedwater systen (05) low to (er'each(iu(F) D lnadvert Open SRV IXEF 22 Fuel Fer'Fute 99 'teF(Iue)e uVrelrre Fru level (Ll) signals fron ccrc ay ten (75). Provide si als o close recirculaticn valves to recirculaticn systen (N). D lnadvert Open SRV COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Open inject(a valves on lcd reacta pressure signals frci core spray systen (75) .and IR systen (7I) tPCl initiaticn signals due to low water level (L1). )I Loss of R( Flow COEF 2-2 Fuel Failure 19 Core Cooling SF Send feedwater systen (05) lrnr reactrr water level signal (L2) to initiate RClC systen (71). 15 Loss Offdite AC COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Send feedwater systen N) low reacta water level signal (l2) to initiate KlC systea (71). 15 Loss Of finite AC (3EF 2-2 Fuel Failure 19 Ccrc Cooling SF Start pueps on low reactcr water level (Ll) signals fron ccrc spray systen 05). Provide signals to close recirculaticn valves to recirculatim systen (N). er e 9 SAFE SN/HM ARMIS Page 214 of $ 07 Prepared: , Bate: 1l RESIMNL )RAT RENNN. SYSTEN ChKkedl Bate: EF)H)SSHN IRev I) C. L Parker ElJE)lT REACT MXEPTABLE SAFE m CPER RESILTS SAFETY ACTIN FIK SEL HJENT TITLE STATE Code Title Code Title m 15 Loss Offdite AC C0EF 2-2 Fuel Failure 19 Core Cooling SF Bpen injecticn valves cn lw reactor pressure signals fron are spray systen 05) and M systen f74) LKI initiation signals due to'lw uater level ILI). 22 Loss Shutdwn Cool KO 2-2 Fuel Failure 19 Ccrc Cooling )SF Naintain water level ILPCI). 23 6 Ctrl Fail~ CHIEF 2-2 Fuel Failure 19 Core Cooling SF Send feeduater systen IN) lw feactcf uatef level signal IL2) 'to initiate RCIC systen I7)).'tart 24 Rod brop Accident ijF 3-2 Fuel Failure 19 Core Cooling SF pueps cn lw reactm uatlf level ILI) signals fron ccrc spray systen 175). Provide signals to close recirculaticn valves tn rKlrcula'ticn systea IN) ~ 24 Rod Brop Accident BF 3d Fuel Failure 19 Core Cooling SF Bpen injecticn valves cn lw reactor pressure signals froa ccrc spray systems OR nl ()(()y)(es (7(l ()C( initiation signals due to lw uater level ILI). 25A Large Break Inside CCEF 3-2 Fuel Failure 19 Ccrc Cooling SF Start puap cn prinary cmtainant systea N) high dryuell pressure signal uith lw reactor pressme signals froa core spra {75). Prrhde s'lose recirculaticn es to fKlfcula'tlcn s 25A Large Break Inside COEF 3-2 Fuel F Core Coding SF OppingecCie'valk~ w eyc P".; =, prI)ssure'.Signal: fros, 'c spray '. ':.Syiten'f75) and &fsysten 174) LPCI '4 * ' initiation signai pressure. due to highgcyuetI g I I 25A Large Break Inside CCEF 3-2 Fuel 'lure 19 Core Cooling SF 'Start puap cn Iw reactN'ater level IU) signals frca ccrc spray systen f75). Provide signal to close recirculaticn valves to recirculaticn systen IN). ~ ~ ~g 0 0 SAFE S)61HN NNLYSIS Page 215 Preparedi RES INNL fEAT RBML E er SY STD( Checked: BF)H)S83<$ 8 Nev I) C. N. Parker BM REACT MXEPTlRE SAFE N (PER RERlTS 'SAFEIY ACTIN FBC SE(). EIM TITLE STATE Code Title Code Title CXE 25A Large Break Inside (XEF 3-2 Fuel Failure 19 Ccrc Cooling SF Open injecticn valve cn Iow reactcr pressure signal frow ewe spray systen (75) and RN systen (74) tPCI initiation signal due to Iow water level (Ll), 258 Intern Brk Inside CBEF 3-2 Fuel Failure 19 Ccrc Cooling SF Start puaps cn low reactcr wats level (LI) signals frow ccrc spray systen (75). Provide signals to close recirculation valves to recirculaticn systea (N). ZB Intern Brk Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Open injecticn valves m low reactor pressure signals free ccrc spray systen (75) and R% systea (74) IRI initiaticn signals due to low water level (Ll). 258 Intern Brk Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Start puaps cn prinary ccntainant systen (M) hinh drywell pressure signals with Iow reactor pressure signals froa ccrc spray systen 05). systems Provide signals to close recirculation valves to recirculaticn systen (68). 258 Intern Brk Inside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SF Open injecti ve on low reactcr signals frow coredgwa (35) an~ pw1k,pl) initiqysQndp'(ue th)be ' jV' '~PP~pM" 258 Intern Brk Inside (KEF 3-2 Fuel F urpglMgeeokng.SF- %ii'vlde4gnal. (that a R% '.,'>-'unning) to wain ste en (Ol) for the autonatK depressN izatim ~MS). 25C Snail Break Inside C(EF 3-2 Fuel Failure 9 rCooling SF Start puaps on Iow reactor water level (Ll) signals frow ccrc spray systea 05). Provide signals to close recirculation valves to recirculatim systea (M). g ~ ~ 'I SAFE M(T(OI( NNLYSIS Page 2)6 nfl&~ Prepared: Date: Z RESIDUAL )EAT E)itttAL E. S6TE)( Checked: Date: BRH)S63&8 IRev 1) C. L Parker SYS(E)( 74 GM REACT (N(XEPTABLE SAFE OR CPER RES(LTS SAFETY ACTIN RK SE(). aERT TTTLE STATE Code Title Cade Title (IE 25C Saall I'eak inside (XEF 3-2 Fuel Failure 19 Core Cooling SF Open injection valves m low reactor pressure signals fraa cl e spray systss 1151 sst MR systss 1st! typal initiaticn signals due to low water level (Ll). 25C Snail Break inside CDEF 3-2 Fuel Failure 19 Care Cooling SF Start puaps cn prinary ccntainaent systen (64) high drywell pressure signals with law reactcr pres@re signals fran care spray systen (75). Provide signals to close recirculation valves to recirculatim systen (6B). g.l 2X Snail Break inside QEF 3-2 Fuel Failure 19 Care Cooling SF Open injecticn valve an law reactor pressure signals fra ccrc spray systea (75) and RN systen (74) LFCl .) g I initiaticn signals due to high dr pell pressure. 25C Snail Brea'k inside CKF 3-2 Fuel Failure 19 Care Cooling SF Provide signal (that a II pimp is running) to aain stean systen (01) fcr the autoaatic depressurizitim systen (A(jS). 27 Pipe Break Outside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SF Start puaps cn low react level (L)) rca ccrc spray ). Provide signa t~%, $ 27 Pipe Break Outside (XEF 3-2 Fuel Failure 19 li((g',Sf 'Open:.iojectiqnnybrt4 cn law reactor s ~ ' prdssurb signa)s fran care spray systea (75) and R)R LPCI initiat due to low water ve (Ll). 32B Low Reservoir (XEF 2-2 Fuel Failure 19 Ccrc Cooling SF Send feedwater systen (()3) low reactcr water level signal (L2) to initiate RC(C systen (71). 32E Fire NL 4-4 Fuel Failure 19 Care Cooling Design of safety features ast include Appendix R fire cmslderat(mss SAFE )HJTm emiS Page 217 of $ 0) ~f RESIDUAL )EAT. RE)SVAL ST STER SRHSM~ EVEHT IRev I) Prepared: Checked: E C. )L Parker Oate: 0ate: II ~ REACT NCXPTAOLE SAFE N IFER KkLTS 'SAFETY ACTIN RK SG). EVEHT TITLE STATE Code Title Code Title fmE ISO ISO Shutdao COEF 2-2 Fuel Failure 19 Ccrc Cooling )fSF With a reactor low pressure peraissive and RCIC/)FCI loss due to vessel depressurization, aanually start cne IN puap in lKI aude to naintain RPV water level. ISD ISO Shutdae IXEF 2-2 Fuel Failure 19 Core Cooling SF Send FW systee N) low reactor water level signal IL2) to initiate RCIC systen (71). HSD Horaal Shutdae COEF 2-2 Fuel Failure 19 Ccrc Cooling HSF With a reactcr Iow pressure peraissive and RCIC/)FCI loss due to vessel depressuriaaticn, aanually start me IN puao in LKI aude to ~aintain RPV water level. HSD Hcraal Shutdooo COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Send FW systen (M) low reactcr water level signal IL2) to initiate RCIC systea f71). SSD Safety Shutdown COEF Z-2 Fuel Failure 19 Core Cooling HSF With a reactcr Iow pressre peraissive and IPCI loss due to vessel depressurilatlon) nanually start cne RN puap in LKI aude to aaintain RPV water level. 2 N Hain Sta Isolation COEF 2-1 Rad Release )solaticn signal tripped on low ~e)+LQ A ) q i ptigffkl',Tph ccn)jndeat

' I;systei,N)~~.!$ )~e(y'ac)ion'~ps Q.

 ; ~ expyted tp.:adur, 'hut iAdt a requireeent for this event since there is no fuel failure. 05 Loss of Vacuua 2-1 Rad Release IN Qff/ COEF SF isolation signai tripped on low water level (0) signal froa priaary ccntainaent systea N). This safety action is expected to occur, but is not a requireaent for this event since there is no fuel failure. h ~ ~ I ~ ~ 'I SAFE SHIM'( ARALTSIS Page 2)8 Prepared: Date: -0 RESI NNL }EAT ROD%. SUSIE)( Checked: Date: SRHM~ (Rev I) STSIBl 74 EV8(T REACT 9(A(XEPTABLE SAFE CPER RES(lTS SAFHY ACTI% RK SEQ. EVENT TITLE STATE Code Title Code Title CODE D lnadvert Open SRV CDEF 2-1 RadRelease 20 RPVIsol SF tripped on Iow water level (Q) signal froa priaary containaent systea (6l). Ibis safety acticn is expected 'to occur) but is not a requireeent fcr this event since there is no fuel failure. 2-1 Rad Release 20 lsol M c r~.isolation signal 14 Loss of F(( Flow (XEF RPV SF tripped cn low water level (Q) signal froa priaary containeent systeu N). This safety acticn is expected to occur, but is not a requireeent fcr this event since there is no fuel failure. 15 loss Off-Site AC Nl 2-1 Rad Release 20 RPV Isol SF M isolaticn signal tripped on low water level (Q) signaI froa priaary ccntainuent systea N). Ibis safety actim is expected to occur, but is not a requireeent for this event since there is no fuel failure. M~h'. F(( Ctrl Fail~ CDEF 2-1 Rad Release 20 RPV Isol SF M isolation signal tripped on low water level (Q) signai frou naent eu N). This safety action is expected to occur> b((t js ugly'g~, )) PQ fh(s~t Qn'.5,', ( ~ 'h e'"di}o 4dI j fl',; v j 24 Rod Drcp Accident DF 3-1 Roleap )20 aRPV'lsolc ' M isolation signaI tripped on low water I s Luary contalllsst systee N). 2'arge Break Inside CDEF '5-I Rad Release 20 RPV Isol ~ SF M Rf~g isolation signal tripped on low water level (Q) signal froa prisarv containaent systea N). SAFE )filHN AHAL61S Page 219 47 Prepared: Bate: < RES)OIL )EAT RE)SVAL F E ert. S6TEN Checked: Bate: BRH)SS3~ (Rev I) S6TE)f 7l, EVEN REACT IACCEPTABLE SAFE N IVER RESILTS 'FHY ACTICH RK SG). EVEHT TITlE STATE Code Title Code Title COBE 7 25B Intern Brk Inside COEF 3-1 Rad Release 20 RPV Iso) SF M isolation signai tripped cn low water level (Q) signai fron prinary containnent systen N). fcE 2X Sea) I Break Inside COEF 3-1 Rad Release 20 RPV Isol SF M isolaticn signai tripped cn low water level IQ) signal free prinary containnent systen N). E 250 Brklnside+mtLRad COEF 3-1 Rad Release 20 RPV Isol SF M isolatlcn signai tripped on low water level IQ) signal fron priaary ccntainaet systen N). fGE gf~ 27 Pipe Break Outside CDEF 3-1 Rad Release 20 RPV Isol SF M isolaticn signai tripped cn low water level (Q) signal fron priaary ccntainaent systen N). 07 Shutdn Cool )falfun AtD 2-2 Fuel Failure 22 Rest Homal )I Turn off or reduce M Shutdoe Cooling to stop undesired cooling. 22 Loss Shutdown Cool AC0 2-2 Fuel Failure 22 Rest Hcrnal )f Restore reactcr tecperature to ncrnal by restcring shutdcen heat rexnral. EcE ZA Large Break Inside COEF 3-1 Rad Release 2b Est Pri Cent SF M reel] pressure sinnal fron the Prinary Ccntai tr. ZB Intern Brk Inside GEF 5-1 Ital R &@ '>,kragedd'hi drywell pressure signai fron the Prinary Ccntainan systen N) ~.E .C, 5C Snail Break Inside 'XEF 3-1 Rad Rel st Prl C0At SF M " isolation signai tripped cn high drywell pressure signal fron the Prinary Ccntainnent systen N), 'st x/ 250 Brklnsid~tNad CCEF 3-1 Rad Release 26 Pri Cont SF M isolaticn signal tripped cn high drywel1 pressure signal fron the Prinary Ccntainnent systen N). SHJMII NQLYSIS 7 SAFE RESIDUAL )EAT ANNAL Prepared: P E. C. 220 c D ts ~i/ Zg SYSIE)l Dcecked: Date: . BF)HIS63~ IRev I) C. L Parker SYSIE)I 74 EUENT REACT IlfACCEPTABLE SAFE OR IRR RESILTS SAFETY ACTIN flK SEO. EUENT TITLE STATE Code Title Code Title CODE '5 Loss Offdite AC ALL 2-1 Rad Release 27 Est Sec Cont Suppm t secondary ccntaineent fcncticn. ZD RrklnsidHmt4Rad COEF 3-1 Rad Release 27 Est Sec Cont Suppart secondary ccntainaent fcmctian. 26 Fuel Handl.Accidnt A 3-1 Rad Release 27 Est Sec Cent Support seccndary ccntalneent functi on. 33 Loss Fuel Pao'l ALL 2-1 Rad Release 27 Est Sec Cont Support secandary antaicaent function. OIR T/6 Trip-BP Fails F 2-4 Cent Stress 30 Cant Cooling )ISF Provide suppressicn pool cooling function. 01C PresReg Fail<lose COEF 24 Cont Stress 30 Cent Cooling)SF Providesuppression pool cooling functicn. 03 )Iain Stn Isalaticn COEF 2-4 Cant Stress 30 Cent Cooling )SF Provide suppressicn pool ccoling function, 05 Loss of Uacuue COEF 2-4 Cant Stress 30 Cant Cooling)SF Providesuppressicn pool cooling functicn. I@ Pres Reg Fail+en COEF 24 Cent Stress 30 Cant Cooling %F Provide suppression pool caaling function. 12R Inadv Open-Bypass COEF 2-4 Cant Stress 30 Cant Cooling lSF Provide suppression pool coo 13 Inadvert Open %V COEF 2H Cant Str s 30 Cant Cooling )ISF Provi@ qprcesj~h+4ohjg u 15 Loss Off-Site AC COEF 2H Cent Stres t>6plingp5( a Pr'cncide su'pprhsicn pool cooling ~ fLNctlN. 15 Loss Off-Site AC COEF 24 Cant Stress A Cant KF on e dryweII cooling lif necessary) by supplying torus water through the ccntaineent and tees spray node. This wade requires that the R)R~I is running and that 2Q

0 cae coverage systee, 03).

is indicated I feedwater SAFE SWMN NQLYSTS Page 221 nf M7 Prepared: RESIDUAL IEAT RENNAL F SYSTEM)4 Checked: Date: BF)H)SS3~ tRev 1) C. L Parker SYSTEN 74 BSIT REACT MXEPTABLE SAFE N CPER REStLTS SAFETY ACT)% RK SEQ. EVENT TI1LE STATE Code Title Code Title IE 22 Loss Shutdao Cool CD 2-4 Cont Stress 30 Cont Cooling )SF Renove decay heat fron Tcrusmter using Pcol cooling. 24 Rod Drop Accident DF 3H Cent Stress 30 Cont Cooling I)SF Provide suppression pool ccoling function. 25A Large Break inside COEF 3-4 Cont Stress 30 Cont Cooling ))SF Provide suppressicn pool cooling function. 258 )ntern Brk inside CDEF 3-4 Cont Stress 30 Cont Cooling )SF Provide suppression pool cooling functicn. 25C Saal) Break inside COEF'H Cont Stress 30 Cont Cooling )SF Provide suppression pool cooling functico. ZD Brk)nside-CcntQhd CDEF 3-4 Cent Stress 30 Cent Cooling )SF Provide suppression pool cooling function. 25D Brklnsid~ttRad CDEF 3-4 Cont Stress 30 Cont Cooling)SF Cont. sprayaodeof M allwsrapid cco) doe of priaary cent. Ti provides for N ataosphere nixing to assist cont. atnospheric diluticn systen iB4). Cent. spray pernissive requires that M~1 be running ) 2I3 ccrc coverane is indicated (FN, 3). 25D Brklnside-ConttRad CDEF 3-4 Cent Stress us spray alloiis cooldoo of the wiwli sni aids in Trre prie i s  ;:, ';. '":, i", gh f'2j) if'Vlecigh n inMiakd l4eednater systea,'3). 25D Brklnsid~tLRad CCEF 3-4 Can Stress 30 Cont Coo)in SF s containnent by renovino care heat to the suppressicn pool via uater fleeing out of the break. This node also indirectly assists drywall pressure ccntrol by ccndensing stean remaining in the dr yell. '$o7 SAFE SHiHXN ANALYSIS Page 222 Prepared) RESlONL %AT RBSVAL E SYSTEN Checked: Date: TF)H)SS3~ IRev 1) C. ll. Parker SYSTE)1 74 BENT REACT lNCCEPTABLE QFE OR OPER RESH.TS SAFE)Y ACT1% FlK SE9. EM TlTLE STATE Code Title Cade Title CNE 27 Pipe Break Outside COEF 3H Cant Stress 30 Cant Cooling lSF Provide suppression pool cooling Reservoir functian. 328 Low COEF M Cant Stress 30 Cant Cooling ISF Provide suppressicn pool cooling function. 32E Fire COEF 4% Cent Candtns 30 Cant Cooling Design of safety features aust include Appendix R fire ccnsideratians. lSD 1SO Shutdao COEF 2-4 Cant Stress 30 Cont Cooling %F Provide suppressicn pool cooling function. HSO Homal Rutdae COEF 2-4 Cant Stress 30 Cont Cooling %F Provide suppressicn pool cooling functicn. ,0 SSO Safety Shutdown COEF 3H Cont Stress 30 Cont holing ))SF Provide suppression pool cooling functicn. SSO Safety Shutdoo GXF 3-4 Cent Stress 30 Cont Cooling )SF Provide containaent and torus spray axles, if Mcessvy. These andes require that RN~l is running and that 213 ca e coverage is indicated Heeduater systea, 03). 28 Shtdn Froa Outside COEF 4-1 Local Shut 39 Cooldn Outsd )1 Suppressicn pool cooling eade-28 QADI@ Shtdn Froa Outside ~ +~t ~+ Nl 4-1 Local Shut 3q 39 '4Ce+g Cooldn Outsd )1 Suppressicn pool teaperature gv/cd~'eg Jt ann(( ~~~a Shutdcee Cooling )kde-is 'il+cacrf~),Sye)a, C4 e+% Q Avc 4ea4~$ 4. a er reactcr vessel s/sP~ is bein 100 si 9 . MVgF P 28 Qtdn Fran Outside kL 4-1 Lacal ',. '...:,: i "dhe rM'c9'driiny shutdown ceaiing. 1 320 Ear thqua'ke Nj. 4-3 Equip F '1 44 Seiseic Shtd Desi n of ve capability to provide shutdoe follauing a design basis seisaic event. ~ ~ P SAFE SHJTOM QNLYSIS Page 223 of 1 Preoared'. bate: ~Ll~< E ORE SPRAY SYSTE)( Eck Omked( Bate; BF)H)S83~ (Rev 1) C. )(. Parker EVE)(T REACT MXEPTABLE SAFE OR (PER RES(LTS . QFElY ACTIN RK SHL EVENT TITLE STATE Code Title 'Code Title CXE '03 )(ain Sta Isolatim COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide flow path fcv'lternate source of water to RCIC systea (71) froe suppressim pool N). 0$ Loss of Vacuua (XEF 2-2 Fuel Failure 19 Core Cooling Provide flow path for alternate source of water to RCIC systen (71) froa suppression pool (M). 12A Pres Reg Fail+en COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide f(ow path fcr alternate source of water to KIC systee (71) fron suppressicn pool N). 128 Inadv Open-Bypass COEF 2-2 Fuel Failure 19 Core Cooling Provide flow path fcr alternate scarce of water to KIC systea (71) froa suppressim pool N). D Inadvert Open SRV COEF 2-2 Fuel Failure 19 Core Cooling Provide flow path for alternate source of water to KIC systen (71) sim p l N). D Inadvert EEF 2.2 Fpl+apu) FL Eya~hel~ e) (03) low Op [)IP) f 0, '(~j(ff C r l;-< u roc 'y I signals (Ll) and I(tw gdicgg~rqsur signals Hroa / feR~ter sy ea (0 ) and recirculation syst N)) to RH( ~@wadi 13 lnadvert Open %V C(EF 2-2 Fuel Failure 19 Ccrc Cooling SF Start punps on feedwater systea (03) low reactcr water level (Ll) signals. D Inadvert Open SRV (XEF 2-2 Fuel Failure 19 Core Cooling SF Open injection valves on feedwater systea (03) cr recirculatim systea (68) low reacta'ressure signals in conjunction with low water level (Ll) s(gnaI froa feedwater systee (Ã). lnadvert Open SRV COEF 2-2 Fuel Failure 19 Core Cooling SF Provide standby diesel generatcr (82) start sionals m feedwater systea (03) Iow reactcr water level signals (Ll). This safety actim nay occur, but is not required by this event since ncraa) ac is available. SAFE SHJHXN AN.YSIS P e224of 307 Prepared: C(RE SPRAY SYSTE)I F Owcked: bate: SF)H)SH~ (Rev 1) C. IL Parker GM REACT MXEPTAKE SAFE OR KR REKTS SAFETY ACTlS( RK SER. BM TITLE STATE Code Title 'ode Title GEE '4 Loss of F)( F)ow C0EF 2-2 Fuel Fail'ure 19 Core Cooling Provide flow path for alternate source of water to RClC systea (71) fry suppressicn pool (6l). 15 Loss Nfdite AC COEF 2-2 Fuel Failure 19 Ccrc Cooling Provide flow path for alternate surce of water to RCIC systen (71) fron suppressim pool N). 15 Loss Nfdite AC COEF 2-2 Fuel Failure 19 Care Cooling SF Send feedwater systen (03) low reactor water level signals (Ll) and low reactcr pressure signals (fry feedwater systea (05) and recirculatim systea (N)) to M systea (74). 15 Loss Nf<ite AC COEF 2-2 Fuel Failure 19 Core Cooling SF Start puaps cn feedwater systen I(6) / low reactor water level (Ll) signals. 15 Loss Nfdite AC COEF 2-2 Fuel Failure 19 Ccrc Cooling SF Open injectim valves on feedwater systen IN) cr recirculatim systen N) low reactcr pressure signals in cm signal froe feedwater systen ((LT). 15 Loss N finite AC ne J~ ': 2-.2>fael'galere v'cork'cNjrg'!p~ ~&briva ": ~q )p ')@lilt IL1). f Fqb($ a tiff ter(fiiilkhpll 25 F)( Ctrl Fail~ COEF 2 ol pa à alternate source of water to RClC systea (71) fron suppressim pcol (64). 24 Rod brop Accident OF 5-2 Fuel Failure 19 Core Cooling SF Send feedwater systea (N) low reactcr water level signals IL1) and low reactor pressure signals Ifron feedwater systen (03) and recirculation systen IN)) to M systen (74). 24 Rod brop Accident OF 5-2 Fuel F>>iur>>9 Ccrc Coo>>ng SF Start punps m feedwater systen N) low reactor water levl IL1) signals. ~ ~ > ~, SAFE SNllHN ANALYSIS Page 225 of Qc7 Prepared: (XRE SPRAY SYSTEN E. E ert Reeked Sate: BF)H)883~ (Rev I> C. )L Parker EVENT REACT l(AKEPTABLE SAFE N OPER RES(lTS QEHY ACTIN FIK SEO. EVENT TITLE STATE Code Title Code Title GEE 24 Rod Orop Accident OF 3-2 Fuel Failure 19 Core Cooling SF Open injectim valves m feedwater systea (03) cr recirculaticn svstea (68) Iw reactcr pressure signa)s in cmjmctim with)ow water level (LI) signal froe feedwater systee (03). 24 Rod Orop Accident OF 3-2 Fuel Failure 19 Care Cooling SF Provide standby diesel generata'82) start signals cn feedwater systea (03) Iow reacta water level signals (Ll). ZA Large Break Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Send Iow reactcr pressure signal (froa feedwater systea (03) and recirculaticn systen (68)) to M systen (74). 25A Large Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Start puaps cn priaary ccntainmt systen (64> high drywell pressure signal with feedwater systee (03) cr recirculation systee (68) low reactcr pressure signal. 25A Large Break Inside COEF 3 in ection valve cn feedwater systea (03) cr rec)rcu a i 25A Large Break Inside COEF ~9Core Cooling SF Provide standby diesel generatcr (82) systes (64) high dryweII pressure signal ~ 25A Large Break Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Send feedwater systea (03) low reactor water level signal (Ll) and Iow reactcr pressure signal (fron feedwater systee (03) and recirculaticn system (68)) to 88 systen (74). 25A Large Break Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Start puaps cn feedwater systea (03) low reactcr water level (LI> signal. 9W ~ SAFE SNT(ON N(ALYS)S Page 226 Prepared: NK SPRAY SYSTE)( Checked: dste: 8RHM~ (Rev 1) C. L Parker I EVENT SGL EVENT 71TLE REACT IVER lNKEPTABLE STATE Code RESETS Title SAFETY Code ACT)9l Title SAFE FlK C00E 5A Large Break inside COEF 3-2 Fuel Failure 19 Care Cooling SF Opm injectim valve on feedwater systea (N) cr recirculatim systea (N) lead reactcr pressure signal in cmjunctim with lod water level (U) signal froa feed<<ater systea (N). Z'4 Large beat inside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide standby diesel generator (82) start signal m feed<<ater systea (N) lo<<reactor water level signal (Li). Zik Intern krk Inside low reactor,wP)~ sj s (U) and als (froa e~ fp err' ."<<) ", ~<~ <)',I a v '- - 'rrecircutititjnsys'sha LN) to M systea (7I). 258 intern Brk inside C0EF 3~2F puca-lg-~Eoot~ Ka~rpuaps on feed<<ater systea LN) low reactor water level (Li) signals. 258 lntera Brk Inside C0EF 3-2 Fuel Failure 19 Ccrc Cooling SF Open injectim valves m feedwater systea (03) cr recirculation systea (N) lo<<reactcr pressure signals in cmjunctim with lo<<water level (U) signal froa feedwater systea (03). 258 intera Brk inside COEF 3-2 Fuel Failure 19 Core Cooling SF Provide standby diesel aeneratcr (82) start signals cn feed<<ater systea (03) lo<<reactcr water level signals (Li). 258 lntera Brk inside COEF 3-2 Fuel Failure 19 Core Ccoling SF Send lo<<reactor pressure signals Hroa feed<<ater systea (N) and recirculat(m systea (68)) to IR systea (li). Send priaary ccntainaent systea N) high drywell pressure signals to ifCl systea (T3). 258 Intera Brk inside COEF 3-2 Fuel Failure 19 Core Cooling SF Start puaps m priaary cmtainaent systea (6l) hioh drywell pressure

0 signals with feed<<ater systea (N) or recirculation systea (N) low reactor pressure signals.

~ ~ ~ SAFE SWTO(IN ANALYSIS Page 227 of w7 Prepved: CIRE SPRAY SYSTE)I E.. Ec Checked: Oate: BF)H)SS3~ IRev 1) C. L Pvker I BENT SEL SENT TITLE REACT (M(NXEPTABLE (PER RES(LTS STATE Cade Title SAFETY Cade ACTIN Title SAFE FlK (IE 25B Intern Brk Inside COEF 3-2 Fuel Failure 19 Care Caoling SF Open injectim valve cn feedwater .Systen (03) or fecifcu)at(m spiN (N) lw reactl pressure signals in cmjunctim with high drywell pressure signals fran prinary cmtainaent systen (6I). 258 Intern Brk Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide standby diesel generatcr (82) start signals m prinary ccntainaent s ten IM) hi dr ll esture sagnalso 258 Intern Brk Inside 2 iFu) 'kaIIPejk; gp~igj ( c g ~ ~y sigetkis:."(tIlaf cpg ean ) a depress&I?atim systen (AOS 25C Snail freak Inside ue ai ure P Care &l~ng reactor water level signals (L1) and lw reactcr pressure signals (fran feedwater systen (03) and rectrculat(m systen (N)) to R)R systen (74). 25C Snail Break Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Start puaps cn feedwater systen (I) lw reactcr water 'level (Ll) signals. 25C Snail Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF . Open injecticn valves on feedwater systen (03) cr recirculaticn systen (N) lw reactor pressure signals in cmjmcticn with lw water level (LI) signai fran feedwater systen (03). 25C Snail Break Inside (XEF 3-2 Fuel Failure 19 Core Cooling SF Provide standby diesel generator (82) start signals on feedwater systen (03) lw reactcr water level signals ILI). 25C Snail Break Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Send lw react(r pressure signals (fran feedwater systen (03) and recirculatim systea IN)) to M systen (7l). Send prinary cmtainaent systen (Q) high drywell pressure signals to IfCI systen (73). SAFE 9$)IO(N QQLYSIS 3o'7 Prepared: Oate: 1 e m '.? CIRE SpimY SYSTE)I EF)H)SS3~ IRev 1) Checked) C. L Parker Oate: BM I SEQ. BM TlTLE 88)CT IMXEPTARE OPER STATE Code REM.TS Title SAFEtY Code ACTIN Title .SAFE FIC CNE 25C Snail Ereak Inside COEF 3-2 Fuel Failure 19 Core Cooling SF Start puaps cn prinary ccntainant systen N) high drywell pressre signals with feel)water systea 103) cr recirculaticn systen (N) lw reactor pressure signal s. 2X Snail Break inside COEF 3-2 Fuel Failure 19 Core Cooling SF Open injectim valve cn feedwater systen (03) cr fecirculatim systen N) lw reactcr pressre signals in ccnjunctim with high drywell cmtainaent systea N). ~ e 25C Saall Break inside $ Apl Faille gP (ere Cbolin'g SI., ~RA'Q'.,4tahdbf;dI'esel,:gene@ter(IB2) ',",','..., >.", ~start signds cn priaary'cmtainapnt t', <'k 4 '. i; ':,:;. '.s)jsten'IMI: high~}l'pressure'ignals. 25C Seal 1 Break inside 3~4Muco4-puaps are running) to naia stean systea (01) for the autoaatic depress'iaatlm systen IAOS) ~ 27 Pipe Break Outside IXEF 3-2 Fuel Failure 19 Core Cooling SF Send feedwater systen (03) lw reactor wats level signals IL1) and lw reacts'ressure signals Hron feedwater systen (03) and recirculatim systea Ib8)) to IR systen (70, 27 Pipe Break Outside COEF 3-2 Fuel Failure 19 Core Cooling SF Start puaps m feed<<ater systen N) low reactor water level IL1) sigaals. 27- Pipe Breal Outside COEF 3-2 Fuel Failure 19 Core Cooling SF Open injectim valves m feedwater systea (03) cr recirculaticn systen IN) lw reactcr pressure signals in conjmcticn with lw water level IL1) signal froa feedwater systen (03). 27 Pipe Break Outside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide standby diesel generator (82) start signals cn feedwater systen N) low reactcr water level sigaals (Ll). 31 SAFE SWARM AIALYSIS ORE SPRAY SYSTDI Prepared: P 229 Ilats ~7 > ~ lF)HM~ IRev 1) Reeked: 1)ate: 7 C. L Parker BM REACT IIACCEPTASLE SAFE OI 9%R RERLTS 'AFEIY ACTllM FIK SEL EUM TlTlE STATE Code Title Code Title CNE 32S Low Reservoir GKF 2-2 Fuel Failure 19 Ccrc Cooling Provide fits path fcr alternate smrce of water to RC)C systen I71) froe suppressim pool IM). lSI) lSO Shutdown GXF 2-2 Fuel Failure 19 Core Cooling Provide fits path fcr water to RClC systen I71) frcn suppressicn pool I6f). IXEF 2-2 Fuel Failure 19 Ccrc Cooling Provide flow path fcr water to RClC systee I71) frow suppressim pool IN. SSI) Safety Shutdown CKF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide backup to fHI I7$ ) - tPCI ~ ode) if necessafyo ZI) Srklnside+ntSad Sop port 03 )lain'Stw lsolatim P Release+:Est+ j;Cmt~$F.~ fpfcrn iso)atimpctimfp) ~ receiving tow.watfr;levy Ith) ~ Isolatlm sagnal frow ti)a Pff nary 'Cmtaihaent'systea 'tQ).'ys safety actlm Is espected to ocos', but is l.oqgireeeMx>~nt since there is no fuel failure. 05 bss of Uacuue IXEF 2-1 Rad Release 26 Est Pri Cont SF Perfornisolaticn acticnls) upon receiving low water level IQ) isoliticn signal fron the Prieary Ccntainaent systen 16l). This safety acticn is expected to occur, but is sot a requirecent fcr this event since there is no fuel failure. 13 lnadvert 0pen SRU Rad Release 26 Est Pri Cmt SF Perfcra isolation acticnls) upon receiving low water level IQ) isolaticn signai fra the Prinary 'mtaimat systen 161). This safety actim is espected to occur, but is aot a requireeent fcr this event since there is no fuel failure. ll loss of f)I flow Rad Release 26 Est Pri Cont SF Perfcre isolatim actimis) tpn receiving low water level IQ) isolatim signai froa the Priaary Cmtaltw%t systen I6I) This safety actim is expected to occur, but is sot a requireeent'fcr this event since there is no fuel failure; N SAFE SN)T(M( NNLYSIS Page 230 of Prepared: C. ()ate: (XRE SPRAY SYSTB( E. Ec Checked: 1)ates MKSS3~ (Rev 1) C. )L Parker BE)(T REACT MXEPTAILE SAFE N PER RES(LTS SAFETY ACTI% RK SEL SE)(T TITlE STATE Code Title Cade Title GEE finite

15. Loss Of AC NL 2-1 Rad Release 26 Est Pri Cmt SF Perfa'n isolatim actim(s) pe receiving Iow water level (Q) isolatim signal fron the Priaary Cmtaineent systen (M). This safety actim is expected to ocnr, but ls aot a require+at fcr this event since there is no fuel failtre.

23 F)( Ctrl Fail~ (XEF 2-1 Rad Release 26 Est Pri Cont SF Perfcrn isolatim acticn(s) upon receiving low water level (Q) Iso)akim signal frow the Prinary Cmtainant systen (M). This safety actim is expected to occur, but is fcr this t since there is no fuel fail e. 25 Rod hop Accident @let+( h'urge.@M ~"4Mijri~t~atpw~>  !.', ~ . Qsolitim signil'frat()e Cmtaimnt systen N). 25A Large Ereak Inside -MR%lease 26 St'Fi'Kn receiving Iow water level (Q) 1solatim signai fron the Prinary Cmtaimnt systea (6l). 25A Large Ik eak Inside 0)EF 3-1 Rad Release 26 Est Pri Cont SF Perfcra isolatim acticn(s) upon receiving high drywell presme isolaticn signal froa the Prieary Cmtainaent systen N). 258 Intern h'k Inside (XEF 3-1 Rad Release 26 Est Pri Cont SF Perfcrn isolatim actim(s) upm receiving low water level (Q) isolatim signal frow the Priaary Contain+at systen N). 258 Intern Brk Inside CSEF 3-1 Rad Release 26 Est Pri Cent SF Perfcrn isolatim actim(s) upon receiving high drywell presare isolatim signal free the Priaary Cmtainaent systen (6l). 25C Snail h'eak Inside (XEF 3-1 Rad Release 26 Est Pri Cent SF Perfcra isolatim actim(s) upm I receiving low water level (Q) isolatim s(goal froa the Prinary Cmtaineent systee N). SAFE SHJMN AN.YSIS Page 231 M~~ Prepa ed: Sate: Cn(E SPRAY SYSTE)( E Decked: Bate: BF)H)SB3~ (Rev I) C.JLP k UM IEACT IMXEPTAKE SAFE OR (PER REKTS;SAFEIY ACTTCI( RlC SE(). Be(T TlTLE STATE Code Title Code Title CNE 25C Snail Treak inside IXEF 3-1 Rad Release 26 Est Pri Cont SF Perfme isolatim action(s) upon receiving high drywel1 pressre isolatim signal froa the Priaary Containaent systea (M). 25D Brklnside+mtlf(ad C(EF 3-1 Rad Release 26 Est Pri (s) upm receiving Iw water 1 l (Q) ip)atpp I((I t e Priaary ...eP;;~L, tf';lg'Y~",'," 250 Brkinside4mt 3-1 s) upm high drywel1 'eceiving em'e isolatim s Priary tainaent systea (M). 25D .Brklnsid~tNad COEF 3-1 RadRelease 26 EstPri Cont Supportpriearycmtaimnt functim. 27 Pipe Break Outside CKF 3-1 hd Release 26 Est Pri Cont SF Perfcra isolatim action(s) upm receiving Iw water level (Q) isolatim signal fra the Priery Cmtaineent systee (M). '.~A,~ p~> ccX 15 Loss Off-Site AC ALL 2-1 Rad Release 27 Est Sec Cmt Support sec mtainaent fmctim. 25D Brklnsid~tÃad D)EF 3-1 Rad Release 27 Est Sec Cent Support seccndary cmtainaent functim. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont Supper t secmdary cmtainamt fmcticn. 33 Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont Support secondary containaent fmcticn. 28 Shtdn Frow Outside (XEF l-l Local Shut 3g Cooldn Outsd Provide fin path fN'lternate smrce of water to RCIC systee (71) froe supprgssim pool (g), P~~~ Pi~m ~Ice,)~ 32D Ear thquake Nl 4-3 Equip Fail 4I Seisaic Shtd Design of safety features aust hav(f & %A~ capability to provide shutdown follonng a design basis seisaic event. r SAFE )ITEN NOLYSIS Page 232 Pry ared: C. 0 Oate: CfRAMNT MRTIS E. SSEI I'eeked: 0ate: 'NHM~ (Rev 1) C. )L Parker STSIE)f 76 BENT ISA SNXPTASLE SAFE OR NER IKSI.TS . SAPETT ACTlN RK SB. SHIT TlTLE STATE Code Title Code Title GXE 250 Srklnsi~thRad IXEF & Cmt Stress 23 Lin 0isturb L The Prinary Cmtainant is inerted with nitrogen to displace aygen. 03 Nain Stn lsolaticn Rad Release 2b Est Pri Cmt SF Perfcra isolaticn acticnis) upon receiving lw water level fQ) isolatim signal fron the Prinary Ccntainnent systen N). This safety actim is epected to occur, but is not a requireaent fcr this event since there is no fuel failure. 05 Loss of Vacuun Pri Cent SF Perfcrn isolaticn actim(s) upon r isolaticn signal fron the Prinary , ~ > ii 4.,l i;,"'..jnot"a,r~eatnt,'for'i]his eject sincVth~er8 is no fueE: faUge. 13 lnadvert Open SRV fXEF 2-1 Rad Release 2~st~~HF ~ecfzy isolation acticnls) receiving lw uater Keel isolatim signal fran the Priaary Cmtainnent systea N). This safety actim is expected to occur, but is aot a requirenent for this event since there is no fuel failure. 14 Loss of FN Fice CCEF 2-1 Rad Release 2f Est Pri Cmt SF Perfcrn isolation acticnfs) upon receiving lw uater level tQ) isolatim signal fron the Priaary Cmtainnent systea IM). This safety action is eapected to occur, but is aot a requirenent fcr this event since there is no fuel fai)ure. 15 Loss Of finite AC ALL 2-1 Rad Release 2b Est Pri Cont SF Perfcra isolation actimfs) upon receiving lw uater level (Q) isolatim signal fron the Prinary Cmtainaent systen le). This safety icticn is apscted to occur, but ir not a require+at fcr this event since there is no fuel failure. SAFE BiAlXNA%LYSIS Pag 233 of A ted'te: Prepared: Bate: I'. CC)fTAIQSIT 1%HTII6 SSIBI 4P BF)H)SS3~ {Rev 1) C. L Parker BM I SEL EVENT TITlE REACT CPER MXEPTARE RESETS STATE Code Title . SAFEIY Code ACTIN Title SfE flK GIE 23 flf Ctrl Fail~ G)EF 2-1 Rad Release 2b Est Pri Cmt SF Perhra isolaticn actimls) upm receiving Iow water level IQ) isolatim signal froa the Prinary Cmtainant systen N). This safety acticn is expected to occur, hut is not a requiri!t fcr this event since there is no fuel failure. 2f Rcd hop Accident If 3-1 Rad Release 2b Est Pri Cmt SF Perfan isolatim acticnfs) upon receiving low water level IQ) isolatim signai fra the Priaary Cmtainaent systen N). 25A Large Break Insi t Pri Cant SF Perfcra isolatim acticnls) upon receiving Im wa er eve IQ) isolatim signal frow th Priaary, 25A 'arge Break lnsid a3-I 44 ReIAsf Ck ht",5'i"Cmt!SF," .'Perfcri i4 atIicfL icn s) upon receiving high drywell esslre,. isolatim signal frow t Prinary 258 Intern Brk Inside G)EF 3-1 Rad Release 2b Est Pri Cant SF Perfcrn isolaticn actim(s) upm receiving lcw water level iQ) isolaticn signaI fron the Priaary Cmtainaent systen N). 25B Intern 1kk Inside G)EF 3-1 Rad Release 2b Est Pri Cmt SF Perfcrn isolaticn acticn(s) upm receiving high dryweII press!re lsolatim signal fron the Priaary Cmtalna8lt systen N) ~ 25C Snail Break Inside CKF 3-1 Rad Release 2b Est Pri Cont SF Perfora isolaticn acticnis) upon receiving low water level IQ) isolatim signai free the Priaary Cmta)AK%t systen N) ~ 25C Snail Break Inside GXF 3-) hd Release 2b Est Pri Cont SF Perfcrn isolaticn acticnts) upon receiving high drywell pressure isolaticn signaI fron the Prlaary Cmtainant systen N). SAFE SHlliSH ANALYSIS Prepared'. Date: 4 GNTAIQBIT MRTII6 SYSTE)I BF)H)SS3~ IRev 1) Checked: C. IL Parker Pate: ~41 7 SYSTE)I 76 BE)IT REACT MXEPTABLE SAFE OR 5%R RESETS . SNYlY ACTIN FIK KI. BE)IT TlllE STATE Code Title Code Title GEE 25D Brklnside+mtSad lXEF 3-1 RadRelease 26 Est Pri Cont SF Perfcre isolatim actimls) upon receiving low water level IQ) isolatim signal fron the Prinary Containant systee I6l). 259 Brklnsi~tSad EKF 3-1 Rad blease 26 Est Pri Cont SF Perfcre isolaticn actimis) qxn receiving high drywell pressure isolatim signal froe the Prinary Cmtainaent systen N). 27 Pipe Break Outside IXEF 3-1 Rad Release 26 Est Pri Cont SF Perfwe isolation acticnls) upm receiving low water level IQ) isolaticn signal fron the Priaary fC< 15 Loss Of finite AC NL 2-1 Rad Release 27 Est Sec Cont Support ~ary cmtaineent functim. 25D frklnside+mtIRad GEF 3-1 Rad Release 27 Est Sec Cent Support secmda y cmtainesnt 26 Fuel Handl Accidnt AI 05PP Pfei4 $ gP Rally 'tn. 33 Loss Fuel Pool Cl ILfa. fmction. SAFE SITMN ANLYSlS '(N(ASTE SYSTH( Prepared: QO7 nie: ~r t'7 Chwked: Oate: P NH)SS3~ (Rev 1) EUBIT ~ REACT (NACCEPTASLE SAFE OR , OPER KBLTS 'AFETY ACllSI RK SE(). EUENT TlTLE STATE Code Title Code Title (IE O3 )(ain Sta lsolatim COEF 2-1 Rad Release 26 Est Pri Cmt SF Perfae isolaticn action(s) upcn receiving )ow>>ater level (Q) isolatim signal froa the Priwary Ccntainant systea (M). This safety actim is expected to occur, but is not a requireeent fcr this event since there is no fuel failure. 05 Loss of Uacuua (XEF 2-1 RadRelease 26 Est Pri Cmt SF Perfcra isolaticn actimls) upm .receiving low water level (Q) isolaticn signal froa the Prinary Cmtainaent systee (64). This safety actim is expected to occur, but is not a requireeent for this event D Inadvert Open SRU Gp 2 l iied Reieeee+6 ieiai Peti e(eegere iieieiiee y:ija(eei v y fety actim is expected to occur, bu is lot a f if since there is no fuel failure. / H Loss of FN Flow CKF 2-1 Rad Release 26 Es't Pri Cont SF Perfcre isolatim acticn(s) upon receiving lw water level (L3) isolaticn signal frow the Priuary Cmtainaent systen (M). This safety action is expkted to occur, but is ~ ot a requireaent fcr this event since there is no fuel failure. 15 Loss Off-Site AC ALL 2-1 Rad Release 26 Est Pri Cont SF Perfcre isolation actim(s) upon receiving )ow water level (Q) isolaticn signal froa the Priaary Cmtainaent systea N). This safety acticn is expected to occur, but is not a requireaent fcr this event since there is no fuel failure. 23 R( Ctrl Fail~ COEF 2-1 RadRelease 26 Est Pri Cont SF Perfcra isolaticn acticnls) upon receiving low water level (L3) isolaticn signal froa the Priaary Containeent systee (N. This safety actim is expected to occur, but is ~ ot a requireaent fcr this event since there is no'fuel failure. v ~ 4 SAFE )IJINN( NQ.YSIS Page 236 QO7 Prepared: C Date: Zl MNSTE STSTH{ E Checked Date: 'BF)HEST (Rev I) C. IL Parker BM REACT MXEPTABLE SAFE N lPER RES(1TS . SAFEIY ACTI(N F(K SEL SENT TITLE STATE Code Title Code Title GEE '2l Rod Drop Accident DF 3-1 Rad Release 26 Est Pri Cmt SF Perfw'n isolaticn acticn(s) upon receiving low water level {Q) isolaticn signal frow the Prinary Cmtainaent systen (M). 25A Large Break Inside GKF 3-I Rad Release 26 Est Pri Cent SF Perfcrn isolaticn action(s) qxn receiving low water level (Q) isolaticn signal frm the Priaary Cmtainmt systen {M). 25A Large Brea'k Inside (XEF 3-1 Rad Release 26 Est Pri Cont SF Perfoa isolaticn acticn(s) qxa receiving high drywell presme isolaticn signa) fron the Prinary CmtalnaHlt sys'ten {6()~ 0 258 Intern Brk Inside GXF 3-1 Rad Release 26 Est Pri Cont SF Perfan isolaticn acticn(s) upm receivi aso atim sional free the Priaar Cmtainaot systen {+f ) !5 ~ f 258 Intern Brk Inside 'E ),, tP> t) i@fjn)('solitiie.acQm{s)'~ l V' '" ' { 'asolaticn signal fron the Prinary Cmtainaent systea {6l 25C Snail Break Inside t SF Perfcra isolaticn actim(s) upm receiving low water level (Q) isolaticn signal froa the Prinary Cmtainaent systen N). ZC Snail Break Inside CDEF 3-1 Rad Release 26 Est Pri Cont SF Perfcra isolaticn acticn{s) upon receiving high drywel I pressure isolatim signal froa the Prinary Cmtainaent systea N). 25D BrklnsidH'mtNad (XEF 3-1 Rad Release 26 Est Pri Cont SF Perfa ~ isolatim actim(s) upon receiving low water level (Q) isolaticn signai froc the Priaary Cmtainaent systen (M). -0 25D Brkinsid~tNad (XEF 3-) Rad Release 26 Est Pri Cont SF Perf(rn isolation acticn(s) upm receiving high drywel I pressre isolatim signaI froe the prinary Cmtainaent systee N). gaol SAFE )STBM NMIS Page 2v7 of Prepared: ~.C.E Oate: 4 AANA)STE SYSTE)I E Necked: Bate: NH)SS3~ IRev 1) C. )L Parker BM REACT MXEPTABLE SAFE OR !PER RESILTS,'AFETy ACTifN RM! SH). PM TlTLE STATE Code Title Code Title GEE 27 Pipe Break Outside C0EF 3-1 Rad Release 26 Est Pri Cent SF Perforn isolatim acticnis) upon receiving )ou uater level IL3) isolatim signal fron the Priaary Cmtainee~et yeast IM). 4 15 Loss Offdite AC NL 2-1 Rad Release 27 Est Sec Cmt Support secmdary cmtainaent fmcticn. (g~~ n 250 Brklnside-CcnttRad C0EF 3-1 Rad Release 27 Est Sec Cont Support secondary containeent functim. 26 Fuel Ifandl A:cidnt A 3-1 Rad Release 27 Est Sec Cont Support seccndvy ccntaineent fmctim. 27 Pipe Break Outside CBEF 3-1 Rad Release 27 Est Sec Cont Provide cmtrol roon indicatim of high uater level in the Reactor Building+ fe+~a(~ laaL 4Hahe/'(mfa ) g,o+~ c~ ~~ 27: Pipe Break Outside CBEF 3-1 Rad Release 27 Est Sec Cent SF Puep out Reactcr Building Floor 0rain Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cent ~fi~~llc/a-/mP~~ Suep on high level.Cg Support secmdary b~~' contaimni fmcticn. SAFE 9eliaa ANALTSIS Q>1 Page g3 ate: e Prepared: C M FILL PKL CNLI% h E. IGIMRALII1%SWEN checked: Sate: RHSM~ (Rev 1) C. L Parker BM REACT MXEPTAKE SAFE N SQ RESII.TS . QfHY ACTIN RK SN. BM TITLE STATE Code Title Code Title GEE '15 Loss Offdite AC Nl 2-1 Rad Release 27 Est Sec Cont Support secmdary cmtaineent funitim. 25D Srklnside-Canted CHIEF 5-1 Rad Release 27 Est Sec Cmt Support secmdary cmtaineent functim. ) 2b Fuel Handl Accidnt A 5-1 Rad Release 27 Est Sec Cent Support secmdary cmtaineent fmction. ,2 Loss Fuel Pool CI ALL 2-1 Rad Release 27 Est Sec Cmt Support secmdary containeent functim. 15 Loss OH<ite AC ALL 2-1 Rad Release Q FuelPoolCool N Provide fuel pool cooling shen power and cooling water ve available. -0 4 SAFE SH)TIQN ANALYSIS P 239 of FKL )NDL1% A)0 SlUNK SYSTEN SF)H)SS3~ Iev 1) Checked: E. C C. IL Parker Date ~1 BENT REACT MXEPMLE SAFE N 9%R RERI.TS 'AFElY ACTlCff RK SEL Oer TTTLE STATE Code 'Title Code Title CNE 11 Control hd Reeovl A 2-2 Fuel Failure 23 Li~ Disturb L Refueling interlock switches <<e used to liait errcrs. 2b Fuel Handl Accidnt A 3-1 Rad Release 23 Lie Disturb SF The refueling interlock equipaat is used with the M systew I85) rod blocks to prevent a ccntrol rod withdrawal errcr during refueling. 15 Loss Off<ite AC Nl 2-1 Rad Release 27 Est Sec Cont Supp<< t seccndary ccntaineent fmcticn. ZG Brklnsid~tNad CCEF 3-1 Rad Release 27 Est Sec Cont Suppcrt seccndary ccntaineent fmtim. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont Support seccndary ccntaineent function. 33 Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont Support seccndary ccntaineent fmx:ticn. 32C Ta nado ALL 4-3 Equip Fail 40 Struct Integ L hfueling platfcres are to be tied dao under tornado conditicns. Mith this provisicn, the refueling platfcres aust withstand tcrnado design loads (passive), 32D Earthquake NL 4-3 Equi eg ion of safety features est have capability to povide shglse if'$I)%fj '"afig4. ise~ic. r>.. ~ $ l p~< ~+I) i3 SAFE QUINN ANALYSIS Page 20 D1 pry v~: e Date: STARDBY DIESEL SBKRATN E. SYSIE)I Checked: Date: RHEA~ IRev I) C. L Parker EVENT REACT INCCEPTABLE SAFE OR PER REStLTS 'AFEIY ACTIN FIK SEL EVENT TITLE STATE Code Title Code Title CIXE 'l5 load~mt Open SRV CKF 2-2 Fuel Failure 19 Core Cooling SF Diesel g8leratcrs s'taft m low reactor water level sionals IL1) free are spray systen 175). Ibis safety actim nay occur, but is not required by this event since nornal ac is available. 15 Loss Off-Site AC COEF 2-2 Fuel Failure 19 Care Cooling SF Diesel generatcrs stvt m low reactcr water level signals ILI) frcw cae spray systee I75). 2f Rod Drcp Accident EF 5-2 Fuel Failure 19 Core Cooling SF Ihesel generatcrs start m Iow reactl water level sianals ILI) fran ccrc spray systen 175). 25A Large Break Inside COEF 3-2 Fuel Failure 19 Ccrc Cooling SF Ihesel generatcrs start m prfnary cmtainaent systen I6I) high dryweII pressure signai. 25A Lvge Break inside CDEF 5-2 Fuel Failure 19 Care Cooling SF Diesel generatcrs start on low reactor water level sianal ILI) fran ccrc spray systen 175). 258 Intern Brk Inside CDEF 5-2 Fuel ailure 19 Core Cooling SF Diesel generatcrs start on Iow , ~, , seactgr yterc~l.@gnais ILIA'io~: i j icre;pre,'s)stan:'.It5) "." .;.;,". ) g.( 25B Intern Brk Inside CDEF 5-2 Fuel ai ur4 l &e QIITcgI SF ~'Itck'I pen'craters start m prinvy ccntainaent systen N) high dryweII 25C Snail Break Inside COEF 5-2 Fuel Failure 19 Core Ccoling SF Diesel generatcrs stvt on low reactm water level signals ILI) froa care spray systen f75) 25C Seal I Break Inside CDEF 3-2 Fuel Failure 19 Care Caoling SF Diesel generatcrs start cn priavy ccntaicment systea N) high drywell pressure signals. 27 Pipe Brea'k Outside COEF:-2 Fuel Failure 19 Core Cooling SF Diesel generatcrs start on low reactcr water level sinnals ILI) free care spray systen 175). I v,( SAFE QUEM AHALYSIS Page 24I oQSL~ Preoared: Date: QNHY DIESEL 6E%RATN E. SYSTEN Checked: Date: ~IS /7 BF)HSS3~ (Rev I) C. )L Parker BSIT REACT MXEPTABLE SAFE IR CPER REStlTS 'SAFETY ACTI% RK SEO. BIE)IT TITLE STATE Code Title Code Title CODE ZE Fire ALL W Fuel Failure 19 Core holing Design of safety features ant include Appendix R fire consideraticns. SSD Safety Shutdae COEF 3-2 Fuel Failure 19 Core Cooling SF Supply eeergency poser to uata supply systees for ccrc cooling, if required Nl P~gQ gcdr'@sell fuel ~~4m ~ 15 Loss Offdite AC 2-1 Rad Release 21 Rest AC Pow N pool and cooling ~~ ~ ~wee lor pool aakeup)@ken coolie ~~~I~ '+p 4+~6 15 Loss Offdite AC IXEF 2H Cent Stress 3D Cont Cooling NSF Provide ac pacer to suppcrt ccntaimnt cooling functicns taanual or auto start cn loss of offwite pe%) ~ ZC Saall Break Inside CIXF 3-l Cont Stress 3O Cont Cooling SF Provides standby peer toccntainaent cooling systeas. 2% Brklnside+ztQhd CDEF H t ling SF The diesel generatcrs start cn loss of $ .16 0 bus voltage Or@all hig)) 'g'. j ".(b:iigainQti ORol)ngxvsystens.. "4 32E Fire t Condtns ~A Cont Cooling Design of safety features oust re ccnsideratlcns. t SSD Safety Shutdose CKF 3H Cont Stress 30 Cont Ccoling SF Supply eeergency p~ to R}ft systen Ilk) suppression pool cooling, if required. 28 Shtdn Froe Outside ALL H Local Shut 39 Coo)do Outsd N Required fcr operation of systees if ncraal ac pouer is lost. 32D Earthquake Nl I-3 Equip Fail Q Seisaic Shtd Design of safety features ast have capability to provide shutdae following a design basis seisaic event. QG7 SAFE S)mlHXN ANALYSlS Page 2f2 Prepared: hte: ~+( HNAM)fTATMSP)HIE E. Ec D]LUTIN SYSTEM Checked: Date: P 7 DF)H)SS3~ fRev 1) C. L Parker MH REACT MXEPTADLE SAFE N IPER RERLTS QFHY ACTIN flK SED. MIT TITLE 'TATE Code Title Code Title GEE 25D 8rklnsid~tNad CDEF 3H Cent Stress 23 Un Disturb NSF The prinary cixitainnent atnosphere is diluted with nitrogen after a LKA to liait the anount of hydrogen and oxygen presents 25D Brklnside+atNad IXEF 3H Cont Stress 23 Un Disturb N The prinary ccntainnent atnosphere is vented to SSTS N) tn naintain pressure bel~ a predeternined liait. D3 Nain Stn Iso)atim CDEF 2-1 Rad Release 2h Est Pri Cont SF Perfora isolation actionis) upon receiving low water level (Q) isolation signal froa the Prinary Ccntainnent systen (Q). This safety action is expected to occur, but is not a reguirexent fcr this event ~ '5 Loss of Vacuun CDEF 2-1 Rad Release 26 Est Pri Cont SF since there is no fuel failure. Perforn isolation actionis) upon receiving low water level (Q) isolation signal fron the Priaary ~ Ccntaimnt systen Sl). This safety acticn is expected to occurx bg~ not a requir is event ere is no fuel failure... 13 lnadvert 9N 2-1 Rad Release Est Pri Cont SF P~foxa) M@or,. a'ctionis).open "; Dpen CDEF fwl ~+ ~ yl '~)'> rjCei4ng'lowkiter",1b4l Tg) '. "- 0 ': ';iiolabm'j iqr44froa 'th"e Priaary ~ x" Ccntainaent systen N). This safety vxj> acticn is expected to is not a gquireeSAcr this event since there is no fuel failure. ll Loss of FN Flow CDEF 2-1 Rad Release 2b Est Pri Cont SF Perforn isolation acticnis) upon receiving low water level iQ) isolation sinnal fron the Prinary Containnent systen 164). This safe'ty action is expected to occur, but is not a requireaent fcr this event since there is no fuel failure. (WT(XN ANALYSIS 5o7 SAFE Page 213 Prepared: Date: il GNAMN ATN)SP)GK DILUTIDN SYSTEM (heckedc Date: BRH)S83-048 (Rev 1) SENT REACT MXEPTABLE SAFE 0( IFER RESETS SAFEIY ACTI(N FlK SEL SENT TITlE STATE hde Title hde Title GEE 15 Loss Dff&te AC ALL 2-1 Rad Release 26 Est Pri Cont SF Perfora isolaticn acticn(s) upon receiving Iow water level (Q) isolaticn signal froa the Priaary Containaent systea N). This safety acticn is expected to occur, but 1's aot a requireaent for this event since there is no fuel failure. R( L'trl Fail~ CDEF 2-1 Rad Release 26 Est Pri Cont SF Perfora isolation acticn(s) upon receiving Iow water level (Q) isolatim sional froa the Priaary hntainaent s)wtea N). This safety action is expected to occur, but is pot a requireacnt fcr this event since there is no fuel failure. 2l Rod Drop Accident" 3-1 Rad Release 26 Est Pri Cont SF Perfora isolatim action(s) upm receiving Iow water level (Q) isolatim signal frca the Priaary Ccntainaent systea (6l). 2M Large Break Inside CDEF 3-1 Rad Release 26 Est Pri Cmt SF Perfora isolaticn action(s) upm receivi 1 ) 'acon signal froa the Priaary CmtaiAKot systea+ ~ c t ZA Large Break Inside 3-1 Rad 1~,.26')st Prf hnl SF.. Perfori!jsplat(cn.'actimts) upon,; i CDEF f c-'a:~ ".. "-', "rect'ising',hcghAiyweIT ]@assure xrx5 Q "~-'olitccn sional fron the Priaary hntainaent systea (6l . U'ad 258 lntera Brk Inside Re Perfora isolation actim(s) upm receiving low water leal (Q) isolatim signal froa the Priaary hntainaent systea N). 258 Intera Brk Inside, CDEF 3-1 Rad Release 26 Est Pri Cont SF Perfora isolation actim(s) upm receiving high drywell pressure isolaticn signal froa the Priaary hntainaent systea (6l). 25C Saall I'eak Inside CDEF 3-1 Rad Release 26 Est Pri Cent SF Perfora isolatim acticn(s) upm receiving Ioec water level (Q) isolation signal froa the Priaary hntainaent s)stea N). SAFE S)URN ANALTSIS 24 '3b 7 Page Prepared: Date: QNAIQefT ATN)SP}fRE DILUTllN SYSIE)I Checked: Date: 7 SF)H)SS3~ (Rev I) C. L Parker BM REACT MXEPTABLE SAFE N IPER RESLLTS QFHY ACTIN FIK SE9. BW TITLE STATE Code Title Code Title GIE-ZC Seal) Break Inside CDEF 3-1 Rad Release 26 Est Pri Cont SF Perfora isolatim actimfs) upm receiving high drywell pressure isolation signal froa the Prieary Cmtainmt systen N). 25D Brklnside+mtlftad IXEF 3-1 'Rad Release 26 Est Pri Cent SF Perfore isolatim actimls) upm receiving low water level (L3) isolaticn signal froe the Prieary Cmtainaent systee (bl). 25D Brklnsid~ttRad IXEF 3-1 Rad Release 26 Est Pri Cmt SF Ps fore isolation actimls) upm receiving high dryweli pressure isolatim sional froa the Priaary Cmtalneent systee 16l). 27 Pipe Break Outside COEF 3-1 Rad Release 26 Est Pri Cont SF Perfore isolaticn actimis) upm receiving low water level IL3) Cmtainaeqt systen recuc serg~ ~, isolatim signal froa the Priaary &wHN). 15 Loss Dff-Site AC Nl 2-1 Rad Release 27 Est Sec Cont Support secmdary cmtalfl%%t functicn. 25D Brklnsid~t Sad CDEF 3-1 Rad Release 27 Est Sec Cont Support secmdary cmtalneent functim. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont ary cmcalneent fnocti cn. IW, I) l r 1/ if ') Jlp I, ~ '. l ) 33 Loss Fuel Pool Cl ALL 2-1 Rad lease 27/st Sfc)<ba+ p >.Rp'ph,'t;sejmdjry;cmt'iinednt t. ij v'I ki"i s k Rt S)ITIQN AN.YSIS sf~7 ~ SAFE Page 24$ Prepared: Date: CNTIK RDD DRIVE E. Checked: ate: SRHM~ IRev 1) C. L Parker EVM REACT IIACCEPTABLE SAFE N 8%R RESILTS SAFETT ACTICff RK EEL EVENT TITLE STATE Code Title Code Title CDDE DIA Generator Trip F 2-2 Fuel Failure 17 Scraa Scraa sional fron reactor protectim systen N) will activate the ccntrol rod drive systea to insert rods. Scraa function cnly. DIB T/G Trip-5'ails F 2-2 Fuel Failure 17 Scraa Signal fron RPS199) will activate the Ccntrol Rod Drive to insert rods. Scraa function mly. 01C PresReg FaiH:lose 'F 2-2 Fuel Failure 17 Sera Scraa signal froe reactcr protectim systen N) will activate the ccntrol rod drive systen to insert rais. Scran functim only. 02 Turbine Trip F 2-2 Fuel Failure 17 Scraa Scraa signal froa reactor protecticn systen l99) will activate the ccntrol rod drive systen to insert rods. Scraa functim only. 05 Nain Stn Isolation DF 2-2 Fuel Failure 17 Scraa Scru signal froo reactor protectim systea l99) will activate the ccntrol rod drive systen to insert rods. Scran functicn only. 04 Close Cne ISIV F 2-2 Fuel Failure 17 Scru Scraa signal fron reactor protecticn systea r od dr'ert N) will activate Scraa function cnly. the c rais. 05 Loss of Vacuua 2- Fuel Failure 17 Serg,, SF.;., Scrakslgnjl~!froi.'qeactcI protjctIm; i! i .r,"'-4yastbi, lggP will activate tbe,'c)I&oi > sz4<').',<,".';" <;; ', '~";rod,/rivi sI)stei to insert rods. Scraa function mly. 06 Loss F)f Heater 2-2 uel Failure 17 Scraa gna froo reactor protectim systen f99) <<ill activate the ccntrol ra! drive systen to insert rods. Scran functim mly. 07 Shutdn Cool Nalfun D ":2 Fuel Failure 17 Scraa Scraa signal froa reactor protecticn systea N) will activate the control rod drive systen to insert rods. Scraa function mly. t SAFE BM N SNIMW AhNLYSIS CMHl R00 SRHKM~ DR)YE IRev I) REACT IM MXPTABLE RESILTS Prep ved: Checked: SAFETY ACTICÃ 5AFE RK Pane E. 24 C. II. Parker of Date: Date: $3 tF SED. BENT TITLE STATE Code Title Code Title CXE '8 Inadv Puwo Start DF 2-2 Fuel Failure 17 Serac 5F Scraa signal frow reactcr protectim systee I99) will activate the cmtrol rcd drive systea to insert rods. Scraa functim only. O9 Rod Withdraw Errcr DF 2-2 Fuel Failure 17 Scraa SF Scrae signal froe reactor protectim systea f99) will activate the cmtrol rod drive systen to insert rods. Scraa functim only. 15 Pres Reg Fail+en OF 2-2 Fuel Failure 17 Scraa SF Scraa signal fron reactcr protectlm systen 199) will activate the control rod drive systea to insert rods. Scran functim only. 12B Inadv Open-Bypass If 2-2 Fuel Failure 17 Scru SF Scraa signal fron reactor protectim systea 199) will activate the ccntrol ~ rod drive systea to insert rods, Scraa functicn only. 13 Inadvert Open SRV DF 2-2 Fuel Failure 17 Scrae Scraa signal froa reactor protectim systea f99) <<ill activate the cmtrol rod drive systee to insert rods. Scraa functicn cnly. Ik Loss of FN F)ow IF 2-2 Fuel Failure 17 Scraa SF Scraa signal fron reactor protection systea I99) will activate the cmtrol rod drive systea to in cn yo IS Loss Off<its AC 17 Scraa SF Scra~ajg+ frg rea r.n <~Xl P,p', ~~. ~< 5 @tie PP) ')i]$ pctb ate,'t4 icnthl y >+ ',"i .."',;.:gRip'lsystei'$o ihsg'rb kifsY Y r7 s 19 Recirc.Pao Seizure 4mB'silurs 17 Scraa SF Scraa signal $ cacti)r protecticn -sysf6i (99) will activate the ccntrol rod drive systen to insert rods. Scraa function only. 20 Recirc Fail-incr F 2-2 Fuel Failure 17 Scraa SF Scraa signaI fron reactcr protecticn systee 199) will activate the cmtrol C', ~i,: ~ rod drive systea to insert rods. Scran function. cnly. SAFE QUIlNN AHALYSIS Page 247 QO7 ONE ROD DRIUE Prepared: E. zC Dale: ~A/ Reeked: Date: BF)1~~ IRev I) C. L Parker EUE)IT REACT MXEPTABLE SAFE OR OPER RESISTS 'SAFEIY ACTIN FlK SEQ. EUE)IT TIRE STATE Code Title Code Title CNE 2), Startup Idl Recirc If 2-2 Fuel Failure 17 Scraa Scran signal froo reactor protectim systen N) will activate the control rod drive systeo to insert rods. Scran functicn only. 23 F)f Ctrl Fail-Nax If 2-2 Fuel Failure 17 Scraa Scran signal froa reactcr protectim systea N) will activate the control rod drive systeo to insert rods. Scraa function only. 24 Rod Drop Accident lf 3-2 Fuel Failure 17 Scraa Scran signal fron reactm protectim systea N) will activate the cmtrol rod drive systen to insert rods. Scran function mly. ZA Large Break Inside DF 3-2 Fuel Failure 17 Scraa Scrae signal froa reactor protectim systea N) will activate the ccntrol rod drive systea to insert rods. Scraa functim only. 25B intern Brk Inside DF 3-2 Fuel Failure 17 Scraa Scran signal froe reactor protectim systea N) will activate the control rod drive systen to insert mls. Scran fmction cnly. ZC Snail Break Inside DF 3-2 Fuel Failure 17 Scraa SF Scraa signal froo reactcr protecticn systen N) will activate the cmtrol tee t~EFT rons. Scran function only. 27 Pipe Break Outside DF 3-2guggailure 17eScrfk ~~:.< SP ."~ The $ crai4scharee Uolte'est)e P5",'-. ', ','I'.v,c:--::;,.:e 1'.Mlh@~ jhow'de:a;sig 'I tj th acconodate a scraa is not available. 27 'ipe Break Outside DF 3-2 Fuel Failure T)IF05Gitrol Air Header oust be ~ mitcred to provide a signal to the RPS Systen if sufficient pressure is aot aaintained in the header to assure proper yositiming of the control rods, SAFE 9STDOW ANALYSIS $ 07 Page 24 'reoa ed: < GNR9. ROD DRIVE E.C c Checked: SRHSL~&8 (Rev I) MHT REACT MXEPTADLE SAFE DR 5%R RESETS SAFEIY ACTIN FIR SED, DEHT TITLE STATE Code Title Code Title G)OE '27 Pipe Dreak Outside OF 3-2 Fuel Failure 17 Sera<< SF Scraa signal fron re ctor protectim systen 199) <<ill activate the cmtroI rod drive systen to insert rods. Scran functim mly. 30 Overpres Protect OF 2-3 Syst Stress 17 Scraa SF Serac signal froa reactor protectim systen 199) <<ill activate the ccntrol rod drive systen to insert rods. Scran functicn cnlv. 32B Lo<<Reservoir OF 2-2 Fuel Failure 17 Scraa SF Sera<< signa'I fron reactor protectim systen 199) <<ill activate the ccntrol rod drive systen to insert rods. Scran fmctim only. 32E Fire ~ OF 4-4 Fuel Failure 17 Scraa Design of safety features <<st include Appendix R fire cmsideratims. HSD Hcraal Shutdose COEF 2-2 Fuel Failure 17 Scraa SF Nanual scran iif needed) fron reactcr protecticn systen 199) <<ill activate the CRD systea to insert control rods. Scran functim only. 25D Srklnsid~tMbd COEF 3-1 Rad Release 20 RPV Isol Support lPI isolation functicn. 06 Loss FM Heater F 2-2 Fuel Failure 2 Rest Homal N Inserted to extent needed, after r reductTon, to reduce po<<er to initial load line u i see Slj,.370.~~; T. 07 Shutdn Ccol Nal fun D 2-2 FP ureg2 fest:,&nal 'tl,~';T)pÃtgr "s actims tq'inseCt Cmtrab x '4 j+i ~ -',JM.'to fntigatl'po<<er increase. ~~ji ~ a 5 Rod Nithdra<< Error F 2-2 uel Failure 22 Rest Homal )I Operatcr to ch ahern rec for the <<ithdra<<al error and noraal rated ccnditims. 'estore 17 Trip I Recirc Puao COEF 2-3 Syst Stress 22 Rest Hcrnal )I Insert control rods ias needed) to avoid scraa durinn restart of the tripped recirculatim puap. SAFE QST)XN ANALYSIS Page 249 $ o7 Prepared: Date: CMRI. Im OR19E E. er Checked: Date: SRHM&8 (Rev 1) C. H. Parker EIJEHT REACT MXEPTABLE SAFE CR ae SmTS SAFETY ACT)a m SEO. 'IEHT T1TLE STATE Code Title Code Title CODE 18, Trip Both Recirc. F 2-2 Fuel Failure 22 Rest Hcraal H Insert ccntrol rods prier to restart of first recirculation puap. 31 Bundle Errcr OF 2-2 Fuel Failure 22 Rest Horaal H gcn discovery of abnoraal power peaking fcr location of failed fuel), use a aodified rod pattern to restcre ccnditions within n~nal liaits. D9 Rod Hithdraw Error F 2-2 Fuel Failure 23 Li~ Disturb L The CROS inhibits further novenent of ccntrol rcd given block signal froa RBH. 11 Control Rod Reaovl A 2-2 Fuel Failure 23 Lin Disturb L The velocity liniter design is such that adjacent fuel bundle asseablies ~ ust be withdraw prior to ccntrol rod reaoval. 'l Ccntrol Rod Reaovl A 2-2 Fuel Failure 23 Lin Disturb L Refueling interlocks initiated by Reactcr Hanual Control Systeala subsystea of the control rod drive systea) are used. Rcd Blocks aust be operati cnal ~ 2l fed Dry A:cident OF 3-2 Fuel Failure 23 Lia Disturb L Provide control rod sequence ccntrol restrictions to $i~ci aaxiau~ag with. kcidnt 11aRelgk Sg Lzi'hSSr)'jd:-' k Ne'65 2b Fuel Handl ~ fxx: ~."' 'I '! i. >%de $ 'RoII )fmiockS p bri)0th)!.rjfuelpg)zpterfacks syttea'aI e fS withdrawal error during refueling 19 Recirc.Pap Seizure 0 reduce power and initiate shutdown , per procedures. Expected path of action, but is not required. 328 Low Reservoir F 2-1 Rad Release 24 Pow Reduce H Hanually insert control rods to reduce power and steaa flow to aain condensere 0)A Senerator Trip F 2-1 Rad Release 2f Est Pri Cont F Perfcxa isolaticn acticnfs) upon I receiving scraa signal fron the Reactor Protection Systen f99). QUlHN AIIALYS IS JOT SAFE Page ZO Prepared: C Date: ONION RDD DRI'llE E t Checked: .Date: TF)H)S63~ IRev I) C. L Parker BM REACT MXEPTAME SAFE ITI 9%R RESETS SAFETY ACTISl RK SEQ. SENT TITLE STATE Code Title Code Title NK OIC PresReg Fail<lose DF 2-1 Rad Release 26 Est Pri Cmt SF Perfva isolation actimls) uom receivino scraa sional fron the Reactor Protection Systea N). 02 Turbine Trip F 2-1 Rad Release 26 Est Pri Cont SF Perfor ~ isolatim acticnis) upon receiving scran signa) fron the Reactor Protectim Systea N). N Rain Stn Isolation DF 2-1 Rad Release 26 Est Pri Cont SF 'Perfcrn isolation actimis) upon receiving scraa signal fron the Reacts Protection Systea N). Ol Close Dne lSIV F 2-1 Rad Release 26 Est Pri Cont SF Perfora isolation acticnis) upon receiving scraa signal fron the Reactor Protection Systea Igg). 05 Loss of Yacuua DF 2-1 Rad Release 26 Est Pri Cont SF Perfcrn isolation actionis) upon receiving scraa signai fron the Reactor Protection Systea N). 06 Loss F)1 Heater F 2-1 Rad Release 26 Est Pri Cont SF Perforn isolation actimis) upcn receiving scraa signal fron the Reactor Protection Systen Igg). 07 Shutdn Cool IIal fun D 2-1 Rad Release 26 Est Pri Cont SF Perfora isolation actimis) upon receiving scraa signal froa the Reactor Protection Systea N). 08 lnadv Pusp Start DF 2-1 Rad Release 26 Est Pri Cont SF Perfora isolation acticnis) s>gnal fron the Reactor Protection Systen Igg). , 3 Og Rod llithdra~ Erra lF 2-1 Rad 1 L 12A Pres Reg Fail+en DF "-I Rad Re ease 26 Est Pri Cont SF Perfora~iso a~ receivinn scraa signal fron the Reactor Protection Svstee N). 128 lnadv Dpen~ass ~ If 2-1 Rad Release 26 Est Pri Cont SF Perftrn isolatim acticnis) upm receiving scraa signal fron the ' Reactor Protection Systen N). ~ ~ r r ~ ~ ~ 'I ~ SAFE S)IIIXN AHALYSIS GNKL ROD BF)H)SS3~ DRIVE (Rev I) Prepared: Checked: C. Page 251 C, L Ec Parker Iefe: Date: ~ BELIE)(T REACT MCEPTABLE SAFE OR 9%R RERlTS SAFETy ACTIQl RK Kl. BM TITLE- STATE Code Title Code Title GEE Inadvert Open SR(( DF 2-1 Rad Release 26 Est Pri Cont SF Perfcra isolatim action(s) upm receiving scraa sional froe the Reactor Protection Svsten N). Il Loss of F)(FIw, OF 2-1 Rad Release 26 Est Pri Cmt SF Perfmn isolaticn acticn(s) upm receiving scraa signal fron the Reactcr Protectim Systea (99). 15 . Loss Off<ite AC OF 2-1 Rad Release 26 Est Pri Cmt SF Perfm a isolation acticn(s) upon receiving scraa signal free the Reactcr Protecticn Systea N). 19 Recirc.Pop Seiaure DF 2-1 Rad Release 26 Est Pri Cont SF Perfora isolation acticn(s) upon receiving scraa sional fry the Reactcr Protection Systen (99). 20 Ret;irc Fail-Incr F 2-1 Rad Release 26 Est Pri Cont SF Perfcra isolaticn action(s) upon receiving scraa signal fra the Reactcr Protection Systea N). 21 Star tup Idl Recirc OF 2-1 Rad Release 26 Est Pri Cont SF Perfcra isolation action(s) upm receiving scraa signal froe the Reactor Protection Systea (99). R( Ctrl Fail~ OF 2-1 Rad Release 26 Est Pri Cont SF Perfa'I isolatim acticn(s) upm receiving scraa signal froe the Reactcr Protectim Systea (99). 2f Rod Orcp Accident DF 3-1 Rad Release 26 Est Pri Cent SF Perfma isolation acticn(s) upon receiving scru signal frca the Reactcr Pr 'an.M) . 25A Large Break Inside OF 3-1 Rad Release 26 Est Pri Cont SF Perfora isolationgctjm(s).upon; : eh, -,r, erefeieekj eerie/i fhff f'r'ia'hhe "': h 258 Intern Brk Inside 3"I a CPS(. Perfcra isolation actim(s) upm receiving scraa signal froo th >m ystea N). 25C Saall Break Inside OF 3-1 Rad Release 26 Est Pri Cont SF Perfcra isolation action(s) upon receiving scraa signai froe the Reactm'rotecticn Systea N). )UIMN AHALVSIS, Q 47 SAFE Page 22 Prep ved: Oate: GNIKL ROO ORI'lK E. Ec BF)HSB3~ Oecked: nae: ~N'E7 IRev 1) C. )L Parker BM REACT lNKEPTAKE SAFE N 5%R RERLTS SAFETY ACTIN FIK SEQ. BM TITLE STATE Code Title Code Title COBE '25D Brklnsid~tSad 3-1 Rad Release 26 Est Pri Cont SF Parfum isolaticn acticnls) upm receiving scran signal fron the Reactor Protecticn Systen N). 27 Pipe Break Outside 3-1 Rad Release 26 Est Pri Cent SF Perfan isolatim acticnls) upon receiving scraa signal frm the Reactcr Protection Systen N). 30 Overpres Protect EF 2-1 Rad Release 26 Est Pri Cont SF Perfcrn isolation action(s) upon receiving scran signai fra the Reactor Protection Systen N). 32B bw Reservoir BF 2-1 Rad Release 26 Est Pri Cmt SF Perfcrn isolation acticn(s) upm ~ receiring scraa signal free the Reactor Protection Systen N).

0 15 Loss Of flite AC Nl 2-1 Rad Release f

I 27 Est Sec Cont - 'fonctiat; -- Support seccndary ccntainant 259 Brk)nside+mtNad COEF 3-1 Rad Release Sl-Est'.Sec Cant: ". 'upport secmdar~ cmt))inca@. q., ",'fmcticn. -., - ':,'":" i';l;",.'27 ~ I lg 7 I 26 Fuel Hand) Accidnt A 3-1 Rad Release Est Sec Cmt Support secmdary'cmtiinhent" " C J functicn. v'3 Loss Fuel Pool Cl ALL 2-1 Rad Release 27 Est Sec Cont Support secmdary cmtalnnslt functicn. 25A Large Break Inside COEF 3-2 Fuel Failure 31 Stop RodEjec L Housing supps t physically in place to prevent rod ejection lpassive). 258 lntera Brk Inside COEF 3-2 Fuel Failure 31 Sty> RcdEjec Provide passive cmstraint to avoid ccntro rod ejecticn. 1CRO Housing Support). 25C Saal1 Break Inside COEF 3-3 Syst Stress 31 Stop RodEjec L Provide passive cmstraint to avoid cmtrol rod ejection. KRO Housing Support) 21 Rod Orop Accident BF 3-2 Fuel Failure 32 Lia Ins Rate L Control rod velocity Iiaiter liaits (0 rod drop rate. (Passive Functim) SAFE SHJO(SH( AN.YSIS Page ZQ of Prepared: Oate: e ONE TF)HKQ~ R00 ORIOLE (Rev 1) Checked: C. )(. Parker Oate: BE)IT REACT IN(XEPTASLE SAFE OR OPER RESETS SAFETY ACTTOI flK KL BM TTTLE STATE Code Title Code Title GIE '28 Shtdn Froe Outside DF 4-i Local Shut 38 Shutdn Outsd SF SEQSKK A, STEP 2: All rods are inserted autoaatically cn loss of poar to reactcr protectim systea (gg). 28 Shtdn Froa 0utside OF 4-1 Local Shut  % Shutdn Outsd SF %LB', STEP 2: All cmtrol rods are autonatically inserted m scrae signal fry reacta protecticn systen (gg). 28 Shtdn Fron 0utside OF 4-1 Local Bet 2 Shutdn 0utsd A SERBKE C, STEP 2: All rods m autoeatically inserted in groups or clusters due to trip of scran pilot va)ves. 28 Shtdn Froa 0utside GXF 4-1 Local Shut à Cooldn Outsd 8 GS cooling eater supply,nay be provided to supplement the reactcr coolant inventory supplied by the t8 $ 444W M ~ ' WW'4K Aha as()s RClC 01) systee (not required). ~IM ~4 CaCt ~ let;~e~bI]s~~ os) ticn ~g 29 Shutdvn 8/0 Rods OF 4-2 Shut Rods 42 Shutdn )(ORod s indicatim to identify i 329 Earthquake ALL 4-3 Equi F'ail '-"44 Seisaic Shtd; sign'of;safety features~) have.:,: capability to.'provide shutikwn foll(aring a design basis seisaic t. ' 'l '$07 SAFE SWTINN Q(ALYSIS Page ZR 'Prep ved: DIESEL STARTI% AIR E. . SYSTE)( 9RHKN~ (Rev 1) Checkedy C. L Pvker Date: >P'EACT MXEPIAKE SAFE 5%R REQLTS SAFEIY ACTISi F1K STATE Code Title Cade Title CIRCE '13 Inadvert Open %V CDEF 2-2 Fuel Failure 19 Ccrc Cooling SF Provide diesel starting air to diesel generatm systen starting m Iaw reactor water level (Ll) signa)s. This safety action nay cccur, but is not required by this event since ncrnal ac is available. 15 loss Dffdite AC COEF 2-2 Fuel Failure 19 Care Ccoling SF Provide diesel stvting air to diesel generata systen starting m law reactor water level (LI) signals. ?f Rod Drap Accident DF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide diesel starting air to diesel generator systen starting on law reactcr water level (Ll) signals. 25A Large Break Inside CKF 3-2 Fuel Failure 19 Care Cooling SF Provide diesel starting air to diesel (0 generator systen (8?) starting m high drywell pressure signal. 25A Large Break Inside CUEF 3-2 Fuel Failure 19 Care Cooling SF Pravide diesel starting air to diesel gmefatar systen stvt(ng m law reactor water level (Ll) signal. 258 Intern Brk Inside COEF 3-2 Fuel Failur 19 Care Cooling SF Provide diesel starting air to diesel j E <genecntcc ~ten star)i'o Ip ." .".'+e4ctcr. water,4evi ) (J I)'. signal's,? 'v 258 intern Brk Inside (XEF 3-2 Fuel Failur 19 Ccrc Caoling SF h'ovldb d1&l'kvting-'air to d!ese1 .. generatcr systen (8?) starting m 25C Snail Break Inside CDEF 3-2 Fuel Failure 19 Ccrc Cooling SF Provide diesel starting air to diesel generatcr systea startiog on law reactcr water level (Ll) signals. 25C Sea)I Break Inside (XEF 3-2 Fuel Failure 19 Care Caoling SF Provide diesel stvting air to diesel genera'tm'ysten I?) stvting cn high drywell pressure signals. 27 Pipe Break Outside CDEF 3-2 Fuel Failure 19 Care Cooling SF Provide diesel 'starting air to diesel genera'tar systen s'tarting m low reactor water level (Ll) signals. Page Zi5 Preparah DIESEL STARTIIS AIR STSTE)( Checked: 8F)H)883~ (Rev 1) C. )(. Parker BM REACT 'N(A(XPTARE SAFE N SB REM.TS 'AFETT ACTIN RK SE(). BM TITLE STATE Code Title Code Title GEE '3?E Fire Nl M Fuel Failure 19 Core Cooling Design of safety features aust include Appendix R fire colQ derati cns SSD Safety Shutdoe (XEF 3-2 Fuel Failure 19 Ccrc Cooliog SF Support diesel generatcr systea (8?). 15 Loss Off&te AC IXEF 2-4 Cont Stress 30 Cont holing P Provide support systea (8?). to diesel generatm 250 Srklnsid~ttRad CKF 3H Cont Stress 30 Cent Cooling SF Provide diesel starting air to diesel generator systen (82) ~ . 3?E Fire CKEF 44 Cont Candtns 30 Cont Cooling Design of safety features ast include Appendix R fire Nathrabaa gg;<eg SSD Safety Shutdoo CKF M ebs';30.'>Cent hdlxng gP, j'.Support diesgl ~ens'atm systee (8?). 28 Shtdn Froa ()utside ALL 4-1 al'hut-'39. Cooldn Putsd;gC., Pr'mride,diesel 'starting air 'to d(es(l'. ihr systen,(8?) if ncraal ac:,' 'ouer is lost. 3?D Earthquake ALl 4-3 Equip Fail 44 Seisaic Shtd ~ 'esign o~fsa e y features.m capability to provide shutdoun follaring a design hasis seisaic event. SAFE SUINNI NOLYSIS Page 256 of Qo 1 Preps ed: Oate: RAMATIN NNITN1% E C Ec SYSIBI Oecked: bate: IF)H)S&0f8 (Rev I) C. IL Parker BE)fl REACT MXEPTARE SAFE N SB RESILTS SAFEIY ACTIQI fUC SG). BM TITLE 'STATE Code Title Code Title GEE 2$ RNI Orty Accident IF 3-2 Fuel Failure 17 Provides aain stean line high radiatim signal trip to RPSN) fcr 24 Rod Orop Accident IF 3-1 Rad Release 20 Rpg lsol SF Provides nain stean line high radiatim trip signal to react+ protectim systen (gg). 31 Sundle Errcr 2-1 Rad Release 22 Rest kraal )l I)onitcr aain stean line radiatim, nmitcr plant offgas radiaticn levels so that cmtrol rod/pour changes can be sade as necessary to stay within na'nal operating linits. 03 Nain Stn Isolaticn CEEF 2-1 Rad Release 26 Est Pri Cont SF PerfN ~ isolaticn actimls) upon receiving low water level (L3) isolaticn signal free the Prinary Cmtainant systen lbh). This safety actim is plpected to occur) but is not a requireaet fa this event, ince there is no fuel failure. 05 Loss of gacuun C0EF 2-1 Rad Release 26 Est>pra~SF Perfcrn isolaticn acticnis) upon ..receVv'ing Ite:waterlepl IL3) 'solatlm,'sign'al'froeithiQ ., Containeent. sgten.';I+. ~ilhis safety: actlm ls Npecte'd-'to occur) but is,: not a requireeent for this event " since there is oo fuel failure. 13 Inadvert Open 9N CKF 2-1 Rad Release 26 Est Pri Cmt.SF Perfan isolatim actimis) upon receiving low water level IL3) isolaticn signal fron the Priaary Containaet systen N). This safety acticn is expected to occur, but is aot a requireeent fcr this event since there is no fuel failure. Loss of F)l flow NEF 2-1 Rad Release 26 Est Pri Cont SF Perfan isolaticn acticnis) upon receiving Iow water level IL3) isolatim s<gnal froe the Pr>nary Cmtaimnt systen N). This safety actim is <<rpected to occur, but is not a requir %lit for this event since there is no fuel failN e. SAFE SNIIIOII AN.YSIS Page 257 qf Prepared: C. hste: ~el& r7 RMIATISI )OIITN1% E SYSIHI Qjecked: 9ate: SF)H)S63~ (Rev I) C. L Parker EIM REACT NCXP?ASLE SAFE IR PER RERLTS SAFElY ACTIOI FUC SEL acr TITLE STATE Code Title Code Title CKE 15, Loss 8f finite AC Ni 2-1 RadRelease 2b Est Pri Cont SF Perfcrn isolation actimfs) upon receiving Iw water level (Q) isolaticn signal fron the Prinary Cmtainaet systen IM). This safety actim is expected to occur, but is not a requireant fcr this event since there is no fuel failure. 23 FN Ctrl FaiH4x CHIEF 2-1 Rad Release 2b Est Pri Cent SF Per fora isolation actimfs) upm receiving Icw water level IQ) isolaticn signal frca the Prinary Cmtaineent systea lb4). This safety action is expected to occur, but is requireaet fcr this event 'ot a since there is no fuel failure. 2f Rod Erop Accident EF 3-1 Rad Release 2b Est Pri Cmt 0 Perfmn isolaticn actimfs) upm receiving low water level IQ) isolatim signal fry the Prinary Cmtainnent systen IM). 24 Rod Srop Accident lf 3-1 Rad Release 2b Est Pri Cont SF Provides stean line high radiatim signal to RPS N) for stean line drain and saaple line isolaticn. 25A Large Break Inside 0)EF 3-1 Rad Release P+ pegt gp pefiere tsutsttse setilu  !::: fec@ving,,> water level (L3) f, 'solktim".signaI, froe the Prinary '; < Cmtainmt systea>>IM),~, s ' ax a p 25A Large Sreak Inside NEF 3-1 Rad Release gh Eshpej gsut gp purples tsutetiru s:tisulsl'hepal '. Cp s 'h'>> ~use'ug high rtrprett pressure s 'solation stgnDVma-the Priaary Cmtainaent systen N). 258 Intern 8rk Inside CSEF 3-) Rad Release 2b Est Pri Cont SF Perf urn isolatim actimfs) upm recexvxng Iow water level IQ) Isolatim signal froa the Prinary Cmtainaet systen (M). 258 Intern 8rk Inside CCEF 3-1 Rad Release 2b Est Pri Cmt SF Perfcrn xsolatxm actxonls) upon (0 receiving high drywell pressLre isolaticn signal fron the Priary Cmtaimnt systen ffA). RI S)SINN 3+1 SAFE NQLYSIS Page 258 of Prepared: Date: RADIATIQI )MTNM E. Ec SYSTE)l Checked: Date: NH)SS3~ IRev I) C. L Parker BBIT REACT MXEPTARE SAFE N 6%R RESILTS QFHY ACTIN RM', SEL BENT TlllE STATE Code Title Code Title GXE 25C S>>all Break Inside GKF 3-1 Rad Release 26 Est Pri +fera isolaticn actimfs) upm receiving low water level IQ) isolatim signal frm the Prinary Cmtainant systen f6l). 25C Saal1 Break Inside IXEF 3-1 Rad Release 26 Est Pri Perfor>> isulatim actimfs) upm high dryweII presscre 'eceiving isolaticn signa) fro>> the Friary Cmtaimat syste>> IM). 25D Brkinsid~tNad GKF 3-1 Rad Release 26 Est Pri Perfcrn isolation actim(s) upm receiving low water level IL3) isolatim signal froa the Pri>>ary Cmtaimnt syste>> (M). 250 BrklnsidHmtLRad IXEF 3-1 Rad Release 26 Est Pri Perfcra isolatim acticnls) upm receiving high dryweII pressure isolatim signal fro>> the Prinary Cmtaimnt systen IM). 25D BrklnsidHmtNad - COEF 3-1 Rad Rel 26 Est Pri Cont 'tes Prinary CmtainMlt Isolatim a s te>> 6I m high J q: i',zgiatim in>>ain stean i 4,,W /, '..", eject. Off~, Dff~as vent pipe, ;cmtain'. Uenf.'Pi+age;: process liguids, cmtain>>ent.exhaust.", 27 Pipe Break Dutside CDEF 3-1 Rad Release 26 Est Pri Cmt"SF. Perfora isolation actimis) apm: rkeMaggaw water level IL3), isolatim signal"froe-thR Prinary ~~+~~~. f4'p~pp Cmtainant syste>> N). 258 Intern Brk Inside CDEF 3-1 Rad Release 27 Est Sec Cont Estahlish>>ent of secmdary cmtaimnt is required seekpp>>>>4>> 4Q Sirkbg M~'>>4', 26 Fuel Randl Accidnt A 3-1 Rad Release 27:. Est Sec The react~ building ventilatim radiaticn anitcring systee sases high rabat)cn to Init)ate secmdary cmtainaent isolatim signaI sent IQ). to'yste>> $ 07 SHIIT NQLTSIS hye ZB ~/E SAFE Prepared: tata: RAOIATIN )9ilTNIIS E. STSIEN Dtecked: Bate: 3 7 If)H)SS3~ Nev I) \ REACT llCQHABLE SAFE [PER BERLTS SAFETT ACTIN RM,'ade STATE Code Title Title Cm)E Loss Fuel Pool CI kL 3-I hd Release 27 Est Sec Cent SF The reactcr building ventilaticn fadiaticn %Alters sense high radiaticn to initiate secondary contalneent lsolaticn (N 2f Rod Brop Accident 3-5 Pers Ovsrexp 36 Cent Bay Env SF Provides signal cn high radiaticn in ccntrol bay inlet ducts to air conditicning systen l31) to isolate the air supply to the cmtrol bay.. 2 5D Brkinsid~tNad COEF 3-5 Pers Overexp Cant Bay Env SF A signal fra the Process Radiaticn Nani ter in the reactcr building <via systen 6I) cr ccntrol bay air int es ccntrol bay 26 Fuel Handl Accidnt A 3-5 Pers exp 36 Cont Bay Env SF Excessive rNii$ian" Ii'sendedfid reactcr buil'Ling'ekaust and iaaf is nnt ta prtaary swtaiannt'ystsa,'Sr WIt' tQ~-control bay envircneental I isatatira. Hiiiiynliattntia antral bay air inlet initiates a signal --~ ( / / directly to SAC Isysten 3l). 27 Pipe Break Outside CKF 3-5 Pers Overexp 36 Cant Bay Env SF The Radiaticn Nanitming Systen ast provide a signal to the AC Systen I3I) if above ncrnal radiatim levels are detected in the AC inlet ducts to the KL c.< Loss Fuel Pool Cl ALL 3-5 Pers Overexp 36 Cent Bay Env SF ~e<<I.,~ 4~111 Provide excessivezradiatim signal to prinary ccntaimnt systen N) to initiate ccntaineent actions. SAFE SaT1001 A)e.ySIS 260 w7 IBJHKH )9IITDR1% cF E STSJBI Checked: Date: At IRH)S6~8 (Rev I) C. IL Parker EUE)IT REACT MXEPTASLE SAFE N IPB REStLTS SAFETT ACTIN RK SEL EUer TInE STATE Code Title Code Title ME '1A Generate Trip F 2-2 Fuel failure 17 Scraa SF Provide APN neutrm flux trip signaI to reactcr protecticn systea 199) failmfe logic. 018 T/8 Tri~ Fails F 2-2 Fuel Failure 17 Scraa SF Provide APN neutrm flux trip signal to reactcr protecticn systea N) fai I~e logic. 01C PresReg Fail<lose F 2-2 Fuel Failure 0 Scraa SF Provide APN neutrm flux trip signal to reactcr protecticn systea 199) fail~fe logic. 01C PresReg Fail<lose D 2-2 Fuel Failure 17 Scraa SF Provide IN neutrcn flux trip signai to reactcr protecticn systea N) isee section 7.6.2). 02 Turbine Trip F 2-2 Fuel Failure 17 Scraa SF Provide /RE neutrm flux trip signai to reactcr protecticn systea N) failmfe logic. t6 thin Sta Isolatim D 2-2 Fuel Failure 17 Scraa g Provide IN neutrm flux trip signal to reactcr protecticn systea N) (see sectim 7.6.2). 0I Ouse Cue)SIU 2-2 Fuel Failure 17 Scraa SF Provide N%N neutrcn flux trip signal to reactcr protection systea 199) failmfe logic. 05 hss of Vacuua DF 2-2 Fuel Failure Provide NN neutrm flux trip signal or s ea N) faiimfe logic. F 2-2 Fuel Failw Scraa,'"'; 'SF; ";Provide:F81'neutrmVIil Plpwignal ' toreactcr jrotecti'm!sjit'ea,'199)", I r 1 ~ g / ~ 06 bss FN Heater F 2-2 Fuel Failure 17 Scraa V-.W~APN theraal pcaar signal to reactcr pro~tee i~tea499)jiaC g 1 07 Shutdn Ccol )lal fun D 2-2 Fuel Failure 17 Scraa SF Provide IN neutron flux trip signai-(I to reactcr protectim systea N) (see sectim 7.6.2). 1 Sera% emiS of t.:~l SAFE Page 261 Prepared: f lGAHN )NIHRIIS E. ar STSIEN Checked: Date: >P 7 fRHKM~ IRev 1) C. L Parker BM REACT INCCEPTARE SAFE N 5%R RERlTS 'SAFETy ACTIQI RK SEL EVEHT TIIIE STATE Code Title Cede Title CIE M lnadv'uap Start F 2-2 Fuel Failure 17 Scran SF Provide APRN neutrm flux trip signal to reactcr protectim systen 199) failmfe logic. N Inadv Puep Start D 2-2 Fuel Failure 17 Scraa SF Provide IRN neutrm flux trip signal to reactor protecticn systen 199) (see sectim 7.6.2). N Rod lfithdrapp Error DF 2-2 'Fuel Failure 17 Scran SF Provide IRN neutrcn flux trip signal to reactcr protectim systen f99) isee sectim 7.6.2). D9 Rod NitMratd Error DF 2-2 Fuel Failure 17 Scraa SF Provide APRN neutron flux trip signal to reactm protection systen 199) failmfe logic. Pres Reg Fail+en DF 2-2 Fuel Failure 17 Scraa SF Provide IRN neutron flux trip signal to reactcr protection systen 199) isee sectim 7.6.2). 12S Inadv Dpen~ass . D 2-2 Fuel Failure 17 Scraa SF Provide IRN neutrm flux trip signai to reactcr protecticn systen l99) (see sectim 7.6.2). D lnadvert Dpen %V DF 2-2 Fuel Failure 17 Scraa SF Provide IRN neutron flux trip signaI to reactor protection systen 199) (see sectim 7.6.2). 20 Recirc Fail-Incr F 2-2, Fuel Failure 17 SF Provide APRN neutrm flux trip signal tcr protectim systen f99) failmfe 21 Startup ldl Recirc DF 2-2 Fuel Failure I i fail~fe logi I 21 Star tup ldl Recirc DF 2-2 Fuel Failure 17 Serai'Provide IRN neutrm flux trip signal fo >eactcr.protection systen 199) Isee sectim 7.Rll. 9 Ctrl Fail-Nax DF 2-2 Fuel Failure 17 ip provide fpfdl ueutrro flux trip uipvot to reactcr protectim systen 199) failmfe logic. SAFE SATIN ANUSIS Page 2&2 of $ ,07 Prepared; C Date: )8JIRN lQIITNIIS E. SYSTEN Checked: Dte: 4Si7 SRHSQ~ IRev I) C. L Parker EVM REACT INGXPTNKZ SAFE DR IFER RESILTS SAFEPf ACTIN RK SEL EVENT TlTLE STATE Code Title Code Title CXE '24 Rod Drop Accident DF 3-2 Fuel Failure 17 Scrae SF Provide 1RN neutron flux trip signal to reacta protection systen Igg) Isee section 7.&.2). 2$ Rod Drop Accident TF 3-2 Fuel Failur 17 Scran P rovide APRN neutrcn flux trip signal otection systen Igg) failmfe Ingle. P 30 Overpres Protect OF 2-3 Syst Str s': 1) . $ n:an  !'. SF,': Provi0eeN'peutraq.flux trip signal !toPeactj')i'~ticn;cy'sfen W) 25D Drklnside-CantNad IXEF 3-1 Rad Release 20 RPV lsol ..Support RPV isolaticn functim. "; .' g 0& Loss F)I Neater F 2-2 Fuel Failure 22 Rest Nornal APRNs nmitor paar TSrel~mII provide florreferenced alarn si if reached. 07 Shutdn Cool Nalfun D 2-2 Fuel Failure 22 Rest Nornal N Operatcr's actions to ~ange IRN if needed to avoid scfaL

3) Dundle Error TF 2-2 Fuel Failure 22 Rest Hcrnal N IJpcn discovery, use a nodified rod pattern to restore pcxar to ncraal; Utilixe LPRN to sense abneaat pae.

peak logo N Rod Iiithdrau Error F 2-2 Fuel Failure 23 Lie Disturb SF The NIS coepments designated fcr rod block nmitcr function provide a ~ trip signal to the Reacts Nanual Control Subsysten of the CRDS. 29 Shutdn )I/O Rods DF g-2 Shut No Rods l2 Shutdn kRod 1Visplay pwer/neutrm flux m the APRN/IRN/SRNs in the ccntrol roon as the event is identified and as barm shuts doe the reactcr. ~ ~ 'L ~ SAFE SH/HQM A)(ALYSIS Page 2Q +& ~ ~J 0 Prepveh Oate: TRA)8E((S1% 1%3K E. E fMK SYSTE)( Checked: Oate: SF)HKQ~ (Rev 1) C. L Parker SYSTD( 94 BM REACT MXEPTABLE SAFE N IfER RES(LTS SAFETY ACTIN RK SH). BBIT TITLE STATE Code Title Code Title CKE 250 SrklnsidHmtNad GKF >I Rad Release 20 RPY Isol Support RPV isolatim functim. OT ((ain Stn Isolaticn IXEF 2-1 Rad Release 26 Est Pri Cont SF Initiate (M IJJ4+liH) PIYVf7 TIP withdrawal>m low water I level (Q) signal fron prinary cmtaNwlt systen (6h) This safety acticn is expected to occur, but is not a require+et for this event since there is no fuel failure. 05 Loss of Vacuun COEF 2-1 Rad Release 26 Est Pri Cont SF Initiate TIP witMrawal~ Iow iota'evel (Q) signai free prinary ccntainaent systen N). This safety acticn'is expected to occur, but is not a requlfeeent for 'this event 'ince there is no fuel failure. 15 Inadvert Open COEF 2-1 Rad Release 26 Est Pri Cont SF Initiate TIP )( ~ ~ll7 low water withdrawals level (Q) signal frow prinary cmtaineent systen N). This safety SN'k actim is wpected to occur, but is not a requirement fe this event since there is no fuel failure. Loss of F)( Flow (XEF 2-1 Rad Re ease 26 'ate TIP with'drawal on low water jl level (Q) s>g ,l(f'. cmtainaent systen (~jkbj usp ~ N). This safety s m>s ~ Cince'there'is"'no'fuel faalurR; r;,'." r 15 Loss OHRite AC ALL 2-1 Rad (QIQ3F~-Est.lh {mt SF - '.'+~4'lk Initiate TIP witMrawal~m low water, ~ 'I@el-ftQ-iqaaLJro~ainary cmtaineent systee (M). 5i's actim is expected to occur, but is not a requirenent for this event since there is no fuel failure. x &~ W(rl 23 F)( Ctrl Fail~ (XEF 2-1 Rad Release 26 Est Pri Cont SF Initiate TIP withdrawagm low water level (Q) signal fron prinary cmtainaent systen (N. This safety , acticn is expected to occur, but is not a requirenent for this event since there is no fuel failure. y(a J is~(ski~) 9<< 4 "7 SAFE SHJIMN ANALYSIS Page 26l 9~l .C Date: TRA)8JEI(SI% I)ORE E.C Ec PR(IE SYSTE)( Checked: Date: 5$ BF)H)SS3~ (Rev 1) L Parker SYSTE)( gl BE)(T REACT II(A(XEPTABLE SAFE OR lPB RES(ITS SARY ACTI(I( RK SEL EVENT TITLE STATE Code Title Code Title C(NE 2t Rod Drop Accident IF 3-1 Rad Release 26 Est Pri Cmt SF Initiate TIP withdrawai<cn low water level (Q) signal fran priaary ~ cmtalnnent systen (M) ~ ZA luge Break Inside COEF 3-1 Rad Release 26 Est Pri Cent SF Initiate TIP withdrawa)~m low water &Q4ty level (Q) signal froa priaary cmta(A%At systen (bl) ~

a. ~fvtalgg 25A Large Break Inside COEF 3-1 Rad Release 26 Est Pri Cont SF initiate TIP withdrawai>m high dryweII pressure signal frm prinary cmtainannt systen (bl) ~

25B Intern Srk Inside (XEF 3-1 Rad Release 26 Est Pri Cent SF Initiate TIP withdrawal cn low wa'ter (Q) signal froa priaary 'evel ccntainaent systen (bl). 25B Intern Brk Inside C(KF 3-1 Rad Release 26 Est Pri Cmt SF initiate TIP withdrawal m high r <+Q4P dryweII pressure si fran prinary ccntainnent systen (bl). ~ gG QQpy 25C Snail Break Inside COEF 3-1 Rad Release 26 Est Pri Cont SF Initiate TIP withdra+cn low )fater level (Q) sional fran prinary cmtalnRllt systen (bl), 2X Snail Break Inside CDEF 3-1 Rad Release Initiate TIP withdrawal<en high l fran prinary cmtainaent systen (6l) 250 SrklnsidHmtNad 3-1 Rad Release t afoot' I hqitiake'Tip',)withdrew'al~'llep~a

v. ' ~ ~ (I "-level,(Q) signal "fJon piinarjy." d .T )! <<!

ccnhinaent 'N).".;...,;,'(EF 25D Srkinsid~tkRad CDEF 3-1 hdRelease L 26 fs~&f Cont-%- lnitige TIP systen withdrawalpmhigh g g6 QJn drywell pressure s]ghd-fron~inary. ccntainaent systea (bl). 27 Pipe Break Outside COEF 3-1 Rad Release 26 Est Pri Cont 9: Initiate TIP withdrawal~ low water )(. g~~tg7 level (Q) signal fran priaary <g cmtainaent systen (bl). ~ w'~~y~ p (~ (s l~~~) y~~gy (0 SAFE, QJTCHN ANALYSIS Prepared P 2b5 of o ~'. ~/ e RECIKRATIW FUN CMRCL 8F)HHV~ STSTEN (Rev 1) SySTEN 96 Checked: E. C. )L Parker Date: BENT REACT MXEPTA8LE SAFE N ERR RESILTS SAFETy ACTIN RK SGl. MfT TITLE STATE Code Title Code Title GEE

06. Loss F)) Heater F 2-2 Fuel Failure 22 Rest Hcrnal N Run back recirculation flw per SIL 570 instructims - Nanual Ccntrol Node Ibefcre insertin9 control rods).

Sritch to nanual ccntrol node if in autonatic flw cmtrol node. IRi 16 Recirc Fail~ F 2-2 Fuel Failure 22 Rest Homal N Use nanual procedures to restcre ncrnal balanced recirculaticn flw and reactcr pwer ccnditims. 21 Startup Idl Recirc CDEF 2-2 Fuel Failure 22 Rest kraal N Establish initial pwer level prqmrly, and balance both loops after puap start. Rl 19 Recirc.pep Seizure F 2-2 Fuel Failure 24 Pw Reduce N Use ncrnal procedures to reduce and ( gi aaintain flw in active loop. Expected path of operatcr actim, but is not required. 328 Lw Reservoir F 2-1 Rad Release 24 Pw Reduce N Nanuaily reduce ccrc f1 w to reduce pwer. Provide si9nal to recirculaticn systen (68). +01 SAFE SHJlHN ANLYSlS Page 2b6 Prepared: Date; REACTOR PROIECTI(5 E. STSTE)I Checked: late: ~P RH)SQ~ (Rev I) C. L Parker BE)IT REACT MXEPTABLE SAFE N IVER 8EBLTS SAFETy ACTIN RK SEL EVE)IT TTTlE STATE Code Title Code Title CNE OIA 6enerata Trip F 2-2 Fuel Failure 17 Scrae SF Provide scran signal to the cmtrol rod drive (CRD) systen (85) m cain turbine cmtrol valve fast closure trip signa) froa turbine cmtrol systee N7) and naln stean systea (01) signa) indicating turbine first R.l stage pressure > 30L 01A 6eneratcx Trip F 2-2 Fuel Failure 17 Scrae SF Provide scran signal to the cmtrol rcd drive (CRD) systea (85) on neutrm nonitoring systen (92) APR)I seutrcn flux trip signal. 01A 6eneratcr Trip F 2-2 Fuel Failure 17 Scraa SF Provide scraa signal to the trol rcd drive ) ,0 uater systea ((G) high vessel p ess e tctp<~g y,! r 018 T/6 TripW Fails 2-2 Fuel Fa(lurq.g<~ar,"..."; g Provide scras'.sign@~'hi;t4e'yen r'od,h'ive'ICRObiysten ($) m 'n ol turbine ccntrol valve fast closur <<'l~'y trip signal free tur ~ ain stean systen (01) signal indicatinn turbine first stage pressure > 301. OIB T/6 Trip-5'ails F 2-2 Fuel Failure 17 Scraa Provide scraa signal to the cmtrol rcd drive (CRD) systea N) cn neutron acn(tcring systen (92) APR)I aeArm flux trip signal. 01B T/6 Trip-BP Fails F 2-2 Fuel Failure )7 Scraa SF Provide scraa signal to the ccntrol rod drive (CRD) systea N) on feeduater systea (OS) high reactcr vessel pressure trip signal. OIC PresReg Fail<(ose F 2-2 Fuel Failure 17 Scraa SF Provide scru s(anal to the cmtrol rod drive (US) systea (85) cn neutrcn acnitoring systes (92) APR)I neutron flux trip signal. SAFE S)I(MD ANALYSIS $ 07 Page ?f7 Preoared: Date: D~> fl REACT(m PROTECTIN E. E SYSTE)( SF)HSD3~ Checked: Date: ~56 (Rev 1) C. )(. Parker EUEHT REACT NA(XEPTASLE SAFE OR VER RES(LTS SAFETY ACTIN FBI: SEL EUENT TITLE STATE Code Title Code Title (IE '0!C PresReg Fail<lose F 2-2 Fuel Failure 17 Scraa Provide scraa sional to the cmtrol rcd drive (D(9) systen (85) m feeduater systen (03) high reactcr vessel pressure trip sional. 01C PresReg Fail<lose D 2-2 Fuel Failure 17 Scran Provide scraa sinnal to the cmtrol rcd drive (DS) systen (85) m neutrcn ncnitcring systen (92) IR)( neutrcn flux trip signal. 02 Turbine Trip F 2-2 Fuel Failure 17 Scran Provide scraa signal to the cmtrol rcd drive (Dm) systen (85) cn neutron nmitcring systen (92) NN neutrm flux trip signal. 02 Turbine Trip .F 2-2 Fuel Failure 17 Scraa SF Provide sa as signal to the ccntrol .rcd drive (DI) systen (85) cn feeduater systes (03) high reactl vessel pressure trip signal. 02 Turbine Trip F 2-2 Fuel Failure 17 Scru SF Provide scraa signal to the ccntrol rod drive (I)) systen (85) m sionals froa aain stean systen (01) indicating aain turbine stop valves < 90l open and turbine first stage pressure > 30I'. 03 ((ain Sta Is laticn F 2-2 Fuel Failure 17 Scraa Provide scraa signal to the cmtrol rcd drive (DS) systea (K) cn aain stean systen (01) )SIUs < 901 yen trip sional (if in M node). 03 )(ain Sta Isolation . D 2-2 Fuel Fail e 17 Scran SF Provide scraa s(goal to Hie cmTrrT ( , >, rqI drive (DI) systen (85) m (l(D) k'iit@'~i)Cr@g tsyStatn,dp) Cl)lp; ~ 04 Close One )SIU F 2-2 Fuel Fai ur e 17 Scraa Provide scraa signai to &e ccntrol'od drive (I)) systen (85) m n rm 'gaysten-Q?)-4%1( neutrcn flux trip signai. QUEON Qo) I SAFE ANALYSIS Page 2b8 of Prepared: C Date: ~el REACT(R P(ITECTIN . Ec STSIEN Checked: Datee BF)H)S85~ (Rev I) C. N, Parker EYENT REACT 0(ACCEPTAELE SAFE (I 5%R RES(I.TS SAFEIY ACTI(N RK SEQ. EYEHT TITLE STATE Code Title Code Title IE 05 Loss of Yacuun DF 2-2 Fuel Failure 17 Scraa SF Provide scras signal tr the ccntrol rcd drive (CRD) systen (85) on neutrcn ncnlta ing systea (92) APRN neutrcn flux trip signal. 05 Loss of Yacutl DF 2-2 Fuel Failure 17 Scru SF Provide scru sinnal to the ccntrol rcd drive (DI) systen (85) on feeduater systen (05) high reacta vessel pressure trip signal.. 05 Loss of Yacuun DF 2-2 Fuel Failure 17 Scru 6(ven Iotl condenser vacuun signal Hroa systen 02) initiate signal to (I (85) fcr scru (extra, anticipatory sinnal). This safety acticn is expected to occur, hut is not a requirenent for this event. 05 Loss of Yacuua F 2-2 Fuel Failure 17 Scraa SF Provide scraa signal to CRD systen (85) on signals froa nain steaa systen (01) indicating aain turbine stop valves < g0'. open and turbine first stage pressure ) 302. 06 Loss R( Heater F 2-2 Fuel Failure 17 SF Provide scraa signal to the control 'od drive (Dm) systen (85) on neutron nonitcring systen (g2) APRN neutrcn flux trip signal. 06 Loss F)( Heater F 2-2 Fu I Failure 17 Scru SF Provide scraa signal to e c tea+ eee ~ ~, p Dc>:.;...,,.rod drive (N9) systea (85) on

)teiItjontmitpig's)ILtl(a (~

. thernaI.'pttDDer signa) . (utIper(llew jik~di f1 oDD ref ecetlce'd 'sejooiI(t ail'>"; > I 07 Shutdn Coo) Nalfun D 2-2 F SF Provide scru signal to the control r rTvKfR-systen-(SQ neutron aaita ing systen (92) IN( neutron flux trip signal. 08 Inadv Fuq Start F 2-2 Fuel Failure 17 Scru SF Provide scru sianal to the ccntrol rcd drive (Dm) systen (85) on neutron nonitaing systen (g2) APRN neutron flux trip signal. ' SAFE 95(INN A)(ALYSIS P e2b of Prepared: Date: REACT(R MTECT19( E. C Ec STSTE)( RI~Hli(8 (Rev I) Checked: C. L Parker Date: 0 g EVE)(T REACT MXEPTABLE SAFE N SG RES(I.TS 'AFETT ACTIN RK SEL M(T TITLE STATE Code Title Code Title GEE 08 Inadv Pump Start D 2-2 Fuel Failure 17 Scru SF Provide scru signal to the control rod drive (0I) systen (85) cn neutral anitoring systen (92) llw neutrcn flux trip signal. RI N Rod Withdraw Error EF 2-2 Fuel Failure 17 Scru SF Provide scraa signal to the ccntrol rcd drive (ND) systen (85) cn neutron anita ing systen (g2) IN( neutrcn flux trip signai. Og fhxl )fithdraw Error EF 2-2 Fuel Failure 17 Scru SF Provide scru signal to the uatrol rod drive ((I) systen (85) on neutron ncnite ing systea (g2) N5 aeutrcn flux trip signal. 12A Pres Reg Fallen OF 2-2 Fuel Failure 17 Scru SF Provide scru sinnal to the ccntrol rod drive (ND) systen (85) on aain steu systen (Ol) ISIVs < gOI open trip signal (if in RN node). 12A Pres Reg Fail+en OF 2-2 Fuel Failure 17 Scran SF Provide scru signal to the ccntrol rcd drive (ND) systen (85) on neutrcn ncnitorinq systen (92) Iiw aeutrcn flux trip signal. 128 Inadv Open-Bypass F 2-2 Fuel Failure 17 Scru SF Provide scru signal to the ccntrol rod drive (ND) systen (85) on nain steu systen (01) )(SIVs < %C open trip signal (if in RS node). 128 Inadv Open-Bypass 2-2 Fuel Failure 17 Scru SF Provide scru signa o c p  !~ ~ rcd drive (ND) systen (85) on "..nebtjon kjigiilnII,qst+g2) pg ~ c F '-'".-...,::prrrtrk.'llrri'irr'ii;r'riinrh ~ L 2- Failure l7 Scru Provide scru sional to Se'contt67 s D Inadvert Open SRV Fuel SF rcd drive (ND) systen (85) on (L5) trip signal. Inadvert Open 90 EF 2-2 Fuel Failure 17 Scran Provide scraa signal to the ccntrol <<drive (ND) systea (85) m n<<rcn nonitcring systen (g2) IN( neutrcn flux trip sional. SAFE QUINN NQLYSIS Page 270 of $ 07 Prepared: Oate: Z. P REACT(m PROTECTIBl E.. E SYSTEN Checked: Date:. +> 7 8F)H)883~ (Rev I) C. N. Parker BM REACT (IACCEPTAOLE SAFE OR PER RES(LTS SAFElY ACTIN RK SEQ. BG(T TITLE STATE Code Title Code Title (IE , 'l3 lnadvert Open SRV CF 2-2 Fuel Failure 17 Scru NSF Nanual scran (per procedures) nay be g.l ) required if suppressim pool temperature becones too high. Ii Loss of FN Flow 2-2 Fuel Failure 17 Scraa SF Provide scran sinnal to the central rod,drive (DI) systea (85) cn feedwater systen (03) low water level (L3) trip signal. 15 Loss Offdite AC OF 2-2 Fuel Failure 17 Scran SF Provide scru signal to the cmtrol rod drive ((I) systen (85) cn feedwater systen (03) low water level (L3) trip signal. 15 Loss Off-Site AC F 2-2 Fuel Failure 17 Scraa SF Provide scran signal to the ccntrol rod drive (CRO) systen (85) m aain turbine control valve fast closure trip signal fron turbine ccntrol systee (47) and nain stean systen (01) signal indicating turbine first stage pressure > ~XI. IS Loss Off-Site AC OF 2-2 Fuel Failure 17 Scran SF Provide scraa sinnal to the control rod drive (DS) systen (85) m loss of noraal ac power - failsafe design. 19 Recirc.pop Seizure OF 2-2 Fuel Failure 17 Scran NSF Initiate nanual saran (if necessary) to shutdown plant. RJ Recirc Fail-Incr F 2- SF Provide scraa sional to the control rod drive (CRO) systen (85) on neutrcn nonitcring systen (92) APRN 21 Star tup Idl Recirc OF 2 ue Qi4W IX ISA'a '~~",: SF,c; 8'ovldd, g'ka,'signaI tII',)he".cmtggJ ', rod drive (CRO) systen (85) on neutrm ncnitcring systen (92) N%N signah 21 Startup Idl Re irc OF 2-2 Fuel Failure 17 Scraa SF Provide scraa signal to the cmtrol rod drive ((I) systen (85) m neutron ncnitoring systen (92) IRN neutrcn flux trip sional. SAFE SNJHXN AMALYSIS 27) of po7 P Prepved: Date: REACTOR SYS)E)I PROTECTIN BF)H)SS3&8 (Rev 1) Checked: E. C. Ec L Parker Date: f fP BM fMT MXEPTABLE SAFE 8( (PER RES(LTS SAFETY ACTIN FIK SED. M)IT TIRE STAlE Cade Title Cade Title CKE '23 F)I Ctrl Fail-)Iax F 2-2 Fuel Failure 17 Scran 'F Provide rad drive scru signal to the central (CRD) systen (85) cn sionals fran nain stean systen (OI) indicating aain turbine stop valves ( 901 open and turbine first stage pressure ) RL 23 FM Ctrl Fail~ DF 2-2 Fuel Failure 17 Scran SF Provide scraa signal to the control rcd drive (CRD) systea (85) cn feedwater systen (03) high reactor vessel pressure trip signal. 23 F)1 Ctrl Fail~ DF 2-2 Fuel Failure 17 Scran SF Provide scraa signal to the ccntrol rad drive (CRD) syste: (85) on feedwater systea (03) low water level IL3) trip signal. 23 F)) Ctrl Fail~ DF 2-2 Fuel Failure 17 Scraa SF Provide scraa signal to the ccntrol rad drive (CRD) systes (85) on neutrcn anitaring systen (g2).APR)I neutron flux trip s(gnal. 24 Rod Drop A:cident DF 3-2 Fuel Failure 17 Scran SF Provide scraa signal to the control rod drive (CRD) systen (85) on neutrcn nanitcring systen (92) 1R)I neutron flux trip signal. 2k Rad Droo Accident EF 3-: Fuel Failure 17 Scraa SF Provide scraa signal to the ccntrol rad drive ID%) systen (85) on aeutrcn flux trip signal. pl- f x~~xi p~~~ o ~ 0'cpi$ 1 p~ ~ j 2i, Rod Drco Accident DF -3 I alitiA,'.)~'~Sc'raw ': ."', SF .'-;Provl~e scran'sig~~(;~h <i"q . If' ~x':dcave'(CRD); 'syste'a IN acetral j a'n~radi'gati line high radiatic" trip signal. ZA Large Break inside DF 3-2 Fuel Failure 17 Scran SF Provide saran signal to the central rod drive (CRD) systen (85) on feedwater svsten (03) ]or water level (L3) trip sional. SAFE SWTMN N(ALYSIS 'Pqe 2 he~'~ 'Prepared: %C Date: 'Lr REACTOR PROTECTION E. crt SYSTEN Checked: Date: NHKN~ (Rev I) C. L Parker EVEHT MXEPTAKE 'E REACT SAFE OR  %$ RESH.TS SAFETY ACTIC)( RK SE(). EVENT TITLE STATE Me Title Code Title Nt8iE ZA Large Break Inside DF 3-2 Fuel Failure 17 Scran SF Provide scran sinnal to the ccntrol rcd drive (CRD) systen (85) on prinary containnent systen (64) high dryweII pressure trip signal. 258 Intern Brk Inside DF 3-2 Fuel Failure 17 Scran SF Provide scran signal to the ccntrol rod drive (fX9) systen (85) cn feedwater systen (03) low water level (L3) trip signal. 2R Intern Evk Inside DF 3-2 Fuel Failure 17 Scran SF Provide scras sinnal to the central rcd drive (DE} systen (85) cn priaary ccntainnent systen (64) high drywell pressure trip signal. 0 ZC SIall Break Inside DF 3-2 Fuel Failure 17 Scraa SF Provide scran signal to the control rcd drive (CRD) systen N) on feedwater systen (03) low water level (L3) trip signal. ZC Snail Break Inside DF 3-2 Fuel Failure 17 Scran Provide scru sinnal to t e ccntrol ) on (6$ ) high I. 27 Pipe Break Outside DF 3-2 Fuel Failure 17 Scraa SF Provide scraa signal to t ccntrol feedwater systei (03)Iow water level (L3) trip signal.'F 27 Pipe Break Outside DF 3-2 Fuel Failure 17 Scraa Provide scraa s(goal to the control rod drive (0 9) systen N) on high Scran Discharge Volune signal froa C(I systen N). 27 Pipe Break Outside DF 3-2 Fuel Failure 17 %ran SF Provide scran signal to the control rod drive (08) systen N) cn low CRD Control Air Header pressure signal froa CRD systen (85). J 27 Pipe Break Outside F 3-2 Fuel Failure 17 Scraa SF Provide scraa sinnal to'the ccntrol rod drive (05) systen (85) cn aain st~ sy t~ (0)) )SIVs < g0T q e trip signal (if in fdN node). SAFE SHJIHN ANALYSIS Page 273 of 301 Frepared Bate; REACTN P(ITECTIN ~E. SYSIE)( BRH)SS3448 (Rev 1) Checked: ~ C. 1(. Parker er Bate; >3 7 SYSTE)( 99 BM REACT INCCEPTAKE SAFE OR 9%R RES(LTS SAFETY ACT1O(( RK SEO, BENT TITLE STATE Code Title Code Title COOE 30 Overpres Protect OF 2-3 Syst Stress 17 Scraa SF Provide scraa sinnal to the ccntrol rcd drive (01) systea (85) on neutron mitcring svstea (92) N5 neutron flux trip sional. ~A 'verpres Frotect OF 2-3 Syst Stress 17 Scraa SF Provide scraa sional to the ccntrol rod drive (05) systea (85) on feedwater systea (03) hinh reacts vessel pressure trip sinnal. voir 2-2 Fuel Failure lens. 328 Low Reser CF 17 Scraa SF Provide scraa signal to the ccntrol rcd drive (CRB) s)=-tea (85) cn condensate systea (02) 1 ou condenser vacuua sinnal (extra backup sional). This safety action nay occur, but is not reouired by this event. 328 Lw Reser voir F 2-1 Rad Release 17 Rraa %F Initiate nanual scraa if noraal rcd inserticn is not sufficient. Lcw Reservoir F 2=" Fue! Failure 17 Scraa SF Provide scraa signal to CRO systea (85) on sinnals frca aain stean s stea (01) indicatin aain turbine c'est derat stop valves <M cpen ( turbine first stage pressure >301. is safetv required bine trip v dgcur gk a; P~ Fire EF 4-4 Fuel Failure 17 Scraa Oesign of safety featur aust HS0'oraal Shutdaal COEF ."-2 Fuel Failure 17 Scraa le )(anual scraa (if needed) before aain stem systea (01) isolation, 12A Pres Reg Fail-Open F 2-1 Rad Release 20 W lsol SF Provide N()(ode sianal to systea 64 for lou stean line pressure isolaticn interlock.. 1 ~ 128 lnadv Open-Bypass F 2-3 Syst Stress 20 RPV lsol SF Provide RN )@de signal to systea 64 for lm stean line pressure isolatim interlock. SAFE SNTMN AN.YS IS Pane 274 $ 01 Prepared: C. Oate: ~'z REACTCR PROTECTIN E. SYSTE)( Checked: bate: IF)(~<48 (Re; I) C. L Parker BSIT REACT IMXEPTA8LE SAFE R !PER FES(LTS SAFETY @TIN RK SEO. Ev'W TITLE STATE Code Title Code Title COOE Inadvert Open %V 2-1 Rad Release 20 RFV Isol SF Provide RN Me sional to systea 64 fcr Iow stean line presswe isolation interlock. 24 Rod Oroo accident OF 3-I Rad Release 20 RPV lsol SF Provide radiat:on nonitcring systen (g0) aain stean line hinh radiaticn trip signal to Prinary Containaent systen (64) ~ 25A Large Breal Inside iia-l Release 2D IPI Idol SF Provide R(N ((ode sigral to systen 64 fcr Iow stean line pressure isolation interlock. This safety action aay occur, but is riot renuired. ZC Sea)I freak Ins)de PEF '.-1 Rad Release 20 RPV Isol SF Provide R(N )(ode signal to systea 64 4~~wlq fa Ines stean line presswe isolaticn interlock. Ihis safety action nay occur, but is not required. 27 Pipe Ereak Outside Rad Release 20 RFV Isol SF Provide QN )bde sional to systea 64 fcr low stean line pressure isolation interlock. This safetv action nay occur, but is not required. ll Control Rcd Rail A 2-2 Fuel Failure 23 Lia Distwb )(SF Reactor node switch is used for 'REFILL'ode. OIA Generate Trip F 2~ Fuel Fa uce SF Send signal to open rec r culation puap 'ao,br ea 68) cn n b r al e fast s ca b e control y en ) aain ste systeo (01) signal indicatirig > turbine first stane 0(8 Tl8 Trip-IP Fails F e Pou Reduce SF Send signal to open recirculation puap aotcr breakers (systen 68) on aain turbine ccntrol valve fast closure signal free turbine control systen (47) and 'aain stean systea (01) sional indicating >'I turbine first stage pre:sure. SAFE SWTNN ANALYSIS Pane 275 of Prepa ed: C REACTCR PROTECTIBl E. SYSTEM Checked: SFN-0SS3~ (Rev 1) C. )L Parker E(IENT REACT MXEPTASLE SAFE (8 Sent REBLTS 'AFETY ACTIC(( FlK SEQ. BENT TITLE STATE Code Title Code Title COOE 02 Turbine Trip F 2-2 Fuel Failure 24 Pcs Raiuce SF Send signai to open recirculaticn pwp actor breakers (svstea N) on signals froa aain stean systen (01) indicating nain turbine stop valves < 90X open and >30X turbine first stage pressure. 05 Loss of Vacuua F 2-2 Fuel Failure 24 W Reduce SF Send sianal to open recirculation owp outcr breakers (systeo f()) cn signals froa aain stean systen (01) indicating cain turbine stop valves < g0X open and >30X turbine first stage pressure. 14 Loss of Fi( Flou CF 2-2 Fuel Failure 24 Pa< Reduce Provide RPS ATU pceeer to feebcater systen (03) Iou water level (L2) instrwentaticn for trip of recirculaticn pwps. This safety action is expected to o:cur, but is sot a requireaent for this event, 23 F)( Ctrl F il+ax F 2-2 Fuel Failure 24 Pa'educe SF Send signal to open recirculation pusp aotor breakers (systen N) on signals froa nain stean systen (01) irAicating oain tur stop valves < >AX tur 'ne first stage pressures k 0verpres fvolect Syst Stress 24 Pou R vl t feeduater t (0 N, a tor ressure n rua a ion fcr trio recirculation pumps. Los Reservoir -2 Fuel Failure 24 Por Reduce signal to coen recirc. puap aotcr breakers (systen 68) on signals fron systen I indicat(ng aain turb. stop valves <AX teen L >~AX turb. first stage pressure. This safety action nay occur, but is requ(red only if a Ic 'vacuw turb. trip occv 5 SAFE S(ITD(N ANALYSIS Paae 276 Prepared: C Date: REACT% PROTECTIN E. E SYSTE)( Checked: Date: IAH663~ IRev 1) C. L Parker. 886 REACT INCCEPTAKE SAFE OR OPER RES(US SAFETY ACT@[ RK SE(). BE)(T T1TLE STATE Code Title Code Title CODE 03 )(ain Sta lsolaticr CDEF 2-1 Rad Release 26 Est Pri Cont SF Provide low water level (Q) signal frow feedwater systen (03) to priaary cata(nant systea IM) for initiation of L3 isolaticns. This safety actin is expected to occur, hut is not a requireeent fcr this event since there is no fuel failure. 05 Loss of Yacuun CDEF 2-1 Rad Release 26 Est Pri Co t SF Provide low water level (Q) signal frow feedwater systeu (03) to priaary cantainnent systea (ba) for initiation of Q isolaticns. This safety action is expected to occur, but is not a requireaent for this event since there is no fuel failure. 13 Inadvert Open SRV CDEF 2-1 Rad Release 26 Est Pri Cont SF Provide low water level (Q) signal froa feedwater systen (03) to priaary cmtainaent systsw (ba) for initiation of L3 isolations. This safety action is exoe:ted to occur, hut is not a ~equireoent for this event since there is no fuel failure. )5 Loss of F)( Flow CDEF 2-1 Rad Release 26 cEst fri Cent SF Provide low water le el IQ) sional syst ( 03) to priaary cNtainaent systea I ) for initi i Dj 's 1 This fe . ec ed 'to occur) t for this ent t e is no fuel failure. 15 Loss OH-Site AC 2- ele 26 Est Pri 2',! SF Provide 1 L3) signal eedwater sy<ten (03) to priaary containaent systen (M) fm initiation of Q )so) ations. This safety action is expected to occur, hut is not a requireaent for this event since there is no fuel failure. 23 FM Ctrl Fail~ CDEF 2-1 Rad Release 26 Est Pri Cont SF Provide low water level IQ) signal frow feedwater systen (03) to prinary containaent systen (6() for initiation of.L: >solaticns. This safety acticn is expected to occur, hut is not a reouirene.t for this event since there is no fuel failure. ~ ~ QFE SHUTDW A((ALYSIS REACT0R PROTECTIBI Prepved; Page 277 E. of aisle: ~/ SYSTE)( Checked: Date: SRHM~ (Rev I) C. )(. Parker BE)(T REACT IMXEPTABLE SAFE 0R SB RESW.TS 'AFHY ACTllf( RK KL SENT TITLE STATE Code Title Code'itle GXE 24 Rod Dry Accident . DF 3-1 Rad Release 26 Est Pri Cent SF Provide lo<<<<ater level (L3) sional frca feed<<ater systen (03) to priaary cmtainaent systen N) for initiation of L3 isolaticns. 24 Rod Drop Accident DF 3-1 Rad Release 26 Est Pri Cont SF Provide radiaticn aoni toring systen (90) nain stean line hioh radiaticn trip sional to Priaary Ccntainnent systen (64). 25A Laroe Break Inside CDEF 3-1 Rad Release 26 Est Pri Let F Provide Io<<<<ater level (L3) sional froa feed<<ater systea (03) to priaary containaent spten (64) fcr Initiatim of L3 Isolaticns. 25A Laroe Break Inside CDEF 3-1 Rad Release 26 Est Pri Cont SF Provide high dry<<el I pressure signal froa priaary ccntainsent systen (64) to priaary containaent systea (64) fcr initiation of high dry<<ell pressure isolaticns. Intern Brk Inside CDEF 3-1 "Rad Release 26 Est Pri Cont SF Provide lo<< <<ater level (L3) signal frcn feed<<ater systea (03) to prinary ccntainaent systen (64) for initiation of L3 isolations. 25B Intern Pc'k Inside CDEF 3-1 Rad Release 26 Est Pri ry<<e 1 pr sur e signal frco prinary ccntainnent ys tea (64) I At' tea N) hi g)'r 11 2Z Snail Break Inside ease 26 Est Pri Cont SF Provide Io<<uter level (L ) signal fran f to priaary containaent systea (64) for initiation of L3 isolations. 2X Saall Break Inside CGEF 3-1 Rad Release 26 Est Pri Cont O'rovide high dry<<eII pressure signal froa priaary ccntainaent systea N) to priaary ccntainnent systen (64) for initiaticn of high dry<<eII pressure isolaticns. 1 ~ ~ SAFE SHJWi( ANALYSIS Page 278 Prepared; C Date: zc REACM PROIECTIN E. Ec SYSTE)( Checked: Date: BFN~<48 (Rev I) C. )(. Parker PIE)(T REACT MXEPTABLE SAFE OR IM RESKTS QFETY ACTIO)( RK SEO, 886 TITLE STATE Code Title Code Title CNE '50 BrklnsidH'mtbRad COEF 3-1 Rad Release 26 Est Pri Cont SF Provide low water level (Q) signai frow feedwater systen IN) to prinary cmtainaent systen (64) fe initiaticn of L3 isolations. ZD Brklnsid~tNad CKF 3-I Rad Release 26 Est Pri Cmt SF Provide high drywel I pressure signal froa priaary cmtainaent systen N) to priaary ccntainant systen N) fcr initiaticn of high drywell pressure isola'tions. 2o RPY Xsol 27 Pipe Break Outside COEF 3-1 Rat( Release 46-E~ef-Sevk SF Provide low water level (Q) signal \ $ 41~3 frow feedwater systen (03) to prinary /tx ~~/g1 yG ccntainxent system (6() fcr initiatim of 9xso)atims. ~'+ 03 )(ain Sta Isolation COEF 3-5 Pers Overexp 36 Cont Bay Env SF Provide low water level (L3) signai frow feedwater systea (03) to priaary cont. systea N) fcr initiatim of ccntrol bay isolatim. This safety acticn is expected to occur, but is aot a requireaent for this event since there is no fuel failure. 05 ~, Loss of Yacuua CKF 3-5 Pers Overexp 36 Con ovide low water level (Q signai froa feedwat $3I t pri nary st inttla im of ba i . This ety ii c s 't occur x tis aot a requirenent for this t since there is no fu 13 Inadvert Open SRV CKF 3-5 Pers Over y v SF Provide lcw water level (L3) signal froa feedwater systen lN) to priaary cmt. systen N) for initiatim of 'ccntrol bay isolatim. This safety actim is expected to xcur, but is sot a requirexent for this event since there is no fuel failure. Ik Loss of FR Flow COEF 3-5 Pers Overexp 36 Cont Bay Env SF Provide low water level (Q) signai fron feedwater systen (N) to priaary cont. systen N) for initiaticn of cmtrol bay isolation. Ihis safety action is expected to occur, but is not a requireaent fcr this event since there is no fuel failure.' SH)TMN 3(x1 SAFE ANALYSIS Pf epared( lhte: M~zl REACTN PROTECTIN ~ SYSTE)( Checked: Date: BF)HK63~ (Rev I) C. )L.Parker BENT REACT NACCEPQRE SAFE OR HER IESU.TS 'AFETY ACTIN RK SE(). BENT TITLE STATE Code Title Code Title CODE '5 Loss Off<ite AC Nl 3-5 Pers Overexp 36 Cont Bay Env SF Provide low water level (Q) signal frow feedwater systen (03) to prisary cont. systen N) for initiatim of ccntrol bay isolaticn. %is safety actim is expected to occur, but is not a reguireaent for this event since there is no fuel failure. 23 FN Ctrl Fail~ CDEF 3-5'ers Overexp 36 Cmt Bay Env SF Provide Iw water level (L3) signal froe feedwater systen (03) to prinary cent., systen (M) for initiaticn of control bay isolation. This safety actim is expected to occur, but is not a requirenent for this event since there is no fuel failure. 24 Rod Drop Accident DF 3-5 Pers Dverexp 2 Cont Bay Env SF Provide Iow water level(Q) signaI frow feedwater systen (031 to prisary cmtaiwlt systen N) for >nltiat>m of ccn asol at 1m. 25A Large Preak inside CGEF 3-5 Per Bay Env SF Provide low water level g. signal fr s en 4) to prinary t (6lh~ f i ti bury i latim. 25B Intern Brk Inside P exp Cont Bay Env SF Provide low water level (Q) ignaI fron en (03) to pnaary ccntainaent systen'6l) fw initiation of cmtrol bay isolaticn. ZC Snail L.eak Inside COEF 3-5 Pers Overexp 36 Cent hy Env SF Provide Iow water level (Q) signal froa feedwater systen (03) to prinary cmtainnent systen (6() for initiation of control bay isolation. 250 Brklnsid~tNad CBEF 3-5 Pers Overexp 36 Cmt Bay Em SF Provide low water level(Q) signal frow feedwater systen (031 to priary cmtainaent systen (6() fcr initiation of ccntrcl bay isolation. 27 Pipe Break Gutside CKF 3-5 Pers Overexp 36 Cent Bay Env SF Provide low water level (L3) siyal fron feedwater systea (03) to prinary'mtalnaent systen (6l) for initiaticn of ccntrol bay isolatim. SAFE S)I/)TON ANN.YSIS P 280 Prepared: REACTCR PROTECTION E. SYSIE)I Checked; BRH)S83~ (Rev I) BM I SE9. SENT TITLE REACT IRR lNKEPTABLE RESILTS STATE Code Title QFETY ACT]SI Code Title QFE flK ONE Shtdn Fron Outside OF l-I Local Shut 38 Shutdn 0utsd II 5ERKMX A, S)P I: )Ianually open bus AS circuit breakers to initiate CRD 185) rod insertim and )SIV clue signal thrmgh systen M. 28 Shtdn Froe Outside OF l-I Local Shut 38 Shutdn Outsd )I SEQKMX Bs SlP I: Open power supply brealers fm RPS ~ uter generator sets at l80v outcr ccntrol center to initiate CRO N) rod insertim and ISIY closure signai through systen H. 28 Shtdn Froe Outside OF. 4-1 Local Shut 38 Shutdn Outsd )I 5BNIZ c, slP I: gcifceit f~G ~)ters to the scraa pilot valves at RFS bus. AI14r@N>an't be open 'to assure insertim of all ccntrol rcds. 29 Shutdn )I/O Rods OF l-2 Shut No Rods l2 Shutdn)loRod Provide RPS ATU power to feedwater systea 103) Iow water level (L2) instrueentatim fcr trip of recirculaticn puaps. 29 Shutdm )I/O Rods OF l-2 Shut No Rods l2 Shutdn )IoRod Provide RPS All) power to feedwater systen H6) high reactcr pressure instrueentaticn fcr trip of recirculatim puaps,

33) Earthquake ki l-3 Equip Fail 4 Seisaic Shtd Oesign of safety features est have capability to provide shutdoo following a design basis seisaic event.

SAFE S)STION ANALYSIS CRA)KS M H)ISTS IRH)SS3~ IRev 1) Reeked: ~f:~l Page 2f)I E. C. C. L Pvker of go1 Date: P P'1 SYSTE)I 111 BE)IT N REACT IIACCEPTARE 'AFE 'AFETY ACT1$ ERR KkLTS RK SEL 6E)IT TTTLE STATE Code Title Code Title GXE '.32C Tornado Ni I-3 Equip Fail 0 Struct 1nteg L Reactcr building crane is to be tied dao under tcrnado ccnditicns. )lith this provisicn, the crane 1st eithstand full design loads (passive). , ~ (0 ~ V ~ I ~ ~ ~ SAFE SHIM'NALYSIS Page 28? of2bk~ Prepared: Date: ELECTRICAL E. er Ihcked: Date: SF)H)S63~ (Rev 1) C. L Parker t~ BM REACT MXEPTAOLE SAFE OR IPER RESETS SAFETY ACTIOaf RK SEQ. BM TIRE STATE Code Title Code Title CEE finite '15 Loss Of AC COEF 2-2 Fuel Failure 19 Ccrc Cooling SF . OC cmgol pouer is needed. ER 250v 8"~15 d fordi~i g %antral, breaker control for laud shedding and breaker aperaticn. XE Fire Nl 49 Fuel Failure 19 Ccrc Cooling Design of safety features aust include Appendix R fire ccnsidlratlcns. ISO ISO Shutdse COEF 2-2 Fuel Failure 19 Ccrc Cooling Support shutdoe ccrc cooling functicns. NSD Noraal Shutdao COEF 2-2 Fuel Failure 19 Care Ccoling Suppart shutdoe care cooling functi ons. SSD Safety Shutdoe COEF 5-2 Fuel Failure 19 Core Cooling Suppm t shutdoe ccrc cooling functicns. 14 loss of F)f Flan t  % 2-2 Fuel Failure 24 Pcs Reduce Open recirculation lU6 drive outcr breakers l?00, electrical systen) cn feeduater systen (N) Iau uater level iL?) signal. This safety acticn is expected to occur, but is nat a requireaent for this event. 30 Overpres Protect COEF 2-5 S t Stress 24 Pre Red breakers 1200, electrical systen) an feeduater s 1 ) igh 15 Loss Off-Site AC Nl ";-4 lin .. c 1 s n ed. gG and I dc far d>esel generatcr W lcantrol, brealer central for load >ng rea er operation. 3?E Fire COEF 4-6 Cont Candtns 30 Cant Cooling Design of safety features aust include Appendix R fire consideratians. ISD ISO Shutdoo CKF 2H Cant Stress 30 Cont Cooling Support shutdown cmtainaent cooling , ~ functi cns. t ~ ~ ~ ~ ~ ~ SAFE SM/IHN ANALYS)S Page 2S3 >07 Prep ved: hte: ~+I/+ ELECTR)CAL E Dmked: Date! OF)H)SS3~ (Rev 1) C. IL Parker OBIT REACT UQCCEPIAKE SFE DR CPER RESETS SAFETY ACTlQI FBC SEQ. 6EHT TITLE STATE Code Title Code Title CODE )8) kraal Shutdoo CDEF M Cont Stress 30 Cont Cooling Support shutdaa cmtaineent cooling functicns. SSD Safety Shutdoo COEF 3H Cont Stress 30 Cmt Cooling Support shutdcao cmtainaent cooling functims. 2S Shtdn Froe Outside Ni 4-l Local Shut 3g Cooldn Ou'tsd )l hckup Cmtrol Center - Essential services are emitcred froa the back~ ccntrol panel l25-32). Also, actual cperatim is carried mt fry the applicable shutdown cr RNN boards. 29 Shutdn M/0 Rods EF 4-2 Shut Ho Rods 42 Shutdn NoRod Open recirculatim NS drive eotcr breakers l200) cn feeduater systen (03) low eater level (L2) signal. 29 Shutd~ N/0 Rods OF 4-2 Shut No Rods 42 Shutdn HoRod Open recirculation )V6 drive notor breakers l200, electrical systea) m feeduater systee (03) high reactor pressure signai. 32D Earthquake ALL 4-3 Equip Fail 44 Seisaic Shtd Design of safety features est have capability to provide shutdoe folloring a design basis seisaic event. ~TSiS P~ 284 ~ 307 ~ EQttggTT~ N6 'repg ~~ ed; SECNJTy Syph ate: E. Iev l} Checked: hte: BM N RQCT ~faitE QFE BBT TlTtE STATE Code Title ~ ~max pgg Code Title CNE 5 Shtdn Froa Outside AU. 4-1 Local Shut 3V Ccoldn Outsd li Rypart systea required to cardinate the vari as activities to be annal ly perforeed outside the control roon. p ' ~ SAFE Sema Ae.YSIS Qc7 P e 285 Prpv~: H)ILDMS AHD STRXllRES E. E Checked: BF)HM~ IRev 1) C. )L Parker GM REACT MXEPTABLE SAFE CR , 6%R REBLTS 'AFETY ACTIN FIK SH). DM TITLE STATE Code Title Code Title CODE 250 Brklnsi~tiRad CDEF 3-1 Rad Release 27 Est Sec Cont L Secondary ccntainaent leakage rate criteria ast be aaintained by the reactor building. Personnel air locks <<st be operable. 26 Fuel Handl Accidnt A 3-1 Rad Release 27 Est Sec Cont L Seccndvy ccntainnent leakage rate criteria aust be aaintained by the reactcr building. Loss Fuel Pool CI Nl 3-1 Rad Release 27 Est Sec Cont L Seccndary ccntainaent leakage rate criteria aust be aaintained by the reactor building. 32A Flood Nl 4-3 Equip Fail 40 Struct lnteg )I Reactcr building uater-tight access doors/flood gates should be closed uhen the lake level is at 558 feet and rising, ALL 4-3 Equip Fail 40 Struct lnteg )I Rahcaste building uater-tight access dccrs should be closed ken the lake level is at 558 feet and rising. Tcrnado Nl 4-3 Equip Fail 40 Struct Integ L Pending IE Bulletin 80-11 nods., selected reactcr building interior doors are blocked en naming. Reactor building exteri concrete nails and steel C r f ths an I s 32C Tcrnado 4-3 Equip Fail 40 Struct Integ L Reactor building blonat panels cr reac cr building siding and internal conpartant nails Ipassive). 3X Tornado NL 4-3 Equip Fail . 40 Struct Integ L lhder tcrnado design loads, upper secticn of stack aust fail befcre the )ocr secticn reaches ultinate load capacity (passive). SAFE Seed ANALTS)S NILD1%S NiD STRXTNES NHKQ~ (Rev 1) Checked: Page 286 .Ec C. )L Parker oy hte: ~ Its Y~>>P1 PSIT REACT SAXEHARE SAFE N CPER RESLLTS 'SAFETy AC119l RK SEO. BM TITLE STATE Code Title Code Title CNE 32C Tcrnado ALL 4-5 Equip Fail 40 Struct Integ L Mater tanks ast withstand full design loads Ipassive). 3Z Tcrnado Nl 4-3 Equip Fail 40 Struct lnteg L Radwaste building aust confcrn to Class 1 Iseiseic) design criteria under tornado loading conditicns Ipassive). 32D Earthquake ALL 4-5 Equip Fail 40 Struct lnteg 'esign of safety features ast have capability to provide shutdown following a design basis seisaic g~ events SAPE SHVZDORK AMLYSIS t Prh pared: F,F page 287 Date: of / 07 BFNWSG3-048 (Rev 1) Checked: C. M. Parker Date: ~>t 7 aaaassaaaassstaaaaaaaaaaaaatsaaaaaassssaaaaaaaassssssaaaaaaaaaatsaaaaassaaaaaastaaaaaaa PAGES 288 - 291 ARE INTENTIONALLY OMITTED hMLTSIS M& ShFE SHUTIX%8 Prepared: Checked: BFNWSG3-048 (Rev 1) C. M. Parker aosaaaa-asasaaaeaassaasascaasasssssaasaggssaasssasaassssaaaowassaaaaasaasssasssscsaoo 8.5 Manual Shutdown Actions The following table lists the manual actions which have come out of the individual event and standard shutdown sequences. The list obviously does not include all steps that come from an in-depth analysis of all emergency procedures; however, it should provide a cross check of many of the key actions that are involved in shutdown. Some of the actions are associated with manual restoration of normal plant conditions (Safety Action 22 in the analysis), and manual restoration of ac power (Safety Action 21). They usually represent normal plant procedures rather than emergency steps. They can be separated from the required safety steps by noting the event that utilized them, and by the safety code designation given (MSF for the safety shutdown items). I~ ~ SAFE BIJIION NN.YSIS Page ?g3 of307 e )IANNL SAFETY NHSQ~ FlKTIOI GEE NCLYSIS IRev 1) Prepared: Uerified: Date: Date: ~ S'7 REACT 5%L NV STATE SSA EUE)IT EVE)IT TlTLE SAFETY Code ACTIN Title TUA SYS g). SYSTE)l Nf I SAFE FIK F 01B Tl6 TriyH'ails 30 Cant holing 23 MSN %F Squirt M systen 174) suppressicn pool cooling node. F 01B T/8 Trip-F Fails 30 Cant Caoling 67 EEN %F Support M systen I74) spy'essim pool cooling node. F 01B T/8 Trip& Fails 30 Cent Cooling 74 SR %F Provide suppressim pool cooling fmctim. CDEF 01C PresReg Fail<lose 30 Cont holing 23 Mcaf %F Support SR systen I74) . suppressim paul cooling node. CDEF 01C PresReg Fail&use 30 Cant Cooling 67 EEL %F Support SR systen I74) sqqressim paul caoling node. , CDEF 01C 03 Pr Hain FaiH:lose 30 Cm I %F Provide suppressian pool ten I74) CDEF i;" . f ng 1 cooling cade. CDEF 03 )lain S tn lsolatim 30 Cent Cooling 67 EEC)I I74) sappressim ool cooling node. 'stee CDEF 03 )fain Sta lsolaticn 30 Cant Cooling 74 SR %F Pronde suppresssm pao 1 cooling fmcticn. CDEF 05 Loss of Uacuue 30 Cmt Cooling 23 RS %F Suppart SR systee I74) suppressim pool ccoling node. CDEF 05 Loss of Uacuun 30 Cmt Caoling 67 EEDI %F Suppart SR systen I74) aypressim pool cooling sade. GEF 05 Loss of Uacuua 30 Cm't haling 74 SR %F Provide suppressim pool cooling functim. F 06 Loss F)l Heater 2? Rest Homal 47 TlRB1% CNIRI. )l Load deaand setpaint is used to cRain reducticn ken in autcnatic node of recirculaticn f1<a cmtrol. 4 &4hhl~ qO F 06 Loss F)I Heater 2? Rest )fcrnal II inserted. tn ectsnt needed> after reorculat>cn redact>mt to reduce pauer to initial load line see S1L 370. SAFE BlJIIOH ONLYSIS Page 29'f%7 Prepared) Date: NAIAL SAFETY mTIW )IE JNLYSIS 1RHKEHHS )Rev )) Yerifiedi gati: v /~t REACT SSA TN SAFE SEJL APP SAFETY ACTIN SYS SYSTEN RK STATE BENT BENT TIRE Code Title ML NE AXE 'F 06 Loss F)J Heater 22 Rest Harna)  % REIRC FUN C)JTRL N Run hack recirculatim f)w per SIL 370 instructims-Nanual Cmtrol Node lbefore inserting cmtrol rods). &retc to eanua) cmtrol node if in autonatic f)au cmtral ~ a44 ACD 07 Shutdn Cool Na)fm 22 Rest Ncrna) 7$ II N Turn off cr reduce M Shutdaun Cooling to stop mdesirsd cooling. 0 07 ~ Shutdn Cool Na)fun 22 Rest Naraa) 85 05 N Dperato's actims to insert Cmtrol Rad to nitigate pauer increase. 0 07 Shutdn Cool Nalfun 22 Rest Haraa) g2 )BJT)ol )NITI N Dperatl"s acticns to ~ange F N lnadv Puap Star avl ic I an )NIP ret 0 ~ a). 1 F 06 lnadv Puap Start ' 73 IPCI ind)cat)m so opefatcf can stap systen and return to Icfna) e F 0g Rad )JitMrau Errcr 22 lhst Naraal N Dperatcr to change rod pattm to ccrrect far the uithdrawa) error and restcre ncraa) rated conditions. A 11 Cmtrol Rcd Reeavl 0 Ue Disturb gg JIEACHR Ri)KTN NSF Reactor node snitch is used fcr 'REF)KL'ode. It CDE )2A Pres Reg Fai)~ 20 RPY iso) 1 NA))i STENI N 0peratcr senses blordcan and initiates aanua) NSIV closure lf excessive depressuriaatim cannot be stapped uhen he)au ncrea) )aw pressure )ieits and not in RN sade. CDEF )2A Pres Reg Fail~ 30 Cant Cm)ing 8 RHSJ Suppcrt M systee I7$ ) aqyessim pool cooling ecde. Q t 't ~ ~ 9)UOHN ~..~l. SAFE ANALYSIS Page 295 of307 0 NMNL SAFETY IF)HKS3&8 (Rev I) FlKTIN GXE ASLYSIS Prepared: Verified: htn ~K'p REACT SSA CPER. lfP SAFEIY ACTI% STATE EVENT EVENT TIT1E Code Title CP@ (~ Rerky@I~W ~>4+ 'COEF 12A Pres Reg Fail~ 30 Cont Cooling 67 EG( suppressim pool cooling node. COEF 12A Pres Reg Fail~ 30 Cant Cmling 7I M Provide suppressim pool cmling functim. ~~~(~) )C~fg] GE 128 Inadv Open~s 20 RPV lsol I SIN STEAN Close ISIVs>m indicaticns of lou turbine inlet pressrte and/cr excess bypass valve cpenlng ~ (XEF 128 Inadv OparBgass 30 Cont Cmling 23 MSN Suppet M systee (7f) sqqressim pool cooling code. CKF 128 Inadv Open~ass 30 Cent Cmling bf EG( Suppcrt M systee (74) sqqression pool cooling node. GEF 128 Inadv Openness 30 Cont Cooling 7f M Provide aqpressicn pool (XEF 13 lnadvert Open SRV Rest kraal I WIN STEAN N t to I COEF 13 lnadvert Open SRV 30 n .. j., suppression pool cooling node (XEF I3 lnadva t Open SRV 30 Cent Cmlin cr M systee (74) sqqressicn pml cooling node. COEF I3 Inadvert Open SRV 30 Cont Cooling 74 M Provide suppressim pool cmling functicn. OF 13 Inadvert Open SRV 17 Scraa Nanual scrae (per procedures) ~ ay be required if suppressicn pool teeperature becoees tm high. ~ 0'g1 COEF 15 Loss Off-Site AC 30 Cent Cooling 23 RHSN Support M systee (7$ ) suppressim pml cooling node. COEF 15 Loss Of flite AC 30 Cont Cooling 23 MS(( Suppcrt M systee (7f) dryuell spray ecde of dryuell cooling (if necessary). 4t ~ ~ s SAFE BlHMNI ANALYSIS Page 2gb ofM IIANNL SAFDY RKTII NK AIIALYSIS nte: ~/ NHM~ tRev 1) Yerified: ni ~Ma~ 5%L STATE N'AFETY IS)CT SSA BM 6HIT TlllE Cade ACTIIN Title COEF 15 Loss NHite AC 50 Cant Cooling 67 EDI )SF5qyrt IR systee I7$ ) suppression pool cooling wade. CtEF 15 Loss Nf&te AC R Cant holing 67 EDI, II Suppart drywell cooling fcnctian by IKDI systee f70'(~ cr Ift systen I7f) (reprmrt )+C All 15 Loss Nf<ite AC Q Fuelpaolhol 67 EDI Attach a fire hose to the EDI to aaintain the water )eve) in the fuel pool. ~& COEF 15 Loss NHite AC R Cant Cooling 67 ECM ISF Provide support to diesel gtl8'atcr sys'tee l82) ~ II Provide drywell cooling An power and cooling water are available. COEF 15 Loss Nfdite If: R hnt holing 7$ IR ISF Provide suppressicn pool cooling functicn. COEF 15 Loss Offdite AC AA Cont holing 74 M ISF Provide dr ling lif necessar ) b ying torus water through the cntaineent

e. This es at the 1 and that 2Q care coverage s Indi ca'ted

<feedwater systen, ). ALL 15 Loss Of Hi e 6 Fuelpoolho) 78 flEL PKL GXLl% I) Provide fuel pool cooling when power and cooling water are ava I 1 abl e. ALL 15 Loss Nf<ite AC 21 Ihst AC Paw 82 MESH. EBEN'I 8~A A<.P~ kuel pool and frywell cooling pool asap)yaw+~~ > COEF 15 Loss OfHite AC R hnt Cooling 82 NESEL %%MTR ISF Provide ac power to supfuz't ccntaineent ccoling fmu:ticns Ieanua) cr auto start an loss of afrite power). ~f SAFE 9STHN NQLYSIS Pa9e 297 of507 Preparii" h)n NAIlIAL SAFElY RKTIN CIXE NNLYSIS %HKH~ IRev I) Verified: 8 te: ~ REACT SSA TVA SP. APP SAFEIY ACTIN ~ SYS SYSTEN STATE EVDIT B)DIT TITLE Code Title I8. Nf F 16 Recirc Fail-Beer 22 Rest kraal 96 IKIRC FLN MR. N IIse nanual procedures to restcre narnal balanced recirculation flow and reactor paer canditicns. IXEF 0 Trip I Recirc Puap 22 Rest Hareai N IKIIQIATIN N Restart puap and restcre the systen for ncraal aperatian via areal procedures. IXEF 17 Trip I Recirc Puep 22 Rest Noreai N Insert control rods Ias needed) to avoid scran during restart of the tripped .0 CCEF 18 Trip Both Recirc. 22 Rest.Naraal N IKIROUTIN N recirculation puep. Restart puep start praced F 18 Trip Both irc. 2? Rest Iiaraal I ol s ior to r irs r rculatian t:pp BF 19 Recirc.pap Se zure 24 Pou Reduce 85 IXD N Nanually insert cont ol rods ver rostrate shutdae per procedures. Espected pa'th of action) but is not required. F 19 Recirc.Pap Seizure 24 hw Reduce 96 IKIRC FLNMRL N Use ncrnal procedures to reduce and aaintain flw in active loop. Evpected path of Operator act!an, but is nat requIreL BF 19 Recirc.Pep Seizure 17 Scran 99 REACKR NHECTN NSF Initiate aanual scran Iif pPeriAvpas a.a M)l%CV)) )D shlltdlMI p) Rl) " Plant Procedures CDEF 21 Startup ldl Recirc 23 Li~ disturb 68 RECIRDUTIQl N Require earn up of Laap befcre puep start. 'BEF 21 Startup Idl /hcirc 22 Rest IIarnaI N IKIRQLATIII N ~ Establish initial pouer level properly, and balance both loops after puap start. SAFE 94JIMN NN.YSIS I Page 29() of307 NA)4)AL SAFHY FlKTI(I ANALYSIS s>>e: ~&iE 4 IRH)SS3~ (Rev 1) Verified: Date: 7 REACT SSA TVA SAFE IRR. APP SAfHY ACTISN SYS SYSTEN RK STATE EVENT EVENT TlllE Code Title )L N% GEE 'SEF 21 Startup ldl Recirc 22 Rest Noma) 96 RECIRC RS CNTRL N Establish initial pauer level properly, and balance both loaps after puep start. CD 22 Loss Shubbo Ccol 19 Care Cooling NSF Naintain pressure ) ow enough fcr M systen if necessaryi uith Relief Valve(s).- CD 22, Loss Shutdae Cool 30 Cant Cooling 23 MS)( N Suppart operaticn of M(74) pool cooling. ACD 22 Loss Shuts Cool 22 Rest Noma) N Support tpntion of M(74) Reactor Shutdoe<ooling. 0 CD 22 Loss Shutdown Ccol 30 Cont Cooling 67 EEOI ation of M(74) paol caolin . f A(3 22 Losi Shutdown 2? ~ tr e>> ~ tion of M(74) wmli)q. Q ACD 22 Loss d 1 9 e ling 67 EED( Suppcrt cper ion of IR(74) Cant Cao]ing 74 IR NSF Reeove decay heat fran Taros~ter using Pool cooling. ACD 2? Loss Shutdoe Cool 19 Core Cooling 74 M NSF Naintain uater level (lJCl). ACD 2? Loss Shut&an Ccol 22 Rest Ncrnal 74 M N Restore reactor tenperatLe e to mcanal by restoring shutdoe beat f eeoval.'F 24 Rod Oreg Accident 30 Cant Cooling 23 R)f8( NSF. Suppcrt M systen (74) suppression paol cooling node. SF 24 Rod Drop Accident 30 Cont Cooling 67 EED( NSF Suppart M systen (74) suppression pool cooling node. SF 24 Rod Drop Accident 30 Cant Cooling 74 M ISF Provide suppresslls pool cooling function. CSEF 25A Large Sreak inside 30 Cent Cooling 23 MS)( )8'uppart M systen (74) suppressicn pool cooling node. gi ,0 ~ I t~ SfE )ITBIN A))ALYSIS !NNNL SAFETY If)HM~ RKTINI GEE )Rev I) A)fALYSIS ~,.'. Verified: Page 299 of507 cF s<<w Bate: ~ AEACT 5SA TVA SAFE IVER, APP SAFETY ACTINI SYS SYS)Bl FQC STATE, EVENT EVE)fT TIIIE Cade Title )L Nf GEE . CBEF 25A Large Break Inside 50 Ccat Cooling 67 EH)f Suppart M systen f7l) suppression pool cooling node. G)EF 25A Large Break Inside 30 Cent Cooling 7l SR Provide suppressicn pcol 'cooling fuxticn. CBEF 25B Iatern Brk Inside 30 Cant Cooling 25 SRS)f Suppart M systen (7l) suppressicn poal coo)ing node. CCEF 258 Intern Brk Inside 30 Cant Cooling 67 EEC)f Suppcrt SR systee f7l) suppressacn pool ccoling ecde. CBEF 25B Intern Brk Inside 30 Cant Cooling 7l M Provide suppressicn pool cooling function. CBEF 2X hall Break Inside 30 Cant Cooling 23 )NS)f 5uppm t SR systen i7l) .0 CDEF 25C Snail Break Inside 30 Cont Cooling 67 EEC)f )8'sten suppressicn pool cooling noie. suppressicn p f7l) I ccoling node. Rt CEEF 25C Snail Break Inside 30 7l SR icn pool ti G)EF ZO Brkl id li g Suppart SR syst (7l) suppressicn oo ling ecde. CCEF 250 Brklns, tQIad Std Eknt Cooling 67 EEC)f Support M systea f7l) aypressicn pool ccoling node. pi \ GEF Zb Brllnsid Cent Cooling 74 SR Provide suppression pool cooling functicn. COEF 259 Brklnside-CcnttRad 30 Cent Caoling 7l M Cent. spray node of SR alias rapid ccoldoo of prinary cent. L provides fm B)f ateosphere eixing to assist cent. atnospheric diluticn systen (Bl). Cent. spray peraissive requires that M~I be running b 2Q care coverage is indicated fF)f, 5). Rl fXEF W ErkinsidH'mtNad 50 Cant Cooling 7l SR Tcrus spray a)loss caoldoe of the uetall and aids in Tcrus pressure ccntrol. Tms spray pernissive requires that M ~ IPCI be runn)ng and that 2Q ccrc coverage is indicated 'ffes<<h<<ater systen, 8). SAFE QUINN NNLYSIS Page 300 ofR7 ))AN)N. QFHY RKflW mE NOLYSIS Preparedi C F seel~ IRH)SQ~ IRev I) Verified: Bate: W JFF REACT SSA SER. APP SAFHY ACTIQI STATE EVE)II EVDIT TITLE Code Title COEF 22 Brkinside+atIRad 25 Lie lYisturb Q ClO The prinary ccntaineent ateosphere is diluted uith nitrogen after a UXA to Iiait the aaant of hydrogen and oxygen presents COEF 25D h'klnside+ntIRad 25 Un Olsturb The prinary containeent ateosphsre is vented to SSTS (65) to naintain pressure belou a predeternined Iiait. COEF 27 Pipe Break Outside 19 Core Cooling I NN STENI I)SF The NS Safety Relief Valves (SRV) est open reeote, ~ anually fra the nain ccntrol roon to reduce reactm . CEEF .27 Pipe Break tside 5O Cont Cooling 2 RHKN I)SF pressures &gert M systee si pal ~ 7$ ) ing node. COEF 27 Pipe Break s 75) I v c $1 Ing eodee COEF 27 Pipe Break ~tside R Cont Cooling 74 M Provide suppressim pool cco Lng unc lena COEF 28 Shtdn Froe Outside 59 Cooldn Outsd I l)AIk STENI )ISF NS SRVs are used to depressuriae the reactcr pressure vessel. OF 28 Shtdn Froe Outside 2 Shutdn Outsd I )IAI)I STENI SERBKE C, STEP 3: All %1Vs (S'( are nanually closed to isolate dc ~ ~ I 14 sHpt I) fC'ED)~/Q CDEF 28 Shtdn Froe Outside 59 Cooldn Outsd SIIN STEAN IIanual deactivation of SRVs, and )SIU test circuits. Nl 28 Shtdn Fron Outside 39 Cooldn Outsd 2 R)IS) II . M Service lfater System required fcr operatim of M in all nodes. , SAFE 9lIIMW AN.YSIS IINIAL SAFETT FIKTIN GEE AN.YSIS ERHI553~ Nev II Prepared: Yeriffeh C'~r. Pa9e 301 d307 ~ ISQ S5A T9A %EL APP QFETY ACTI% . SYS STAIE 69IT MIIT TITlE Code Title ML Ag, g8 SL4l g 44fd'e ~iLAZ Nl Fra 7P gsoool psalm p~ ug)vip awi>> coafw/ ~. eye Ltr+ O 28 Shtdn Outside 39 Cooldn artsd Suppcrt systen relnured for lIR I74l and Oiesel Senerator (82). NI. 28 Shtdn Fron Outside 39 Cooldn Outsd 68 FEINllATIIPI II Positicn itecirculaticn valves to pereit tKI node of 10R, 28 5htdn Free Outside 39 Cooldn artsd 69 BGI ltanual isolaticn capability outside nain ccntrol roa. CTEF 28 5htdn Fron Outside '9 Cooldn Outsd 70 RKOI Provide dryvell cool)n9 pover and ccolin9 water are ~ available, CXEF 28 Shtdn Fron Outside 39 Cooldn Outsd 71 KIC h lhintains nrnal vater reactor vessel >bile pr sure above 100 psi9. ccntrol fra ,j 3, r overridin9 writ ed off es by at the back up control ter to sel QEF 28 Rtdn Free Outside 39 Cooldn Outsd 74 M h Suppressia pool coolin9 node - 5uppressla pool tt%)erature +i...).+~.~.~(l is aaintaioed belo I) ~ its. ~p fits A~fl>te46g ~4i0 fa( R egg Nl ZII 28 Shtdn Fra Outside 39 Cooldn Outsd 74 SR II cw'44i eaA+44ees Shut on Coolin9 lixle-fey i + DV covn 4~ sAA4srra itanually put canto cperaticn after reactcr vessel is bela 100 psl9. ALL 28 Shtdn Fra Outside 39 Cooldn Outsd 74 RHI II tPCI nile d BR svsten nay be re9uired to supply nakeup uater fe'he reactor dLe'in9 shutdan coolinu. NI. 28 5htdn fra Outside 39 Cooldn Outsd 82 OIESQ. SOGNTTR II Itequired fe operat>a ol systees if ncrnal ac peer is lost. CCEF 28 Shtdn Frcn Outside 39 Cooldn Outsd 85 05 II 05 coolin9 nate supply nay be provided to suppleeent the reacts'oolant inventory applied by the KIC l71) systee IM't regulfeoie SAFE )ITMN AWLYSIS Page Ã2 of%7 , kC Bate: < . NAM)AL SAFHY NHKQ~ (Rev FlKTISI ME I) ANLYSIS Verified: Bate: el 0'r REACT SSA TVA SAFE SER. AFP SAFHY ACTIISI SYS SYSTEN FIK STATE EVE)II EVENT TITLE Code Title M). Nlf GEE N9ENIS A~ t8 QtaL 6 Ocgth>>%~~09sa( tt CRy fss&JII lewl1)iw e+44lifp wesfrictC ~w <<aoatel ma~ ~ If 28 Shtdn Frm Outside 38 Shutdn Outsd 99 lMTGI RROTECI)l N KRKMX A, STEp li Nanually M+ open bus AI3 circuit breakers to initiate GI I85) rod insertim and NSIV closure signal through sys'ten 6I ~ EF 28 Shtdn Fr<<a Outside 2 Shutdn Outsd 99 REACTOR PIITECTN N SENXE 8, STP Ii Open power supply breakers fcr RPS RAN'eneratN'ets at 45v enter ccntrol center to initiate Gm 185) rod insertim and ISIV closure signai through Systee 6l. IF 28 2!dn Frco Outside 38 Shutdn Outsd 99 REACTN PROTEClN N SHlSKE C, STEP li r<<<<~~ie~&4reekers to the krak jilo) va)ves at RPS bus. All aust be <yen to Rt assure loser ion ccntrol rods. t w Nl 28 Shtdn Fr s e k ol t vi e emit<<red roe k~ cmtrol panel I5- . Also, s arrl ed t froa the app licab le shutdown cr R)SV boards. ALL 28 Shtdn Fr<<N Outside 9 Cooldn Outsd 244 IXNNICAT)Ik SEC N Support systee required to cardinate the various activities to be aanually perfcrned outside the cmtrol roonr S utd~ BIO Rods a Shutdn NoRod Q STM)BY LIP CNlHl N initiate standby liquid cmtrol systee injectim given indicatim of inconplete shutdoe with rcdslSystee 85 l 9?). Nanitcr injectim of the soluticn in the reactcr. Send signal to RKU(69) to isolate. (0 EF 51 Bundle Errcr 2? Rest kraal . 66 OFF~ N Ccntrol plant of fleas process to assist ccntrol rodlpower changes as cfuvatim is adjusted to stay within ncrnal ~ operating liaits. SAFE MTHN ANALYSIS Page 303 of301 ~,.~: kX F~~~,: ~z NARNL SAFETY flKTINCME IRHSQ~ IRev I) ANALYSIS Verified @tan ~+1 REACT SSA TN %EL APP SAFElY ACTIN SYS SYSTEN STATE BENT BENI TITIE Code Title )S. ICE IF 3I Sundle Errcr 22 Rest kraal N ~ discovery of abncraal power peaking Ior locatim of failed fuel), use a aolified rod pattern to restcre cmditims within baal liaits. If 31 lhndle Errcr 22 Rest kraal 90 RAOIATIIN )OIITN N Ncnitor aain steaa line radiatim, amitcr plant of fleas radiation levels so that ccntrol rod/power changes can be aade as necessary to stay within ncraal cperating liaits. 0 If 31 rrcr 2? Rest Naraal Upcn discovery, use a aodified rcd pattern to restore power sense Utiliae lRN to aal power peaking. v alve closure or I I try to prevent water ch flooded the base of the ack, standby gas , and off gas building, froa flowing into the radwaste building. 40 Struct Integ M SLIjSS 0 STRXT N Reactcr building watertight access doers/flood gates should be closed when the lake level is at 558 feet and rising e f0 Struct lnteg 3M MSS I SlRXT N Radwaste building watertight access doors should be closed ken the la'ke level is at 558 feet and rising. GXF 328 Reservoir Cmt Cooling MSN M systea Ill) 0 Lnw 30 23 NSF Support . ~ appressicn pool cooling aode. IXEF '2$ Low Reservoir 30 Cent Cooling 27 CM CIRC NATER N Nanua)ly break vacuua to prevent backflow into the fcrebay SAFE M)MN ANALYSIS Page 3N of307 Prepared: Dates lNNNL SAFETY FR]IN GXE A)NLYSIS ]RHKS3~ Nev I> Veri fi 0te: REACT SSA 1N SER. APP SAFETY ACT]9) SYS STSTE)f STATE SENT SENT TITLE hde Title NL N% ~ IXEF 32D bw Reservoir 30 hnt Cooling 67 EEN &gert RN systen f74) suppression pool cooling sade. Pi) lXEF 32S Lau Reservoir 30 Cant holing 7$ R)R Provide suppressicn pool cooling functicn. F 32D Lou Reservoir'k Pow Reduce lhnua))y insert control rods to reduce power and stean f]or to cain condenser. F = 328 Lau Reservoir 2f Pow Reduce 96 REC]RC FUN CW)RL )i Ranua))y reduce ccrc f]<a to reduce pauer. Provide signal to recirculation systen (68). ir 17 Scrae ln)tla e nanua ran if All 33 loss Fuel 1 I uel ] 'tiV I c ncreal rod insert cn is not '5'I to the EEDI 'to naintain t ter ]eve] in the fuel pao]. CDEF ISD ISO Shutdcne 30 Cent holing 1 SIN STUN hntrolled eanua] I depresQI]aaticn of RP9 using S)Ns, if suppression pool teeperature approaches a))unable ]inits. If necessary, provide autoeatic depressN']aat]an systen tADS) ~ IXEF ]SD ]SO Shutdtae 30 hnt Cooling 23 R)f8f Support RN systee (76) suppression pool cooling node. Support drpk]l coo]lng fustian. tXEF ISD ISO Shutdoe 30 Cont Cooling,67 EECM Suppart RN systen 17') suppressicn pool cooling function. Suppcrt dryue)) coo]ing function. CDEF ISD ISO Shutdoe 19 Core Cooling 67 EEN Suppart M 176) tPC] operaticn. CKF ISD ]SO Shutdoe . 30 Cant Cooling 70 RKN Provide dryue)) cooling Sea foNer and cao]>ng eater are avai]able. SAFE Mirlae) ANALYSIS Page 305 of307 Prepared) Date: INNNL SAFETY RKTIDN IXK AN.YSIS NH)553~ (Rev I) Uerl fiedL >>>>>>>>L ~%~ P1 REACT IfER 5SA APP STAlE BENT BENT TIT1E SAFETY Code ACTIN Title TI) SYS 1L SYSTE)I NK I SAFE FIK IXEF 150 150 BwtdoLhL 19 Core Cooling 74 R)ft )hth a reactor IOLL pressLve psraissive and RCICI)fCI loss due to vessel depresQe'Izatim) Mlually start me M puhp in tPCI ~ ode to haintain RPV LLate level. IXEF 150 ISD ShutdoLs) 30 Cmt Cooling 7$ M Provide suppressicn pool cmlieg fmctim. CDEF NSD kraal BLutdoeL 30 Cont Cooling I )ttLIH STEA)l Cmtrolled auwal dapresQFi?atlm of W us)fig Sws, if suppressicn pool hes alloLLable lihits. lf RI necessary, provi autohatlc aa ten IADS). I l6 and stean tahe drains after fra reactor prot tim ed). IXEF NSD Horhal BLutdon 30 Cent Cooling 23 R)RSN Suppcrt R)I systee I7f) expression pool ccoling node. Suppcrt dryLLe)I cooling fmctim. CDEF NSD kraal ShutdLhaL 30 Cmt holing 67 EECN SuppLrt R)R systeh 174) suppression pool cooling functim. Suppmt dryuell cooling functicn. CCEF HSD Horhal ShutdoLsL 19 Core Cooling 67 EECN Support R)R {70 IPCI ~ opsratim. GKF NSD Normal BwtdoLLn 19 Core Cooling 67 EECN SuppLrt fdl I7f) tJCI operatia. CEEF NSD Noraal ShutdLhsL 30 Cont Cooling 70 IKCN Provide dryLLell cooling Lban foer and cooling LLatar are available. CDEF )8) kraal QwtdoLLn 19 Core Cooling 7f R)R %F Nith a reactcr IcLL presara peraissive and RCICI)fCI loss due to vessel depressor la at) N, hanually start cne R% Ruhp Ih tPCI code to haintain RPY uater f~~eI. ~> pcs. 4 ~~~ SAFE Bfl)TIQN A)NLYSIS SNNL SAFETY NHK6RNS ffhv I) RKTIN fmE A)NLYSIS Yerified 'age 306 oU07 Sate s ~ tn ~v~ -<> IEACT SSA TVA 5%L APP SAFHY ACTIQf SYS SYSTBI STATE BM BM TITLE Code Title I6. N% GKF )fSO kraal Shutdae R Cont Cooling 7$ QR %F Provide suppressim pool cooling functicw. GXF )8) kraal Shutdmm 17 Scraa %F ))annal scraa fif needed) fron reactcr protection systea fgg) )dll activate the fm systa to insert control rods. Scrae functim cnly. COEF )fSO kraal Shutdoe I 99 fEACTIR fRHECAl %F )fanual scrae fif needed) isolatlsl. GP SSO Safety Shutdoe ,if i teeperature approaches le liaits. If necessary, provide au ma >c depresQKlaatlcn systee NS) I GXF SSO Safety Shutdown 19 Care Cooling I NI)f STEN %F Oepressmiae using NS valves to establish lw presare ccrc cooling functicns. CKF SSO Safety Shutdoo 30 Cant Cooling 2S MS)f %F Support BR systea f75) sqpessim pool cooling node and/or dryuell spray node. fXEF SSO Safety Shutdoa 30 Cont Cooling 67 EEN %F Scppart 8)R systee f75) aypressicn pool cooling code and/a cataineent spray node, and support diesel generator , oparatim. NEF SSO Safety Shuthae 19 Core holing .74 NR %F )hth a reacts IN pressm'e peraissive and )KI loss due to vessel depressm iaatim), aanually start one SR pusp in LPCI node to na1nta14 IPI water level. CCEF SSO Safety Shutdown 30 Cont Cooling 7$ J)R %F Provide suppression'pool ccolieg functicn. QFE MfMINN.YSlS 1NHNL SAFElY RKTIQI GEE NOLYSIS NHM~ Ohv 1) Va'ified! REACT $ %R. STATE N'AFElY SSA GM SENf TlllE Code ACT1% Title VO STS ML SYSIBI IN% IXEF SS0 Safety Shutdoun R Cont holing 7I M lSF Provide ccntainaent and tcrus sfe'ay Kldesg if necessarye These andes require that NHRl is nnning and that 2Q ccrc coverage is indicated ffeeduatsr systes, OQ. SAFE SHOTKNH MbLYSIS hppend~ Ih Page 1 of 9 Prepared: FC Fc Date: ~/P'F7 Generator Trip rt BFN-OSG3-048 (Rev 1) Checked: E. C. Ec B. J. Cheek Doe~: ~g7 aeaa- eaaaaaaaaaaaaaaaaaaeaaaaaaaaaaaaaeaaaaaeeaaaaaaaaaea aaaaaaaaaaaaeeae

1.0 DESCRIPTION

OF THE EVENT A generator trip results when the unit is separated from the grid, losing its electrical load. In order to avoid unit overspeed, the controls imnediately initiate fast closure of the main turbine contxol valves. These valves close rapidly and result in pressure changes in the vessel which can cause a power increase transient. Termination of steam flow to the main turbine also stops steam supply to the feedwater heaters, which in turn will gradually increase core power (if shutdown has not occurred). This appendix is based on the previous issue (reference 6e) plus new inputs and comments subsequently received. It addresses the normal case in which the turbine steam bypass valves respond as designed. Licensing submittals also cover the case in which the bypass is assumed to fail. That degraded event is covered in Appendix 1B of this report. This event is onl applicable to reactor state of operation F with the turbine rolling and the generator connected 'd This state of operation can be divide ee parts for the a ke of analysis: Fl) power urbine-generator trip scr b p s setpoint, F2) capaci y slid 3)) valve For gen t i

                             ~

below h e scram bypass set ox cj 8 8 p p, ass valve elow the bypass at power levels above the turbi nerator t:rip scram b ass setpoint (about 30 reactor is imnediately scramme and the n pump motors are tripped. A generator trip fr a power leveL below the total bypass capacity will result in no inxnediate change in reactor operation, since no scram should occur and pressure control will be automatically transferred from the turbine control valves to the bypass valves.' small, gradual power increase'ay occur depending on the degree of feedwater heating that is discontinued. The sequence is somewhat different if the initial power level is in . the small "window" between the scxam bypass setpoint (about 30 percent) and the steam bypass valve capacity (about 25 percent). The )ZAN scram and recirculation pump trip are bypassed. Reactor steam flow is diverted to the bypass valves, which will open fully because reactor power is higher than their capacity. Two factors characterize this event (assuming no operator action). Fixst, since more steam is being produced than can pass through the bypass valves, pressure and power will rise. Second, because extraction steam is no longer being supplied by the turbine, feedwater temperature will decrease, contributing to an additional, but a more gradual power increase.

SAFE SHUZD%% b5bLWXS kypcndxa IA Page 2 of 9 Prepared: Date: 7 Cenerator Tei.p E. C.Ece Checked: Date: ~e>> 7 BFNWSG3-048 (Rev 1) B.. Cheek aaasaaasaaasasassasssssssaaassasasssassaaasaasasaasasssassaaasassssssssssas 2.0 EVENT CATEGORY This event is classified as an Abnormal Operational Transient. 3.0 PLANT PARAMETERS AND SAFETY CONCERNS 3.1 Pressure Concerns The generator trip event and the resultant fast control valve closure cause a combination of effe ts on the reactor vessel pressure. The first is an acoustic pressure wave reflected back up the main steam lines into the vessel, caused by the sudden shutoff of steam flow at the turbine inlet. The second is a more gradual, although still rapid, integral increase in vessel pressure caused by the mismatch between core power azd steam flow. Protection is needed to avoid e r mary system pressure (criteria 2-3 in xn the main b dy of the report). p.,~)) X/ 3~2 Re c tivit and Inc react r c ses collapse of ste voids in the

                                                                                      ~     ~

rea or coolant..Thxs ef fee t in the core thr to increase rea si aspe

                'r, tivity and power. Decreas but much more er temperature has a ual effect. These power increase s of the event may challenge the fuel cladding integrity
          =

(criteria 2-2) ~ Reference 6f recognized this criteria for the turbine generator trip event early in BFNP design for all 3 units. 3.3 Vessel Water Level Behavior The pressure transient coupled with power and flow changes also affects water level, but normal controls should be suf ficient. Long term heat removal may also be needed for complete shutdown. LOS) I+eel (Oj I'foie fin ad ~+ceW+ 4&S +1p oc4ar 4+q~ Q

4. 0 EVENT MITIGATION 4.1 A licable Safet Actions 4.1.1 Fuel Protection (2-2)

Above the turbine-generator trip scram bypass setpoint> trip of the recirculation pump motors must be initiated (with scram> from the fast control valve closure (reference 6h) . The pump

                                                                            ~  e.".e.

Shm SHOTO+a dewar.mzs Cca ator Trip QqDemdxx Prepared: Ik E. C. Ec rt Page 3 Date: llof l 9 7 Checked: Date: ~13/ 7 BFNWSG3&48 (Rev 1) B.. Cheek aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa motor trip causes a rapid decrease in core flow and power (Safety Action 24 <<Power Reduction), protecting the fuel cladding (especially at end of core life) from the impending pressure-driven power excursion (reference 6d). Above the turbine-generator trip scram bypass power level, scram (Safety Action 17) is initiated from the control valve fast closure. This assures that control rods will already be moving in when the pressure effect, occurs, azd is needed (together with the pump trip) to ensure that &el thermal limits are not exceeded (references 6d axd 6g). Below the turbine-generator trip scram bypass setpoint, a high neutron flux or high pressure scram may be initiated c&pacity is exceed ed by core power production. if bypass valve The scram limits core power and helps assure fuel cladding integrity. 0 '4. 1.2 Reactor Pressure Protection (2-3) The turbine bypass valves relief valves must operate will be opened and (if needed ) safety briefly dur y ta ges of a generator trip at hi h order to provide RP p reseure relief c @on 18) and limit the sev rjtg f e trans 4.1.3 Normal h g+ For ge e a or ripe above the scram bypass set e path to no al shutdown is the scr procedure. Reactor water 1 vel should be maintained by the feedwater system, preesur control by the turbine controls and the main condenser is available as a heat sink. For generator tripe at low power levels (within the bypass capacity), the plant continues to operate with steam passing to the main condenser through the bypass valves (Safety Action 22, Restore Normal) ~ The operator supervises the turbine-generator operation, especially preserving house load which may remain on the unit or may be transferred to offsite power. Loss of feedwater heating occurs, but the loss hae less effect on feedwater temperature since low power extraction steam flow rates are less than those at high power. The power increase i should be monitored, especially f the bypass valve position is near full open imnediately following the trip. In that case> the operator may insert some rods to reduce power and avoid scram in ant icipati on of .restoring the unit to the, line (Saf ety Action 22).

SAFE RHJZD%8% dMLYSIS d~mndiz ld Page 4 of 9 Prepared: Date: Generator Trip E. C. E ert Checked: Date: E i/ 7 BFNWSG3-048 (Rev 1) B. J. Cheek ssaa aaaaassaaaaaaaaaaaaaaaaaaaaaaassaaaassaassssaaaaaaaassaaaaaassaassassaaaaaaaaaaa 4.2 S stems Re uired Table 01A-1 summarizes the required system functions. It was collated from event~nique actions and standard sequences from section 7.6 of the main body of the report (Tables 7.6-6, -8, -9,

            -12, -13, -26).

4 .2.1 Power Reduction (24) The power reduction action is performed by sensing generator trip (System 47). The signal goes to the RPS .(99) and the Recirculation System (68). This action is only active if power is above the logic circuit bypass setpoint (about 30X power). The bypass logic uses the turbine first stage pressure signal provided by the Main Steam System, 01. 4.2.2 Scram (17) From Control Valve Past Closure In order to prevent the control valve fast closure transient from causing core po~er to exceed thermal limits, control valve fast closure is sensed by the Turbine System (47) and a signal sent to'rip the RPS System (99) so that control rod insertion (System 85) is initiated quickly before the pres nsient caused by the valve closure can have scant effe t on the core. This, generat scram (together wit/ recirculatio otor trip) assures t Ae fr sponse is mitig ted before the p e excessiv lyaf t The contr levels be op a e f tel scram is deactivated at urbine-generator trip sc ass setpoint er (about 30 ercent power), iden the pump trip bypass. The permis ive for thi utomatic scram bypass is from turbine first stag ressure sensors (Main Steam System, 01). 4.2.3 APRM Flux Scram (17) For partial power cases where the control valve fast closure scram is not in effect, power will rise when the generator trip pressure transient reaches the vessel. The APRM signal from the Neutron Monitoring System (92) will signal an RPS (99) scram trip if neutron flux exceeds the setpoint 'value. (No credit is taken for the flow referenced setpoint).

0 - Sly SEUTK5% Ceaerator Tri.p MbLMIS hppendzx 1A Prepared: E. CD Eck t Page 5 Date: / of

                                                                                                     /

9 7 Checked'- Date: ~19 BFN-OSG3&48 (Rev 1) B.. Cheek aaaaaaaaaaaaaaaassaaaaaaaaaaaaaaassaaaaaaaaassassaassaaaaaaaaaaassaaaaaaaaaeeeaaa 4.2.4 Pressure Scram (17) Vessel pressure may trip the high pressure scram during a partial power generator trip. This signal is provided by the Feedwater System (03) to the RPS (99) ~ 4.2.5 Pressure Relief (18) B ass A generator trip opens the turbine steam bypass valves (Systems'7 and 01) to mitigate the severity of the pressure disturbance and provide a steam path to the condenser for heat removal axd normal recovery from the trip (Reference 6i defines this design of this f'unction). The potential failure of this system is addressed in Appendix 1B (Turbine-Generator Trip With Bypass Failure). 4.2.6 Pressure Relief (18) - SRV A generator trip from high power may result in a pressure transient which exceeds some or all of the safety relief valve (SRV) opening setpoints (Main Steam System Ol) ~ These valves will cycle open manentarily to limit the r em pressure, passing steam thr discharge pi es (System 10) ppression pool (System 64)~ 4.2.7 Resto e Normal ( ) p lsag ir,~ At low reacto r e+$ ~g a pass oapaoity) th e iS controlled by the bypass valve normal operation can be restored b xng the turbine generat r. This in e urbine Control System (47), aK the byp ves (Main Steam System, 01>. Manual insertion of the control rods (System 85) may also be utilized to avoid scram and/or prepare for resynchronization of the unit. 5e 0 LONG TERM SHUTDOWN CONSIDERATIONS 5.1 Normal Shutdown The sequence provided in section 7.6 of the main body of this report for normal necessary (Table non-isolated shutdown would be followed 7.6-2). However, in this event,. it is often if 0 expected that restoration of the unit to the grid will be possible and full cooldown will not be necessary. RI

%8K SHOED(5% dMLXSXS               ~xmdxx Ih                            Page 6    of  9 Prepared:         5'C.               Date.

Ce ator Trip Checked:. E..Eeet Datet I 3 P7 BFNWSG3&48 (Rev 1) B.. Cheek 'QCI~RQSRQRRCCCRDQQl5%NCJKRNQRDRCSM~RQRQRNÃCtSSRRRSRRRQQCEXFOR%>>It <DCCCR<+>>++>++> 5.2 Safet Shutdown Normal, long-term response, as discussed in 5.1, utilizes all systems normally available to the operator. If any (or all) normal control equipment is assumed to be unavailable the reactor if

                                                                              ~

can still be shut down to cold conditions necessary, by utilizing only safety equipment as given in the safety shutdown sequence in section 7.6 (Table 7.6-3) starting in this event from. a non-isolated condition. 6~ 0 REFERENCE S

a. Browns Ferry Nuclear Plant TSAR, Chapter 14; System Safety Ana lys is.

b. Browns Ferry Nuclear Plant FSAR, Appendix G; Safety Operational Analys is.

c. Browns Ferry Nuclear Plant FSAR, Chapter 7; Responses to AEC Questions.

d. Browns Perry Reload Licensing Report; Unit 2, Cycle 6, April, 1985.

         'FN A,

e. Safe Shutdown Analysis, Appendix lA (Revision 0), B. J. Cheek and V. G. Blanchette, Jr.,'FNWSG3&48, June 28, 1986.

f. NEB-RAC-1593 (B45 860630 061), BFNP DEAR, Amendment 7 estion 4.

g. NEB-DLH-1013 (B45 860628 725), BFNP SER, Sec ion III.A (page 15), June 6, 1968; Aidit'rbine/Generator ) ip (anticipatory) scr a>l $

h. NEB-RAC-1 5 860624 016), TVA l e 0b'ptem8jr 25, 1978 from J. E. Gillelatd,' to-S'.' ) Png~j cle 2, Addendum t Qe Wipe 'Qh )> Addition f T/G trip recir pg ot

i. GEC-GDC-10 1 ( 05 9

                                     -     87,   GE  document 22A11             @axon   6),

Design Spec 'cation Turbine Press ator and Bypass. 7.0 NOH-APPLICABLE Sections 4.2 aal 4.3 in the main body of this report show results of an assessment of licensing references and FSAR questions. The items which have been identified as likely SSA (Rev 1) inputs for this event are given in Tables 4.2-1 and 4.3-1. These references were either used above in this appendix or they have been evaluated atd not utilized as described below.

SCAPE RiOTK5% h%LT,MX5 Jlypendix Page 7 of 9 Prepared: 4 FC Daze: ~// Ceaerator exp E.. Ece BFNWSG3-048 (Rev 1) Checked: Date: ~/9 f7 B, . Cheek aaeaaaaaaaassaaaaaaaassaaaaazsamazszaagggaeII~~acsaaaaaaeaSsaaeazSaIIaaeeeINwaaaaaaaeaa 7.1 FSAR, Volume 7, Question R10.2 - Although identified as potentially applicable to Turbine/Generator trips with SRV actions, this question really addresses the aix-actuated, ADS mode of the valves (not the pilot-operated mode used in overpressuxe protection cases). Pl

SfE SHJIEON N)ALYS)S APPE)011 01A Page 8 of J

                                                                  ~~so                 FX~

Senerato Trip NH)583~ Iev 1)

                                                                  ~<<                         ~<<~z~                      r7 TSE   01A -1 ISCT SKQHAfkE                            TVA                   SAFE KR.        RESllTS       SfETY ACOOI     S5       SSTBI        FBC STATE  Me Title          h)e    Title    NL        IN%        'GEE F      2-2 Fml Failure 17 Scree            1  SIN SlENl        fI: Provide >PL turbine first stage presar>>

interlock signal to reactw protecticn systen 199) fail~e logic. F 2-2 Fuel Fallacy 0 Scru SF Provide high reactor vessel pressure trip signal to reactcr protecticn systm f99) fail~ logics F 2-2 Fuel Failure 17 Scree 47 TNSBK GNRL  % Provide nain turbine control valve fast clmra trip signal to reactor protectim systee 199) fail~fe logic. F 2-2 Fuel Failure 0 Scree $5 N0 SF Scree signal free roacbr protectia system N) nill activate the cmtrol rud drive systen to" insert rods. Scrae fenctim only. F 2-2 Fuel Failure 17 Scrae 92 Kl5$ NNlTN SF Provide N%N neutron flui trip signa) to reactor protectim systen i99) fail~ logic. F 2-2 FLNl FailNe 17 Scran 99 lSCHR PROIKTM  % Provide scrae signal to the cmtrol rod drive lOS) systee (85) cn naia turbine-cmtrol valve fast clare trip signai frm turbine cd Micating turbine first stage presswa > XC. F 2-2 Fuel Failure 17 Scrae F 2-2 Fuel Failure 17 Scrae 99 IBCHR P8UTETN F Provide scrw signal to the cmbol rod drive m sys reactor vessel pressure trip signal. F 2-3 Syst Stress 18 Pres @lief 1 NTN SlM EF SHs open m high reacta pressure. F 2-2 Fuel Failwe 18 Pres Relief 1 NNSI8N Hain turbid bypass valves open m turbine control systen N7) generator trip signal. This safety actim nay mar, but is not required hy this avet. F 2-3 Syst Stress 18 Pres blief 10 1611M VIS h W Provide path fcr cain'stean systen f01) %Vs stean hloedoen to soppressim pool N).

SfE Mmawl M.TS)S ARBOII 0$ Page ~ 9of 9 cF C.'rt Senerate iRH)SSRHS Trip Iev 1) Oected>- F LJ. SSIi ~<~ j TARE ON 8 IBCT INCCEPTARE TN SAFE 5%R. RERLTS SfHY ACT)lf S5 SISIE)) F)M', STAlE Code Title Code Title I). le II)0E F 2-2, Fuel Failure 1S Pres Relief 47 llRSDK NSK Send siyal to c4ea and astrol aain stean systen l0)) bypass valves cn generator trip. This safety acticn nay this event.

                                                                                               ~,     bet is not rmapkred by F      2-5 Syst Stress   18   Pres hlief     64 'PRl CC)fTA)IHT         Accept %Ys    stem btuebn       ffra boiler vents  and Mns     systen, 1M to ayressicn pool.

F 2-2 Fuel Fail n 22 hst krna) 1 ))A)N S)EN S)gna) fran Turbine Control 147) to bypass va)ves provides ncrnal control after trip fron )on initia) ponery avoid scraL Tb)s safety act)co F 2-2 Fuel Failw'e 22 Rest )turnal 47 TNS)lf GIfHK After Seneratcr Trip, signal bypass valves to aaintain presara coetro). This safety acticn

                                                                       . nay'ccÃ'y but is not required by this evente Provide HtC bebine       first stage yrasare interlock signal to reactor protect)co systen l99) failMe logic.

F 2-2 Fuel Fai)wv 24 Pw hduce 47 TNS)%, GNRL 'F Provide cain turbine ccntrol valve fast clare signal to reactor protecticn system 69). F 2-2 Fuel Fail+a 24 Pw hduce N RECIRQLATltw L 6pen redrculaticn peep eotlr brea)mrs sl%PS f99) s)goal due to naia turb)ne contr as c san and > 30l brbine first stage pressre. Coastdoun <<It be faster than assueed r I F 2-2 Fuel Failure 24 Pon hduce 99 Saul signa) to quan recirculaticn peep eobr breaters fsysten 68) m cain turbine ccotro) systen 147) and nain stean systen (01) sill indicating > Rl ttvbine first stage pressure. F 2-1 Radhlease 26 Est Pri Cent 85 D5 Perf em isolaticn acticnls) upon receiving scf& signal fry the Reacts Protecticn Systen f99).

Tmb~~erator Trip Pith bypass Failure

                                 """," cc Prepared:

Checked: r~ E. C. Eck rt Page Date. Date: 1 of 7f

                                                                                 > 3 7

BFNWSG3&48 (Rev 1) B.. Cheek aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa 1.0 DE SCRIPTION OP THE EVENT The turbine-generator is coupled as a unit, aalu therefore, shares many similar abnormal operating transients in common. The generator trip occurs mostly from loss of electrical load (e.g., fran lightning strikes). A turbine trip (stop valve closure) can be initiated from any of several unit trouble sources (e.g. moisture separators). The normally expected events are described in Appendix 1A (Generator Trip) and Appendix 2 (Turbine Trip). Either event, generator or turbine trip, initiates fast closure of the turbine inlet valves. This appendix is based on the previous issue (reference 6c) plus new inputs and comments subsequently received. It addresses the occurrence of either event with the additional failure of the turbine steam bypass valves to respond. Turbine control valve, fast closure or stop valve closure with bypass valve failure causes a rapid pressurization transient and isolates the reactor. Analysis of the event with this assumption is currently required by the NRC to be included in the PSAR and Reload Submittals. The thermal-hydraulic effects on the core are more severe than with the bypass operating, although failure of the bypass system is not expected. The event is represented by the generator trip with bypass failure in the latest reload submittal reference 6a, Browns Ferry Nuclear Plant Reload Licensing Report Unit 2, Cycle 6. A turbine-generator trap with bypass failure is only ' applicable ip operating state P w ere the turbi r o ~ ' j 2.0 j The BPNP PSAR and r load submittal conservatively treat this event as an Abnormal Opera ional Transient. 3.0 PLANT- PARAMETERS AFFECTED AND SAFETY CONCERNS A loss of generator electrical load from high power conditions (or turbine trip) produces significant response of several parameters. Appendices 1A and 2 describe this response for the expected event. Areas where bypass failure increases the severity of the event are described here:

a. Reactor vessel pressure rises more rapidly because bypass flow is not available. Nuclear system stress may be in excess of that allowed for transients by applicable industry codes (2-3).

Appendix 15 Prepared: Pith Bypass Pai1ure Checked: Date: / BFN-OSG3W48 (Rev I) B. J. Cheek aaooaooaoaaaaaaaaaaoaaaooaaooaaaaaaoaaaoaaaaaoaaaaaaaaaaaoaaaaaaaaaoaaaaaoa

b. The more rapid pressure rise creates a greater core void fraction reduction, potentially causing a significant power increase.

Excessive fuel failure (2-.2) could be calculated as a result of the transient. Reference 6d recognized this criteria for the turbine generator trip event early in BFNP design for all 3 units.

c. Since the main condenser is available, but no longer accessable, the reactor is essentially isolated and decay heat removal must be.

through the relief valves to the suppression pool. Suppression pool cooling is needed to remove decay heat while keeping pool temperature within its limits (2W) .

                   <<< <<>) ~'isl~kiW~ a 4 C knaw~rCkia,s pip ocooav gik'~k ~f ~rp~t<~.

4. 0 EVENT MITIGATION 4.1 A licable Safet Actions

 ~       4.1.1    Fuel Protection (2-2)

Power reduction (Safety Action 24) to assist the scram in ' is px. nee xg mitigation of th power) conseque e events Scram (S f ion of the cont er to reduce th p v0C' urbances. 4.1.2 RPV Ove essure Protection (2-3 Pressure ref (Safety Action 18) is needed to keep the pressure transient within the required code limits. 4.1.3 Containment Protection (2W) Containment cooling (Safety Action 30) is needed to accommodate the temperature increase in the suppression pool. 4.2 S stems Re uired Table 01B-1 summarizes the required system functions. It was collated from event~nique actions and standard sequences from section 7. 6 of the main body of the report (Tables 7. 6-6, -8, -9,

                -12, -26, -34).                                                             Rl

append~ 15 Prepared: Date'- 7/ Tarbxa~aerator Trip E. C. Ecke 'Pith bypass Taxlare Checked: Date-BFNWSG3&48 (Rev I) Cheek aaaaaaaassaaaaaaaassaaaaaaaaaaaaassssassaassaaaaassaaassaaaaaaaaaaassaaaaaaaaaaassssaa 4.2.1 Power Reduction (24) Together with the quick power reduction provided by scram, the recirculation pump motor trip provides fuel protection (references 6a, Browns Perry current Reload Licensing Report, and 6g, which documents the commitment to the NRC to utilize this feature) ~ This analysis requires the specified initiation logic timing and the rapid coast down characteristics assumed in the analysis for the recirculation pump trip. Quick power reduction (24) is accomplished by sensing the turbine-generator trip. Equipment of the Main Steam System (Ol) senses stop valve closure while a load rejection (presented here) is sensed by the Turbine System (47). The signal is sent to the Reactor Protectio'n Syste'm (99) and to the Recirculation System (68), where the recirculation pump motor breakers are tripped open. This feature is only active if the reactor is above 30 percent power, using the permissive logic on the turbine first stage pressure (System 01). 4.2.2 Scram (17) Scram is 'accomplished by sensing the turbine-generator trip. (References 6e ard 6f established this scram)- A load rejection is presented here axd in the current reload submittal (reference 6a) It is sensed by equipment in the Turb e If

                                    ~

Control System (47) ~ turbine re, the main steam system ( se the closure of the stop v ives. Thl. s si al is sent to the Reactor Pr 'on ) azd Control od Drive is a ac 1 only above 30 p p te (01). Below the 30 pecent power permissive, senso u ron Monitorin System (92) ure sensors in the Feedwater provide the input signal for scram to the Reactor Protection System (99). I 4.2.3 Pressure Relief (18) This Safety Action is accomplished solely through the operation of the pressure relief valves in the Main Nuclear Steam System (Ol). (Since the turbine bypass valves are'ssumed to fail in this situation.) The SRV discharge piping is provided by System 10 for proper quenching of the steam in the suppression pool (System 64) ~

ShTK SHUZM%% hMLMI5 kppendzx 15 Prepared: Tabb exp Vith ~sass Fai.lure Checked: BFNWSG3-048 (Rev 1) B. J. Cheek nn tn enà R D seen o tt Q'te tent en N en ennl c tent tn % ns tn en % D ma tn sn aa tn nnnn zsR R sn $ ~% enne ct% sn R en ntns ne R% R ns ts Rue D % en toln % su%ense

                                                                                                      $                                                                   w nt 4.2.4             Containment Coolin                               (30)

This Safety Action is accomplished by manually initiating the suppression pool cooling mode of the Residual Heat Removal system (74) The operator would initiate this system following indication of high pool temperature (sensed by the Primary Containment System, 64). Support is needed from Systems 23 (RHRSW) axd 67 (EECW). 5~ 0 LONG TERM SHUTDOWN CONSIDERATIONS

5. 1 Normal Shutdown Although the main condenser should be available as a heat sink, the valves leading to it are all postulated to be inoperable. The feedwater system is available during the initial part of the event, but cannot indefinitely draw its supply from the unreplenished condenser. i Therefore, f the steam path to the condenser cannot be restored, shutdown to cold conditions will be conducted using the Normal Isolated Shutdown sequence from Section 7.6 in the main body of the report.

5.2 Safet Shutdown Normal, long term response, as given 'above> utilizes all systems normally available to the operator. If any (or all) normal control equipment is assumed to be unavailable, the reacto shutdown to col x. xzx.ng only safety equipment s given in the Sa ty Shutdown sequence given in Section 7.I ting in this event fr

6. 0 REFERENCE S

a. Browns Ferry Nuclear Plant Current Reload Licensing Report TVA"RLR&02, IRevision 1, Unit 2

b. Browns Ferry nuclear Plant FSAR 14 5 1 1

c. BFN Safe Shutdown Analysis, Appendix 2 (Rev. 0), H. A. Greaves and S. K. Mehta'BFNHSG3&48, June 28, 1986.

d. NEB>>RAC-1593 (B45 860630 061), BFNP DEAR, Amendment 7, Question 4.

e. BFNP, FSAR, Volume 7, Question Q+.7, Credit for T/G Trip Scram.

f. NEB-DLM-1013 (B45 860628 725), SER, Appendix A, Section III.A (page 15), June 6, 1968, Addition of T/G trip (anticipatory

-0 g. scram). NEB-RAC-1263 (B45 860624 016), TVA letter dated September 25, 1978 from J. E. Gilleland to H. R. Denton (NRC), BFNP Unit 2, Cycle 2, Addendum to Supplemental Submittal (NED0-24095), Addition of T/G trip recirculation pump motor trip.

o kppeukix lI5 Page 5 of 7 Prepared: Date: Pith bypaaa Pai1are Checked: E.. Ecke

                                                                             ~

Yl'ate: 3/' BFN-OSG3"048 (Rev 1) B.. Cheek aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa

7. 0 NON-APPLICABLE INPUTS Sections 4.2 atd 4.3 in the main body of the report show results of an assessment of licensing references and PSAR questions. The items which have been identified as likely SSA (Rev 1) inputs for this event are given in Tables 4.2-1 and 4.3-1. These references, were either used above in this appendix or they have been evaluated and not utilized as described below.

7.1 PSAR, Volume 7, Question R10.2. Although identified as potentially applicable to Turbine/Generator trips with SRV actions, this question really addresses the air actuated, ADS mode of the valves (not the pilot-operated mode used in overpressure protection cases). 7.2 PSAR, Volume 7, Question U4.2 - This question atd response addresses the old SRV configuration (with two unpiped, spring safety valves). It is obsolete - replaced by current reload analyses for the current configuration. 7.3 NEB-WRB-1050 (B45 860623 644), TVA letter dated October 15, 1982 from L. M. Mills to H. R. Denton (NRC), Technical Specification changes (unit 2), Replacement of Two Safety Valves with Two-Stage Target Rock Safety/Relief Valves. This reference primarily discusses ASME overpressure protection covered in Appendix 30, not 1B. 7.4 NEB-RAC-1439 (B45 860626 645), TVA letter dated December 5, 1978, J. E. Gilleland to T. A. Ippolito (NRC). This reference addressed overpressure protection (event 30, not 1B it 1 (Unit 2 is covered by the item discuss

SAFE BSMN ANLTSIS NVE)I)lI 018 Page 6 of 7 F Turbine/Seneratar Trip E.C. ert uith Bypass Failure Dechdx lRHKQ~ (Rev 1) LJ. TSE 01B -1

    %4CT   WXEPIARE                            VN                     SfE

56. RES(LTS SAFETY ACT)(N SS QS(EN RIC STATE Code Title Code Title )L NK (IE F 2-2 Fuel Failwe 17 Scran 1 )NN STEAN SF Provide >301 turbine first stage pressure interloct signal to reactor protectim systea (99) fail~e logic.

F 2-2 Fuel Failure 17 Scrae tf Provide high reacbr vessel pressxn trip stgnal to rector protectim~ (99) fail~ logic. Provide nain turbine ccntrol valve fast clare F 2-2 Fuel Failee 17 Scraa 47 T(R))lf UNTIL trip signal to reacbr protectim systa (99) fail~ logic. F 2-2 Fuel Failure 17 Scrau 85 Gm S(gnal frm IFS (99) exll activate the Control RN( irive to insert rods. Rrm functim mly. F 2-2 Fwl Failure 0 Scraa 92 KAHN )9(IR SF Proride API( neutrm flux trip signal to reactor

                                                                           ,protectim systen (99)     fail~e logic.

F 2-2 Fwl Failwe 17 Scree 99 REACT(R PROTECTN SF Provide scr systen (ED m fast elmore' o the cmtrol naia f brbi 'e rod'& cmtrol

                                                                                                                          > 30K F     2-2 Fuel Failure    0   Scrae                                              scrae signal   to the ccntro (CR0) systen (ED m                   bring systen (92)            m flux trip signal.

F 2-2 Fuel Failure 17 Scrae 99 Provide scraa signai to the cmtrol rod drive t(m) systen (8Q m fsedeater ~an (03) high reactcr vessel pressN e trip signal. F 2-3 Syst Stress 18 Pres Relief 1 5N STM SF S(ws open cn high reactcr yraart, 2-3 Syst Stress 18 Pres Relief 10 H)liER-NTS I NN Provide path fcr naia stean systa N) SSs stean blodoe to suppressim pool N). 2-3 Syst Stress 18 Pres Relief Q PM GRADIENT Accept Sfms stean bloudouo (free bciler vents and drains syst~ 10) to suppressim pool. '0 Y 2-2 Fuel Failure 2$ Pou Rsduce 1 INK STEN( F Provide >301 turMoe first stage inter)cct signal to reacbr protectim (99) fai)mfe logic, pressure system

SAFE MIMNANTIS Turbine/Senerata OPEN)ll 0)B Trip Page Prspared3 7 of 7

                                                                                             ')ate eith  Bypass  Faille NHSEHIS        IRev 1)                                                      B  l.

TRRE 01$ "I IEACT QNKEPlARK TVA Sff 5%R. RESKTS SAFETY ACTIIN STS SySIDI fQC STATE Code Title Code Title NL NK CNE F, 2-2 Fuel Failure 24 Pou Reduce 47 TNBllfQNNfL SF Provide naln tNhise cmtrol valve fast claws signal to reactcr protectia systen W). F 2-2 Fuel Falsie 24 Pen Reduce 6B RECIRCRATIN 1. 0pen recircelatim peep outcr breath'm iPS N) signal doe to nain turbine cmtrol valve fast clam and > 35 turbine first stage prassee. Coastdous cost he faster than asseeed in reload analysis to LMt severity of the event. F 2-2 Fuel Failme 24 Pou Reduce 99 REACT% ROOECT)I F Sand signal to cpen recirculatim pusp outcr breakers Isysten N) m nain tsrhine ccntrol valve fast clue signal froe turbine control systen I47) and naih stean systen I01) signal Indicating > 30L turbine first stage proser'. F 2H Cmt Stress R Cent Cooling '5 MSI 18: Support M systen I74) sypresaim pool cooILeg

                                                                        ~ odeum F      M    Cont Stress   30   Cont Cooling  64 PRI QNAINKN           Provide appressia peal twperature and level indicatim to suppwt       RIN~ I74) syyessim pool cooling eodt~ mls'o    vn~ ~o//<~,         F
                                                                                                                    ~ Va /Pp F      M    Cmt Stress    30 Cont Cooling    67 EQI             %F Support NR systen I74) suppressim pooL cooling
                                                                        ~ oder F      2-4 Cmt Stress     R    Cont Cooling  74 RIR             1%F    Provide soppresstm pml cooling fuacUm.

3am saVXnam aae.mzs Pressure Regalator Failure Closed hppendxx IC Prepared:

                                                             .E c.r~

Eckert Page Date: 1 of Y 6 i-8> 87 Checked: Date: BFNWSG3&48 (Rev 1> ~ ~ rice

   +CCt~WMWMMM MMMWI MME&&SXMCtSSMMCtWMIICtSRCSCIIWMSIClMCRWCtSXDWWCRlRCSCtSCW M  MC5MCtWMMCXCIMM~ZRMMMMZR 1 ~ 0 DE SCRIPTION OF THE EVENT The turbine pressure regulator normally provides continuous system pressure control by modulating the turbine control valves (or, during star tup, the turbine bypass valves). Two different bounding cases are possible    if a regulator fails. One is when either of the two regulators malfunctions in the 'open direction suck that the pressure initially decreases. This event. is described in Appendix 12A. The malfunction when the regulator fails in the closed position so that steam flow is reduced and the pressure initially increases is the event described in this appendix.                It   is baaed on the previous SSA issue (reference 6c) plus new inpu s and comments subsequently received.

Either of the cases can only occur when the turbine is being supplied with steam: reactor states C, D, E, axd F (in state A the vessel head is off). The FSAR (references 6a 6 6b) address this situation together with the pressure regulator f'ailed-open. This event has been documented in this appendix since the results of the event per the FSAR are similiar to a turbine/generator trip. 2.0 EVENT CATEGORY This event is considered to be an Abnormal Operational Transient, although the failure of both regulators in the closed direction is considered to be very infrequent.

3. 0 PLANT PARAMETERS AND SAFETY CONCERNS The initial result of the pressure regulator closure is to increase the main steam line pressure. In the case where the backup regulator functions properly, the initial (small) rise in pressure will cause the transfer of control to the backup regulator. Po steam flow along with main steam There are no safety concerns.

pressure will then 'eturn to n ormal.

                                                                                        ',r In the unli             ent that    the ba        P                         ab e,     the comple te clo u re    of the                                                    ont oiling stroking rat               t                                        ypass), caus      s a  rise in pressure                  r       e         h  t   is expected to reac                scram setpoint of t The   following       c cep table                must be considered:

a. -

Potential el damage (unacceptable result 2-2). - the increased . ~ reactor power (before scram) can challenge fuel thermal margins.

kppend~ 1C Page 2 of 6 Prepared: Date.. 1 7 Presaare Regalator Q.~CW kert /- FSL1ILxo Clos ect Checked: ~ , Date: BFNWSG3&48 (Rev 1) B. F. rice aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa

b. High nuclear system stress (unacceptable result 2-3) - the closure of both pressure regulators will increase the pressure in the main steam lines azd back up to the reactor pressure vessel thus placing a pressure >tress on the primary system.

c. Containment stress (unacceptable result 2W) must be considered when the pressure relief system operates, the release of the high presssure steam into the suppression pool will cause an increase in the suppression pool temperature, challenging its operational lI.RI.tss

d. In addition, reactor water level will decrease due to the but normal level control will be resumed by the pressure'ncrease, feedyater system. (No safety concern). <o~ /~l C~>)i>o~<4s'~ ~M Cssss sssssboykgg l+/ssss) sss

0) DC~ Ass+ 4+ ~f ~see( . fCK WW/Pg

4. 0 EVENT MITIGATION 4.1 Safet Actions Re uired If the secondary regulator takes over properly when the controlling regulator fails, no safety actions are needed. The following paragraphs identify the safety actions which must be taken to mitigate the worst case safety concerns with both regulators failed in the closed direction.

4.1.1 Provide Fuel Protection (2-2) To prevent excessive fuel damage, reactor scram should be initiated (Safety Action 17). 4.1.2 RPV Ove ressure Protection (2-3) To prevent excessive nuclear system pr re rise fol t e closure of the pressure regula , pressure r ef . (Safety Action 18>. 4.1.3 Containment Pr tection (2- ) To remove energ e e d e pression pool hereby keep containment i ns m.thin accepta emits, containment cool ng afety Action s needed.

PEcssure Wgulator ert Page 3 Date. 'I of

                                                                                               /

6 -0 Fai.lare Closed BFN-OSG3-048 (Rev 1) aaaeaezeeaaaazs Checked:

                                                           . gE.

race Date. aaasaaaaaaaaaaaaasaaaaaaaaaaeasseeaeaeaaaasaaaeaemaaassaaazaaassaaaaa l"9'7"-~7 4 ~2 S stems Required Table 01C-1 summarizes the safety actions aA system requirements related to this event. They were collated fran event~nique actions azd applicable standard sequences fraa section 7.6 of the

               ".main body of the report (Tables 7.6-6, -7, -8> -12, -34).

4.2.1 Reactor Scram ( Sa fet Action 17) From normal, high power conditions the power will increase to the APRM setpoint (Neutron Monitoring System, 92) ax@ send a signal for reactor scram tv the Reactor Protection System (RPS, System 99). At intermediate power levels, the scram will come from reactor high pressure, sensed by the Feedwater System (03) ~ The RPS signals the Control Rod Drive System (CRD, System 85) to initiate the rapid insertion of the control rods (Scram). If the power is low, (e.g. in the STARTUP mode range) scram would be initiated via high neutron flux (Neutron Monitor System, 92) as measured by the IRM ins trument s. 4.2.2 Pressure Relief ( Sa fet Action 18) Following closure of the turbine valves, the Safetv-Relief Valves (S/RVs) will automatically open aA cycle as necessarv to maintain the reactor vessel pressure within safe limits. The S/RVs are part of the Main Steam Svstem (1). The discharge piping (System 10) axe the suppression pool (System 64) support this action. 4.2.3 Containment Coolin (30) Suppression pool temperature and level monitors, which readout in the control room, are provided as part of the P Containment System (64). Wh'en the suppr pool tempera ure approaches the thermal operatio erature limit e System (74) must be aligned into n t cooling mode oops maximum r 1 n a a single failu e backup) ~ t te thi mode are als

kpyenda 1C Page 4 of 6 Date. Pressure Regulator Paxlare Closed Checked: E~ ert Date. p>> BFN-OSG3-048 (Rev 1) B. F. Price aasassaaasaaaaaasaasssssaaassaaaaassasaasaasaasaaaasassaaassassasaasassasss

5. 0 LONG TERM SHUTDOWN CONSIDERATIONS 5~1 Normal Shutdown Although the main condenser should be available as a heat sink, the controls for the valves leading to it are all postulated to be inoperable. The feedwater system is available during the initial part of the event, but cannot indefinitely draw its supply from the unreplenished condenser. Therefore, path to the condenser cannot be restored, further shutdown if the steam actions will be conducted using the Normal Isolated Shutdown sequence from Section 7.6 (Table 7. 6-1) in the main body of the report.

5 ' Safet Shutdown Normal, long term response, as given above, utilizes all systems normally available to the operator. If any (or all) normal control equipment is assumed to be unavailable, the reactor can still be shutdown by utilizing only safety equipment as given in the, Safety Shutdown sequence also given in Section 7.6, starting in this event from an essentially isolated condition (Table 7.6-3).

6. 0 REFERENCE S

a. FSAR, section 14.5.1.7.

b. FSAR, Appendix G.5.3.5.2.

c. BFN, Safe Shutdown Analysis, Appendix 1C (Rev 0), H. S. Robbers and B. J. Cheek, BFNWSG3-048, June 28, 1986.

7. 0 NON-APPLICABLE INPUTS Sections 4.2 axe 4.3 of the main body show results of an assessment of licensing references an9 FSAR questions. No items have been identified as likely SSA (Rev 1) inputs for this event.

QfE S(lJMN NCLTSIS Pressure Regulatcr Faille - C)used IRHjS(M(8 IKKT MXEPTAKE (PEL RELLTS (Rev I) APf9bO 01C SAfEIy ACTT(N TN STS TSE 01C STSTS

                                                                -1 Page PropiYIdl Cheetah SIFE F(gC Sof E.

6 Ect F. Pr ce ate gati:

                                                                                                                       ~

STAIE Code Title Code Title IL Nf NK F 2-2 Fuel Failme 0 Scree "SF Provide high reactr vessel presan trip signai to reactcr protectim systee (gg) fail~ log(co (F 2-2 Fuel Failre 17 he (5 59 ~ SF Scrae signal free reactor protectim systee (gg) ekll activate the'ceetrol rod drilal sysba to insert rods Scra+nctim only. F 2-2 FuelFailure17 Scree g2 IHJIIS(IQ(mR Provide it'll oeutrm %lux trip signal to reactor protactim systee (gg) fa(chafe logic. 0 2-2 Fuel Fails e 17 Scree 92 lEUIIIIQ(ITI F Provide IIIeeutran flux trip signal to reecbr protectim systse (99) (see sactim 7A;3. F 2-2 Fuel Failwe 17 Scrae 99 REACII fmTECIN PF Provide scree signal to the control rod drive HR) systee (ED m eeutrm em(tcring systee (92) APIg( neutrm flux trip signal. 2-2 Fuel Failre 17 Scree 99 ISNN P%IECIN SF Prov(de scree signal 'to the cmtrol rod 'Irive ((3') systee (85)'m feedeatr systee (03) high reacta vessel pressure trip signaI. 0 2-2 Fuel Faillre 17 Scrae '8 REACIIR PRIEIN I: Provide scree signal to the cmtrol rod drive ((I) systee (ED cn eentrm anitzing systeo Ig2) 1RN neutron flux trip signal. C(EF 2-3 Syst Stress 18 Pres Relief 1 ININSIEAli SF SRIts reactcr pressoreo C0EF 2-3 Syst Stress 18 PresRHIef 10 Provide ) GEF 2-3 Syst Stress 18 Pres Relief 6( ae hloedmsl (frm hoilr drains systee, 10) to

                                                                                                    ~w ~w Wa Wn 0F     2-I,Rad   Release   26   Est Pri Cont 85                          SF               Rdatim actim(s) upm receivleg       ace a~      ~ra u (3EF   M     Cmt Stress    30   Cmt   holing  23 RHSI        .           IEF Support     RI systee    0$ ) seppressim pool aolieg eodee C(EF   M     Cent Stress   30   Cont Cooling  6( PR1 C(g(TAINB(T               Provide suppressim pool teepratre and level.

(ndicatim to soppcrt IR systee (7$ ) sappresaim 0 pool cool(ng node) P,4o voonl geol'el/ @~i

                                                                                                                            < Qs,o/S/
                                                                   ~      ~

%E SUMM ANLTSIS APPEMIli 01C Page 6af 6 Prepsru C Prescore Re9RIRtor E.C. Failn - Closed Sated NHL%RN thv 1) LF. Pnce ME 01C-1 ND QNXPlARE 1N 5RL KkLIS SFETT ACHN SS SSIBI STATE Cade Title Cade Title NL NK REF,&,Coat Stress R Cmt Coolie 67 EES 5f Support RN systee Oil sqqresaim baal caaliay eadem IXEF & Coat Stress R Cast Coolly 74 RR

                                                          'SF    Proride sepyressica paai caathg hectim.

SAFE SHUTDORH AMLTSIS Append~ 2 Page 1 of 9 Turbine Trip E CD Ec rt Date: ~/ Checked: Date: ~~> BFNWSG3-048 (Rev 1) B. J. Cheek ssaaaaatmassaasaaaaaaaaaaaassssassaaaaaaaaaaaaaaaaaaaaaaaaaasssmssssaaaassassaaaaaaaaaaa

1.0 DESCRIPTION

OP THE EVENT A turbine trip results in the tripping closed of the main turbine stop (and control) valves. These valves close rapidly and result in pressure changes in the vessel which can cause a power increase transient. Termination of steam flow to the main turbine also stops steam supply to the feedwate heypprs~hich in turn will gradually increase core power (if has n~o occurred). This appendix is based on the previous issue (reference 6e) plus new inputs and couments subsequently received. It addresses the normal case in which the turbine steam bypass valves respond as designed. Licensing submittals also cover the case in which the bypass is assumed to fail. That degraded event is covered in Appendix 1B of this report. This event is only applicable to reactor state of operation F with the turbine rolling. This state of operation can be divided into three parts for the sake of analysis: Rl) power above the turbine-generator

 ,0        trip scram bypass setpoint, F2) below the scram bypass setpoint but greater than the bypass valve capacity, and F3) below the scram bypass setpoint and below the bypass valve capacity.

For turbine trips at power levels above the turbine-generator trip scram bypass setpoint (about 30 percent), the reactor is iamediately scrammed and the recirculation pump motors are tripped. A turbine trip from a power level below the total bypass capacity will result in no imnediate change in reactor operation, since no scram should occur and pressure control will be automatically transferred from the turbine control valves to the bypass valves. A small, gradual po increase may occur depending on the degree of feedwater 'hat is discontinued. The sequence is somewhat small "window" be the steam bypass valve oapaeit isb recirculation pump to the bypass valves, w higher than their capac tri diff n the scram b

                                                ~

(assuming no operator ac won). First, sine the eih'd'"' hag'qf t , factors characteriz a g

                                                                             'i initial~

T~ ii'l t b ( because reactor Le e c er lq event steam is being (xe'he and ow is diverted

                                                                                                           +dd produced than can pass t rough the                       s valves, pressure and power will rise. Second, beca                     raction steam is no longer being supplied by the turbine, feedwater temperature will decrease, contributing to an additional, but a more gradual power increase.

.0 1 EVENT CATEGORY This event is classified as an Abnormal Operational Transient.

              ~   ~ ~

ShFE SHUXNSW hMLTSXS kypemdix 2 Page 2 of 9r Wf'repared-Date. ~/7~I( 7 Turbxae Trip E. C. E e Checked: Date: ~ 9 7 BFNWSG3&48 (Rev 1) J. Cheek saaaaaaaassassaaaeseeaaaeaaaaeaeee<+~+~~~+>>+~~~~++~~ 3~0 PLANT PARAMETERS AND SAFETY CONCERNS, 3.1 Pressure Concerns The turbine trip event atd the resultant stop valve closure cause a combination of effects on the reactor vessel pressure. The first is an acoustic pressure wave reflected back up the main steam lines into the vessel, caused by the sudden shutoff of steam flow at the turbine inlet. The second is a more gradual, although still rapid, integral'ncrease in vessel pressure caused by the mismatch between core power and steam flow. Protection is needed to avoid excessive primary system pressure (criteria 2-3 in Section 6) ~ 3.2 Reactivit and Power Concerns Increasing reactor pressure causes collapse of steam voids in the reactor coolant. This effect in the core threatens to increase reactivity and power. Decreasing feedwater temperature has a similar, but much more gradual effect. These power increase aspects of the event may challenge the fuel cladding integrity (criteria 2-2). Reference 6f recognized this criteria for the turbine generator trip event early in BFHP design for all 3 units. 3.3 Vessel Water Level Behavior The pressure trans nt coupled affects water level bu n r ficxent. Long term heat remo 1 complete shutdown. L w /Wd Q.V) tso4km WQ O~f ~+PR&C s~

4. 0 EVENT MITIGATION <~(pr

                                       \

4.1 A licable Safet Actions 4.1.1 Fuel Protection (2-2) Above the turbine-generator trip scram bypass setpoint, trip of the recirculation pump motors must be initiated (with scram) from the stop valve closure (reference 6h) ~ The pump motor trip causes a rapid decrease in core flow and power (Safety Action 24 Power Reduction), protecting the fuel cladding (especially at end of core life) fran the impending pressure-driven power excursion (reference 6d), 1

5hTS SHUTD%% h%bLTSIS kppeax%ix 2 Page 3 of 9 Prepared: Date: // 7 Tarb%ee Trip C. Eck t Checked: E Date'l'/

   ~

BFNWSG3&48 (Rev 1) B. J. Cheek Above the turbine-generator trip scram bypass power level, scram (Safety Action 17) is initiated fran the stop valve closure. This assures that control rods will already be moving in when the pressure effect occurs, are is needed (together with, the pump trip) to ensure that f'uel thermal limits are not exceeded (reference 6d axd 6g). Below the turbine-generator trip scram bypass setpoint, a high neutron flux or high pressure scram may be initiated if bypass valve capacity is exceeded by core power production. The scram limits core power and helps assure fuel cladding integrity. 4.1.2 Reactor Pressure Protection (2-3) The turbine bypass valves will be opened and (if needed) safety relief valves must operate briefly during the early stages of a turbine trip at high power, in order to provide RPV pressure

  ~       4.1.3 relief (Safety Action 18) axd limit the severity of the transient.

Normal Shutdown and Recove For turbine trips above the scram bypass setpoint, the path to normal shutdown is the scram followup procedure. Reactor water level should be maintained by the feedwater system, pressure control by the turbine controls and the main condenser is available as a heat sink. For turbine trips at low power levels (within the bypass capacity), the plant continues to operate with steam passing to the main condenser through the bypass valves (Safet n 22, Restore Normal) ~ The operator superv's ' 'enerat operation, especially preserv transferred to o 'wer. occurs with rbine trips at 1 e~w Loss of e 11 be e iQ effect on f edwate tery p steam flow a)5 ipse .W q, wer. The po r increase sho %t b$ monitored> 'especially position is ear full open imnediately followi if the bypass valv x.p. In that case, t e operator may insert s to reduce power and avoid scram in ant1c1pat1on of restoring the unit to the

                                             ~ ~

line (Safety Action~).

  ~

i SAFE SHOXlX%8 ANALTSIS hppend~ 2 Page 4 of P Prepared: Date: Turbine Trip E. C. c rt Checked: Date: ~~/ F7 BFNWSG3-048 (Rev 1) B. J. Cheek aaaaaaaaaaaaaaaaaaaaaaaaaasaaaaaaaasaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaasaaaaa 4.2 S stems Re uired Table 2-1 summarizes the required system functions. It was collated from event~nique actions and applicable standard sequences from section 7.6 of the main body of the report (Tables 7.6-6, -SP -10, -12, -14', -29) ~ 4.2.1 Power Reduction (24) The power reduction action is performed by sensing the stop valve closure (Main Steam System, Ol). The signal goes to the RPS (99) and the Recirculation System (68). This action is only active if power is above the logic circuit bypass setpoint

             "(about 30 percent power). The bypass logic uses the turbine first stage pressure signal also provided by the Main Steam System,    01.

4.2.2 Scram (17) Prom Sto Valve Closure In order .to prevent the stop valve closure transient from causing core power to exceed thermal limits, stop valve closure is sensed by the Hain Steam System (01) and a signal sent to trip the RPS System (99) so that control rod insertion. (System

85) is initiated quickly, before the pressure transient caused by the valve closure can have a significant effect on the core. This turbine trip scram (together with the recirculation pump motor trip) assures that the power response is mitigated before the pressure induced transient can excessively aff fuel thermal margin.

                                                                                    - fs'4 The stop    valve closur             xs  deactivated at       er below the turbi e -generator        trip scr     b            o'        gout 30 percent powe ) , identi                               b      s permissive for h                                                tur one first   stage pre             so                    ystem, 01.

4.2.3 APRM Flux Scram For partial power ere the stop valve closure scram is not in effect, power will rise when the turbine trip pressure transient reaches the vessel. The APRM signal from the Neutron Monitoring System (92) will signal an RPS (99) scram trip if neutron flux exceeds the setpoint value. (No credit is taken for the flow referenced setpoint).

Sh?K SHUTDOWN hML~ kgqmadxx 2 Page 5 of 9 I Prepared: Date: g~T/~f Toxbime THLp E C~ E k t Checked: Date! 9/ BFNWSG3&48 (Rev 1) B. J. Cheek aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa~aaaaaaa~ 4.2.4 Pressure Scram (17) Vessel pressure may trip the high pressure scram during a partial power turbine trip. This signal is provided by the Feedwater System (03) to the RPS (99) ~ 4.2.5 Pressure Relief (18) -B aes A turbine trip opens the turbine steam bypass valves (Systems 01 atd 47) to mitigate the severity of the pressure disturbance and provide a steam path to the condenser for heat removal aai normal recovery from the trip. Reference 6i defines the design of this function. The potential failure of this system is addressed in Appendix 1B (Turbine-Generator Trip With Bypass Failure). 4.2.6 Pressure Relief (18) - SRV A turbine trip from high power may result in 'a pressure transient which exceeds some or all of the safety relief valve (SRV) opening setpoints (Main Steam System Ol). These valves

• will cycle open aementarily to limit the primary system pressure, passing steam through the SRV discharge pipes (System 10) to the suppression pool (System 64).

4.2 ' Restore Normal (22) At low power (below turbine steam bypass capacity) the reactor pressure is controlled by the bypass valves until normal operation can be restored by reactivating the turbine generator. This involves the stop valve logic in t in Steam System, 01, the Turbine Control System, and' the bypass valves (Main Steam System, 01). 1 inser e control rods (System 85) may also and/or prepare for resynchr 'on tilieed of t o d m 5~0 LONG TERM SHUTDOWN CONS ONS I 5.1 Normal Shutdown The sequence provid n tion 7. 6 of main body of the report for normal non solated shut would be followed, it if -0 necessary (Tables 7.6 ). Ho expected that restorat o r, in this event, the unit to the grid will be possible and full cool own will not be necessary. ie often

SKID SHOaX~ hsaxmxs Thxbxlhe TExp kppendxx 2 Prepared: tc5 E. C. Ecke Page 6 o Date: / 9

                                                                                                  '3/

Checked: Date: ~SP BFNWSG3&48 (Rev 1) B. J. Cheek aaaoooaoaaaoaaaaaaaaaaaaaaaaaomsoaoaaaooaaaaaaaaaaaooaaaaaaaaaaaaaaaaaaaaa Normal, long-tenn response, as discussed in 5.1, utilizes all systems normally available to the operator. If any (or all) normal control equipment is assumed to be unavailable the reactor can still be shutdown to cold conditions, if necessary, by utilizing only safety equipment as given in the safety shutdown sequence in section 7.6 (Table 7.6-3) starting in this event fran a non-isolated condition.

6. 0 REFERENCE S

a. Browns Ferry Nuclear Plant FSAR, Chapter 14; System Safety Ana lys is.

b. Browns Ferry Nuclear Plant FSAR, Appendix G; Safety Operational

       '. Ana lys is.

Browns Ferry Nuclear Plant Questions. FSAR, Chapter 7p Responses to AEC

d. Browns Ferry Reload Licensing Report; Unit 2, Cycle 6, April, 1985.

e. BFN Safe- Shutdown Analysis, Appendix 1 (Revision 0), B. J. Cheek and V. G. Blanchette, Jr.; BFN-OSG3&48, June 28, 1986.

f. NEB-RAC-1593 (B45 860630 61), BFNP D & AR, Amendment 7, Question 4.

g. NEB-DLM-1013 (B45 860628 725), BFNP SER, Appendix A, Section III.A (page 15) June 6, 1968; Addition of Turbine/Generator trip

       '.     (anticipatory) scram.

NEB-RAC-1263 (B45 860624 016), TVA letter dated September 25, 1978 from J. E. Gilleland to H. R. Denton (NRC), BFNP Unit 2> cle 2, Addendum to Supplemental Submittal (NEDO-24095) &ion f T/G trip recirculation pump motor trip.

i. GEC-GDC-1031 (B45 860619 887), G ument 22A1185 (Revi ~ 6),

Design Spenifieation - ne Pressure Rp 1N or a I p ss snl IP I(

7. 0 NON-APPLIOASLZ DPS Sections 4.2 and ~. e!ma~ B . he report show s of an assessment of lic ing) eQnkes a FSAR ques The items which have been id iMed as likely SSA sion 1) inputs for this event are given in Tables 4.2-1 . -1 ~ These references were either used above i th endix or they have been evaluated and not utilized as describ below.

~, ~ ~

- kypend~ Prepared: Date: / 3/ 7 TNxbxDO Tlap Checked: E.. Ecke Date: BFN-OSG3&48 (Rev 1) J. Cheek aaaaaaaassaaaIIeasSIIaaessaaaaINCseaaaaaaaaaaaaaaaaaaeaaeeaeeaWaasSseeaeaINaaaaeaaeaa 7.1 FSAR, Volume 7, Question R10.2 - Although identified as potentially applicable to Turbine/Generator trips with SRV actions, this question really addresses the air actuated, ADS mode of the valves (not- the pilot~perated mode used in overpressure protection cases).

         'I J'
      ~
                  ~ '          . i

QFE I)IIIIIlSLTSIS lRBDIT 02 Page Sof 9 Prep aredx Turbine Trip 'E.C. Qschah nate 4" I7 EF)HS5418 IRev 1) TSE02 -1 REACT QNXEPTARE TN SK IFER. RELENTS Sf'CTIOt, $ 5 STSIE)I FQC STATE Cade Title Cade Title IL NK ted% IXigSIIS F 2-2, Fwl Fail+a 17 Scrae SF Provide > RK turbine first stage prearm inter]act signal to reectcr protectim systee I99) fail~ logic. F 2-2 Fuel Fallm 0 Scree Provide naia tmbine shy valve < 90l qea trip signal to reactcr protectim systee I99). F 2-2 Fuel Failee 17 Serac 3 FEHNTER SF =Provide high reactm'essel to reactor pratectia systee I99)

                                                                                                                 ~efail~  trip  signal logic.

F 2-2 Fuel Failure 0 Scree 8$ 05 SF Scrae signal fran reactcr protectim systee I99) uill activate tbe caetral ral drive systen to insert rods. Scree functim cnly. F 2-2 Fuel Failwe 17 Scrae 92 IBlllQINIOII F Provide N%)1 nestrm flux trip signal to reactor protectim systee I99) fail~e logic. F 2-2 Fw] Failwr 17 Scrae 99 REACII PROIECIN SF Pravide scree signal to the cmtrol rad drive KO) systee NS' eeutrm emitaring systa I92) NI neutrm flux trip signal. F 2-2 Fuel Failure 17 Scrae 99 IEACIN PRSTECIN SF Pravide scrae signal to the cmtrol rod dive III) systee I85)

                                                                                                      'm feedeater systee I03) high reactm vessel pressure trip signal.

F 2-2 Fwl Failure 17 Scrae 99 REACII PIIIECIN SF Provide serac signal to tbe cmtrol rad drive III) systen 185) cn si cating cain ttehine sty valves <

                                                                                %C open and      turbine first st F       2-3 Syst Stress    18   Pres Relief      1 F      2-2 Fuel Failure    18   Pres Relief      1  Il                           n t    ine bypass valves apm m turbine cmtrai systee I47) ttbine        trip signal.

sd xc required by this evwt. 2-3 Syst Stress 18 Pres Relief 10 IeltER NTS hiRN Provide path far cain stean systee I01) Sws F stean blaudaun to aypressim.pool IM). 0

SAFE SUING ASLTSIS NENI1 02 Page 9of 9

                                                             ~

Prepared i Turbine Trip E.C. Echart hecbdt IF)HK&Ol8 Nev 1) Lo TSE02 "I RD INQPTARE TN SER. IRIS SfHY KTIIN $ 5 SSIBI STATE br'itle Mo Title )L NK F 2-2 Fuel Failm 18 Pres Relief 47 llNDE QKRm.. Send signal to open and control cain stean systen i01) bypass valves on tmhine trip. This safety action nay aar, but is not reguired hy this event. F 2-5 Syst Stress 18 Pres Relief Q PRI QKAIMfà Accept SNs stean blwdwn ffron boiler vents and drains systea, 10) to ag~sico fcoi. F .2-2 Fwl Failse 22 Rest Nuraal 1 NN SIM Signal fra Turbine hntrol f47) to bypass valves provides noraal control aft'rip froe lw initial pwer, avoids scraL This-safety act}a nay ocar, but is not required by this event. )g J F 2-2 Fuel Failure 22 Rest)kraal 47 TNSDEIXNRm. Aft'urbine Trip, signal bypass valves to

                                                                    ~ aintain    presare control. This safety acti+
                                                                    ~ ay  ear,     but   is not required hy  Ns    event.

2-2 Fuel Failure 24 Pw Reduce 1 NI)f SIM 'F Provide nain turbine stre valves < %C open trip signal to reactcr protection systen N), F 2-2 Fwl Failure 24 Pw Reduce 1 NIH SIM SF Proride > 30l turbine first stage pressure interlock signal to reactcr prntectim systen f99) fail~ logic. F 2-2 Fuel Failure 24 Pw Reduce N RKIKRATIQI SPNI reclrcuiatiol pimp enter breakers cl RPS f99) signai due to cain turbine stop valves < %C pen and >RL turbine first stage pressure. Codon aust be faster than assueed in analysis to licit severity of . ( S4 + eve 4s OLTI n~ Cg ~py F 2-2 Fuel Failure 24 PwReduce 99 REACTNRHHEI)l SF Send open recirculati ers isystee N) systae I 2-1 WRI~ 2f Estpriut fm am i atiim acti+is) upon ae goal froe the Reactcr System (99).

hMLMXS IIaf of k6 SLICK SHUTÃ%% kppeadix 3 Page 1 Prepared: Date: / 2. 7 ?solatium of hll E~ ck r Hain Steaa Linea Checked: Date: X Z 87 BFN-OSG3&48 (Rev 1) V. G. lanchette Jr. astaaatsaaaaaaaaaaaaaassstaaassaaaaaaaaaaaaaassaaaaaatsaassaast aaaaaaaaaaaaaaaaaaaa 1 0 DESCRIPTION OF THE EVENT When the reactor is at normal operating power, the closure of all Main Steam Isolation Valves (MSIVs) will result in a sudden reduction of steam flow which will create a significant nuclear system pressure increase. A reactor scram results from multiple MSIV closure. After the reactor is isolated and shutdown>core stored and decay heat will cause an increase in nuclear system pressure. Relief valves are utilized to control the pressure level. The reactor vessel water level decreases because the feedwater turbines lose their steam supply after isolation. Core cooling water supply systems are used to maintain reactor vessel water level until normal shutdown level control is established. This appendix addresses this event based on the previous SSA issue (reference 6f) plus new inputs and comments

         'subsequently received.

The main steam line isolation event is defined and analyzed in the FSAR under paragraph 14.5.1.6 of chapter 14 (reference 6a) and in Appendix G (reference 6b). 'References 6c and 6d were reviewed and no material was uncovered which would significantly alter the reference 6a and 6b evaluations. Isolation of all main steam lines is most severe and the resultant pressure rise is'ost rapid in operating state F during power operation. In states C, D, axd E, steam line isolation becomes a lesser or partial case of the potential state F sequence. This event is non-consequential for operating state A. 2 e0 EVENT CATEGORY The BFNP F R addresses this e ato 1 Transient ( efe 3.0 PLANT P 'S ZF P RNS 3.1 Reactor Pressure With the ~actor abruptly isolated through closure of the main steam isolation valves, the core stored and decay heat will casse a rapid increase in reactor vessel pressuxe. The rapid increase in reactor vessel pressure directly threatens the integrity of the reactor vessel and associated primary sy'tem piping (2-3). 1 r ~ r1

                                             ~   '
           ~    ~

e ~

Iaolatxaa oX k11, Hain Stem Xxnee kypendix 3 Prepared: Checked: CC F~ Ecker sass: Dste: > > ll

                                                                                 /~z~/Q
                                                                                         >7 BPNWSQ3<<048 (Rev 1)                              V. G. Blanchette,     r.

aaeaaaaaeaaaaeaaeaesseaaassaeeeeassaaassaeeaaaaaaaaaaeaaeaa aaaassaeaaassaeaaaaa 3 ' Core Reactivit and Power A potentially large reactivity increase and neutron flux peak can occur due to the rapid pressurization and the resultant decreased coolant void fraction. The potential neutron flux and resulting fuel heat flux increase threatens to cause excessive fuel clad temperature and challenges fuel integrity (2-2) ~ 3.3 Reactor Vessel Level The level of the saturated water and entrained steam inside the reactor vessel will initially drop due to the increased pressure, reaching the first group of isolation logic at L3, and providing the first steps in protection against potential radiological release (2-1) or overexposure of control room personnel <3-5) ~ It will then rise due,to the effects of flashing (liquid to vapor) caused by pressure drop from relief valve actuation> axd will eventually again decrease due to the effects of loss of feedwater flow. Lowering of the reactor vessel water level due to loss of feedwater and mass loss through vessel pressure relief valves threatens to uncover the core and therefore threatens fuel integrity (2-2). 3.4 Su ression Pool T erature Suppression pool temperature will rise due to transfer of core stored and decay heat from the reactor vessel through pres relief and core cooling systems actuation. Incre ppressi n pool temperature threatens primary conta overstress (2-3I) 4.0 EVEÃZ MITIGATION 4.1 Re uired Safet 4.1.1 Avoid Puel Pail Upon reactor iso ation, the reactor to be scrammed (Safety Action 1 to rapi ove the nuclear power production from t inputs which contribute to fuel thermal margin (and to the reactor vessel pressure ramp) ~ This

              .must quickly terminate the power transient safely below the amount which could result in fuel failure. Due to loss of normal feedwater, an alternate core cooling (Safety Action 19) water supply to the reactor vessel is -needed to replace the

appendix 3 Page 3 of M~ Date: Xsolation of All Eckert Hain Steam Lines Checked: Dace:Pstt 8 ~ BFNWSG3-048 (Rev 1) V,. G. Blanchett Jre aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaassaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaassaaaaaassaaa water which is lost through vaporization for removal of core stored and decay heat. This must keep the core covered in. this event, thereby avoiding risk of fuel damage. 4.1.2 Avoid Excess Prima S stem Stress (2-3) Pressure relief of the reactor pressure vessel (Safety Action.

18) is necessary to limit the post-isolation vessel pressure ramp to an acceptable value for primary syst: em pressure integrity.

LL P 4.1.3 Avoid Containment Overstress (2- 4lff/f1 Heat which is transferred from the reactor vessel to the suppression pool by vessel pressure relief and through condensation of exhaust steam from the turbine driven water supply systems must: be removed (Safety Action 30) as necessary to maintain the pool temperature below the temperature permitted 'for this plant condition. 4.1.4 Avoid Potential Radiolo ical Release (2-1) In anticipation of any possible radiological difficulties, the initial stages of isolation (Safety Action 20) are to be taken as water level drops below L3. 4.1.5 Avoid Potential Overex oeure of Control Room Person 3"5) l Low water level logic is also used to pro 'nticip t ~

                                         'i protection by initiating             Control Bay            ironment   1 C (Safety Action 36).

4.2 S stems Re uired Table 03"1 summariz e the u nctions. was collated dtos event n applicable andatd sequences from secti t g main body the report

                                    -2 6
                                              ~

(Tables 7.6-7, -ll, ~ , -18, -20 ). 4.2.1 Reactor Scram (Safet Aetio Position switches on the main steam isolation valves, which are a part of the Main Steam System (Ol) provide a signal for reactor scram to the Reactor Protection System (RPS, System 99) V

                                     ~   ~ e

II, g sept sHOTDonm kMLTszs nppetoiiu 3 Page 4 o Prepared: Date: Isolation of h11 E. Ec r Wain Stcam Lines Checked: Date: 4-//-E7 BFN-OSG3-048 (Rev 1) V. G. B anchet asasaassaaaaaasaassasaaaaassaaaasasasasaaaaaasaaaaasasssasaaaaaaasaassassss if valves in three or more main steam lines ere less than 90 percent open and if the reactor mode switch is in the "RUN" position. The RPS signals the Control Rod Drive System (CRD, System 85) to initiate the rapid insertion of the control rods (Scram) . If the power is low and the reactor is not in RUN mode, scram would still be needed, but would most likely be initiated via high neutron flux (Neutron Monitor System, 92) as measured by the IRM instruments (possibly the APRM instruments). 4.2.2 Pressure Relief (Safet Action 18) Following isolation valve closure, the Safety-Relief Valves (S/RVs) will automatically open and cycle as necessary to maintain the reactor vessel pressure within safe limits. The S/RVs are part of the Main Steam System (1). The discharge piping (System 10) and the suppression pool (System 64) support this action. 4.2.3 Core Coolin (Safet Action 19) Reactor water level instrumentation, which is a part of the Feedwater System (03) provides trip signals to the high pressure reactor cooling systems upon low reactor water level (L2) . The High Pressure Coolant Injection (HPCI) System (73) end the Reactor Core Isolation Cooling (RCIC) System (71), which provide single>>failure backup to each other, are actuated to maintain the level during the initial part of the event. Some of the systems which support RCIC and/or HPCI operation are also shown in Table 03-1. 4.2.4 Containment Coolin (Safet Action 30) w! tuppreeeion pool tempereture e vel monitors w

                                                                                      $  au in the control room,'r             ovided as p Containment S s             4) ~  When                s n       1 Alelperatur approaches   th   -normal    o                      t         xt, the RHR System (74) mu     t                             6 o'he   containmen cooling mode (          o                   quired, with on          ing as e i

single failure aVVppP~ A rief descripti

                                                                  'sing        this mode of transmittals) is the RHR (as originally described i~n provided in reference 6g. S                7 the systems which support this mode are als'p     sh~       n Table 03-1.
 ~

SAFE SHOTD%% ANALYSIS Appendix 3 Page 5 of $0 Prepared: Date: Isolation of dll . Eck r Hain Stcaa Lines Checked: nore: p-I/-87 BFNWSG3-048 (Rev 1) V. G. B anchet e, Jr. asaaaaassaaasssaaassaasaaaasasasassasaasaaaaaaasaaasssassassaasassaasaasass 4.2.5 Isolation (Safet Actions 20 26 and 27) Water level is expected to drop below L3 (sensed by system. 03, Feedwater), and System 64 initiates several isolation actions. Table 03-1 shows the systems that support these actions. They are not required for this event since no fuel failure occurs. 4.2.6 Control Ba Environmental Control (Safet Action 36) Low water level (L3) signals are also used to initiate steps for Control Bay Isolation. Systems 03, 99, 64, and 31 participate as shown in Table 03-1. Again, this action is not required for this event since no fuel failure or radioactive material release occurs. 5.0 LONG-TERM SHUTDOWN 5.1 Normal Isolated Shutdown The steps provided in Section 4.0 achieve stable, hot shutdown conditions. Should the su'ppression pool temperature approach upper acceptable limits, manual depressurisation and cooldown may be initiated'he sequence of equipment required is given in section 7.6 of the main body of the report for this standard path if needed (Table 7.6-1). 5.2 Safet Shutdown Normal, long-term response, as given above, ut normally available to the operator. If or

                                                                                     's all) sys n

s control equipment is assumed to be ailable, .Qhq a'qfo can still be shutdown by utilizin y safety equi' gi n in the Safety Shutdown sequ also giv~j . ab

6.0 REFERENCES

e. Broene Ferry N 'eri)114AR, Section ..1.6.

b. Browne Ferry N e Alhnt FSAR> A xx G, Section 5.3.5.2, Event Number 14

c. Browns Ferry Nuc ear Pla AR> Volume 7, Responses to AEC

             , Questions.

d. TVA-RLR-002, Revision 1, APril> 1985, Reload Licensing Report for Browns Ferry, Unit 2, Cycle 6.

e. OE Calculation BFN-OSG3-037, Master Components Electrical List (MCEL) Design Basis, December 17, 1985,

                                            ,4 ~

p1 ~ e e e e

                                              ~     ~ e

0 ShTK SHOTD%% kMLWIS appendix 3 Page 6 of ~

                                                                                   /I gw Prepared:        Fw    Ecker Zsolat,iaa oE h11 Main Steaa Linea BFNWSG3-048 (Rev 1)

Checked: V G. B anchett, Jr. Date.- Z-9/7 saasaaaaaasasaaaaassaasassssasssassaasasssaaaaaaaaas saaaasaccaaaasssssssascc

f. BPN Safe Shutdown Analysis, Appendix 3 (Rev 0), J. C. Kobold and M. C~ Olich; BPNWSG3~8, June 28, 1986.

g. NEB-RAC-1756 (B45 860701 717), TVA letter dated May 31, 1968 from R. H. Marquis to Secretary of the AEC, Exhibit 1, Description of Safety Punctions of RRt (and other) systems.

7. 0 NON-APPLICABLE INPUTS Sections 4.2 and 4.3 of the main body of the report show results of an assessment of licensing references and PSAR questions. The items which have been identified as likely SSA (Rev 1) inputs for this event are given in Tables 4.2-1 and 4.3-1. These references were either used above in this appendix or they here been evaluated and not utilized as described below.

7.1 BPNP, FSAR Volume 7,,Question R10.2 - Although identified as potentially applicable to the MSIV closure event, this question really addresses the air actuated, ADS mode of the valves (not the pilot-operated mode used in vessel pressure protection in this event).

SAFE Sama A)Nf.ISIS APPE)mTI 03 Page 7 of 11 Prep ~) 1solatim of All Nain E. Qeckab ta>ZD-37 MHX&08 IRev 1) V.6. Blanchettex Jr.

  $ EACT  tNCPTARE                           TN 5%L       KRLIS          SAFETY ACTIQI     STS     STSTBI STATE Code   Title       Code     Title    NL       Nf F      2-2 Fuel Failure 17 Scree               56N  STEAN            Provide )SIV < 9R open        trip   signal to reactor protecticn systen 89)        fail~ logic.

IF 2-2 Fuel Failure 17 Scran 85 GS' SF Sera signal fron reactar protectim systen 89) sill activate the cmtrol rod drive systen to insert rods. Scrae fmctim only. D 2-2 Fuel Failure 17 Scraa 92 IBl1HN )QGIIR SF Provide IW neutrm flux trip signal to reactcr protectim systea 89) (see sectim 7.6.2). F 2-2 Fuel Failure 0 Scraa 99 KCHR PHOIEIN SF Provide scrae signai to the cmtnd rod drive IGS) systen (85) m naia stean systen (01) )SIVs

                                                                       < 90I quan trip signal (if in RN node).

D 2-2 Fuel Failure 17 Scraa 99 REACTIR PRDIECIN cf'rovide scraa signal to the cmtrol rod drive (0$ ) systea lfm) cn neutrm wxritoring systen 192) HN neutrm flax trip signal. GEF 2-3 Syst Stress 18 Pres hlief 1 NIN SIGN SF HtVs open cn high reacte pressure. GKF 2-3 Syst Stress 18 Pres hlief 10 101LERVNISbNI Provide path for aain stean systee (0)) SVs stean bloudao to suppressim pool lbl). CKF 2-3 Syst Stress 18 Pres hlief M PRl GNAIST Accept SRVs stean bloudoun ffron boile'ents and drains systen, 10) to suppresaim pool. CKF 2-2 Fuel Failure 19 Core Cooling 1 NIH STM Provide stean fear KIC Ol) turbine. I CKF 2-2 Fuel Failure 19 Core Cooling 1 NIN SKNl Prwide stoa fa lfC1 173) turbine in support of lfCI initiaticn m lm eater level IL2). IXEF 2-2 Fuel Failure 19 Core Cooling 2 CMENSATE Provide ncraa) ly open ua'.er RCIC init'r systen Oi). GEF 2-2 Fuel Failure 19 Ccrc Cooling 2 CMENSAIK Pr nally open ea OS) GG 2-2 Fuel Failwe 19 Care Cooling 3 ffBN 1 1 signal L2) 0 'c fcr K1C syst tia

SAFE MIIIMA)NLYSIS APPEM)II N Page 8 of 11 Isolatim of All )fain E.C. Eckert Stean Uncs (Rev 1) Omckah V.L 8)anchette, Jr. hte: ~Z~ TRE O3 -1 NCT SNXEPTARE TVA 5%L REM.TS SAFETY ACTIIN SYS STATE Code Title Code Title N). GEF 2-2 Fml FaiILre 19 Core Cooling 3 FEHNATER Provide path fm RCIC I71) Bow to the vessel through the feedwatsr spargers. GEF 2-2 Fuel Faille 19 Core Cooling 3 FKNIER SF Provide low reacttr water level signal 62) to IfQ systen 03). IXEF 2-2 Fuel Failmlg Ccrc Cooling 3 FEBNTER Provide path fm'fCI l73) fi to the vessel the feedwabr spargsrs fcr IPCI Initiatim m Im water level 62). IXEF 2-2 Fuel Failure 19 Care Cooling 6l PRI GIITAINSIT Provide alternate sarce water supply fcr KIC systen I71) through Ccrc Spray systen 175) piping frm aqressim pool. Provide suppressicn pool Ieve) Indicatim. Accept RCIC turbine exhaust steaL 2-2 Fuel Failure 19 Ccrc Cooling 9 PRI GNAINBIT Provide alternate'sarce water sqyly fcr IfCI systen 03) free agressim pool fcr IKI initiatim m Ia water level IL2) signal. Accept IPCI turbine eshaust. steaa GEF 2-2 Fuel Failure 19 Core Cooliog 71 RCIC SF Provide HXS NU power to Rf systea ILT) low water level 62) Instruaentatim fcr KIC systen iaitiatim. GEF 2-2 Fuel Failure 19 Core Cooling 71 KIC SFS KIC initiatim m lm reactor water level 62) signal fron M systen I74). G)EF 2-2 Fuel Failve 19 Ccrc Cooling 71 RCIC SF Provide HXS Nll power to F)I systea I03) low water level 62) instrueentatim fa IKI systen

03) initiatim.

IXEF 2-2 Fuel Failure 19 Ccrc Cooling 73 IPCI SFS IPCI initiatim m Ri water level 62) GEF 2-2 Fuel Failure 19 Core Cooling 74 M SF ter cyst I evel signai GEF 2-2 Fuel Failure 19 Core Cooling 75 GNE SPRAY 'de th ternate sorce of IC I) frm suppressi 1 N). REF 2-1 bd Release 20 RPV lsol 8 R)G) ose RN) isolati ves cn Im wabr level

 ~                                                                           IL3) sign          prinary cmtainant systen N).

ety acticn is epected to occm,.but is a requireeent fcr this event since there is no fuel failtre.

t SAFE ISMM NESTS APPBOIT N 1)ate I lsolatim of All I(ain C. Ec Stean Lines (hechda oatn p~~- r IF)&K~ (Rev I) U,L Manchette, 'I %EN -1 REACT INCXHAKE TUA (PER. RES(LTS SAFETY ACTTS SYS SYSIB( STATE Code Title Cade Title N4 NK cC / isolaticn signal tripped a GEF, 2-1 Rad Release 20 RPU lsol 7l RIR SF RIR lw uater level (Q) signal frm prinary cmtainaat systen N). This safety acticn is espectsd to acar, but is not a requireast fcr this event since there is no fuel failm. IXEF 2-1 Rad Release 26 Est Pri Cmt 5 FEB)NTER 5F Provide lw eater level (Q) signal to ra@br pr act(m systen (99) fcr initiatim of LE isolaticns. This safety actim is wpected to ocor, but is not a requireeent fcr this event since there is no fuel failure. GKF 2-1 RadRelease 26 Est Pri Cent 2 CMKLAIR Perfcre isolatim actim(s) upon receiviag lw uater level (Q) isolatim signai frm the Prlaary Containaent systen (6(). This safety 0 actim (s expected to occx) but is aot a requireeent fcr this event since there is no foal fails e. C(EF 2-1 Rad Release 26 Est Pri Cant 6( PR1 GMIAINN 5F ~ LM. (Q) signal isolatim actims frw RPS (99), initiate Q b send priaary/secmdary cmtainaent isolaticn signals to systoes 32, 65, 69, 74, 75, 76, 77, Bl, 90, and 94 This acticn is epected to occur,but is not a requiraaat fcr this event. (XEF 2-1 Rad Release 26 Est Pri Cent 7S ONE SPRAY 5F uater 1 'gnal Perfan isolatim acticn(s)

                                                                                 )nary Cmtainaent systen upon N).

frm the Thi tim is expected CHIEF 2-1 Rad Release 26 Est Pri Cmt 76 Al 1 fva isolaticn actim(s) upon receiviog lw uater level (Q) isol unsent systee N). This safety action is wpected to concur, but is eot a

                                                                              'requireeent for this event since there         is'o   fuel failure.

(XEF 2-1 Rad Release 26 Est Pri Cmt 77 AA(IASTE 5F Perfea isolatim acticn(s) upon receiving lw uater level (Q) isolation s(goal free the Prieary Cmtaineent systee N). This safety act(m is espected to omr, but ls not a requireaet fcr this event since there is no feel failure.

SAFE QUEO(H N(ALySIS NPE)all a( Page 10 of 11 lsolaticn of All I(ain Giectedi . et~ a mzv F)H663~ (Rev 1) V.S. Slanchet te, I TARE (6 -1 i8D INKPTARE 74 (PER. RESKTS SfHY ACTIW S6 SSTE)( STATE Code Title Code TNe )L NNf GE 2-1 Rad Release 2L Est Pri Cmt Sl CA0 Psrf ere isolatim actim(s) upon receiving its water level (Q) isolatim signal frm the Prieary Cmtainaent systen (M). Ms safety actim is expected to occur, but is not a requireeent fm this event since there is no fuel failure. 2-1 Rad Release 2h Est Pri Cont 85 (I SF hr fora isolaticn actim(s) qm receiving scraa signal fran the Reactcr Protectim Systen (99). GEF 2-1 Rad Release 26 Est Pri Cent 90 RAMATIQ( 19(1TN 8: Perfcrn isolatim actim(s) upm receiving lm water level (Q) isolatim signal fran the Priaary Cmtaineent systen N). This safety actim is espected to occur) but is lot a requireeent fm this event since there is no fuel failure. d>>alikia) gLgdt7 2-1 Rad Release 2h Est Pri Cmt 9l TlP SF initiate TlP uithdraual~m los uater level (Q) signal fran prieary cmtainaat systen (bl). This safety actim is apected to occir, but is not a requirmnt fcr this event since there is no fuel failure. IXEF 2-1 Rad Release 2b Est Pri Cent 99 REACTIR PRO)ET)I  % Provide lw uater level (Q) signal froa feeduater systen (0$ to prinary cmtaineent systen N) fcr initiatim of Q isolatims. This safety actim is expected to mar, but is not a requirenent fir this evmt since no fuel failire. GEF 2-1 Rad Release 27 Est Sec Cmt M PRI GNTAIMKS erfm ~ isolatim actlm(s) uater l~l m . i I~i

                                                                                                ~

snot a

                                                                               >'r+ryet
                                                                                   ~

Ar is event since there is no

                                                                               'ailure.

GEF 2-1 Rad Release 27 Est Sec Cent h5 SST plant start cn lw uatsr level (Q) signal fran priaary cmtainnent systen (bl). This safety actim is epected to ms', but is ant a requirmnt fcr this event since there is 0 no fuel failure.

                                                         .r
                                           ~ %t  ~

0 SAFE BUflQN AN.TSlS lsolaticn of All )Lain Tf)HE&OS tRev 1) APPE)$ 1I O3 Page Prq~~x 11

                                                                           'EC of 11 V.L Nancfatte, san sat
                                                                                                               ~
                                                                                                               ~447 TIRE  t)3 "1 REACT  lMNXEPTIRE                          TVA IVER. KkLTS          SAFETT ACTTQI     STS     STSTEN STATE hde Title          hde     Title     )L       NA)E hnt holing     23 R)%%           )SF   Suppart'R)gt systen      04) suppressicn paul cooling
                                                                      %dec fXEF  M    Caat Stress   30  Cent Caoling   64 PRI GNM)tM           Provide sappressim pool teeperature and level icatim 'to support R)R systen 17f) suppressim pool cooling cade>      abc    ~e       Cca/i~g.   &K COEF  M    Cant Stress   30  hnt holing     67 EEDf           %F Suppcrt RN systea 17i) suppressim paul cooling
                                                                      ~ aden lXEF  M    Cant Stress   3O  hat   Cooling 7$  M              KF Provide suppressim pool cooling fmctica.

G6 3-5 Para Overexp 36 Cent Say Env 3 FEHNTB SF Provide lw water level fL3) signa) to IfS systea lgg) far initiatim of cmtrol bay isolatim. This safety actim is expected ta occur, but is not a requireant far this event since there is no fuel failure. xF Air cond. 1AD supply ducts isolate b Eeerg. Pres. Systea Ihntrol Rane Eeerg, Vent. Systee) supplies pres. filtered air to KR m tM. IL3) sly'ran systea M. This safety acticn is expected to accN, but is not a requirewmt fcr this event. REF 3-5 PersOvxrexp 36 hnt Say Env M PRl GMTAINN SF Upan lcxx eater, lwel L3) signal froa HPSPB), send isalatim signal to air cmditiming systen 131). This safety actim is expectai to accN, but is not a requireeent fm this event since there is no fuel failure. CCEF 3-5 Pers Overexp 36 Cont Say Env 99 REACT%PROTEIN e ou eater level tL3) signal fran feedeater systen c ~

                                                                     ,(Q) th s event since there ue    failure.

kgxpeodix 4 Page 1 of 5 Prepared: Date: ~M7 Cloaxxre St~ of'ne ?fain Xsolatioa Valve BFNWSG3"048 (Rev 1) Checked: E. C. Eck

                                                            /

S. K. Mehta r nete: aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaateaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa

                                                                                  ~7

1.0 DESCRIPTION

OF THE EVENT There are four main steam lines which conduct steam from the reactor vessel to the turbine. Each line conducts steam from the vessel (common source for all'our lines) to the equalization header near the turbine where the steam from all four lines joins before passing into the turbine stop and control valves, or bypass valves (reference 6d). Two large isolation valves are located on each main steam line'. one inside primary containment, axd one outside primary containment. Limit switches mounted on each valve detect when valve position is less than 90 percent full open. If the appropriate number and combination of valve closures satisfies the trip logic (reference 6e, sections 5.2.3.4 axe 5.2.3.5) while the reactor is in RUN mode, a scram is initiated. However, a nuclear system transient can occur because of an unplanned closure of one MSIV which will not directly initiate a scram. This event can be initiated by actuator circuitry or operator action. FSAR Appendix G identifies this case as event 15. This appendix is based on the previous SSA Appendix issue (reference 6f) plus new input and comments subsequently received. When a main steam line isolation valve closes, flow in that steam line is stopped, although the isolated steam line both upstream and downstream o f the closed valve remains pressurized due to the steam volume common to the vessel and turbine throttle pressure equalization header. The valve closure increases the effective resistance of the steam flow path to the turbine, resulting in higher pressure in the reactor and increases steam flow in the remaining open lines. As long as reactor power is low enough, below appro ' 90 pe ent power, a single MSIV closure should not

                                               'down                a scram,    in fa~t      it is purposely avoided to prevent                            durin val            lg~

closure testing. Abov xxmately 90 pe tfe e in pressure in the essel caused b r 1 i a scram trip on high eutr 1 e n 14..1.6.2). The only condition e h a i n MSIV causes a significant trans ien pta 0 n ar f'ull power eference 6b, section opera, x.e., greater than 90 pere n ~ ~ ~ 2.0 EVENT CATEGORY The BFNP FSAR addre'sacs this event as an Abnormal Operational Transient.

0 kypcmHx 4 Page 2 of 5 Closure of One Sax@ Prepared: Fc. E. C. Eckert Date: ~la /7 Steaa Xaolatxon Valve Checked: Date: ~I3o g> BFN-OSG3-048 (Rev 1) S. K. Mehta 00000000000000000000000000000000000000000000000000000000000000000000000tt000 3.0 PLANT PAREMETERS AND SAFETY CONCERNS 3.1 Steam Flow and Pressure Closure of one MSIV will transiently decrease steam flow and thereby increase reactor pressure. It is most significant in reactor operating state F. Vessel (and steam line) pressure peaks four to five seconds after onset of the event but should remain below the pressure scram and the relief valve settings (reference 6a, section 14.5.1 ~ 6.2). Steam flow in the three open lines increases, but is expected to stay below the high steam flow trip which could cause full isolation. All other operating states do not involve significant main steam line steam flow, and there are no pressure effects for single MSIV closure in those states. Therefore there are no significant reactor pressure concerns during this event. Normal control is maintained by the turbine pressure controls (whether scram occurs or not). 3.2 Reactivit and Power A single MSIV closure will result in a power increase due to the pressur'e change induced by the closure. At high initial power (high steam flow), the resulting power change may be sufficient to trip the RPS System on high APRM flux. This transient challenges the fuel thermal limits, there fore, the principal safety concern is fuel damage (unacceptable safety result 2-2). 3.3 Reactor Vessel Level This event does not significantly affect level unles sc m results, then the vessel level will undergo t rmal sera level transient. Level control will be the reacto Sp+ ater

                                                                                    ~

system. La ~aNaJCC>) CJsk ae'smQ; m ~~ asm 4.0 EVENT MITIGATION 4t.l A licatle Sa t Action 4.1.1 Provide Fuel 4 n Increasing po adhociated wit e MSIV closure pressure increase mast be detected, ' i f large enough to challenge the fuel, scram mush. be xated (Safety Action 17) ~

hppeadix 4 Page 3o 5 Prepared: Date: / z 7 Cloaure of One lfain E. C. Eckert Steam Zsolaticm Valve BFNWSG3W48 (Rev 1) Checked: S. K. Mehta Date: ~e7 aaaassaaaaaaasaaaaaaaaassaaaaasssassaaaaasaaaaaaaaesaasaaaaaaaaaaasassasassa 4.2 S stems Re uired Table 4-1 summarizes the systems required to perform the required safety action. The required system functions were collated fran event unique action and the applicable standard sequence from section 7.6 of the main body of the report (Table 7. 6-8). 4.2.1 Scram (Safety Action 17) The series of actions described below pertain to reactor operating state F at power levels over 90 percent. In states C, D, E, azd F (less than 90 percent power), the same event does not require a scram, and normal operating conditions can be restored using manual procedures and normal controls (pressure, feedwater). The Neutron Monitoring System (92) provides the high neutron flux (APRM) signal to the Reactor Protection System. When the neutron flux level exceeds the high flux scram limit, the NMS sends a signal to the RPS (99). The RPS provides a scram signal to the Control Rod Drive System (85) The rising neutron

                                                                   ~

flux may have exceeded the lower, flo~ariable scram trip setting, however, the input to this trip logic is the thermal power monitor signal (reference 6g) ~ It responds approximately like fuel surface heat flux (much slower than neutron flux) and is not expected to cause scram. In fact, it helps avoid scram for the partial power testing of the valves. Upon receiving a scram signal from the RPS, the Control Rod Drive System (85) causes all control rods to be rapidly inserted into the reactor. 5e 0 LONG TERM SHUTDORH CONSIDERATIONS

                                                                              'f-  >,'E 5.1   Normal Shutdown                                                           ~

The sequence provided in sects 7.6 for i

                                                                            ~

lated shutdown (Table 7.6-2) e foll y. Ho weve if the scram occurs in s event, unit to the grid will b b hat resto full coold will not on of be necessary. 5.2 Safet Shutdown Normal, long-term re ponse, a xven above, utilizes all systems normally available to th~&perator. If any (or all) normal control equipment is assumed to be unavailable the reactor can

, SAFE RIQX5% hMLTSIS appendix 4 Page 4 of 5 Prepared: Date: ~/~~/I7 Cloeare of One Hain E. C. Eckert Steaa Xsolation Valve BFNWSG3-048 (Rev 1) Checked: S. K. Mehta Date: l~ag ssssaaaaaaaassaaaaaaaaaaaaaaaaaasasaaaaasssssssaaaaaaaaaaaaaaaaasaaaasas aa still be shut down by utilizing only safety equipment as given- in the safety shutdown sequency in section 7.6 starting in this event from a non-isolated condition (Table 7. 6-3) .

6. 0 REFERENCE S

a. Browns Ferry Nuclear Plant FSAR, Chapter 14.5; Analysis of Abnormal Operational Transient s.

b. Browne Ferry Nuclear Plant FSAR, Appendix G; Event 15; Plant Safety Operational Analysis.

c. Browns Ferry Nuclear Plant FSAR,. Volume 73 Response to AEC ll;

                                                                                    'uestions.

d. Browns Ferry Nuclear Plant FSAR, Chapter Figure 47W801-2.

e. Browns Ferry Nuclear Plant FSAR, Chapter 5.23 Primary Containment System.,

f. BFN Safe Shutdown Analysis, Appendix 4 (Rev 0), L. E. Pohl and V. G. Blanchette, Jr., BFNWSG3-048, June 28, 1986.

g. NEB-RAC-1239 (B45 860623 687), TVA letter dated October 15, 1982 from L. M. Mills to H. R. Denton (NRC), (Enclosure 2, pages 3, 4)',

Addition of APRM Thermal Power Monitor to BFNP - Unit 2 ~ 7.0 CBt APPLICABLE INPUTS Sections 4.2 axe 4.3 of the main body show results of an assessment of licensing references and FSAR questions. No items have been identified as likely" SSA (Rev 1) inputs for this event, however, one general item (reference 6g) was found to apply.

Sff SWAN AN.YSIS lsolatim of 0ne Hain APPE)I]E 04 pqgrg> F~ Mt'A E.C. Ec'ka't Sectedi htn ~zl x7 NHKH~ tRev 1) S.y I)shta TSEM -1 KtT MXEPTNKE TN 5AFE CPER. RERLTS SAFETY ACTlW 55 STSTE)) RK STA1E Cade Title aa. Title 16. m Cm F 2-2 Fuel Failure 17 Scrae 8S 05 Scran signal frm reacts protectim systea N) sill activate the cmtrol rod drive systen to insert rods. Scrae fmctim mly. F 2-2 Fuel Failure 17 Scrae g2 IBllMNl96lN SF Provide APRN nentrm flux trip signal to reactcr protectim systen igg) fail~ logic. F 2-2 Fuel Failure 17 Scrae gg REACHR RNTKTN SF Provide scran signal to the ccntrol rod driw lOI) systen lED m neutrm emitcrieg systen f92) APR)) neutrm flux trip signal. F 2-1 Rad Release 26 Est Pri Cent 85 DI 5F Perfan isolatim actimfs) upon receiving scree signai fron the Reactcr Protectim Systen f99). 0

(4 Qqemdix 5 Page 1 of .P6 ~g Prepared: nsse: ~/ of Coadeaser Vaanm E . Eckert Checked-Jr. Date: ~~SQ BFNWSG3W48 (Rev 1) V. G. Blanchette aaaaaaaaaaaassaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa

1.0 DESCRIPTION

OF THE EVEÃZ The main condenser provides the normal heat sink for the last stage of the main turbine is exhausted reactor'team. The steam from the into the condenser. Condenser circulating water is used to condense the steam. The Loss of (bndenser Vacuum event is defined and analyzed in the FSAR under paragraph 14.5.1.12 of chapter 14 (reference 6a) and in Appendix G (reference 6b). References 6c and 6d were reviewed and no material was uncovered which auld significantly alter the reference a and b evaluations. Reference 6h points out the plant Emergency Operating Instructions for loss of condenser circulating water pumping. The plant is assumed in the FSAR analysis (reference 6a) to be at full power when the postulated event occurs. It is conservatively taken to be an instantaneous loss of condenser vacuum. This results in a transient which, while similar to a turbine-generator trip coincident with failure of bypass valves (Appendix 1B), also involves loss of feedwater. The reduction or loss'f vacuum in the main turbine condenser will close the main turbine stop valves, trip the feedwater turbines, are inhibit the bypass valves from opening or close these va'ives if vacuum has esentially all been lost. The turbine stop valve closure will lead to reactor scram and trip the reactor recirculation system pumps. BFNP documents (see references 6e, aced 6f) also show a signal for reactor scram generated directly from the loss of vacuum. This signal is currently in the design of the plant. However, the vacuum switches have been deleted from the technical specification and the signal path, may eventually be deleted. Since it is in the design at this time, it is included here, but with appropriate notations. No credit is taken for this signal in the FSAR evaluation (where instantaneous loss of vacuum is assumed). It would affect any event e gradual loss of vacuum occurs. A loss of vacuum can occ ny time steam pres and is therefore appli able to operate transient becomes a le er i t4 w c the reactor is at low power i need for scram protection in states C 2.0 EVENT CATEGORY The Loss of Condenser Vac event is classified in the FSAR as an Abnormal Operational Transient (reference 6b).

                                                                                            . I+g<-

ShPE SHUN~ MALMXS ~mad~ 5 Page 2 of M Prepared: Date: ~/7O dr7 Loca of Coa8enser Vaea~ E. Eckert Checked: Date: +~3- '7 BFNWSG3-048 (Rev 1) V. G. Blanchette, Jr. gaaaaaaaaaaaagaaaaaaaaagraaaaaggagggaaggaaagsaaaggaaaaaaggaaggaaa aagaaaaaaggaaaaaaaaarga

3. 0 PLAHZ PARAMETERS AND SAFETY CONCERNS 3.1 Main Condenser Vacuum The event results in an instantaneous loss of condenser vacuum.

The loss of proper vacuum and potential pressurization of the condenser threatens the integrity of the condenser and therefore threatens the release of radiological material to the environs (2-1) . 3.2 Reactor Pressure With the reactor abruptly isolated through closure of the turbine, stop valves and the bypass valves, the core stored and decay heat will cause a rapid increase in reactor vessel pressure. The rapid increase in reactor vessel pressure directly threatens the integrity of the reactor vessel ard associated primary system piping (2-3) ~ 3.3 Core Reactivit and Power A large reactivity increase and neutron flux peak can occur due to the rapid pressurization and the resultant decreased coolant void fraction. The potential neutron flux increase threatens fuel integrity .(2-2) ~ 3.4 Reactor Vessel Level The level of the saturated water/steam inside the reactor vessel will initially drop due to the increased pressure. Level will then rise due to the effects of flashing (liquid to vapor) o pressure drop from relief valve actuation and will e ually again decrease due to the effects of loss of feed er flow Loweriog of the reactor vessel water level d o loss feedwater and mass loss through vessel pr ure reli f ng to uncover the core and therefore t r t ~ ens fue l

                                                                                          ~

(2-2).

SAFE SHUTDORR AHALYSIS

                                                                                    ]+ .~g Appendix 5 Prepared:

Loss of Coadenser Vacua Checked: E.. E k r Date: ~gl-Y7 BFNWSG3-048 (Rev 1) V. G. Blanchett, Jr. aasasasaaaaaaaraaaasaaasasaaasassaiasaaasasasassasssassaaasaaassssaasssssas 3.5 Su ression Pool Tem erature Suppression pool temperature will rise due to transfer of heat from the reactor vessel through pressure relief and core cooling systems actuation. Increased suppression pool temperature threatens primary containment overpressure (2-4) ~ ,

4. 0 EVENT MITIGATION 4.1 A licable Safet Actions 4.1.1 Radiation Release (2-1)

Upon detection of low condenser vacuum, the reactor vessel must be isolated from the turbine and the condenser to protect the condenser from pressurization and potential rupture (Safety Action 20, RPV Isolation, and Safety Action 17, Scram). Several anticipatory isolation and control bay environmental actions are initiated, but are not required by the event. 4.1.2 Fuel Protection (2-2) For turbine trips at reactor power levels greater than 30 percent, the recirculation pumps are tripped to rapidly decrease the core power and the rate of the pressure ramp (Safety Action 24, Power Reduction). This action is taken (with scram) to provide full assurance, for the worst case fuel control geometry at the end of cycle conditions, that the transient will be terminated without fuel damage. For this type of isolation situation, the reactor must be scrammed (together with recirculation pump trip ab percent power) to rapidly remove the nuclea 'ion power production input from the heat input thermal margin (and to the rea ch contribu vessel pre g 1 (Safety Action 17, Scr . is termi a e e t at a level safe ow that w , e fu failure. Upon loss of fee wQeQrQnk$ 'tne e cooling wet pply to the reactor vesse (Safety Action 19, Cor o ing) is needed to replace the water which is lost ough vaporization which removes core stored and decaye eat, thereby providing longer-term fuel protection.

SAFE SHUTDOQH hMLYSIS hppendix 5 Page 4 sf A~ Prepared: Date: Loss of Condenser Vacua E C. ck rt -W Checked: Sate: ~>-) BFNWSG3-048 (Rev 1) V. . Blanche e, Jr. aaooaaoaaaaaaaaaaaaoaaaassaaaassaossaassssaaaoassoaatsaoaaaaaaaaaaaaaaastaaaaossaaoo 4.1.3 RPV Overstress (2-3) Pressure relief of the reactor pressure vessel (Safety Action

18) is necessary to limit the post-isolation vessel pressure transient to an acceptable value for primary system pressure integrity.

4.1.4 Containment Overstress (2-4) . IR( Heat which is transferred from the reactor vessel to the suppression pool by vessel pressure relief and through condensation of exhaust steam from the turbine driven water supply systems must be removed as necessary to maintain the pool temperature below the temperature permitted for this plant condition (Safety Action 30, Containment Cooling) ~ 4.2 S stems Re uired Table 05"1 summarized the required system functions. It was collated from event~nique actions and standard sequences from section 7.6 of the main body of the report (Tables 7.6.-6, -8,

            -10, -12, -14, -17,,-18, -20, -25, -29, -34).

4.2.1 Reactor Vessel Isolation (Safet Action 20) The instrumentation that senses low vacuum in the condenser and signals the closure of the Turbine Stop Valves =(TSV) and bypass valves is part of the Turbo-Generator Control System (47). The turbine stop valves and bypass valves are part of the Main Steam System (Ol) as are the valves that achieve the feedwater turbine trip. The corresponding sensors and instrumentation logic for the feedwater turbine trip are part. of the Re or Feedwater System (03) ~ Several anticipatory isolatio ons are also gati~ ct hen water level drops to stinted by gy

                                                       ~ (sensed b      p       t,j~+

j.y a'ignals/acti these L3 ac ions by var P t safety acti ns r ..fey gb thPs event since no fuel failure a @pari'a release occurs 4.2.2 Power Reducti ~et A t n 24) Upon closure the ne stop valve ( when reactor power is greater than 3 percent), the Main Steam System (01) will sense the stop valve closure and signal the Reactor Protection System

SAFE KHJKXNH AKLLYSIS Appendix 5 Loss of Condenser Vacua Prepared: cE. Ec t Page 5 o Date: BFNWSG3-048 (Rev 1) Checked: V. . B anchette, Jr. Date: P~ 500500$ t'C$ $ 0CXlfCROE5CSQ'mm@CES$ 00N@tl55RRQCIQSCI@0105C5'N500000I2ISCR555SISS50kS

             $                                                                         0 1 SsiSSsi (RPS, System 99) to initiate a              recirculation  pump   trip. System 68 (Recirculation) accomplishes              the required actions.

4.2.3 Reactor Scram (Safet Action 17) While the low vacuum scram is installed, it provides the first scram path (sensors provided by System 02, Condensate, logic by RPS, 99, and rod insertion by CRD, 85.) However, the following scram paths are provided for the case where the low vacuum scram is deleted as expected in the future. For initial power levels greater than 30 percent (sensed by System Ol), the Main Steam System (Ol) senses the closure of the turbine stop valves and sends a signal to the Reactor Protection System (RPS, System 99), which signals the Control Rod Drive System (CRD, System 85) to initate a rapid insertion of the control rods (scram) . For low power cases, (less then 30 percent sensed by System 01), where the turbine stop valve closure scram is not in effect, the neutron flux increase is sensed by the APRMs of the Neutron Monitoring System (92) and may reach the scram setpoint and initiate a signal to the RPS (99) and the CRD System (85) to insert the control rods. If high neutron flux scram is t reached for a loss ig reactor pressure (se ed by the Feedwater System, 03) may be the pat of initiatin the scram 4.2.4 Pressure Reli Following, tur ine stop valve closure, the Safety-Relief Valv s (S/RVs) will utomatically open and c c maintain the sure vessel pressure within safe limits. The S/RVs are part of the Main Steam System (01) and the associated discharge pipes are in System 10, Boiler Vents and Drains. The suppression pool (System 64) receives the steam. RI 4.2.5 Core Coolin (Safety Action 19) Reactor water level instrumentation which is a part of the Reactor Feedwater System (03) provides trip signals to the high pressure reactor cooling systems upon low reactor water level (level 2). The High Pressure Coolant Injection (HPCI) System (73) and the Reactor Core Isolation Cooling (RCIC) System (71) which provide single-failure backup to each other for low level transients, are actuated to maintain the level., gl

I4-.~g ShPE SHUTDOWN% ILMIS kppead~ 5 Page 6 o Loss oK Cldeneer Vacant Prepared: Date / gQ 7 E. C. Ecke Checked: Date: A~9-s BFN-OSG3&48 (Rev 1) V. G. Blanchett sasaassssssaasasaaaaasasaasasassaaaasaaaasaaasasaasas aaaasaaaasaaaasaaasas 4.2.6 Conta=nment Coolin (Safet Action 30) Suppression pool temperature monitors, which provide indication in the control room, are provided as part of the Primary Containment System (64) ~ The discharge from the S/RVs, together with the exhaust steam discharge from the turbine driven water supply systems, may cause the suppression pool temperature to approach the normal upper temperature limit. The RHR System (74) will be manually aligned into the containment cooling mode (two loops maxisam required, with one serving a single failure backup).

5. 0 LONG-TERM SHUZDORH 5.1 Normal Shutdown The normal pool cooling process should be adequate to handle decay heat transferred to the pool until, if desired, a normal unit cooldown can be performed as described in the Standard Isolated Shutdown sequence given in Section 7.6 of the main report (Table 7. 6-1) .

5.2 Safety Shutdown The safety equipment shutdo ce given in Section 7 6 of the main repor shutdown able 7. 6-3). used if necessary to ro ' dp 1 te

5. 0 REFERENCE S

a. Browns Ferr u P FSAR, Section 14. 5. 1.2.

b. Browns Ferry Nuclear Plant FSAR, A ection 5.3.5.2, Event Number 16.

c. Browns Ferry ar Plant FSAR, Volume 7, Responses to AEC Questions.

d. TVA-RLR&02, Revision 1, April, 1985, Reload Licensing Report for Browns Ferry, Unit 2, Cycle 6.

e. Browns Ferry Nuclear Plant, Reactor Protection System Elementary Diagram 730E915, Sheets 9 and 11.

f. Browns Ferry Nuclear Plant, Mechanical Control Diagram 47W610-2, Revision 1, Condensate System.

g. BFN Safe Shutdown Analysis, Appendix 5 (Revision 0), V. C. Kobold, and K. K. Fuj ikawa, BFNWSG3>>048, June 28, 1986.

h. FSAR (Volume 7) Question Q13.11, Listing of Emergency Operating Instruction and Demonstration Plans,

Appendix 5 Page 7 of M Prepared: Date: Loss of Condenser Vacua Ecke t Checked: Dete: M~(-g 7 BFN-OSG3-048 (Rev 1) V. . anche e, Jr. seal<>aaamaaaaaaaasaaaaaaaaaaaaaaaaaaaszaamzsacsassaaaaaaaaaacsaaaasssscsaaaaaaeaacease 7~0 NON-APPLICABLE INPUTS Section 4.2 and 4 .3 of the main body of the report show results of an assessment of licensing'references and FSAR questions. The items have been identified as likely SSA (Rev. 1) inputs for this event are given in Tables 4 .2-1 and 4.3-1. These references were all used above in this appendix.

mE 4JTWo) AI)ALYSTS oass of Rodenser Vacuua b &M3<$8 Nev 1)

APPEM)11 05 aspirin Checked: E C. C: Eck EMC9 V.B. Blanchette, Jr. naL Bate:

                                                                                                                      ~

Q+~ TK 05 -1 REACT IMXPTAHE TVA SAFE F<, FP'S ~c ACT)OH STS

    ~TATE  Me      Title       Code     Title   )I.      IAE          COBE 2-2. Fuel Failure   17   Scraa         1 HAIH SEN          SF   Provide    > 30l turbine first stage pressure interlock signal to reactcr.protection systen (99) failmfe logic.

F 2-2 Fuel Failure 17 Scraa 1 HA)H STEAH SF Provide aain turbine stop valve < 90X ooen trip signal to reactor protection systea N). s-2 Fuel Failure 17 Scraa Provide )ou condenser vacua signal to RPSl99) for scran fextra.anticipat~ signal). This safety action is expected to occur, but is not a requireoent of this event. ii -". Fuel Failure 17 Scraa SF Provide high reactor vessel pressure trip sional to reactcr orotection systen 199) fai)mfe louie.

         "t2     Fuel e  Failure  17   Scraa                             SF   Scraa signal froe reactor protection systea           199) ell    activate the control rod drive systea to insert rods. Scraa functicn only.

2-2 Fuel Failure 17 Bran 92 hGlTRCff NWITOR SF Provide protec 'n APRN neutr 199) rip sional failmfe i)wig.

                                                                                                                      %x I

reactcr LV '-2 Fuel Failure 17 Scrao 99 FSCT Provide th 1 r d ive {

                                                                                                                        'n s teo 1           n          1     tn    signal.
           "-2   Fuel Failure  17   Scraa                                       'de scraa signal to th                  rod drive fCHB)  systea    (         eeduater systen 1N) hioh re             sel pressure    trip signal.

2-2 Fuel Failure 17 Rraa y 99 ECTH. SF Siven lou condenser vacuun sional. Hroa systea

02) initiate signal to DI N) fcr scraa fextra,anticipatcry sional). This safety action is expected to occur, but is not a requireoent fcr this event.

2;2 Fuel Failure 0 S:raa 99 FEACTOR PNTECTH SF Provide scran signal to DI systea <85) on sionals froa nain stean systea <01) indicatino oain turbine stop valves ( 90'L open and turbine first staoe pressure > 50K. ,0

Q-E Bil~>> QN.TS)S APFEHDII 05 Page 9 of 14

      '.s   ci p@>>g:~C..                              sate:          P7 Condenser     Vacuun                                                    E.. Eck Checked:                               Da'.e: 4-20~

IF.'~48 (Rev I) V.S. 8lanchette, 3r TATLE 05 -1 FEZT L)A'CEPTADLE IVA SAFE DFER. SEBLTS SAFETT ACTIDH SYS SySIEN FlK

  -:;ATE  icoe    ',i tie        Code     Title     N.        HNK          GEE 2-3'cyst Stress        18   Pres Relief        NAIH STM          SF   SRVs open on    high reactor pressure, 2-2 Fuel Failure 18 Pres Relief              1 NAIH STEAN             Nain turbine bypass valves open on turbine control systen (47) turbine trip signal. This safety action nay occur, but is not required by this event.

Svst Stress 18 Pres Re]ief 10 NILER VHTSb DRH Provide path fcr aain stean systen (01) Sos stean b]wdtae to suppression pool N). 2-2 Fuel F i]use 18 Pres Relief 47 TN8]KCMK Send signal to open and control oa]n stean systea (01) bypass valves on turbine trip. This safety action nay occur, but is not required by this events

  '-3

~  :-2 Syst Stress

                ~net   Failure 18 19 Pres Relief Core Coo]ing
                                                      &I PR]

I GNA]M&I NAIH STEAN Accept SRVs stean b]wdwn (froa boiler vents drains systea, 10) to suppression pool. Provide stean for )P'I (73) and turbine in suoport of IPCI initiation cn lw uater level (L2).

  ;.."EF  i-2   Fuel   Failure   19   Core Cooling       NAIH STEAN             Proride stean for      RCIC (71)   turbine.

Fuel Fai]ure 19 Ccrc Coo))ng 2 G)HDEHSATE en uater supply for )FG systen (73) initiaticn on lw e]>($ ).

  '.DE=:-2      cue] Failure 19       Qre Coo]i   g                                             ly                                 CL FK'-2         Fuel Fai]ure 19       Core Coo]in                          SF   Provide   lw reactor     water level si
                                                                                )fCI systen    (73)
  .."LEF  2-" Fuel Failure       19   Core Cooling                              Provide path for fPCI (73) flw to the vessel through the feeduater spargers for IPC]

initiation on ]w uater ]eve] (L2). 2-2 Fuel Failure 19 Core Cooling 3 FEEDHATER SF Provide lw reactor water ]eve] signal (L2) to M systea (74) logic for RCIC systee (71) initiation.

  .;~":-2       Fuel   Failure   19   Core Cooling     3 FEEDH4TER              Provide path fcr RCIC (71) f]ou to the vessel through the feeduater spargers.

ME Sfmma( AS'LYSIS APPEMDll 05 Page 10 of 14 Date: ~ loss of Condenser Vacuw E.(;. EcLert ChKted: V. D te: A-ZIF-S7 IF)H863&8 (Rev 1) V.B. Blanchette, Jr. / TEE 05 -I FS "T WZCEPTAKE TVA SAFE MR. RESiUS ACTIN F(K STATE Code Title SAFETY Code Title SYS

                                                 )I.  'A%   SYSTE)(

CODE CDEF 2-2 Fuel Failure 19 Core Cooling M PRI CNTAIIBIT Provide alternate smrce water supply fm )FCI systea (73) froa suppressim pool for )fCI initiation cn low water level (l2) signal. Accept )KI turbine exhaust steaa. CDEF 2-2 Fuel Failure 19 Core Cooling 64 PRI CNTAIQB(T Provide a) ternate source water supply for RCIC systea (71) throunh Core Spray systea (75) piping froa suppressim pool. Provide suppression pool level indicatim. Accept RCIC turbine exhaust steaa.

  .KF 2-"      Fuel  Failure   19   Ccrc Coo(inn  71   RCIC            SF    Provide HXS ATU power to Rf systea (03) Irw water level (L2) instruaentation for tfC} systea (73)   initiatim.

CKF ".-2 Fuel Failure 19 Core Cooling 71 RCIC SF Provide HXS ATU piner to FM systea (03), low water level (l2) instruaentaticn for RCIC systea initiaticn. Ci6'-2 Fuel Failure 19 Core Cooling 71 RCIC SFS RCIC initiation on Iow reactix water level ((2) signal froo R)8 systea (74). MF 2-2 Fuel Failure 19 Core Cooling 73 HPCI SFS )PCI initiaticn on FM systea (03) Iow reactor water level (l2) signal. iNF 2-2 Fuel Failure 19 Core Cooling 74 RH( SF Send feedwater systea (03) Iow reactor water level signal ((2) to initiate RCIC systea (71). MF 2-2 Fuel Failure 19 Core Cooling 75 CORE SPRAY Provide flow path for alternate source of water to RCIC systea (71) frow suppression pool (64).

   ;.6    2-1  Rad  Release     20  RFV  Isol       1  SilNSTEA((             Nain turbine                                   water  t    ine v ves close, and bypass valves           open (expect but not required) h            ose         'g     s        he r                  ).

a 2-1 Rad Release 20 RPV Isol i n (t systea 01) upon V:: ccndenser Icw vacuua signal. F:-: Rad Release 20 PPV lsol 47 T(RBIIK CNTRI. signal to.Nain Steaa Systea (01) fm aain turbine stop valve closure. i

                            ~
                       ~~

SAFE MMMN AHA.YSIS APPMIX 05 Page 11 of,15 Prepared: Loss of Condenser Vacuua CEc t Checked: Bate: gQQ~7

         %HES3<4B       IRev I)                                                             V.6. Blanchette, Jr I                                                           TABLE 05  -I 1UZT MXEPTABLE                                 TN CPER.      RERLTS          SAFETY   ACTIN      S6 STATE  Rde     Title       Code     Title      IQ.

2-1 Rad Release 20 RPV Isol 47 TIRBIHE COIIR9. Send signal to aiin stean systea I01) to close turbine bypass valve upon very lou cmdenm vacuua signal. C0EF 2-1 Rad Release 20.8PV Isol 69 RXU SF Close R)GI isolatim valves m Iou uater level IQ) signal froe priaary cmtainaent systeu 164). This safety actim is expected to occur, but is not a requireoent for this event since there is no fuel failure. C CBEF 2-1 Rad Release 20 RPV lsol 74 R)ft SF R)fi isolation signal tripped on Iou uater level IQ) signal froa priaary cmtainaent systee I64). This safety action is expectai to occur, but is not a requireoent fcr this event since there is no fuel failure. 2-2 Fuel Failure 24 Pou Reduce I IIAIH STEN SF Provide uain turbine stop valves < 90X open trip signaI to reactor protecticn system I99). \ F 2-2 Fuel Failure 24 hw Reduce I NAIH STEN SF Provide > 30X turbine first stage pressure interlock signal to reacta protection systea I99) fai life logic. 2-2 Fuel Failwe 24 hw Reduce 6B RECIRQLATIN L 0pen recirculation puap xotcr breakers on RF (99) signal due to aain turbine stop valves < 90X open and >30X turbine first stage pressure. Coastdoun aust be faster than assuaed in reload analysis to Iiait severity of the event. 2-2 Fuel Failure 24 Pou Reduce 99 REACTN PRBIECTH SF Send signal to open recirculation signals froa aain s eau systea I01) indicating aain turbine stop a ves e. C."EF 2-1 Rad Release 26 Est Pri I r t eu ) cr initiation of L3 isolaticns. This safety action ls expected occufx but is s Even xnce there is no fuel failure. 1 MF 2-1 Rad Release 26 Est Pri Cont 32 CNIRK AIR SF Perforn isolatim actionis) upm receiving Iou uater level (L3) isolaticn signal froo the Priaary Cmtainaent systea I64). This safety action is expected to occur, but is not a requireoent for this event since there is no fuel failure,

                                                                            ~I
                                ~ ~ I,

 ~    SCTDCW ~(ALYSIS         APPENDII 05                                   Page Prepared:

12 of Ii Oate: ~Ill Loss of Condenser Vacuua E.,Ek t Reeked: Date:4-Q)~ EA-OM~ (Rev I) V.G. Blanchette, Jr. TEE 05 -I REACT L(NXEPTABLE TN CrEn. RESETS SAFE)V ACTIN SYS STATE Code Title Code Title M). CDEF 2-1, Rad Release 26 Est Pri Cont b1 PRI CWAI)ANT SF ()pon 9L (Q) signal frca RPS (99), initiate L3 isolation actions I send priaarylsecondary containaent isolatim signals to systeas 32, 65, 69, 75, 75, 7bs 77, Bls 90, and 9$ . This action is expected to occur,but is not a requireoent fcr this event.

  'CDEF   2-1   Rad Release    26   Est Pri Cent   75   (XRE SPRAY       SF    Perfcra Isolation acticn(s) upon receiving low water level (Q)      isolation signal froa the Priaary Containaent systea N). This safety action is expected to occur, but is not a requireuent for this event since there is no fuel failure.

2-1 Rad Release 26 Est Pri Cont 76 C(WAIN%)(T IIERT SF Perfora isolation acticn(s) upcn receivino low water level (Q) isolation sional froo the Priaary Containaent systea (M). This safety action is expected to occur, hut is not a requireaent for this event since there is no fuel failure. i& 2-1 Rad Release 26 Est Pri Cont 77 RAD((ASTE Per fera isolaticn acticn(s) upon receiving low water level (Q) isolatim signal frco the Priaary Ccntainaent systea N). This safety action is expected to occur, but is not a requireoent for this event since there is no fuel failure. C";EF 2-1 Rad Release 26 Est Pri Cent Bl CAD Perfaa isolation acticn(s) upon receiving low water level (Q) isolation signal froa the Priaary Containaent systea (64). This safety action is expected to occur, but is not a requireaent for this event since there is no fuel fai Vt 2-1 Rad Release 26 Est Pri Cont 85 Perfora isolatim acti (s) on recei r CH'-1 Rad Release 26 Est Pri Cont 90 ac i () ei water level (L3) isolaticn sional froa the Prioary Containaent systea N). 1his safety - acti ooccur, u )sn a requirenent for this event since there is no fuel failure.

ME 9%A'iope( N4LYSIS IE of

                                                                                                      ~t APPENDIX  05                                 Page              4 Prq,~g:              C                   Date:  ~p Lo>> of Cmdenser iFN~H)4B           (Rev Uacuun l)

ChKted: ~ E.g. Eck V;6, Blanchettei Jr. Date: ~27T7 S~ACT MXEPTABLE

  %G           (DECZLTS            SAFETY  PVIDH STATE  ~w]e        Title        Code    Title

(~J isalnkn) g~~q CD& 2-),Rad Release 26 Est Pri Cont 94 TIP SF Initiate TIP withdrawalpon low water level (LE) signal frco prinary cmtainaent systen (64). This safety actim is expected to occur, but is not a requirenent for this event since there is no fuel failure.

LG 2-) Rad Release 26 Est Pri Cont 99 REACT% RDOECTT) SF Provide low water level (Q) sional fron feeowater'systen (03) to priaary containaent systen (64) for initiation of LE isolations.

This safety acticn is expected to occur, but is not a requireeent for this event since there is no fuel failure. Rad Release 27 Est Sec Cont 64 PRI CNTAINN SF Perfora isolatim acticn(s) upon receivino loDD water level (Q) isolaticn sional froa the Reactor Protection Systea (99). This safety action is expected to occur, but is not a requirenent for this event since there is no fuel failure,

6 2-I Rad Release 27 Est Sec Cont 65 SST SF Initiate SST plant start on Iow water level (Q) signal froa prinary cmtainaent systen (64).

This safety action is expected to occur, but is not a requirenent fcr this event since there is no fuel failure. CF~ 2-4 Cont Stre-s ~A Cont Coolina '5 R)fs( %F Support RHR systen (74) suppression pool cooling NXlee C;EF ".-4 Cont S!ress 50 Cont Cooling 64 PRI CC((TAINN ression pool teaperature and e indicatim to support ygten ) es ooi~ooli egg k s 0 CCEF 2-4 Cont Stress ZO Cont Cooling 7 IW'tippn)tx Phe'ttPDi )rmatm pmi tmiiep

                                                                                                  ~
    ~

fQe'a ~ Qat Stress AA Cent Cooling 7 M I(SF Provide ool cooling function. Ck= 7-5 Pers 0verexp 2 Cont Bay Env SF Provide Iow water level (Q) signal to RPS systen (99) fcr initiation of control bay isolation. This safety acticn is expected to occur, but is 0 not a requirenent for this event since there is no fuel failure. e t:- e

Ac ShUTOQe( A(QLYSIS APPMIX 05 Page 14 of 14 Loss of Condenser Vacuun E.C Ecker lXecked: Bate: MAO~ BF(RSD<4B (Rev 1) V.S, Blanchette, Jr TABLE 05 -1

~T      lNECEPTABLE                         TVA ffER. KS(lTS          SAFETY  ACTIN     SYS     SYSTE)(

STATE Code Title Code Title ML IN(E REF 5-5 Pers Overexp 56 Cont Bay Env 5( AIR C(NITINI% SF Air cond, (AC) supply ducts isolate h Energ. Pres. Systee (Control Rona Eaerg. Vent. Systen) supplies pres. filtered air to )KR on L)4. (U) signal iron systen 64. This safety action is expected to occur, hut is not a requireeent for this event. iOEF 5-5 I'ers Overexp 56 Cont Bay Env 64 FRI CCNAINN SF Upon low water level (Q) signal fron RPS(99), send isolation sional to air conditioning systen QI). Ihis safety action is expected to occur, but is not a requirexent fcr this event since there is no fuel failure.

       ~-5 Pers Dnrexp 36      Cont Bay Env  99 REACTN PROTECTH SF   Provide low water level (U) signal fran feed<<ater systen ((8) to priaary cent. svsten (64) fcr initiation of control hay isolaticn.

This safety acticn is expected to occur, hut is not a requirenent fcr this event since '.here is no fuel'ailure.

ShTE SHOED'ML~ Qxpendxx 6 Page 1 of 8 Loss oX Feedeatcr Prepared: E. ~ Eck rt Dete: ~j/ 7 Heater Checked: Dete: ~/3ll8~ BFN-OSG3W48 (Rev 1) W. E. Overstreet aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa

1.0 DESCRIPTION

OF THE EKNT Primary system condensate is returned to the reactor vessel through a series of feedwater heaters. Heat is supplied to these heaters fran steam extracted fran the main turbine. This extraction steam condenses in the feedwater heaters imparting its heat energy to the feedwater, axe then returns to the main condenser hotwell 'as condensate. In a Loss of Peedwat.er Heater event, it is assumed that the feedwater heating fran a portion of the heaters is lost, either by disruption of the steam supply, or through routing of the feedwater through a bypass around the heater. This loss results in a decrease in temperature of the feedwater which is returned to the reactor vessel. The cooler feedwater results in an increase in core inlet subcooling and a decrease in core coolant void content. Due to the negative void reactivity coefficient an increase in core power results. The Feedwater System is described in reference 6a. The Loss of Feedwater Heater event is defined and analyzed in references 6b, 6c, and 6d. This appendix is based on the previous issue (reference 6f) plus new inputs axd comment s subsequently received. Referenc addresses this transient as Event 18. Per reference xs ev nt is to be considered with regard to the nuclear y operational criteria only in operating state P use significant>fjL ater heating does not occur in er operatin esp o the plant from full po e studied for t %is will give the most severe tr ends two sequence of recove ater Heating eve 2.0 EVENT CATEGORY The Loss of Peedwate Heater eve catagorized as an Abnormal Operational Transien ( nce 6c). 3.0 PLANT PAREMETERS AND SAFETY C)NCERNS 3.1 Peedwater T erature and Reactor Power In the PSAR axd Reload Analysis (reference 6d) of this event it is assumed that the loss of plant feedwater heating capability does not exceed 100 P. This represents the maxiaxaa expected single heater (or group of heaters) which can be deactivated by a single event or operational action.

ShPK SHUTDOWN hMLmXS kppcmadxx 6 Prepared: Loss of Feecbeater Beater Checked: BFNWSG3&48 (Rev 1) W. E. Overstreet aaaaaaaaaaaassaaaaaaaaaaaaaaaaassassassassssaaaacsaaacsaassaaaaaaaassaaaaaaaaaaaassaaa The cooler feedwater flow will result in increased core inlet subcooling> decreased core void content, axd due to the negative void reactivity coe fficient, increased reactor power. At the initiation of this event, the reactor may be in either the manual or the automatic flow control mode. In the automatic mode> the recirculation flow control system responds to the power increase by reducing core flow so that steam flow fran the reactor vessel to the turbine remains essen-ially constant although reactor power does increase somewhat. In the manual flow control mode, no compensation is provided by core flow and thus the power increase is greater than in the automatic mode. The power increase transient threatens the fuel cladding thermal limit. (Safety Criteria 2-2). Reference 6e also establishes that in addition to an increase in total, core power level, the reduction in core inlet enthalpy also leads to a shift in power shape. If the fuel is operating at or near the preconditioned envelope at the time of the event, either the increase in power or the power shape shift can result in PCIOHR (Preconditioning Interium Operating Management Recommendations) violations and consequently potential pellet-clad interaction which could lead to fuel failure (2-2).

4. 0 EVENT MITIGATION 4.1 A licable Safet Acti 4.1.1 Provi el Protectio (

For thy ~ipse +$wg configuration nstration must a 4 +a h potential wors e water heater loss sh, eh by single corn failure or operational error bilk'not result in oss of plant feedwater heat capability in exceeds-5X'00oF. This requires Safety Action 23, Lijit.the-Nagnitude of the Disturbance. In reference 6b (anrl expected to be the basis of reference 6d) this event, is analyzed as not requiring protective action if the reactor is in the automatic flow control mode and it is inferred that automatic protective action may be avoided for some or mast potential cases where the reactor is in manual flow control at the time of the incident. For these cases, since prevention of damage to the fuel is the most important

klqma8xx 6 Prepared: Loss of Fee@eater Heater Checked: BFN-OSG3&48 (Rev 1) W. E. Overstreet sseesteeaaaaeaeaaaaaaaassasseaeaaeaeassaaeeaaeassssaaaeaaassaeesseaaaaaaaaaaaaaeaesae consideration following a loss of feedwater heating, reference 6e recommends the following recovery sequence in order to restore the plant to normal conditions (Safety Action 22):

a. The recirculation flow should be promptly reduced to a level that will reduce the thermal power to at least 20 percent below the thermal power prior to the reduction of feedwater heating (or to miniaam flow if this is reached first). APRM reading may be used to indicate thermal power level during the transient.

b. Additonal flow reductions should be performed as necessary to maintain the 20 percent thermal power margin.

c. Insert control rods to the extent needed to reduce power to the initial. load line. If shallow or fully withdrawn 0 d.

control rods are inserted, they should be inserted continuously to the full in position. Core monitoring using the process computer should be executed prior to anv increase in the power level. Reference 6b reports that for the case where the reactor is operating (at full power) in the manual flow control mode "the neutron flux would have reached within one percent of the scram setting". Therefore, for the most severe event .f ith the reactor at full power anal in manual flow c ol a time of

                         'occurrence) a scram from high neutron                 x mav be       ded as final fuel protection.

4.2 S stems Re uired sg I Table 06-1 marines t nctions. as collate om even e a standard uence fran Section ~ 6 e the report e 7.6-8). 4.2 ~ 1 Limit d of the Die Mce (23) The Reac r Feedwat y~t@m (03), the Condensate Svstem (02), and the F. trac ' Steam System (05) must provide documentation that no si e ction or event could cause a'feedwater rop in excess of 100oF. Reference 6h documents (-0

                                                                                                  'emperature that the      maxiaam expected       change is 52 F.

s I* s'. 1 ~ ~

                                                 ~    ~

hpggeodxx 6 Page 4 of 8 Prepared: Date: Loss Beater of PeedINater Checked: E.. Eckert Date: < 3l BFN-OSG3-048 (Rev 1) W. E. Overstreet aaaaaaaaaaaaaccaeeaeacIaaaccaeaaeacIaaaaaaaaaaaaaaacceccaIccIaaaaacIcaaaaaassaaaIcassa 4.2.2 Restore the Plant to Normal Condtions (22) The APRM readouts fran the Neutron Monitoring System (92) are monitored during the transient for indicatione of total reactor power level. Operator annunciation fran the flo~efezenced high flux alarm setpoint is expected to occur for events near the limiting case. For the reactor in manual, flow control, the Recirculation Flow Control System (96) are used to reduce avd regulate recirculation flow. The flow reduction should be suf ficient to reduce the thermal power to a level of at least 20 percent below the level prior to the 'reductions in feedwater heating, or to minima flow if this is reached first (reference 6a) ~ For the reactor in automatic flow control mode, the load demand eetpoint of the turbine generator controls (System 47) is used to obtain the reduction in recirculation flow.'ubsequent

              'transfer to manual control. using the Recirculation Flow Control System (96) directly is expected.

The recirculation pum ie reduced,I ap$ aintained through motor-generate tontr'ol by b g)o:R oirtnlation yloo Control em (96) . Control r,'~ ~ t al y inserted following the co~ Wlo o 'on

                                                    ~

to the ext eded to reduce o e cent load line any shallow or fully s are inserted

                                          ~

y should be inserted dont nuously to the xn position to avoid further erturbation e power shape (reference 6e). 4.2.3 ram <17) e The scram initiation signal is derived fran the APRM's of the Neutron Monitoring System (92) should they reach the setpoint, reference 6h. The plant has incorported the thermal power monitor feature within the APRM, system 92 logic. Reference 6g points out unit 2 acceptance by the NRC while applying for the same feature for unit 3

                                  "~ ~

referenced scram setpoint, SetPOint~~~

                                               ~

AIC

                                                     ~
                                                   <No credit is taken for the flow-4 '+

the upper limit scram g&~ 4IM4ElfS IIQ ale.>It) The Reactor Protection System (RPS), 99, provides the trip logic for reactor scram given the APRM input from the Neutron Monitoring System.

kppendxz 6 Page 5 of 8 Xoee oE Feedeater Prepared: Ecke t Date. / / Beater Checked: /.i BFN-OSG3-048 (Rev l) ers treet e+~~~~~~~ 3'agtngaaaaeanttearnaenraaanrnetanraanaaanrnrnsasaaai++++++~ The RPS logic provides signal input to scram pilot valves of the Control Rod Drive (CRD) System (85) to accomplish the rapid control rod insertion (scram) ~ Xf p~~ ge yg, / /~al Q.>) <<V8;~ 5e0 LONG W ~+;.~~atV~ TERM SHUTDOWN

                                                  ~

CONSIDERATIONS Ge r. bS.+ pa ~ ~ Vis/W 5 ~ 1 Normal Shutdown The sequence provided in section 7.6 of the main body of the report for normal non-isolated shutdown would be followed necessarv (Table 7.6-2). However, in this event, it is usually if expected that the unit will remain connected to the grid and shutdown will not be necessary. 5 ' Safety Shutdown Normal, long term response> as discussed in 5.l, utilizes all systems normally available to the operator. If normal control equipment is assumed to be unavailable the reactor any (or all> can still be shut down if necessary by utilizing only safety equipment as given in the safety shutdown sequence in section 7e6 (Table 7.6-3) starting in this event fran a non-isolated condition.

6. 0 REFERENCE S

a. Browne Ferry Nuclear Plant FSAR, Subchapter 11.8; Condensate and Reactor Feedwater System.

b. Browne Ferry Nuclear Plant FSAR, Chapter 14; System Safety Analysiep Paragraph 14.5.2.1, Loss of Feedwater He

c. Browne Ferry Nuclear xx, Safety Opera ional aragraph G.5.3.5, Ahnornal Operational Try 'eIItsp Analysis'vent 18.

d. TVA-RLR&0 o A r 0 e e r for Browne Fer 2

e. General El r c Se n ormation Letter (SIL, Number 370> Possi le Fuel Damage From Loss of Feedwater Heateis ~

f. BFN Safe S utdown Analysis, A en o old anil V. G. Bla 3&48, June 28, 1986.

g. NEB-RAC-1607 (B45 860702 045), TVA letter dated January 23, 1984 from L. M. Mills to H. R. Denton (NRC) 5 Enclosure 2 (pages 3z 4) 5 Implementation of APRM thermal power monitor (Unite 2, 3) ~

h.,FSAR, Volume 7, Question U14.1, Fuel Protection During Loss of 100oF Feedwater Heating Transient. -0

ShPE SHUXDHK kMLMIS Loss Beater of Peecheater Prepared: Checked: Z E.

                                                    ~'ate: '//

C Ecker~ Page 6 Date. of 8 8+ BFNWSG3"048 (Rev 1) W. E. Overstreet aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa>>aaaaaaaaaaaaaaaaaaaaaaaaa aaaa

7. 0 NON-APPLICABLE INPUTS Sections 4.2 axd 4.3 of the main body of the report show results of an assessment of licensing references and FSAR questions. The items which have been identified as likely SSA (Rev 1) inputs for this event are given in Tables 4.2-1 axd 4.3-1. These references were all used above in this appendix.

SAFE SHflNN NNLTSIS N8011 06 Page 7 of Loss of Feeduater Heater E.C. ert hta +//~lP HlH)SSMQ IRev I) LE. Overstreet TARE 06 -I REACT MXPTARE TN SfE 5%R. REKTS WHY ACTlW $ 5 STSTE)I RK STATE Code Title Cade Title IL IN% CCXE F 2-2 Fuel Failure 17 Scraa SF Scrae signal frm reacbr pratectim systen I99) sill activate the <<antral rad drive systen to insert rods. Scree functim only. F 2-2 Fuel Failwe 17 Scran 92 KlMIIMTN SF Provide NN neutran flux trip signal to reactor pratectim system I99) fail~e logic. F 2-2 Fuel Failure 17 Scraa 92 I)JJRKNI )91TTN SF Pravide NN thernal paar signal to reactcr protectim systen'199) Afar a~ektq

                                                                                                ~

QC F 2-2 Fuel Failure 17 Scran 99 REACTIR ROOIEl)I SF Provide scree signal to the control rad drive

                                                                              $  5) systen IS5) m eeutrm aaitcring systen l92) N%)I neutrm flux trip signal.

99 REACTN PROTEI)l SF Provide scran signal .to the cmtrol rad drive I59) systen IS5) m neutrm emitcring systen 192) N%ll thernal powr signal lapper linit of fla referenced setpoint Qy):,.:,: F 2-2 Fuel Failwe 22 Rest krna) 57 TIRSJ deeand setpoint aid to is firn obtain reductlm eben in autaeatic r firn F 2-2 Fuel Failwe 22 Rest Nornal 85 nserted to extent needed, after recirculatim reductim, to reduce F 2-2 Fuel Failure22 Rest kraal 92 l8fRQIIQOTW APRIIs anitar paar level, and provide fl~eferenced alarn signal if reached. F 2-2 Fuel Failure 22 Rest kraal 96 REClRC FUM NIRL II Run hack redrculaticn per SlL RO instructims - IINual Cmtral )lode Ibef are inserting cmtrol rods), Snitch to nanual control node if in autmatic cmtrol sade. F 2-2 Fuel Failure 8 Li~ I)isturb 2 NDBSATE L Wkr 4eaeeeaL as+silt desi'gn neets 100 degree F assueption for naxiaa feedeater teeperature drop that cm amr for any single actim or failure of Ri heateris). See also Extractim Stalin) and RIIUD Systees. I r." '

                                                                      ~

SEE BIHQN NOLYSIS %9011 loss of Feeduater Heats 06 Page 8 of 8 E.C. Ectsrt 0 tw ~>~ Chected> cate ~ ~ '7 NH)SS3~ Ohv 1) 1LE. Onrstreet TAftE 06 -1 REACT ONXPTAfLE SA SffE PER. REBLTS SAFETT ACTTS . SS QSIEN RK STATE Code Title Code Title M). NK CNE F 2-2 Fuel Failure 23 Lie hsturb 3 FKNTER L as~it design eeets NO degree F I failure~ assuspUm for eaxiaa feedatxr teeperatm that ca occN fs'ny single acUm of Rl locator is). See also Cmdensate 102) aed Extractim Stean N) system. F 2-2 Fuel Failwe 23 Lie Msturh 5 ETTRACSN L Ieaetrete as~it design nests 100 degree F assueptim for eaxiae feehater teepsratxre dna that ca omr for axy %ogle acUm I failure of FN heaterts). See also St81) and Cmdsnsate(82) System. F 2-1 Rad Release 26 Est Pri Cmt 85 CR0 SF Pxrfaa isolaticn acUmls) upm receiring scree signal frm the Reactcr Protectia Systee tgg).

fc'. Shutdown Cooling (RHBS) kppeadix 7 Prepared: c,F E. C. Eckert Page Date 1 of 5

                                                                                     ~ dP Mal&mctioa (Temperature Decrease)

Checked: S. K. Mehta Date: ~a BFNWSG3&48 (Rev 1) sssasaaaaaaaaaaaassaaasassaaasaaaaaaasaaaaaaaaaaassssasaaaasaaasaaassssssss

  '1 ~ 0 DESCRIPTION    OP THE EVENT A  failure or misoperation of      the reactor shutdown cooling function of the  RHR  heat exchangers could result in an unplanned decrease of the reactor water temperature due to excessive cooling. This appendix is based on the previous SSA issue (reference 6b) plus new inputs ani comment s subsequently received.

This event can only happen in operating states A, C, or D where the reactor is operating at low pressure (less than 75 psig). 2~0 EVE8Z CATEGORY The BFNP FSAR (reference 6a) addresses this event as an Abnormal Operational Transient (section 14.5.2.2). 3.0 PLAÃZ PARAMETERS AND SAFETY CONCERNS In operating states A and C, the only parameter that changes is. the temperature itself since the reactor 'ie subcritical with all (or all minus one) control rode inserted. If the transient occurred when the reactor was critical or near critical (operating state D), a slow reactor power increase could result. Other plant parameters (e.g , system pressure or water'evel) are not expected to change signific a mild depressurization and u No signi ficant safety conce 1 ~ v r e P rotection aaxst be maintained even during e ch a slow power increase (2-2) ~ 4~0 EVENT MITIGAITON 4.1 A licable Safet Actions

                                                          /

In order to limit the extent of the transient on power, fuel, axd vessel temperature, Appendix G of the FSAR (reference 6a) states that the operator should use normal controls to accommodate the power change and restore normal shutdown conditions (Safety Action 22). If no action is taken, automatic scram and termination of the power rise should, occur (Safety Action 17) ~ ,~

                                 +~fix 7 Prepared:

Page 2 Date of3

                                                                              /~~~!7 Shatdoen Cooling (PHRS)                             E. C. Eckert Nal&xmctioa (Temperature          Checked:                            . Date:       '      I Decrease)                                            S. K. Mehta BFNWSG3-048 (Rev 1) aaeaeaaeeeaaeeeaaaaaeaaaaaaaaaeaaaaaeaaaeaaaaeaaaaaaaaaaaaeaaasaaaaaasssssa 4.2    S  stems Re  uired Table 07-1 shows a summary of the actions and systems involved in this mild event. The required system functions were collated from event~nique actions and applicable standard sequences from section 7.6 of the main body of the report (Table 7.6-7).

4.2.1 Restore Normal Conditions (22) The operator will use the normal controls involved in plant startup to avoid scram if this event occurs while the reactor is critical (Re-range the Neutron Monitoring Svstem IRM, (921 and possibly insertion of some control rods using the Control Rod Drive System, (85)). Primary action will also involve reducing or shutting off the RHR shutdown cooling as soon as it is concluded that it is malfunctioning (RHR 74) ~ 4.2.2 Scram (17) Should the reactor power increase more than al controls can accommo ate, scram from the Neutron Monitoring ystem IRM (92) could occur (using the subsequent actions of t RPS (99) azd

5. 0 LONG-T I

Thi s ole event is conc'erned with maintaining long-te shutdown capabi e.

6. 0 REFERENCE S

a. Browns Ferry Nuclear Plant FSAR.

b BFN Safe Shutdown Analysis, Appendix 7 (Revision 0), H. A. Greaves and S. K. Mehta, BFNWSG3&48, June 28, 1986.

7. 0 NON-APPLICABLE INPUTS Sections 4.2 ani 4.3 of the main body of the report show results of an assessment of licensing references and FSAR questions. No additional items were identified as likely SSA (Rev 1) inputs for this event.

SAFE QSMN QN.YSIS NE)ol'l 07 p~, rcF Page 3of 3 a~~ Shutdoo Cooling E.C. Ectert an~ ~i~ aS

              ~

functim (Rev 1) TARE 07 "1 REACT lNKEPTlRE TVA SfE 1PER, REM.TS SAFETY ACHSl SYS SSlEN RIG STATE Code Title Code TiUe )L NE GEE 0 2-2 Fuel Failure 17 Scran 85 ON SF Scran signal frm reactor protecUm systee 199) will activate the cmtrol rod drive systen to insert rods. Scraa fwncUm only. 0 2-2 Fuel Fail~a 17 Scrae SF Provide TNneutrm flux trip signal to roactcr protecUm systls 89) (see secUm 7 L2) ~ IR 0 2-2 Fuel Failure 17 Scrae O'ro ide ~ signal to the cmtrol rod drive 159) systen 185) m neutrm anitmng systww t92) HN neutrm,flux trip signai. KD 2-2 Fuel Failure22 Restkraal 7I II 5 Turn off cr reduce QI Shutdown Cooling to stop undesired cooling. 0 2-2 Fuel Failure 22 Rest Nareal 85 ON 0peratcr's acUms to insert Control Rod tn nitigate power iKf%0$b 2-2 Fuel Fai lac 22 Rest kraal: 92 %SAN NMTN 0parabr's acUms to ~ange lN if needed to avoid scrm 0 2-1 Rad Release 2b Est Pri Cant 85 M F Per fern ieotatim actiaHs) wpm receiving ecru steel frm the 1)eactcr Protectim Systea 89).

Cr ShFK SHOT?X%% hMLYSIS kgb.radix 8 Page 1 of 8 hadvertent Pcaap Start Prepared: E. CD ckert Date: / Y4 7 Checked: I r g Date. '~~7 BFNWSG3&48 (Rev 1) W. E. Overstreet

1.0 DESCRIPTION

OF THE EVENT An inadvertent pump start is defined as an unintentional start of any nuclear system pump which adds sufficient cold water to the reactor during xarmal operation to cause a moderator temperature decrease. This appendix is based on the previous SSA issue (reference 6e) plus new inputs and comments subsequently received. Several systems are available for providing high-pressure supplies of cold water to the vessel for normal or emergency functions. The Control Rod Drive System and the Feedwater System, normally in operation, can be postulated to fail in the high-flow direction introducing the possibility of increased power due to higher core inlet subcooling. The same type of transient would be produced by inadvertent star tup o f either the RCIC System or the HPCI Sys tern. In most of these cases, the normal feedwater flow would be correspond-ingly reduced by the water level controls. The net result is simply a replacement of a portion of the 380 F feedwater flow (at design power operation) by flow at approximately 100 F. Excess flow from the Feedwater System itself is considered in Appendix 23. The transient that occurs is similar to the loss of feedwater heater transient. As in that case, the most threatening transient would occur where minimum initial fuel thermal margins exist (maximum power within reactor operating state F) The HPCI

                                          ~                                             s severe than the loss of             er eater    case  because   its  effect    on mixed feedwater temperat re     will produce    a cha n e             bo   4      F",.

(at Gal 1 power) compared 0 d 1 feedwater heater case (r 6 e 6 i t for discussion of the Lo d 2.0 According to FSAR (reference 6a) Appendix G, paragraph 5.3.5.2, Inadvertent Pump Start is classified as an Abnormal Operational Transient event. 3.0 PLANT PARAMETERS AND SAFETY (ONCERNS The severity of the resulting transient is highest for the largest pump which can cause the abnormal event: the -inadvertent startup of the 5000 gpm HPCI System.

fg Iaadeertent Paap 1)'gb)ca8xx Start Prepared: 8 CcF~::,", E. C. Eckert F 2 ii~~(~ Checked: Date: / 8'l SFN-OSG3~8 (Rev W. E. Overstreet aesaseasseaseeaaaewaaeeacasaaaaaagaseeaeaaneeaeaeaaeaaaeaaeasaasa~<+++~e>aaaaaacaaa 3.1 Core Reactivit and Power The safety concern due to inadvertent pump start is that the decrease in feedwater temperature, increase in reactivity a& increase in neutron flux can cause damage to the fuel (2-2) ~ 3.2 Feedwater T erature and Reactor Water Level The safety concern due to feedwater temperature change is that stresses in the components may exceed their design limite (2-3) ~ It is also undesirable from a system stress point of view to allow the vessel to overfill due to excess inventory supply. X4

4. 0 EVENT MITIGATION

                         , /~ /~</ C~~) i4~4;+ +c    k;im4&~r       ~~~          b,km~.0~,-A ga/h 4.1    Ap   licable Safety Actions 4.1.1     Avoid Fuel Failure (2-2)

Should the power increase be enough to challenge fuel integrity, reactor scram should be initiated to terminate the transient (Safety Action 17). 4.1.2 Avoid Excess System Stress (2-3) A key step to limit 'the potential increase in system stress is to shut off the inavertently actuated system and restore normal ditions Sa ety ctxon tions are eu h that water level rises, threatening to 1 the ve 'P~' A overf'o 2)' s needed. Th o cure at low po initLal, d rger design in luded consideration of thxs occurrence whse i lees severe th n the case where required initiation of these stems occurs du% ing los 4.2 Systems Re uired Table 08-1 summarizes the required system functions. It was

            'collated from event~nique actions and applicable standard sequences from section 7.6 of the main body of the report (Tables 7.6-7, -8, -30> -31).

SKPK SHUIX)%W hMLmIS appendix 8 Prepared: Checked: BFN-OSG3-048 (Rev 1) W. E. Overstreet aameaassssaassassssaaaaassaaaaaaaaaassaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa 4.2.1 Restore Normal eration (22) Indication needs to be provided from the HPCI System (73) for operator action to turn off pump and restore normal operating conditions. Indication is also required from RCIC (71), the other system capable of high pressure injection (even though it is a lesser case). 4.2.2 Scram (17) A power increase above the APRM (or IRM) flux set@oint requires that the Neutron Monitoring System (NMS) (92) provide the signal required to start reactor scram actions. The Reactor Protection System (RPS,'9) provides logic on high neutron flux trip (APRM or IRM) and sends a scram signal to the Control Rod Drive Sy'tem (85) to cause fast control rod insertion. 4.2.3 Water Lev'el Reduction (25) If water level increases to level 8 before manual action can trip the system, the Feedwater System (03) sensors send signals to the HPCI System (73) or RCIC System (71) to automatically shut off the pump. Note that in this case> trip of the feedwater and main turbines would course of the event wo ppendxx 23. 5.0 LONG TERM S

5. 1 Normal I

The seq ence provided in section 7.6 of the f the report r normal non-is . able 7~ 6-2) will be followe expected sary. However, in that scram will not occur and, even this event, it is usually if it does, restoration of the unit to the grid will be possible and full cooldown will not be necessary.

5. 2 Sa fet Shutdown Normal, long-term response, as given in 5.1 above, utilizes all systems normally available to the operator. If any (or all) normal control equipment is assumed to be unavailable the reactor can il I)' I ~ ~

                                         ~   ~

8hPE SHUTDORSE MbL~ bypeadxa 8 Inadvertent kaap Start E. C. Ecker Checked: BFNWSG3-048 (Rev 1) W. E. Overstreet RD~CRDESR%RCRÃmRNR%RRgtSRCReaaRRaRaaRRNCWRRRRRsSStRSRDtlRcIcRRSASCBRN~~RSRRRRRR%%RRRRQNR still be shut down to cold conditions by utilizing only safety equipment as given in the safety shutdown sequence in section 7. 6 of the main body of the report, starting in this event from a non-isolated condition (Table 7.6-3)

6. 0 REPERENCK S

a. Browns Perry Nuclear Plant Unit 1, 2, ami 3 "Final Safety Analysis b.

Report" (PSAR), through amendment 67. PSAR, Paragraph 14.5.2.3, Appendix G.5 ' PSAR, Appendix G, Plant Nuclear Safety Operational Analysis.

c. PSAR, Paragraph 7.5,* Neutron Monitoring System.

d. PSAR, Paragraph 7.2, Reactor Protection System.

e. BFN Safe Shutdown Analysis, Appendix 8 (Revision 0), B. J. Cheek and H. S. Robbers, BFNWSG3-048, June 28, 1986.'.

0 NON-APPLICABLE INPUTS Sections 4.2 atd 4.3 of the main body of the report show results of an assessment of licensing references and PSAR questions. No items have been identified as likely SSA (Rev 1) inputs for this event.

1 SfE )UIIGM NMIS APPEI011 M Preparedx E.C. Ectert 0>>ctah ~< IF)HE+% Ohv 1) LE Onrstreet TSE % -1 REACT MXEPTARE TVA SFK IFEL KKTS SAFEIY ACTIN STS STSIE)I FIM: STATE Code Title Code Title 16. NK CKE If 2-2 Fuel Failure 17 Scrm '5 CRD  % 'era situal frm reactcr protectim systee 199) uill activate tbe ccotrol rod drive system to insert rods. Scree fmcUcn eely. F 2-2 Fuel Failure 17 Scrae IBllHNl90IN 92 SF Provide APR)1 oeutrm flux trip si9nal to reactor protectim systee 199) fail~ lope. D 2-2 Fuel Failure 17 Scrae 92 ISAHN )GIITN IF Provide IR)l oeutrm flux trip signal to reactcr protectim systee W) (sae secUm 7.L3. F 2-2 Fuel Failure 17 Scrae 99 REXAR FROIECIN F Provide scr>> si9oal to the cmtrol rod drive f05) systee l85) cn eestrm aeitcrieg eystee

02) lRN eestrm flux trip sigil.

D 2-2 Fuel Failure 17 Scrae 99 ISClN PRHECDI EF Provide ucrae signal to the ccetrol rod drive 0 F 2-2 Fuel Failure 22 Rest kraal 71 RCIC ll ICRD) systee F85) m eeutroe emitorie9 spkea. l92) IN oestrm flux trip siyeal. Provide'systee start McaUoe so cperatcr sty systeu and rehn to ecreel. ca F 2-2 Fuel Failure 22 Rest kraal 73 IFCI ee start IodicaU stlp systee and ferro')0 ~ Fl 'g y.l. level stol itS) t t op. IF 2-3 Syst Stress lael Provide bi9h reactcr eater 1 'pal 68) to Proride HXS AS poser to F)l systee ID3) bi9b eater level (LS) lfCl trip ImtnmataUm. 2-3 Syst Stress 25 Level Reduce 71 'RCIC Provide HXS AS poser to F)f systee 0$ high eater level L8) RCIC trip iostrueeetatim. IF 2-3 Syst Stress 25 tevel Reduce 71 RCIC Shutoff RCIC Of qeraUo9) m F)f systee IN) hi9h reactcr eater level 6$ situal. lf 2-3 Syst Stress 25 Level Reduce 73 )fCI Setoff )PCI Of cperaUn9) m F)f systeo NQ hi9h reactcr eater level tl8) si9oal. 4 ~,

                                                                  ~   ~

SR SHJMM LYSIS NRRII N lnatvsrtsnt hp Start IBHS&08 Nar I) TRE N -l IBCf QNQXPQKE TN IHL RERLTS Sf'HSI 98 STATE hdtv Title Cab Title IL IF 2-1 RadRelease 2$ Est Pti Get 85 05 ~ 5'ricri iaCaticm acticnid pea rmceiviog scrm signal frme the Reacts fretsctica Sph ~ Ig%. gI

kgqmndxx 9 Page 1 of 7 Prepared: Sate: ~l S/ Conor'od Ri.ah~eeal E. ~ Ecker Q Error Checked: 4 Date: Il>'/7> aaaaaasasaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaCksaaaassaaaaaaaaaaaaaaaaaaaaaaaaaaa 1 ~ 0 DESCRIPTION OF THE EVENT The FSAR (reference 2c) and reload analysis (reference 2a) assume+<WtlwI17 the worst case Rod Withdrawal Error (RWE) to occur if a reactor operator makes a procedural error and withdraws the maxiaam worth rod to its fully withdrawn position, resulting directly in a larger than normal positive reactivity insertion. The focal point of this event is

     - localized to a small portion of the core since the transient causes a heat-flux peak in the vicinity of .the withdrawn control rod only A smaller increase is seen in the total core power causing a mild reactor transient.

The RWE event is addressed in FSAR, section 14.5.3 and FSAR, Appendix G (reference 2c) This appendix is based on the previous SSA issue

                             ~

(reference 2f) plus new inputs and comments subsequently received. The transient can occur during the following operational conditions.

a. Continuous rod withdrawal duri'ng startup (reactor states D and lower power part of state F).

b. Continuous rod withdrawal during power range ope ions (reactor state high F).

I No uniq y actions are required 'n cpe States A, C, azd E because the core is more t~n and%8 ii xd could not achieve c~+~ity Mjtlt the>.g)P+ d 1 o one control rcxl. Special chndidedationr," fons g$ IQ$ th rowel err rs during re fueli (saba)q,A)qac'e ll of this port. 1.1 Con inuous Rod Withdrawa s

                                                   'rtu M8 xn Appendix 1.1.1    Descri tion of the Event Per the FSAR, section 14.5.3.2 (reference 2c), the most severe transient occurs when an out of sequence maximum worth rod is assumed to be continuously withdrawn bg the operator with the reactor being just       criticpl,~yt       ~<temperature. The Rod    Worth Minimizer (RWM)~a      s  RN'     fail to prohibit this withdrawal (as      it should). Sequence control should also be provided by the Rod Sequence Control System (RSCS) up to 50 percent rod density (reference 2g), yet the withdrawal of the worst rod is still evaluated.          It  is assumed that the maxiaam allowable IRMs are in the bypassed condition and the operator has also used the permissive to bypass the rod block functions of the rod worth minimizer.

s

                                        'es

kypeadix 9 Prepared: Sass: ~/V/ Control Rod %KthdraeaI E. C. Eckert Error BFNWSG3-048 (Rev 1) Checked: B. s~~4 E. Kakunda Date: I/>~lf~ K aeassaeeeaaaaeaeeaeeeessaeeeeeaaeeeaeeeeeeaaaaaaaaeaaaaaaaaaaaaaaaeaaeaeaeeee 1.1.2 Event Cate o The FSAR considers this as an Abnormal Operational Transient in the Reload Licensing Submittal (reference 2a) ~ 1.1.3 Plant Parameters and Safet Concerns The following parameters are affected by a RWE:

a. Local reactivity atd power ir. the. vicinity of the withdrawn control rod will increase.

b. Core average power level will increase slightly.

c Fuel temperature in the area of the control rcd could increase if adequate heat transfer was not available. The transient challenges the fuel thermal limits hence failure

 ~        1.1.4 of the fuel (unacceptable result code 2-2) is a safety concern.

Ev nt Miti ation 1.1.4.1 i e r ( -2) t n quired to mitigate the incre sed r activity and power which threaten the fuel is cram (Safety tion 17). 1.1 4.2 S stems Re uired Table 09-1 summarizes the required system functions for RWE during startup (and power) operation. It was collated fran event~nique actions and an applicable standard sequence from section 7.6 of the main report (Table 7.6-7, -8). 1 ~ 1.4.2.1 Scram ( Sa fet Action 17) The increased neutron flux will be detected by the Neutron Monitoring System (92) IRMs that are not bypassed. The scaling arrangement of the IRMs is such that the unbypassed IRM channels generate a scram signal before the detected neutron flux has increased'by more than a factor of about 10. A concern has been raised in SIL&45

 ~                        ~

(reference 2i) about a unique failure mode found for the IRMs. At this time it is unresolved for BFNP as discussed in section 7.6 2 of 'the main body of the report. Table 09-1 shows the expected protection path for the IRMs> but

kpg~ixx 9 Prepared- EX- FC/40 Date: ~Pl' Canerol Rod Vi.thdraeal E. C. Eckert Error BFN-OSG3-048 (Rev 1). Checked: afEc~w Date: ~3t//1 B. E. Kakunda <<~<<<<<<aaaa<<<<<<<<<<<<<<<<+ca><<+<<~~~~~~ also includes the APRM trip (at 15K power, reference 2j) which is presented as the IRM backup for this case in SIL-445 (reference 2i) ~ When the neutron flux signal from the IRMs (or APRHS) exceeds the high flux scram limit, the Reactor Protection System (99) provides the scram signal to the Control Rod Drive System (CRD) (85). On receipt of scram signal from. the RPS, all control rods are inserted at scram speed by 1.2 the CRD (85) to render the rea lou iArC.I CL3) ~Solhfi~ Aeg{'4n. Continuous Rod Withdrawal xn Power e 'nf ~y n e

                                                                          ~

r subcritical. S4 A+ eration

                                                                                      ~~

M+~iM.

                                                                                           +~~
                                                                                          ~/fy 1.2.1    Descri   tio       the Event xnuousl    withdraw the maxiaam xs   in operati       state F (high that   all    al        peak linear power limit ck monitor) a            gnored by the operator and rod h

i hibits al is continu til (introduced the RBM rod block physically furthe movement early in BFNP design (re ferenc The system stablizes at the higher power level. The FSAR (reference 2c> axd the Reload Licensing ubmittal (reference 2a) show that operating limits and the rod blocks have been chosen such that the Minima Critical Power Ratio (MCPR) safety limit is not exceeded-1.2.2 Event Cate orv The FSAR azd Reload Licensing Submittal consider the event to be an Abnormal Operational Transient (references 2a and 2c). 1 .2.3 Plant Parameters And Safet Concerns Figure 14.5-7 of the FSAR (reference 2c) show typical results of an analysis of this event. The increase in neutron and heat flux and hence the total core power is stablized once the RBM system trip occurs and rod movement is inhibited. The Reload O I~

                                                                                           +{ g f//P/

Licensing Submittal (reference 2a) and Core Nuclear Methodology (reference 2b) have evaluated the event from the viewpoint of MCPR and shown adequate margin with the required operating limits, The neutron flux, if allowed to exceed the safety limits can cause fuel failure (2-2) which is a safety concern.

ControI Rod %Kthdrsjaal Qspexxixx 9 Prepared: gcF E.. Eckert Page 4 og. 7 Date: /5/ Error Checked: Date: /~7</g7 BFNWSG3&48 (Rev 1) B. E. Kakunda aaaatsaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaasssstsaaaaaaaaaaaaaaaaaaaaaa 1.2.4 Event Miti ation 1.2.4.1 Applicable Safety Actions Avoid Fuel Failure Prevent movement of the control rod before the operating safety limits are exceeded (consistent with the FSAR and Reload Analysis (references 2a and 2c). This is Safety Action 23, Limit the Magnitude of event. Restore the system to normal operating conditions (Safety Action 22) by reestablishing the correct rod pattern. 1.2.4.2 S stems Re uired 1.2.4.2.1 Limit Ma nitude of Event (23) A rod block signal is generated by the Rod Bloc nitor circuitry of the NMS (92) when the setpoi t is exceeded. The Reactor ntrol function pr ides a rod withdrawa xt signal, thus i +rPer rod awal, CRD System 8 t rol rod mov ent is inh b' higher core

                                                                                       \

1.2.4.2.2 Rest o rmk 4n xtions (22) When he core stabili e high power level (but with an a r shape) the operator aust change the contro rod pattern to bring the core power to normal rated conditions. The CRD System (85) is used for this action. 1.3 Rod Withdrawl Error Durin Refuelin This transient condition is considered in the discussion in Appendix ll, "Removal of (bntrol Rod". 1.4 Lon Term Shutdown Considerations unexpected 1.4.1 Normal Shutdown The sequence provided in section 7s6 in the main bodyof the report for normal non<<isolated shutdown will be followed is considered necessary. However, if if this event, occurs at high it power , it is that scram will not occur. If the event occurs during startup, recovery from scram aai continuation of the startup should be possible and full shutdown will not be necessary.

Appendix 9 Page 5 og 71 Prepared: nate: Control Rod %bthdxaeal E. C. Eckert ~AI(p'ate: Error Checked: ~3z 7 BFNWSG3&48 (Rev 1) B. E. Kakunda aototoaaaaaooa oooooooaoooaoososoosasaaaaosaaasaaaaaaaaaaaaaaasasssssssssas 1.4.2 Sa fet Shutdown Normal, long"term response, as given above, utilizes all systems normally available to the operator. If shutdown is considered to be necessary, any. (or all) normal control equipment may be assumed to be unavailable yet the reactor can still be shut down by utilizing only safety equipment as given in the safety shutdown sequence in section 7.6. 2~0 REFERENCE S

a. TVA-RLR-002, Revision 1, "Reload Licensing Report for Browns Ferry Unit 2, Cycle 6.

b. TVA-EG-047, "TVA Reload Core Design azd Analysis'ethodology for the Browns Ferry Nuclear Plant," Tennessee Valley Authority.

c. d. Browns Ferry Nuclear Plant FSAR, Section 14.5 Browns Ferry Nuclear Plant FSAR, Sections 7.5.8., Appendix G. and 7.16.5.3.

e. NEDO-22245. "Safety Review of Browns Ferry Nuclear Plant Unit 2 at Core Flow Conditions Above Rated Core Flow Durin ycle 5; General Electric.

f. Safe Shutdown Analysis, Appendix ision 0), Kiran Kumar and K. F. Fuj ikawa, BFNWSG, une 28, 1986.

g. NEB RAC 1601 30 046), SSER ti e ription of Contro Pattern Cont o S S

h. NEB-RAC 1751 (B45 2 May 3 1, 1968, i.

j. Descri p t General on.~o f 5ga 8p >C o s lectric '-Service Information September 10, 1986, Blod' EEB-WCW-1256 (B45 8 L, including

                                                 )> SSER 7 '

osure, page 14 RBM. L-445, 1-2~ Added 15X APRM 15, Scram in non-RUN mod s.

3. 0 NON-APPLICABLE INPUTS Sections 4.2 and 4.3 of the main body of the report show results of an assessment of licensing references and FSAR questions. The items which have been identified as likely SSA (Rev 1) inputs for this event are given in Tables 4.2-1 axd 4.3-1 ~ These references were either used above in this appendix or they have been evaluated and not utilized as described below.

3.1 NEB DLM 1013 (B45 860628 725), SER Appendix A, Section III.A (page 15). This old reference discusses the Rod Block Monitor (RBM) feature in very general terms as one of the "new" protection paths on Browns Ferry It adds nothing to the SSA. 'I

Coatrol Error Rod %8thdxeeal hppendxx 9 Prepared: Checked: Fc~ E. C. Eckert j3. nate: ~/~>I l7 BFN-OSG3-048 (Rev I) B, E. Kakunda C'R R ER D IIR C Q Q Q R 0 Q le% 0 D Q Q R 015 RR R 0 N SIRES % KRCtSR W IS % Q 0 R 31 Q 0 CR R '0 R R CtS Q IIR Q CS 0 N CIR 0 ~ R $5 CZ5 RCX EE CR 3$ Cl R 3.2 NEB RAC 1242 (B45 860623 690), TVA letter dated October 28, 1975, from J. E. Gilleland to B. C. Rusche (NRC), Generic RWE Analysis showing acceptable protection using RBM ard "new" GETAB/GEXL limits. Although this again shows credit for the RBM, it is obsolete, rechecked for each cycle (e.g. reference 2a for BFNP Unit 2/Cycle 6). I 3.3 NEB RAC 1475 (B45 860628 705), SSER 8, Section 3 (pages 3 5). This RWE reference is old (BFNP Unit 3, cycle 1) and is superceded by the current reload (reference 2a).

QFE SUTNN NMlS l8900 09 Page 7of 7

                                                                     ~arab E.C. Ectert Errcr                                                           Dacha'.E.

lRHK6~8 tRev 1) Xahaia TRE09 -1 RECT WCPTAKE M SfE SER. RERLTS SKtY 4CTTQl STS STSTB f9C SNTE Code Title Code Title IL NK 'QIE CF 2-2 Fuel Failure 17 Scree Scree signal frne reactor protectiue systee 199) uill activate the control rod drive systee to insert rods. Scrm fmctim mly. LF 2-2 Feel Failure 17 Scree 92 %VII NWON F Provide TI) neutrm flux trip sill to reach@

                                                                         'rotecticn      systee f99) isee sectim  7A.2).
 ~

EF 2-2 Fuel Failure 0 Scrae '92 %UTII NGTN V Prmde NN neutrm flux trip signal to reactcr protectlm systee f99) fail~e logic. EF 2-2 Fuel Failure 17 Scrae 99 IGNR MTETX ff: Provide scree signal to the cmtro) rod drive HR) systea 185) m neutrm ecnitcrieg systsn t

82) 1N neutrcn flux trip signal.

IF 2-2 Fwl Failure 17 Scree 99 8EXHR PRSIKI)i SF Provide scrae signal to the catrol rod dri>> GR) systee t85) m eestrm soaitorieg systee f92) N%N neutron flux trip signal. F 2-2 Fuel Failee 22 Rest Nureal 85 Oo ll Operata to change rad pattm to cIrrect fa the withdraual errcr and restcre noraal rated cmdi ticns. F 2-2 Fuel Failure 23 LieNsturb 85 P5 given'bluet signal

                                                                                                 'fry i8l F      2-2 Fuel Failure 25 Li lf    2-1 bd8elease      2b  Est Pri Cent    85 C)o              SF     Perf rate isolatim gn      oe the hactcr'Protectim    Systea f99).

kML'TSIS sspeesis 10 of 3 I ShPE SHOED(5% Page 1 Prepared: Date.' f/ gael Lsseably Tascrti BFN-OSG3&48 (Rev 1) Checked: R, J. E wood Date: ~IS7 aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa

1.0 DESCRIPTION

OF THE EVENT The Fuel Assembly Insertion event is described in FSAR, Appendix G and FSAR section 14.5.3.4 (references 6a and 6b). This appendix is based on the previous SSA issue (reference 6d) plus new inputs and comments subsequently received. This event occurs when a fuel assembly is loaded in an incorrect location (during operating State A). The core load will be carefully checked during mrmal surveillance. Extremely careful procedures are followed (e.g. reference 6e) to be as certain as possible that all- bundles are loaded correctly to find any error that might be made, correct the bundle location, and avoid operation with a mislocated bundle. Consequences of operating the reactor in other operating states, (e.g., to full power), without having corrected the insertion error are discussed in Appendix 31. The refueling interlocks require that all control rods must be fully inserted before a fuel bundle may be inserted into the core. Any single fuel assembly location error in combination with a withdrawn control rod,is shown in the FSAR to produce a K (effective) less than 1.00, thereby precluding a critical or hazardous condition occurring when the bundle is misloaded or when an adjacent control rod is withdrawn for the 'functional azd subcritical 2.0 EVENT CATE The Fuel As e 1 i al t Opera ional t S 3.0 PLANT PARAME RS AND SAFETY CONCERNS A fuel assem ertion error does insert some positive reactivity, possibly more than if the bundle was properly loaded. However, since the core is designed to remain subcritical in this situation (even with a control rod withdrawn), there is no power response. No safety concerns exist for this refueling case. The core is designed such that it is subcritical under the most reactive conditions that could be caused by an insertion error, even with the stongest control rod fully withdrawn. Therefore, any single fuel assembly can be positioned in any available location without violating the shutdown criteria.

ShPE SHDTDOQR hMLMXS hppendix 10 P repared-

   ?hei kescably Insaetxon Checked:

BFN-OSG3-048 (Rev 1) R. J. wood 1 essaaaoaaaa-aaaaamaaaaaaaaaaaaaaxggsgsamaaaagxssaaaaasaszasaaaasaaaaaaaaessaaaaassaaeaa

4. 0 EVENT MITIGATION 4.1 A licable Safet Actions No safety actions are required. The reactor is shutdown and in the "REPUEL" mode. In this condition, no single fuel insertion error will cause criticality. Any single fuel assembly location error in combination with a withdrawn rod will produce a K (ef fective) lese than 1.00, thereby precluding a hazardous condition occurring even if the strongest control rod is withdrawn for the functional and eubcritical test. The refueling interlocks must be operating correctly to prevent the occurrence of this event in combination with a control rod withdrawal or position error. The refueling interlocks do not allow the refueling platform over the core with a bundle unless all of the control rods are fully inserted; nor is rod withdrawal permitted while the refueling platform is over the core.

4.2 S stems Re uired'o sy tern response is required to mitigate this egest since no safet actions a e 8g; c re loading surve avo id o 1 a ro n o (. ce 6e) de (see to Append

5. 0 LONG-TERM S Since the reactor is already shutdown and no power transient is created, continuation of normal shutdown/refueling'rocedures is suf ficient provided core loading surveillance procedures (e.g.

reference 6e) are followed to locate any misloaded bundles before reactor operation.

6. 0 REFERENCE S

a. Browne Perry Nuclear Plant FSAR, Chapter 14; Section 14.5.3.4, Fuel Assembly Insertion Error During Refueling.

b. Browne Ferry Nuclear Plant FSAR, Appendix G; Plant Nuclear Safety Operational Analysis, Volume 2> Chapter 3, Copy 42.

c. Browne Ferry Nuclear Plant FSAR, Chapter 7; Section 7.6, Refueling Interlocks.

d. BFN Safe Shutdown Analysis~ Appendix 10 (Rev 0), B.'. price and B. H. Koepke, BFNWSG3&48, June 28, 1986.

Shm SHUTDOQR hMLMIS kppead~ IO Page 3 of 9 Prepared: c. Date. Xg 7 Pae1 hssemb1y Tnxxertxcxa ert j Checked: Date: lW BFNWSG3&48 (Rev 1) R.. E lwood aa~aaaaaaaaaaaaaaaaaaaaassaaaaaassaaaaaaassaaaaaaaaaaaaaaaaaaaaaaaaaassaaaaaaa I

e. NEB RAC 1263 (B45 860624 016), TVA letter dated September 25, 1978 from J. E. Gilleland to H. R. Denton (NRC), BFNP, Unit 2, Cycle 2, Technical Specification (Enclosure 2, Appendix A).

7. 0 NON-APPLICABLE INPUTS Sections 4.2 axe 4.3 of the main body of the report show results of an assessment of licensing references and FSAR questions. One item has been identified as a likely SSA (Rev 1) inputs for this event (Table 4.2-1). It was not utilized as described below.

7.1 FSAR Volume 7, Question (Psl0. This question and response really address the issue of plant operation after an undetected bundle loading error. It is utilized in Appendix 31. -0 4

Append~ 11 Page 1 oP 4 Prepared: Con~I Rod Removal C. ert Brrar Checked: Date: ~l3t 8 j BFNWSG3&48 (Rev 1> R. J. Elwood aaaaaaaaaaaaaaaaaaa aaaaaaaaaaaaaaaa aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa 1 e0 DESCRIPTION OF THE EVENT An erroneous manual control rcd withdrawal during refueling or error during control blade removal theoretically has a potential for inadvertant reactor core positive reactivity reaction. This appendix is based .on the previous SSA issue (reference 6f) plus new inputs ard comments subsequently received. The event is analyzed under chapter 14 of the FSAR (reference 6a) aai is further evaluated as event 23 under Appendix G of the FSAR (reference 6b) ~ Reference 6b defines that operating state A is the only state in which this event need be an operational consideration, since the action of a manual control rod removal could only occur when the reactor head is removed and manipulation of the refueling equipment over the reactor core is p'ossible. Core rod withdrawal errors in other operating states are discussed in Appendix 9 of this report.

2. 0 EVENT CATEGORY Thi s xs categorized i f F (reference 6b) as an bnormal Oper e t 3.0 P P E CERNS 3.1

                '.:-'i4     ..

re-'Reactivit r e r of control material from the core has the theoretical p tential of producing a critical nuclear reaction. The critical nuclear reaction could result in fuel failure (2-2'l.

4. 0 EVENT MITIGATION 4.1 A licable Safet Actions The nuclear characteristics of the core assures that the reactor is subcritical even in its most reactive condition with the most reactive control rod fully withdrawn or removed in operating State A. This is confirmed periodically according to shutdown margin testing procedure (reference 6d) axd analytically for each reload core (reference 6e).

4.1.1 Avoid Fuel Failure Durin Removal of a Control Blade The design of the control rod (incorporating the velocity limiter) performs a passive safety action because it inherentlv limits the disturbance (Safety Action 23). It does this by not

- Ih1% SHUTDOQR h%bLTSIS kppendix 11 Prepared: FcF ert Page 2 Date..//7C of 4 7 Control Rod Rascal C F Error Checked: Date. I. 3I & BFNWSG3&48 (Rev 1) R. J. Elwood aaaaaeeeaassaeeaeeeeaeeessaaeaeaaaeeeaaaaaaaaaaaaaaaaaaaaaaaaeaaaaaessaaaeeeee physically permitting the upward removal of any control rcd without the simultanous or prior removal of the four adjacent fuel assemblies, thus eliminating any hazard condition. Therefore, no other safety actions are required for this situation. 4.1.2 Avoid Fuel Failure - Control Rod Withdrawal Durin Refueli When the reactor mode switch is in "REFUEL"; only one control rod can be allowed to be withdrawn. Selection of a second rod must initiate a rod block thereby limiting the disturbance (Safety Action 23) by preventing the withdrawal of more than one rod at a time. This will prevent any condition which could lead to inadvertant criticality due to a control rod " withdrawal error during refueling when the mode switch is in position (references 6a and 6g) ~ No other s actions re - required. All rods are blocke interlocks dur yl

                                                                                  ~g).oa(kg in  order to           any unpla                v         u 4.2   S  stems Re   uir Table 011-1    s      rag k he required system                ons. It was collated fran     e ent~nique actions            ~

4.2.1 Limit the Diat'afe Action 23) The Control Rod Drive System (85) performs the rod withdrawal function. Refueling interlocks initiate rod blocks which prevent withdrawal of mere than one rod at a time. These interlocks become active when the reactor mode switch (part of Reactor Protection System, 99) is in "REFUEL" position and therefore, the mode switch and the refueling interlocks perform the safety action of limiting the disturbance fran involving the withdrawal of one than one control rod at a time (reference 6g). When the fuel is actually positioned over the core, refueling bridge interlocks (Reactor Manual Control subsvstem of (bntrol Rod Drive System, 85, and Fuel Handling and Storage System, 79) together with the reactor mode switch in "REFUEL" (part of Reactor Protection System, 99) pr'ovide a block of any control

                'od drive withdrawal (also utilizing the CRD System, 85)

(re ference 6g) .

Shm SHDZDOM hMLMIS ~radix II Page 3 of 4 Ccmtrol %oct Reaamal Prepared: C. rt Date: ~II 7 Error Checked: Date.. ~9 BFN-OSG3&48 (Rev 1) R. J. lwood eaeeaaaaeaaeeaeaaaeaaaaaaaaeeaeaaaeeaaeeaaeeeaaaaaaeeeeaaeeeeaaaaaeaaaaassea

5. 0 LONG-TERM SHUTDOWN For this postulated event, the reactor is already at cold shutdown cond itione. Rt

6. 0 REFERENCE S

a. Browne Ferry Nuclear Plant FSAR, Chapter 14; System Safety Ana lys is Browne Ferry Nuclear Plant FSAR, Appendix'; Safety Operational Analys is.

c. Normal and General Operating Instruction No. GOI-100-3, Refueling Operations, dated February 12, 1986.

d. Browne Ferry Nuclear Plant. Refueling Test Instruction No BF.RTIW dated August 28, 1984.

e. Browne Ferry Nuclear Plant Licensing Report (TVA-RLR&02, Revision

1) Unit 2, Cycle 6.

f. Safe Shutdown Analysis, Appendix V; C, Kobold, BFNWSG3&48, June 28, 1986.

ll (Revision 0), Kirti Kumar and

g. NEB RAC 1671 (B45 860701 647), D&AR Amendment 6> Responses to AEC Question B-4, Potential Power Excursion during Refueling.

7~0 NON-APPLICABLE INPUTS Sections 4.2 a&=4.3 of the main body of the report show results of an assessment of licensing references and FSAR questions. One item has been identified. as a likely SSA (Rev 1) inputs for this event as shown in Table 4.3-1. This reference wae used above in this appendix. 1 L a s a

                                                                                                    "~ h f ~
    ~    a
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QfE SIMN NLTSIS NBIH 11 FaSe 4 hntrof Rod Reeoval Ech art Satlj ~/ E. lRHK6RNS fRev I) hah@ fjatec ~I~~7 R.}. Efuood j TARE lf -1 I KD INKQPTAIE TN 9%L RERLTS INETY le%: Sm $ 5TEff

    'TATE hde     Title      hde     Title   fL       ICK A    2-2 Fuel FafILre 23 Lie Ijfstorb   7lP RR  fNLII           L    Refufig istsrloct    suftches   ee   used'to fisft emrso A    2-2 Fuel Faflwe25 Lie      Nstsri  $ 5 ON                  L    Tbe velocity lfaitar deahp fs e6 that aijaceet hei hedle <<so&lies ossh ie sftbdrI prie to costrol rod reaoral.

A 2-2 Fuel Faifee25 Uohstorh $ 5 ON .L Reiuef fag foterlocf3 faftfatsi by Reactor lhaai Control Systeata sohsystso of 'tbe cootrol Nl drive systsel are used. Rod'$focts est he oplfati coal e A 2-2 Fuel Failure 25 Us fjfstsrb 99 KCTI PR0IElN %F Reactor sode suftch fs used hr RREL'ude. - ~ 4 j vj ~

                                                                'j I
                      ~            ~

j

She SHDXDOI& h%iI.~Zg dppendix l2d 5E. C. FEckert C Page Date-1 of ~9t.c<. ~ I Preseme I Reyxlator Paxlare Open Checked: Date: p BFNWSG3&48 (Rev 1) ~ ~ arg

                                                                             ~+~&'tntxnt>>>>>>>>>>>>ta>>pn>>>>>>>>>>sa>>tn>>sa>>ntns>>>>tn>>>>a>>>>anat>>ns>sntn>>>>ntstnt<<tn>>

1 no DE SCRIPTION OF THE .EVENT The turbine pressure regulators normally provide continuous system pressure control by modulating the turbine control valves (or, during startup, the turbine bypass valves). Two different bounding cases are possible if a regulator fails. One is when failures occur so that steam flow is reduced 'and the pressure initially increases. This case is addressed in Appendix 1C. The second case which is presented here, occurs when either regulator malfunctions in the open direction such that the pressure initially decreases .This appendix is hesed og, tPs previous SSA issue (reference 6c) plus new inputs and comments subsequently re being supplied v@d. This case can only occur when 4'bineis which is in reactor operating states Cp D, E, axd F (in state A, the vessel head is off) ~ The FSAR, Appendix G, addresses this case as event 25 (reference 6b). When the pressure regulator malfunction creates a failure which results in an excess steam flow, the initial response is for the pressure level to decrease as the steam flow increases. The maximum rate of steam flow analyzed in the FSAR section 14.5.4.1 (reference 6a) is 110 percent of the turbine design flow (limited by the turbine control maximum flow limiter) . This is equivalent to 15 percent of the reactor rated steam flow. There are three resulting sequences which may occur as steam flow depressurizes the reactor, depending on the sta the r ctor when the regulator fails. The first is when normal high r opera-tion of the reactor exists with p reactor in th " " nj e. Then the reactor will depres axd the Ma s 'og;V ives will automatically be ed to stop th isolation, f edwater is lo 1 r t in ok wa er level. The con+ u c ni tia 1 depress ration causes a wate )e9e$ ~ e i ably reaches t level trip. This ca sea trx 8 ormal feedwater ard also trips the main turbine e ly in the event. Th essurization is milder, but still eventuall results in 1 essure MSIV closure ard a low water level indication in t~ actor. The third sequence occurs when there is 'a depressurization from low power, low pressure conditions (e.g., during plant startup) ~ This may eventually result in a high flux scram signal from the Neutron Monitor'ing System (IRM) when manual isolation is initiated to stop the depressurization.

k~~hc 12k Page 2 o Prepared: Date: kKcssurc Rggulator E. C. Eckert Failure Open Checked: Date: ~N~$ 7 BFNWSG3&48 (Rev 1) T. L. Garg aaaeaaaaaaaaaaaaaaaaaaaaeeeaaaaaaaaaeaaaaaaaaea a~aaa 2~0 EVENT CATEGORY The BFNP FSAR considers this event to be an Abnormal Operational Transient. C

   '.0    'PLANT PARAMETERS AND SAFETY CONCERNS Parameters   associated   with the severity of the disturbance are          as follows:

Main steam flow rate increases to above normal. Reactor pressure and temperature decrease to below normal operational values. Rapid depressurization of the reactor (and associated thermal changes) could lead to unacceptable safety result 2-3. Reactor water level rises during depressurization, then drops below normal after isolation. Loss of reactor coolant (potential fuel damage) could lead to unacceptable safety result 2<<2. Reactor pressure increases after isolation to above the normal - ~ pressure range. Reactor pressure above normal after isolation causes additional reactor stress (unacceptable result 2-3) Heat up of the

                                                                        ~

suppression pool could cause unacceptable containment conditions (unacceptable results 2-4). High flux in the reactor core may occur (for the case where the regulator failure happens at a low initial po direct isolation scram is not active.) IRM response to a scr ~ Increase in power could lead to po el damage (unacc e results 2-2).

4. 0 EVENT 4.1 S 10 Th s y actions considered are ba conditions identified in the FSAR. The sequen ved are based on high or low ini ial power 'ns', the possible sequence involving a high wat vel trip was not included (as per the FSAR). This sequence would utilize protection very similar to the feedwater controller failure event given in Appendix 23 in addition to some of the depressurization items listed here.

C

                                             ~  ~
                                        ~   ~

Qqseodxx 12A Page 4 of Jd" Pxosscre Regulate Prepared: E. C. Ecker Dat,e.. ~~7 Failure Opea BPNWSG3&48 (Rev 1) Checked: <r T. L. Garg Date: 0~7

 ~~taDtta~tttataaa~attatatataataaaaatattaataaattataaaaaataaaattaD    aaQ 4.2.2        Reactor Vessel Isolation (Safet                      Action 20)

If the pressure regulators have failed in the open position, the increasing main steam flow.to the turbine will cause the reactor pressure to decrease. If the reactor is initiallv in the "RUN" mode, then the depressurixation will initiate main s earn line isolation at approximately 8gpsi~(reference 6d). The Primary Containment System (64) is responsible for the logic associated with the isolation after receiving a signal from System Ol (Main Steam) when the pressure decreases to the low pressure setpoint. At this point, the signal shall be sent back to the Main Steam System (01) to close the Main Steam Isolation Valves (MSIVs) ard stop the depressurization. When the reactor is not in the "RUN" mode (as in reactor states C, D, anl E), then initiation of the isolation aaxst come fran manual MSIV closure initiation by the operator (i f turbine valve closure cannot be accomplished to terminate the event). De,hatt OCCur kaf a~ ~~iW+. CA@ /CISE CL3) i$0lafcQ ag+M4ima fgCk'~S 4.2.3 Scram (Safet Action 17) ECC~p Closure of the MSIV valves (01) with the reactor mode switch in RUN shall signal the Reactor Protection System (RPS, 99) to signal the Control Rod Drive (CRD 8 ~ ~ ~ c ram. In the se uenc e switch is not in RUN a high flux is d tected (following isolation) b t e toring Syst m (NMS 92 io h shall sign 1 r 4.2.4 Press e f e ion 18) Upon essel isolation increase in pressure. The p e ief portion of the Main Steam System (01) mav be needed to activate the pressure relief valves and vent the steam into the suppression pool (Systems 10 and 64). 4.2e 5 Core Coolin ( Sa fet Action 19) Upon isolation, the feedwater turbines will lose steam anl thus feedwater flow will be lost. This will be sensed by the Peedwater System (03) are when the water level reaches low level (L2) the feedwater shall signal Reactor Core Isolation Cooling System (RCIC, 71) azd the High Pressure Coolant

dppeedds lsd Page 5 of ~" Date: Pressure Regulator E. C. Eckert Zailare Open

                                                                                              ~/l'ate:

BFN-OSG3&48 (Rev 1) Checked:, ~~Md

                                                              ~    ~   arg staaaaastaassssaaatsaaaastaaaasasaaaaaaatsaaaaaaaaaaaaaaaatsaaatsaastassaaatsaaaaaaaasaaaa Injection     System (HPCI, 73) to initiate core cooling functions. HPCI and RCIC shall supply flows to control atd maintain the reactor water level. Either system will perform the function and shall share the single failure capability between them. Some of the support systems for RCIC and/or HPCI operation are also shown in Table 12A-1                  ~

4.2.6 Containment Coolin (Safet Action 30) The release of steam into the suppression pool via the pressure relief function heats the pool. The operator needs to observe suppression pool temperature and level (using the instruments in System 64) axd following the appropriate procedures, manually initiate suppression pool cooling using that mode of the Residual Heat Removal System (RHR, 74). Some of the support systems for this mode are also shown in Table 12A-l. 5.0 LONG-TERM SHUTDOWN 5.1 Normal Isolated Shutdown The steps conditions. provided in Section 4.0 Should t

                                                                   'table, r eseion pool.temperat h  t      shutdown pproach upper acce in t iated.

imits, manual d r u i a ooldown may be The se u o 1 iven in secti 7.6 7. -1). 5.2 Safet Normal, long-tean response, a ove> utilizes all systems normall availabl operator. If any (or all) normal control pment is assumed to be unavailable, the reactor can still be shutdown by utilizing only safety equipment as given in the Safety Shutdown sequence given in Section 7. 6, starting in this event from an already isolated condition (Table 7. 6-3). 6~0 REFERENCE S

a. FSAR, section 14.5.4.1.

b. FSAR, Appendix G.5.3.5.2, event 25.

c. BFN Safe Shutdown Analysis p Appendix 12A (Revision 0), H. S.

Robbers, aA B. J. Cheek, BFNWSG3&48, June 28, 1986.

d. NEB-RAC-1186 (B45 860623 719) p TVA letter dated January 26, 1977 from J. E. Gilleland to B. C. Rusche (NRC), change of Low Pressure Isolation Setpoint to 825 peig.

k Prepared: xm ~~ age 6 Date: of Pzee care Regula~r C. Eckert Paxlare Open Checked: Date: <~~~& BFHWSG3&48 (Rev 1) T. X. Garg aaaeeeaaeeeeaaaaaaaeaaeaaeaaeeeeaaeeeaaaaaaaaaaaeaaeeeaaaaaaaeaeaaaaaaeeeaa 7~ 0 NON-APPLICABLZ INPUTS Sections 4.2 azd 4.3 of the main body of the report show results of an assessment of licensing references azd PSAR questions. The items which have been identified as likely SSA (Rev I) inputs for this event are -given in Table 4 3-1. One is used above in this appendix;

                                    ~

the other was not used as explained below. 7.1 EEB-WCW-1255 (B43 860617 939), SSER 7.2 ~ 1 This reference indicates the installation of equipment to close. the MSIVs froa a high water level signal (Level 8) in all reactor operating modes except RUN. Although this action would impact this event (12A), it has not been included in the required safety actions in accordance with Assumption 3.2 (see the main body of the report) Although it requires further verificati.on, it appears that this feature has been deleted from the plant since no drawings have

                                                                                     ~
          ~ been found which show it, azd no current Tech Spec has been found that requires it.

SlfE 9$ 106N NN.TSIS APPM)II 12A 7 of 9 Page Prepared t g ~~(>IN Pressure Regu) e - Bpen

      %)HjSSRHB IRev ato'ailm I)

Dsckeh E.C. Eckert

                                                                                            ~O T.L   Barg Sam   ~l TARE 12A  "I REACT    IIACCEPTARE                                TUA                    SAFE
  'IFER.       REM.TS               SAFETy  ACIISI     SIS       SISIE)l      FIN:

STATE Code Title Code lith NL NE GK IF 2-2 Fuel Failme 0 Scrm 1 'IIIN SIEltl SF Provide ISIU < %C open trip signal to reactor prntectim systen I99) 'fail~ logic. IF 2-2 Fuel Failure 17 Scran SF Scran signal fron rnacbr protnctim systa I99) illactivate the cmtrol rod drive systen.to insert rods. Scree haction mly. IF 2-2 Fuel Failure 0 Scran Provide NI nautrm flux trip signal to reactor protectim system I99) Isae sectim 7.h.2). IF 2-2 Fuel FaIILre 17 Scree 99 REACTNI PRmECIN O'rovide scran signal to the cmtroI rnd drive

                                                                                     $ $ ) systen    Ilm) cn min stean systen I01) l61Us
                                                                                     <  9IC qmn trip signal    N    4n IN node).

If 2-2 Fuel Failure 0 Scran 99 RBCHR PROIEIN  % Provi* sera signal to the cmtrol rai drive IIm) systen II5) m nentrm'nmitcring system I92) IR)i neubm flux trip signaI. IXEF 2-3 Syst Stress 18 Pres Relief I SINSIM SF SRUs open m high reacbr pressure. GXF 2-3 Syst Stress 18 Pres Relief 10 IOILER UNIS'h 5$ Provide path hr naia stem systen I01) SRUs stean hlondoo to m I IXEF 2-3 Syst Stress 18 Pres Relief Q PRI CNIAINBIT Accept 9Ãs stean hlondonn Ifrcn holler vats 2-2 Fuel Failure 19 Gre Cooli I I[ IXEF C0EF 2-2 Fuel Failure 19 Core Cooli I NON SIENI Provide stsa fcr;)fCI 03) turhine in snppcrt IXEF 2-2 Fuel Failure19 Core Cooling 2 NDBNIE Provide ncrnally tpn eater supply hr NC systen I71). IXEF 2-2 Fue) FaIILne 19 Core Cmling 2 CNSKATE 'rovide nmnally open abater apply for IPCI systnn I73) Initiatim m lm eater level II2). IXEF 2-2 Fuel Failure 19 hre Cooling 3 FEHNTER SF Provide Im reactm sate level signal II2) to RH) systen I74) logic fcr KIC systen I71) lnitiatim.

SAFE SHJTIQN NLYSIS APPMII 12A Page 8 of 9 '. Prepared) >t. ~ht( P ressure Regal atm E.C. Ectert Failure - tjpen Gtectedl oat+ ~4/r ++P IRev 1)

                        'HM~

T.L Sarg TQE 12A -1 REACT lNNXPfARE TN QFE OPER. RERLTS SAFETY ACTISI SYS SYSIE)t RK STATE hde Title Code Title )1. NtE NK tXEF 2-2 Fuel Failure 19 Ccrc haling 3 FEHifATER Provide path far RCIC OI) flaw to tbe vessel thrmgb the feedwater spargers. IXEF 2-2 Fuel Faillee 19 Ccrc Cooling 3 FEHRATER SF Provide I<a reactcr water level signal tQ) to IfCI systen O3). tXEF 2-2 Fuel Failure 19 Care Cooling 3 FEHNTB Provide path for )fCI O3) fin to the vessel through tbe feedwater spargers fm )fCI initiaticn cn Ia wats level tLB. GXF 2-2 Fuel Failure 19 hre Cooling M PRI NiTAINBIT Provide alternate sarce water supply far KIC systen t71) through Care Spray systa l75) piping fra suppressim pool. Provide eappressicn pool level indication. Accept RCIC turbine esbaust steaL CKF 2-2 Fuel Failure 19 hre holing M PRI GftTAIWBIT Provide alternate sarce water supply far )fCI

                                                                          ':systen t75)      fra              'srT~ir            I m m    la water    level tl2) si tttttt Itt'I tltttae     el tt     lf CKF 2-2 Fuel Failure 19                                                                                  to             ]Mt Ia tati    ter  RCI    systen i   tlat>m.

CKF 2-2 Fuel Failure 19 hre Cooling 71 RCIC EFS RC

                                                                                     '       'eactar                 water level     tl2)
                                                                            'signal'fra      RN systen t74).

tXEF 2-2 Fuel Failure 19 Care Cooling 71 RCIC SF Provide HXS ATI) power to Rt systen tN) law

                                                                         .'ater        level tQ) Instruaentatim far IfCI systen O3)  initiatim.

GEF 2-2 Fuel Failure 19 Care holing 73 )fCI SFS )fCI initiatim m Rf systee tN) law reactor water level tL2) signal. tXEF 2-2 Fuel Failure 19 Ccrc Cooling 7I 8% Send feehater systen tN) low reactor water level signaI tl2) to initiate RCIC systen Of). C0EF 2-2 Fuel Failure 19 Care Cooling 75 CÃE SPRAY Provide flaw path fcr alternate Narce of water to RCIC systen t71) fra Rppressica pool tM), SF Provide Ia pressure'ignal 'tin nain stm line at turbine) 'to prinary cmtal~t systen tbl) failmfe logic. IRNt node)

Fai SAFE SUMN ANALYSTS APPE)ND 12A Page 9of 9

                                                                                                .C F~        ~~'      O  tei  M<~~y e

Pr esture Regulator E.C. Eckert lee - Open 7 NHS)RN8 IRev 1) T.L Sarg YARE 12A -1

       )BCT MXEPTA9LE                                TYA                      SAfE IPER.       RENTS           SAFETY ACTlON     STS       STSTB         FIIC STATE  hde     Title       hde     Title     N).        Nf          'KE F       2-1   Rad Release   20  8PY  Isol       1  )I)TN STEAN        SF     Close lslYs and nain stean drain lines m lm presare signal Iin cain stean line at turbine) free prinary cmtainant systen IM).              III    node)

GK 2-3 Syst Stress 20 8PY Isol 1 NTNSTEO) ll Operatar senses blordsa and initiates nanual

                                                                                    %1U    cloare   if escessive    depresariaatim cannot be stapped uhen     helm narnal low presare liaits ano   not in  II   node.

F 2-1 Rad Release 20 8PV lsol Q PRI GNATQSiT SF Send naia stean systee I01) lm pressure sigoal fin aain stean line at turbine) to initiate )STY (01) cl mre. IRIII node) Ct F COEF IF 2-1 2-1 Rad Release 2-T Syst Stress Rad Release 20 23 2b RPV lsol Ua Oisturb Est Pri Cont 99 47 85 REACHI GI PROTECTN TIRSllK IX)IIHL hlL SF

                                                                            'F Provide RIf Node signal to systee stean line pres@re isolaticn inter)act.

valves uith the pressme rQatcr tailed apm 119 per FSN analysis Icmt rocea5usted liQt). Per fare isolaticn actimls) upan 8 fcr

                                                                                    ))asian Stean Flw thraugh turbine plus bypass lm receiving scrae is I+

signal frm the Reactm Protecticn Systen I99). COEF 2-4 Cant Stress AA Cent holi lSF 1 cooling GKF 2H Cent Stress R Cant Caol )evel t ( m paol cooling f

                                                                                                                                       "i>

GEF 2-4 hnt Stress 30 Cmt Cooling 67 ee 04) suppressim pool cool ng

                                                                                ~   undec IXEF    2H    hnt   Stress R    Cmt Cmling 74      IR                 )KF Provide suppressim pool           ueling functim.
                   /{~Ic" 4Q 74'o        3  t~ J    r    5gQPy        6+i~ $ 47        >   ~vJsfj~ f 3
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ShPK SHUZDOiK kMLMIS pppepdia 123 Prepared: Page Date'. 1

                                                                                              /of7f 8 Inadvertent Opexxxag oK                                   E. C    ckert hI1 bylmaa Va1vea                      Checked:                                    Date:    VA/~

BFNWSG3&48 (Rev 1) H. S. Robbers

  ~ 33-eaaaaSaaacSeCacSaSaeeeaa~gggg~~~~~gggg~gzzgaaaaaaaaaaaaeaeezSaaaaaaaaaaaeaaaWaaarzewaa 1  o 0   DE SCRIPTION OF THE EVENT This transient event results in the simultaneous opening of all turbine bypass valves (FSAR, question 14.5 of reference 6c), axe is only of conseqence when the turbine is being supplied steam (operating states C, D, E, aced F). This appendix is based on the previous SSA appendix issue (reference 6d) plus new input ari1 comments subsequently received.

Due to control malfunction, an inadvertent. opening of all bypass valves may cause a sudden drop to the inlet turbine pressure. The pressure drop would be equivalent to the full capacity of the bypass valves (about 25 percent of rated steam flow) In the analysis,

                                                                        ~

(provided in response to question 14.5), the reactor was assumed to be operating at 105 percent power and 100 percent core flow and in the auto-flow control mode. The pressure drop is sensed by the pressure regulator and will caxse a corresponding closure of the turbine control valves. The analysis assumed that the pressure regulator does not fail since it had to be operating properly at the start of the event. Refer to Appendix 12A for the Pressure Regulator Failure " Open event for a related transient. 2.0 EVENT CATEGORY The Inadvertent Opening of All Bypass Valves event is analyzed as an Abnormal Operational Transient.

                                                                 /

3. 0 PLANT PARAMETERS AND SAFETY CONCERNS 3.1 Pressure For the hi h t e s rews drops belo h a o stabilized h ct ~ \ hei~ p egu ator. If th la nt is in the x o control.~ody the action of the pressu e gl setpoint ad ustment (as assume~in the a cause the system pres ure to a out 50 psi below the initial pressure.

For lower power initial conditions producing steam less than the bypass capacity, Inadvertent Opening of All Bypass Valves event will cause a greater depressurization. This would require closure of the Main Steam Isolation Valves (MSIVs). 44-hbe This will cause a subsequent

             ~     pressure rise until the safety relief valves open.

,0

ShFE SHDI3NQH hMLYSIS hppendix 12B Page 2 of 8 Prepared: Date: Imuhrertent Opening of E C. E rt hll Bypass Valves Checked: Date: Zr).le > BFN"OSG3-048 (Rev 1) H. S. Robbers ggsgtmtmsgamgstmtgggtmtmsgtmagssmtmtmggtgagsng ggm:gssg~smtm~tg~mtsmtgagggeggotmggggagsetgantggtgggstgatmaatmtnomtgtgtgtmatgtgtac The depressurization that results from the bypass opening and potential isolation could cause nuclear system stresses in excess of that allowed for transients by applicable industry codes (2"3) . Relief valve flow to the suppression pool after isolation could cause unacceptable containment conditions (2-4) . 3.2 Reactivit Power and Core Flow For the high power case, the reactivity and neutron flux respond mildly to the pressure disturhance hpt sep)le ~au near the inital value. Zn 'the auto flow control c" e anaJgsgeuphe final system pressure is lower (suppressing p oewr) and core fiow is increased to its upper limit (raising power). The net result would have the power settle out near the initial value after a mild transient. The operator would manually attempt to restore the reactor to normal operating conditions as soon as possible since the condenser vacuum may be degrading and a complete unit trip could result, but there is no inxnediate safety concern. The bypass valves should not be closed all at once or else the resulting transient could cause a power transient that approaches scram ~ For the lower power case, which may depressurize sufficiently to cause (or require manual) vessel isolation, actions are needed to ensure that excessive fuel failure will not occur (2-2). 3.3 Reactor Vess er evel

                                                       '                s For the high p                                                                      y e0".('xe8eiehce 6c), the wat                 v         i                           1                    ef:.control of the Feedwate              y                     a                        1. a   lo      occurs Io ss y

of feedwater ccurs and additional'water supply must be prov ded to avoid fuel failure (2-2)..todd l'+tdCpS

4. 0 P~~~ may d EVENT MITIGATION

                                                                                    ~      L dna(A 4.1       A     licable Safet            Actions 4.1.1        Avoid Unacce table                 RPV  Stress (2-3)

Closure of the Main Steam Isolation Valves is required for low initial power cases to minimize stress to the reactor vessel by limiting the amount of uncontrolled depressurization and

Shor, SHUZDom halLMIS Page 3 o 8 Prepared: Date: '7/ Inadvertent Owning oK E. C rt All ~mes Va1vea BFN-OSG3W48 (Rev 1) Checked: Date-. MMm H. S. Robbers aggaasaaasaaaaaccaaasaaosaaeaassassaelaaaoa+++I*~~~~~~~+~~ cooldown (Safety'ction 20, Reactor Vessel Isolation). Pressure relief (Safety Action 18) may be needed to prevent excessive pressurization of the nuclear system occurs. if an isolation 4.1 .2 Avoid Fuel Failure (2-2)'. A Sera'm signal (Safety Action 17) would be needed has occurred to prevent excessive power increase and any if isolation unacceptable fuel consequences. If isolation ard the resulting loss of normal feedwater occur, Core Cooling (Safety Action 19) may be needed to prevent e-cessive fuel failure by avoiding or minimizing the uncovering'f the core. 1.3 Avoid Unacce table Containment Conditions '(2-4) Pressure relief valve flow (due to Safety Action 18). will heat up the suppression pool, requiring initiation of Containment Cooling (Safety Action 30). 4.2 S stems Re uired Table 12B-1 summarizes the .required system functions It was collated from event~nique:actions and applicable standard sequences from section 7.6 of, the main body of the report (Tables 7' 7s lit'2'7~. 181 24'4) ~ 4.2.1 Isolation (20) If the reactor is in -the RUN mode, then'he ' be initiated from a ma w, pressure sensor (Sy tern

01) which inputs into the primary ontainment '( v 6P) log' then back to System 01 f t 'jpaetog is not in th r 't 'e! enough to challe h e c s Ken te isolation u i t ly (Sys tern Ol ) .

4.2.2 Scram (17) System 01 (Main Steam) sends a signal upon closure of the MSIVs to the Reactor Protection System (System 99) to initiate a scram to the Control Rod Drive System (System 85) ~ For high neutron flux due to manual isolation from lower power initial conditions, the IRM (System 92) will initiate the scram signal to the RPS.

kppendxx 125 Page 4 of 8 Prepared: Date: ~//7/ Inadvertent apexnng oE E. C. E rt k11 bypass Va1vel SFN-OSG3-048 (Rev 1) Checked:

                                                          ~   ~        ers Date:   3/a/F 'l aaaaaaaaaaaataaaaeaaaaaaaaaaaaaaaaaaatageaaaaaaaaaaacataaaaaaaaaaaaaaaaaa 4  2 3    Pressure     Relief (18)

The safety relief valves (System 01) may be needed to prevent excessive pressurization of the reactor. They discharge through downcomer pipes (System 10) to the. suppression pool (System 64). 4.2.4 Core Coolin (L9)

                                                                               ~a sS t. d4~4e             fc'Y*

To provide core cooling and to maintain water level in the vessel, the Reactor Core Iaolation Cooling System (RCIC, axxl the High Pressure Colant Iniection System, (HPCI, System System'1)

73) are initiated due to a level 2 low water signal (Peedwater, System 3). Other support systems for HPCI ax@/or RCIC operation are also included in the data table (12B-l).

4..2.5 Containment Coolin (Safet Action 30) The release of steam irito the suppression pool via the pressure relief function heats, the pool. The operator needs to observe suppression pool temperature. axxd level (using the instruments in System 64) azd following the appropriate procedures, manually initiate suppression, pool cooling using that mode of the Residual Heat Removal,'System (RHR, 74). Some of the support systems for this fun'/tion are also shown in Table 12B-L.

                                                  \

5. 0 LONG-TERM SHUTD(SN 5.1 Normal Isola ed Shutdown The steps conditions.

upper accept be initiated. section 7.6 pr o S e 1

                                         ~

he S s o 0 d u's ssurx zation The seq uence of e q xpment required the main body of th e t sh and cooldo is given td hacIx ar i mav if needed,'r n a e'.7.6-1. path 5.2 Safet Shutdown e Normal, long term response, as given above, utilizes all systems normally available to the operator. If any (or all) normal control equipment is assumed to be unavailable, the reactor can still be shutdown by utilizing only safety equipment as given in the Safety Shutdown sequence given in Section 7.6,. starting in this event from an essentially isolated condition (Table 7.6-3)

Page 5 of 8 Prepared: Date: ~~e-/ Openxng of E. . kert kl1 bypass Talkee Checked: Date: u~lg~ BFNWSG3-048 (Rev 1) H. S. Robbers aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa

6. 0 REFEREN(Z S

a. Browns Ferry Nuclear Pla& FSAR, Chapter 14; System Safety Ana lys is.

b. Browns Ferry Nuclear Pla~ FSAR, Appendix G; Safety Operational Analys is.

c. Browns Ferry Nuclear Plazr. FSAR, Volume 7, Question 14.5.

d. BFN Safe Shutdown Analysis, Appendix 12B (Revision 0), K. K.

Fuj ikawa and J..A. Clem, BFNWGS3&48, June 28, 1986.

7. 0 NON-APPLICABIZ INPUTS Sections 4.2 axd 4.3 of the main body of the report show results of an assessment of licensing references and FSAR questions. The items which have been identified as likely SSA (Revision 1) inputs for this event are given in tables 4.2-1 and 4.3-1. One was used above in this appendix; the other was not used as explained below.

7.1 EEB-WCW-1255 (B45 860617 939), SSER 7.2.1 This referenced indicates the installation of equipment to close the MSIVs from a high water level signal (Level 8) in all reactor operating modes except RUN. Although this action would impact this event (12B), it has not been included in the required safety actions in, accordance with Assumption 3.2 (see the main body of the report) Although it requires Rrther verification, it appears that this

                                                                                      ~

feature has been deleted from the plaW since no drawings have been found which show it,'zd no ecz. cation has been equx.res it. I >.:

SAFE QUMN NM.YSIS APPMIT 128 Page 6of 8 Inadvertent Bpening of Prepared> CC. Ecb 0 tv ~i~/ E.C AII Bypass Valves Sacked> san ~>~ RHK6RNB IRev I) H.L Robbers TARE 12$ -) IGCT INGXPTAKE TVA SAFE 5B. RERLTS SFETY ACTlll STS STSTDI FIM: STATE Code Title Code Title NL NK mE F 2-2 Fuel Failee 17 Scrae 1 SON SIGN SF Provide NSIV < %C open trip signal to reacta protecticn systen I99) fail~ logic. 2-2 Fuel Faihre 17 Scrae 8$ CR0 SF Berm signal fm reactor protection systen I99) uill activate'he ccntrol rod drive system to insert rods. Scran functtcn only. 0 2-2 Fuel Failure 17 Scraa 92 KHAN )MTN  % Provide TRN neutron flux trip signal to react~ protectim systen 199) Isee sectim 7.L2). RI F 2-2 Fuel Failure 0 Scrae 99 ISCHR PRMEIN SF Provide scrae signal to the ccntrol rod drive KRB) systen I85) cx) nain stean systa I0)) ISIVs

                                                                       < %C cpen trip signal Iif in RIM node).

0 2-2 Fuel Failure 17 Scrae SF Provide scrae signal to tbe cantrol rod drive . IN0) systen (85) cn neutral eonitcring systee

                                                                        <92) TR)I neutrcn flux trip signal.

GXF 2-3 Syst Stress 18 Pres Relief SF SRVs open on high reactcr precare. CDEF 2-3 Syst Stress 18 Pres Relief '10 H)llNVNISbEO Provide path fcr nain stean systen I01) Ws stean bloudoun to suppressicn pool N). IXEF 2-3 Syst Stress 18 Pres Relief 64 PRl DNAIIEN Accept %Us'stean bloudoun. Ifrcn boiler vents aed drains systen, 10) to ~ressicn pool. 2-2 Fuel Failure 19 Core Cooling Provide stean fcr RCIC f71) txbine. 2-2 Fuel Failure 19 Ccrc Cooling SIN SKN IPCT initiation cn lm uater level IL2). CKF 2-2 Fuel Failure 19 Car OXF 2-2 Fuel Falhre 19 Care Provide ncrnally open uter supply fcr 1 GXF 2-2 Fuel Failure 19 Ccrc ing 3 RBNKR 'F systen I73) ini Provide loe reactcr uater level signal A2) to RN systen I7$ ) logic fcr RCIC systen I71) initlaticn., L2)

BUIIMN(ALYSIS SAFE APPBOD 12S Pr~g, 7 of 8 5 5 Mi'jt~~Pi'~l Inadvertent opening of E;C. Eckert e AII Bypass Valves NHjS63~ REACT (Rev MXEPTAELE I) IVA TARE 12S -I hackab SAFE H.S. Rubber s <<a~i~~lr CPER. REM.IS SAFEIY ACT18( SYS SYSIE)( RK STATE Code Title Code Title ML NtK mE (XEF 2-2 Fuel Failure 19 Ccrc Cooling 3 FEHNTER 'rovide path for RCIC (71) flow to the vessel through the feedwater spargers. (XEF 2-2 Fuel Failure 19 Core Cooling 3 FEHNIER SF Provide low reacta water level signal (L2) to SG systen 03). CKF 2-2 Fuel Failure 19 Ccrc Cooling 3 FENAIH( Provide path fcr HCI 03) flow to the vessel through the feedwater spargers fcr )Kl initiatim cn low water level (L2). (XEF 2-2 Fuel Failure 19 Core Cooling M PRI GRAINS(T Provide alternate sarce water supply for RCIC systen (71) through Core Spray systen (7$ piping free suppressim pool. Provide suppressim pml level (ndicatim. Accept RCIC turbine exhaust stean. 2-2 Fuel'Failure 19 Ccrc Cooling M PRI Cn(TAINB(T Provide alternate serce wats apply fcr )fCI systen (73) frm suppresslm pool fcr )PCI initiaticn m Iot water level (L2) signal. Accept IPCI tebine exhaust stean. (XEF 2-2 Fuel Failure 19 Core Cooling 71 RCIC SF Provide HXS RIll power to 8 systen ((6) Im water level (I2) instruaentatim f(r RCIC systen initiatim. CKF 2-2 Fuel Failure 19 Core Cooling 71 RCIC > iatim m low reactor wat I signaI frm RH( V4) CKF 2-2 Fuel Failure 19 Core ing 71 RCIC 1w t taticn fcr )PCI syst nl a'tlco (XEF 2-2 Fuel Failure 19 Core Cooli 7S IPCI SFS )PCI

                                                                                      'ysten water level (L2) signal.

(N) Iow reactm CKF 2-2 Fuel Failure 19 Core Cooling 74 R)f( SF Send feedwater systen (($ Iow reacts water level signal (L2) to initiate RCIC systen 01). REF 2-2 Fuel Failure 19 Core Cooling 75 C(RE SPRAY Provide flow path fa'lternate source of water to RCIC systen (71) fron suppressim pool N). 2-3 Syst Stress 20 RPV lsol 1 I(AIH SIEN( SF Provide low pressure signal (in nain stean line at turbine) to prinary cmtaiomt systen (6() failmfe logic. (R(N node)

  ~

SfE 9$ MN 4N.SIS Inadvertent Cpening of All 8ypass Valves lRBDII 128 Page 8of 8 F C. Ect

                                                                                                  'cF
                                                                                                                  ~~i
                                                                                                                  ~~a~l~r IF)H)S83~     (Rev  I)

'I TARE I28 -) 4'.: ~ IEACT QQCCPTARE TYA SFE QB. KILTS SAKIY ACTICN SS SSTBl F(K

        ,  STATE hde Title           Code     Title     ML       NE          IXK F     2-3 Syst Stress     20 iPY    Isol       I )NN STEAN        F    Qose ISIYs and cain stean drain lines on Im prusure signal (in nain stean line at teAine) frm prinary cmtainant systen Sl). QN node)

GK 2-3 Syst Stress 20 RPYlsoi I (a ~b>Mls NINSTEAN N Cl ose %IYs~ (ndicatiahs of Im turbine inlet pressure and/a escess bypass valve opening. F 2-3 Syst Stress 20 8PY lsol Q PAI (Xf(TAI)KNT SF Sek nain stean systen (01) Im presare signai (in nain st~ line at brMne) tn initiate lSIY (01) cloare. NN node) F 2-3 Syst Stress 20 SPY Isol 99 REACT% PHHECIN SF Provide Rj(!hde s(goal to systen Q fa Im stean line proser@ isolatim interlat. If 2-1 Nad lhlease e 2b Est Pri hnt 85 05 F Perf<rn isolatim actim(s) qua receiving sera signai fron the hactv Protectim Systen (%9. CKF M Cont Stress 30 Cmt Cooling 23 MSN & Support M systen (74) aypressim pool cooling nodeo QEF 2H hnt Stress 30 hnt Cooling M PM t9llAIQ86

                                                                                                  ~

Provide suppressim pool teeperature and level indicatim to pool cooling nodeq l5R system (7$ ) suppresaicn eisa i ~ ~l/~~ ss/s /gj (XEF 2H Cmt Stress 30 hnt holing 67 ES )SF Support'Ng( systen (7I) suppressim pool cooling

                                                                                 ~ odeum (XEF   M    hnt  Stress   30   hnt    Cooling 74  IN              ISF Provide suppressim pool cooling       functim.

P) 3elekQ 5 Rf

kppen8a Prepared: 13 Page pare: 1 of ~ ic'pcs

                                                                                ~/~(Jg Xaadvertent Opeaung o h Safety/Relief Valm f        Checked:

E. C. Eckert pat:a: ~~+~ BFNWSG3&4S (Rev 1) T. L. Garg aaaaaaaaaaasaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaasaaaaaaaaaaaaaaaaaasasaaa 1 pO DE SCRIPTION OP THE EVENT The safety/relief valves (opened by either an electrical pneumatic signal or steamline pressure) are designed to provide a means of limiting pressure excursions in the reactor or to depreseurize the primary system prior to the uee of low pressure pumps for makeup to the vessel. These valves are mounted on the main steam lines inside the drywell (primary containment), ard discharge to the suppression pool where the deposited heat energy is removed by other systems. The event of concern in this analysis, the inadvertent opening of a relief or safety/relief valve (IORV), can be initiated by false electrical "open" signals, improper opening pressure setpoints, vessel pressure transient s, mechanical failure, or aanual "open" signals fran the main control room. The appendix ie based on the previous SSA appendix issue (reference 6e) plus new inputs and comments subsequently received.'his single valve opening event assumes a stuck open valve (reference 6c, 14.5.4.2), and thereby results in a loss of coolant situation with steam flow rates of anv size up to the rated'apacity of the failed valve. This event is terminated by closure of the stuck open valve or by depressurization of the reactor vessel to atmospheric pressure. The inadvertent opening o ety/relief valve 's ficant in reactor operatin es C, D, E, or P opening of safety/ 'alve is also x. e p n he Browne Perry F AR (refere 2~0 EVENT CA

                                \

The Inadv btent Opening of a Saf ref Valve event is classified as an Abn rmal Operati ancient (reference 6d, G.5.3.5.2). 3.0 PLANT P TERS AND SAFETY CONCERNS 3.1 Pressure and Te erature The IORV event effect on pressure is similar for all initial pressurized conditions, however, the timing of the depressurization and the amount of suppression pool heating depends on the initial operating conditions, the flow rate through the stuck open valve, azd the length of time the valve is open. In all cases where the valve cannot be closed, complete depressurization of the reactor primary'ystem occurs over a period of time.

~dverteat OpenxLxxg of kppexedxx 13 Prepared: E. C. Eckert Page 2 Date: lof M 7 h Safety/Belief Valve Checked: Daae: WWa 7 BFNWSG3-048 (Rev 1) T. L. Garg aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa The BWR operates as a saturated system. Therefore,. the primary coolant, reactor internals, axxl the vessel will experience temperature decreases according to the physics of a depressurizing saturated system. The change involves thermal streses on the vessel axd appurtenances (Safety Concern 2-3). 3.1.1 IORV at Power When the reactor is at power with feedwater operating, an IORV event will not significantly affect pressure. This is because the pressure regulator system will sense the nuclear system pressure decrease caused by the open valve and will automatically close the main turbine control valves to compensate for the steam lost through the safety/relief valve. Eventual closure of the valve and/or unit shutdown is necessary. 3.1.2 IORV at Turbine Roll With the vessel at normal operating pressure and APRM power less than the relief. valve steam flow capability, the IORV event will result in depressurization 'of the reactor vessel (Safety Concern 2-3) . This may result in a vessel isolation vessel pressure drops below 825 psig (reference 6f) with the if mode switch in RUN. 3.1.3 IORV t Startu 1 An IO depre pter a

                          'djPghj u            1       t         )  i  1 also a                   s 1                              nder certain combi   apices-'f           ia condition,    a power   transient     IXl Neutr n Monitoring System         (NMS) scram ma     r 3.2   Reactor         e       vel The   effect of the      IORV   event on water level is dependent on the operating status of the plant           when the event takes place, and whether or not the Feedwater System is being used. In the simplest case, with the reactor at power ard feedwater operating>

level will only change imperceptively when the valve fails open (reference 6c, figure 14.5-9) ~ This is due to the large deliverv capacity of the Feedwater System and its capability to respond quickly. The worse case for level effect would include a core with a recent long, high power history (a large inventory of fission products), and the'vent occuring at low power without

I4 Z SAPK SHUTDNR hMLYSIS Appendix 13 Page 4 of inadvertent Opening of Prepared: ~/'~7MK~'ate: E. C. Eckert h Safety/Belief'alve Checked: Date: ~/i>g~g2-BFNWSG3-048 (Rev 1) T. L. Garg aaaassassassaaaaaaassaaaaassssass aaaaaaaaaaaaaaaaaaaaaaaaaaassaaassaaaaaaaaaaaaaaaa 4' EVENT MITIGATION 4.1 Safet Actions A licable 4.1.1 Avoid Excess RPV Stress (2<<3) Although vessel isolation (Safety Action 20).may occur as a result of steamline low pressure or vessel low level, the low pressure initiation of the isolation is not expected to occur immediately with an IORV event at power since pressure control will be maintained by the pressure control system. The pressure decrease isolation is a safety action expected for lower power IORV events. 4.1.2 Avoid Fuel Failure (2-2) Fuel protection will be provided by initiating reactor scram (Safety Action 17). The IORV event will probably result in a reactor low level scram if the valve opens when feedwater is not operating. If feedwater is operating, a manual scram may be required by technical specifications if the suppression pool temperature becomes too high. An Intermediate Range Monitor (IRM) scram may initiate the shutdown if the IORV auses a power transient. Core cool ng (Safety Aetio ) provide f e t R v I compl icat ' l. e ce. During a n e e tang and no isolati n), the main c ndenser would continue to act as a heat sink f r removal of steam produced in excess open valve can 'gmu~h ater Syste m'would provide coolant makeup. When the event occurs with the Feedwater System not operating, the task of coolant makeup would require operation of either the RCIC or HPCI System, depending on the flow rate caused by the valve's stuck open position. If level falls to the point where a low level (Ll) isolation (Safety Action 20) is triggered, the main steam isolation valves will close preventing loss of inventory in addition to the open valve. The water supply to the reactor in the high pressure isolated condition must be maintained by the RCIC or HPCI Systems. Vessel inventory makeup can also be supplied by the Core Spray System or the RHR System'(LOCI mode) from the torus when the vessel pressure has dropped low enough that RCIC/HPCI operation is no longer effective.

 ..                                                                                                  of A ~~

c SKULK SK1IBHRf hMLYSIS Inadvertent Opc.ning of kppendxx 13 Prepared: <c Eckert E. C. Page 5 naze: ~/u h Safety/Mief Valve Checked: nate: WW/ 7 BFNWSG3&48 (Rev 1) T. L. Garg SIR ZSDQRXSRSRSRCCRCRCRRCSZRCXCSWQCRQ!%SR'R CRCCZRRSSCRSRRCNC$ 1SCSSCCCWQRI5RRKE~CRCk%~ÃCSRR'0 ~%SCI2RZtZS%$ $ 4.1.3 Avoid Unacce table Containment 'Conditions (2W)

                                                                     >Mj+ a      eorpQc &gee  lso4 i   )

The main condenser will no longer be ~vai~Ae~as a heat sink. This is significant if the core has a large inventory of fission products since decay heat that would have gone to the main condenser before the isolation will have to pass to the suppression pool. Thus isolation can lead to higher loading on the suppression pool cooli=g systems. Heat transferred to the suppression pool by the ION'vent axd'y operation of HPCI (System 73) or RCIC (System 71) must be removed by the suppression pool cooling system mode of RHR in order to maintain acceptable pool conditions. This system must be manually lined up ard started from the main control room (Safety Action 30). 4.2 S stem Functions Re uired

                 'able        13-1 summarizes the required safety system functions. It was   collated from event~nique actions axd several applicable standard sequences from section 7.6 of the main body of the report (Tables 7. 64, -7, -15, 18, -20, -22, -23, -24, ard
                    -34).

4 '.1 Scram (17 Sensors i d e t seQ Yp w ive level and l. n iD (CRD) Hyd a 1 nt 5 to i sert cont 1 rods (ref rence 6b, 7.2).'The scram safety action may al be initiated as a becomes too high as indicated by the Primary Containment Isolation System (64), or as a Neutron Monitoring System (NMS, 92) scram transient. if the IOTA event induces a power 4.2.2 Containment Coolin (30) The RHR System (74) is manually initiated to cool the suppression pool. The operator initiates cooling given suppression pool temperature indication from the Primary Containment Isolation System and Level (64). Some of the , support systems for this mode are also shown in Table 13-1 ~ Although not shown in the table, it is possible that manual opening of additional SRVs may be performed according to pool temperature limited procedures.

SAFE SHUZN%% ILMIS kypeadix 13 Page 6 of

                                                                                       ~7t of'repared:

Date: Inadvertent Opening E. C. Eckert A Safety/Relief Valve BFN-OSG3W48 (Rev 1) " Checked: Date: f4' T. L. Garg aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa 4.2.3 Core Coolin (19) The IORV event occuring when feedwater is not operating mav required operation of HPCI (73) or RCIC (71) to supply vessel coolant -(reference 6a, 6.4 axd reference 6b, 7.4.3.2.2) ~ As vessel pressure drops below the effective range of RCIC and HPCI, they are tripped, axd the inventory supply is continued by use of the RHR (LPCI mode), System 74, azd/or Core Spray, System 75. The plant is capable'of transferring from high pressure to low pressure coolant supply near 105 psig (reference 6g') ~ If manual initiation is not performed, automatic initiation will occur when reactor level drops to level 1 (shown 'this way in Table 13-1 with other Ll core co'oling a:tions, but manual initiation should avoid initiating all the Ll actions>. Logic and permissives are provided by Systems 03, 68, aced 75. Some of the other support systems for these functions are also shown in Table 13-1 ~ Reference 6h discusses the core cooling aspects of this event coincident with failure of the high (FW, RCIC, azd HPCI). 30,mi 't@ate

                                                     'wn manual depressuri oblant supplies that the oper'ator has over ion, and the low essure coolant supplie                                   '.n,pd'equate core 1(           bi% 13-1 since thev oling.           s a                          I.                        l. dPion       hows the f              r       )i             a   ia   le
                                                                                 ~

using equipme'pt already icu t n ese shutdown analvses. 4.2.4 se 1 Isolat 1.0 Certain IORV sequences lead to NSIV closure (System 01) on comma% from System 64, after low steam line pressure is reached (s ensed by System 01) . Several isolation actions will take place associated with reactor level dropping to L3 (sensed by the Feedwater System (03) and initiated through System 64 to the rest of the systems given in Table 13-1) . If the event degrades to the point where water level drops to the level 1 setpoint, the Main Steam Isolation Valves will be closed (if<qt not already shut) ard the remainder of the isolation functions I+~ also initiated. Since it is expected that water level'ill be kept above Ll, these actions are not included in Table 13-1.

kppeodix 13 Prepared: CD. F~naee: ~I! Inadvertent Opening of E. C. Eckert h Safety/Belief Valve Checked: Date: ~WMit BFNWSG3-048 (Rev 1) T. L. Garg aaaaaaaaaaaaaaaaaaaaaaaa aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa 4.2 ~ 5 Control Bav Environmental Control (36) Steps toward Control Bay isolation and the supply of emergency, pressurized air are initiated from the low water level signal (L3, Feedwater system, 03) via the RPS (99) and system 64 to system 31, the control bay HVAC system. Table 13-1 showgthe funct1ons 17akag aH ~~&44 otic<<r 4+ac.~f ~~ <zc~]> ]~.f"

5. 0 LONG-TERN SHUTD(NN 5.1 Normal Shutdown The steps provided in Section 4.0 provide mitigation during the progress of the event and carry it to a low pressure poi.nt where the SRV can no longer remain open. The equipment norma>.lv expected to maintain ani/or complete the shutdown is given in Section 7.6 of the main body of the report (Table 7.6-1> ~

5.2 Safety Shutdown Normal, long term response, as given above, utilizes all systems normally available to the operator. If any (or all) normal control equipment is assumed to be unavailable, the reactor can still be shutdown to cold conditions by utilizing onl y equipment as given in the ence given in S ction 7.6, startin xs event from an essentially isolated condition (T ble 7.6-3).

6. 0 REFERENCE S

a. Brown's Ferry uclear Plant FSAR, Chapter 6; Core Standby Co ing Systems.

b. Brown's Ferry Chapter 7; Control azd Instrumentation.

Brown's Ferry Nuclear Plant FSAR, Chapter 14; System Safety Ana lys is.

d. Brown's Ferry Nuclear Plant FSAR, Appendix G; Safety Operational Analys is.

e. BFN Safe Shutdown Analysis, Appendix 13 (Revision 0), V. G.

Blanchette, Jr. ard T. L. Garg, BFNWSG3-048, June 28, lq86.

f. NEB-RAC-1186 (B45 860623 719), TVA letter'ated Janaury 26, 1977 from J. E. Gillelaxd to B. C. Rusche (NRC>, Change of Low Pzessuze Isolation Setpoint to 825 psig.

%0K SN71N5% hMLYSIS Appendix 13 Page 8 of ~4t'- Inadvertent Opening of Prepared: Date: Il7/~ h Safety/Belief Valve BFNWSG3&48 (Rev 1) Checked: T. L. Garg ssse: ~r" aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa aaaaaaaaaaaaaaaaaaaaaaaaaa

g. NEB-RAC-1334 azd -1347 (B45 860626 012 and 025), TVA letter dated January 22, 1975 from J. E. Gillelaxd to A. Giambusso (AEC),

Tech. Spec. Change of Threshold for Initiation of Low Pressure Coolant Supply Systems (122 changed to 105 psig).

h. NEB-RAC-1424 (B45 860626 629), TVA letter dated March 26, 1986 from R. Gridley to D. R. Muller (NRC), Discussion Showing adequacy of Current Design to Accomodate Stuck Open Relief Valve or Small Break Events Even with Failur of HPCI.

7.0 NON-APPLICABIZ INPUTS Sections 4 .2 aced 4 .3 of the main body of the report show results of an assessment of licensing references and FSAR questions. Some items have been identified as likely SSA (Rev 1) inputs for this event and are given in Tables 4.2-1 and 4.3-1 ~ These references were either used above in this appendix or they have been evaluated and not utilized as described below. 7.1 Reference GsEC-RRG-1033 (B45 860621 987) was not utilized since it provides caution and recommendatins (SIL-177) concerning

           'surveillance testing with equipment out of service. It is              aimed at avoiding depressurization and/or high suppression pool temperature problems before they could develop into this event.

SAFE 9UIEON ANALTSIS OR%II D Page 9of 16 '- prgp+g: .'c. oats ~l'<IP7 Inadvertent Opening of E.C. Eckert S/R Valve 8F)H)SQ~ IRev I) P7 T.L Sarg TNT D 1 REACT lMXEPTARE TVA OPER. RESILTS SAFETY ACTION SS STATE Code Title Code Title ML IF 2-2 Fuel Failure 17 Scran SF Provide low water level IL9 trip signal to reactor protectim systen I99) faille logic. If 2-2 Fuel Failure 17 Scraa Q PRI CINAINM Teeperature semrs in the suppressim pool sqqiy Infvnatim needed for the operatcr to nake a decisicn tn nanuaily scraa. If 2-2 Fuel Failure 17 Scraa 85 CRO SF SQan signaI ffon reactor protectim systen 199) will activate the ccntrol rod drive systen to insert rods. Scran functim mly. OF 2-2 Fuel Failure 17 Scran 92 IBAD NNITIR SF Provide IR)I neutrcn flux trip signal to reactor protecticn systen I99) I see sectim 7.6.2). OF 2-2 Fuel Failure 17 Scraa 99 REACTOR PROIECTN SF Provide scran signal to the cmtrol rod drive IDS) systse I85) m feedwater systen IM) Iow water level (L9 trip signal. OF 2-2 Fuel Failure 17 Scrae 99 REACIIR PROTECIN SF Provide scran signal to the control rod drive ICRO) systen 185) cn neutrcn xaitcring systen I92) IR)i neutrm flux trip signal. OF 2-2 Fuel Failure 17 Scras 99 REACTN PROTECIN )SF manual scran tper procedures) nay be required if suppressim pool teeperature beccnes too high. COEF 2-5 Syst Stress 18 Pres Relief 10 OOItER 0 ORN Provide path for wain stean systen 101) eau r stuck open SRV to suppressim pool COEF 2-5 Syst Stress 18 Pres Relief M PRI A ep I drains systea, 10) fna stxk open SRV to essim pool IM). IXEF 2-2 Fuel Failure 19 Core Cooling 1 NIN SIEA)I Provide stean fcr RCIC 171) turbine. COEF 2-2 Fuel Failure 19 Ccrc Cooling 1 NIN SIM Proride stean for )Kl IH) turbine in suppcrt of IKI initiatim cn low water level IL2). COEF 2-2 Fuel Failure 19 Ccrc Cooling 2 CMENSATE Provide ncrnally open water sopply for RCIC systen I71). (0

SE QlJIHN AMIS APPEM)II 13 Page 10 of 16 Prepared) Stn ~II P7 Inadvertent Opening of E.C. Eckert S/R Valve EF)H8~ (Rev I) Checked> T.L M< Sarg 0atea ~4;~7 TAELE 13 -1 REACT MXEPTAELE TVA (PER, RESETS SAFETT ACTIS( STS STATE Code Title Code Title ML COEF 2-2 Fuel Failure 19 Ccrc Cooling 2 CREME Provide ernal ly open water supply fcr (Kl systen (73) initiatim m low water Level (L2). (XEF 2-2 Fuel Failure Ig Core Cooling 3 FEBNIER Provide lm reactcr water level signal (L2) to, M systea (74) logic for RCIC systen (71) initiatim. COEF 2-2 Fuel Failure 19 Core Cooling 3 FEBNTER P>>side path for RCIC (71) flow to the vessel through the feedwater spar gers. GE 2-2 Fuel Failure 19 Ccrc Cooling 3 FEBNlER SF Provide lmreactcr wats level signal (L2) to IfC1 systen (73). (XEF 2-2 Fuel Failure 19 Ccrc Cooling 3 FEBNTER Provide path fcr IPCl (73) flow to the vessel through the feedwater spargers fcr HCI initiation cn low water level (L2). SF Provide lm reactcrwater level signals (Ll). Provide low reactor pressure signals to ccrc spray systen (75) and RN systen (7l). Provide low reactcr water level (Ll I L3) pernissive signals to nain stean systen (01) fcr A0S. CKF 2-2 Fuel Failure 19 Ccrc Cooling 18 FlKL OIL. SF Provide diesel fuel oil to diesel generator systen (82) starting m Im reactcr water level (L1) signals. This safety actim nay occur, but is not required by this event since nor (KEF 2-2 Fuel Failure 19 Core Cooling 64 e ly Q(l ( Pop eve indica ion. Accept RCIC turbine exhaust stean. CCEF 2-2 Fuel Failure 19 Core Cooling 6( GNAIHSlT Provide alternate source water supply fcr SKI systen (73) fron suppressim pool fcr IRI initiatim m low water level (L2) signal. IXHTAIQE6

                                                             'W Accept IPCI turbine eshaust
                                                                 ~ 3jpcrt~nre 5e8+)~d'RI stean.

spray systea (75) by providing roon cmling with service water frm EED( (67) in support of HXS actims at low reactcr water level (Ll).

SAFE SINTMN N(ALYSIS APPEHDIZ IS Page 11 of 16: sqirei C E. htn ~+>0( Inadvertent 0pening of S/R Valve ERHKHm (Rev I) Checked: E.C. Eckert

                                                                                          ~~

T.L Sarg u f Oatz ~grani TAKE D -I

\ ~ .

REACT (N(XEPTAKE TVA 5ER. RES(lTS SAFEIY ACTIN SYS SYSTE)1

       'TATE    Code   Title       Code    Title     ML           NE CKF 2-2 Fuel Failure 19        Ccrc Cooling  M    PRI   GNAINSIT        Provide     tms     uater supply for  M     (74) and Core Spray (7$ systens        in support of HXS acticns at lw reactor uater level (Ll).

COEF 2-2 Fuel Failure 19 Core Cooling 67 EEC)I SF Provide cooling water to diesel generatm systen (((2) starting cn lw reactm uater level (L1) signals. This safety actim nay occur, but is s~ required by this event since ncraai ac is available. I

                                                                                                                        <~~K'ARa            c(

CKF 2-2 Fuel Failure 19 Ccrc Cooling 67 EEQ( SF Provide cooling uater to M~1 (7I)>and@ra spray roon coolers (6() in support of HXS actims at lw reactcr uater level (LI). COG: 2-2 Fuel Failure 19 Care Cooling N RECIRQLATIO( SFS Provide lw reactcr presslre pernissive signals to ccrc,spray systen (7S) and M systee (7$ ) fm HXS actions at lw reactcr uater level (L1). CKF 2-2 Fuel Failure 19 Core Cooling N RECIRQUTI(jq SF Close recirculatim discharge valves fcr IJCI injectim m signals froa M systen (7i) indicating IPCI initiatim signal d to lw uater level (LI), w reactcr g A pressure pernissive signal+ f~ lqge C0EF 2-2 Fuel Failure 19 Ccrc Cooling 71 RCIC SF Provide HXS AS pwer to R( systen (Q) lw uater level (12) instrueentatim fcr RCIC systen initiaticn. COEF 2-2 Fuel Failure 19 Ccrc.Cooling 71 RCIC er eve L2) signal froa M systen (7i). I I CKF 2-2 Fuel Failure 19 Ccrc Cooling 71 ta.: cr mystes nti ( I COEF 2-2 Fuel Failure 19 Core Cooling 71 IC SF Provide HXS ATU (6)-"' uter level (LI) and R( systen and recirculatim systes (N) lw reactor pressN'e instr uaentatim. CKF 2-2 Fuel Failure 19 Core Cooling H IPCI SFS )PCI initiation cn Rf systee (N) Iw reactcr .4 uater level (L2) signal.

SAFE QITMN NQLYSIS APP8$ 1'l D ' Page 12 of )6 Prepared> 5C E Date: ~ 'l Inadvertent S/R Valve EF)H)SQ~ Opening (Rev I) of E.C. Ectert T.L 6arg D te: ~~P TADLE D -1 REACT MXPTA8(E TVA SAFE (FB. RES(LTS SAFHY ACTI(I SYS SYSTB( FIR STATE Code Title Code Title ML, Nf GIE CKF 2-2 Fuel Failure 19 Core Cooling 74 RN SF Send feedwater systen (N) low reactor water level signal (L2) to initiate RCIC systen (71). COEF 2-2 Fuel Failure 19 Core Cooling 74 RN SF Start pueps m low reactor water level (Ll) signals froa core spray systen (75). Provide signals to close recirculaticn valves to recirculatim systea (N). CKF 2-2 Fuel'Failure 19 Core Cooling 74 M SF Open injection valves m low reactcr presume signals frm ccrc spray systen (75) and M systen (74),LPCI initiatim signals due to low water level (Ll). COEF 2-2 Fuel Failure 19 Ccrc Cooling 75 GIE SPRAY Provide flow path fcr alternate scarce of water to RCIC systen (71) frm suppression pool (64). 2-2 Fuel Failure 19 Core Cooling 75 GIE SPRAY SF Send feedwater systea (N) low reactcr water k level signals (LI) and low reactcr pressure signals (froa feedwater systen (N) and recirculaticn systen (N)) to R)a systen (74). CDEF 2-2 Fuel Failure 19 Ccrc Cooling 75 GIE SPRAY SF Start puaps on feedwater systea (N) low reactor water level (LI) signals. CKF 2-2 Fuel Failure 19 Core Cooling 75 GIE SPRAY SF Open injectim valves m feedwater systen (N) cr recirculaticn systen (N) low reactcr presske e signals in cmjunctim with low water level (LI) signal frow feedwater systen (N). CKF 2-2 Fuel Failure 19 Core Cooling 75 ORE SPRAY ovide standby diesel generatce (82) start signals on f n s 1 I n r k 1 e~ CKF 2-2 Fuel Failure 19 Ccrc Cooling 82 DIESEL SEM7(A F Diesel generatas st s ) frm ccrc spray systen (75). This safety action nay occur, but is not required by this event since ncrnal ac is available, CDEF 2-2 Fuel Failure 19 Ccrc Cooling 86 DSL K)( START AIR SF Provide diesel starting air to diesel generabr (-0 systen starting cn low reactcr water level (Ll) signals. This safety acticn nay occur, but is

                                                                               ~ ot required by this event since anal ac is available.

QUINN ANLTSIS )3 of )6 SVE APPENII 13

                                                                                   <~~'      '~~:

hts ~t/ Inadvertent Opening of S/R Valve RHSN-0(8 Nev I) Checkedl T.L C. Eckert Sarg httt ~f TAOLK 13 -1 RFXT lNCCEPIARE IVA (PER. RES(LTS SAFETY ACTIOH STS STSIH( STATE Code Title Code Title M). NNK C(REF 2-1 Rad Release 20 RPV lsol 1 NIHSTEA(( SF Provide Ion pressure signal (in nain stean line at turbine) to prinary cmtainnent systen N) fail-safe logic. SlN node) C(6 2-1 Rad Release 20 RPV Isol I NIHSTEA(( Close )SIVs and nain stean drain lines m Iw pressure signal (in nain stean line at turbine) fron prinary cmtainnent systen N). (RN node) GEF 2-1 Rad Release 20 RPV lsol 6I PRI GNIAINENT SF Send nain stean systen (01) Im presa'ignal (in nain sbnn Line at turbine) to initiate lSIV (01) closure. Wl node) CKF 2-1 Rad Release 20 RPV iso( 6g SKU SF Close R)C) isolaticn valves m Im water level (L3) signal fron priaary cmtainmnt systen ((4). This safety actim is expected to occur, but is not a requirenent fcr this event since there is no fuel failure. GEF 2-1 Rad Release 20.RPV,lsol'l RCIC lou stean supply pressure signai closes stean supply line isolation valves. GEF 2-1 Rad Release 20:RPV Isol,=: 73')FCI sure sl al closes stean supply line isolatim valves. COEF 2-1 Rad Rel 20 signai tripped cn level: PJ }~i prl nary c al t s))sten Bee Thl safety acticn is expected to occur) but is not requirenent fcr this event sin I failure.

                                                                       ~ ~

Rad Rel 20 Isol 99 REACT% PWKIH SF Provide R(N Hode signal to systen 64 hr Iou stean line pressure isolatim interlock. CKF 2-3 Syst Stress 22 Rest )kraal I NIH SIEAH H Atteept to reclose stuck open SRV. 06 2-1 Radhlease 26 Est Pri Cmt 3 FEEEHATER SF Provide Ion uater level (L3) signaI to reactlr protection systen (gg) fm initiatim of L3 isolaticns. This safety actim is expected to occur, but is not a requirenent fir this event since there is no fuel faillre. GG'-1 Rad Release 26 Est Pri Cont 32 CNIRB. AIR SF Perfcrn isolatim actim(s) upm receiving Ion uater level (L3) isolatim signal fron the >r Prinary Cmtainaent systen (M). This safety actim is expected to occur, but is not a requirmot fm this event since there is no fuel failure.

BUIIXN AN.YSIS 9 ix / SAFE Inadvertent Opening of APPEHDI'l Page (rasa EC 14of16 Fc~ g(p ~t't/W S/R Valve SF)H)SM~ (Rev 1) Checked! E.C. Eckert T.L eC Sarg I)ate: ~~~ P r TARE D -I REACT ~lKE TVA . QfE CPER. RES(1TS SAFETY ACTIN SYS SYSTE)( F(K STATE Code Title Me Title S. NK GXK COEF 2-1 RadRelease 26 Est Pri Cont 64 PRI GRAIN%MT SF ~ LM. (Q) signal fron RPS (99), isolatim actims h send primy/secmdary initiate Q cmtainant isolatim signals to systens 32f 65f 69x 7l) 75, 76f 77$ N) 9() f and 91. This action ls expected to occltybut is not a requlreeent for this event. (XEF 2-1 RadRelease 26 Est Pri Cont 75 (XNE SPRAY SF Parfvn isolation actim(s) upon receiving Iow water level (Q) isolatim signa) frm the Priaary Cmta(neent systen (6f). This safety action is expected to occur, but is not a requireeent fcr this event since there is no fuel failure. I GXF 2-1 Rad Release 26 Est Pri Cont 76 GKAINBl7 MRT SF Perfcrn isolatim actim(s) upon receiving low water level (Q) isolatim signaI fron the Priaary Cmtaimnt systee (bl). This safety actim 1$ expected to occur y but is not a requireeent fu'his event since there is no fuel COEF 2-1 Rad Release Est Pri Cmt 77 RA(NSTE a i I tion 'Q~) q eiving low (ig si ra the is safety t t ocul, i not a requirenent for this event since e is no fuel failure. CKF 2-1 Rad Release 26 Est Pri Cant 8f M SF Perfaa isolation actim(s) upm receiving low water level (Q) isolaticn signal froa the Pr(nary Con'tainaent systea (6I). This safety action is expected to occur, but is not a requireeent fnr this even't since there is no fuel failure, (F 2-1 Rad Release 26 'Est Pri Cmt 85 CR() SF Perfcra isolatim actim(s) upon receiving scran signal fron the Reactcr Protectim Systen (99). i'-( Rad Release 26 Est Pri Cont 90 RNIATIQ()MTN SF Perfcrn isolation acticn(s) upon receiving low water level (Q) isolaticn signal fron the Prinary Cmtainaent systen (6l). 'This safety actim is expected to ocor, but is not for this event since there is no fuel a'equireeent failure.

Sensa aW.YSIS APPE)mu O of 16 '- SAFE Page Prepared: 15 Satei ~If Inadvertent Opening of S/R Valve 8F)HM~ (Rev I) (heckedx WA T.L W~ F C. Eckert Sarg

                                                                                                                        ~P~~P p 8'te:

TARE D I REACT INKEPTAKE TVA SAFE (PER. STATE Code RESILTS Title SAFElY ACT19( Code Title SYS

                                               )L SYSTE)(

lN% 'E FIK

                                                                                                    /, d      s-j.h-') u<-h4 COEF, 2-1   Rad Release    26   Est Pri Cant  91 TlP                SF    Initiate      T1P  withdrawal~m low water lwel (Q) signal froa prinary cmtaimnt systen N).

This safety actim is expected to occur, but is not a requireeent far this went since there is no fuel failure. CKF 2-1 Rad Release 26 Est Pri Cant 99 REACTI PROTECTS SF Provide low water lwel (Q) signal fran feedwater systen ((m to prieary cantainaent systen N) far initiatim of Q isolaticns. This safety actim is expected to.accur, but is not a requirenent fm this event since there is no fuel failure. CKF 2-1 Rad Rel'ease 27 Est Sec Cont 61 PRI C(ITAI8%N SF Perforn isolatim acticn(s) upm receiving I~ water level (Q) isolatim signal fran the Reactor Protectim Systen (99). This safety action is expected to occur, but is not a requireeent fm this event since there. is no fuel failure. CIA 2-1 Rad Release 27 Est Sec Cont 65 SST SF initiate SST plant start m low water lwel (Q) signal fry prinary cmtaimnt systen N). This safety actim is expected to occur, but is seven sine there is no fuel failure.

                                                                                                                   ~

1 CKF )-1 Cmt Stress AA p t I 1) tssim p I cooling

                                                                                                'l CKF 2-1                                            I GNAMNT              Provide suppressim pool teeperature                 level
                                                                                                               ~

Cmt Stress 30 indi catim to resslcn poo cooling node> d(ISei ceoling. Kt-'<

                                                                                                                                           /~

CKF 2-1 Cmt Stress 50 Cmt Cooling 67 EEDI I(SF Suppart RII systsn (71) suppressim pool cooling

                                                                             ~ adee CKF= 2-1   Cant Stress     AA  Cmt Cooling 71 RIf(                 ISF Provide suppressicn pool cooling             furictim.

CKF >5 Pers (herexp 2 Cont Pay Env 5 FEBNTER SF Provide low water level (Q) signal to RPS systen (99) fcr initiation of control bay isolation. This safety artim is expected to occN', but is not a requireeent fm this event since there is no fuel failure.

SAFE SIUIHN NNLTSIS APPEail D Page

                                                                   ><(((((

16 of lk r-c

                                                                                          ~

(

                                                                                             /
                                                                                            ~~    L oat(: ~z/J(   /I'g inadvertent  Opening of                                                    E.C. Ectert S/R Valve                                                      Chectedt                           Oatex  ~+x~X'P' NHE63~        IRev I)                                                      T.L   Sarg TARE 13  -1 REACT   MXEHARE                           IVA                   SAFE KB.      RERLTS          SAFETY ACHII     STS     STSIE)1       RK STATE Code    Title      Code   Title     M        Ntf          mE COEF  3-5 Pers Overexp 36 Cont   hy   Env  31 AIR C95ITIM%     SF     Air cond. IAC) supply ducts isolate 4 Eeerg.

Pres. Systen Kcntrol Roca berg. Vent. Systee) supplies pres. filtered air to IKR cn tM. IQ) signal fron systee Q. Ibis safety actim is expected to occur, but is not a reguireNNt fcr this event. GfF 3-5 Pers Overexp 36 Cmt Say Env M PRl GNAIQBIT SF Upa low water level IQ) signal frcn RPSN), send isolatim signal to air cmditicning systee 131). Ibis safety actim is expected to occur, but is not a requireaent fcr this event since there is no fuel failure. LP 3-5 Pers Overexp 36 Cant Say Env 99 REACTORPROTECI)i SF Provide low water level IQ) signal frm feedwater systee IN) to priaw'y cent, systen IM) fcr initiatim of cmtrol bay isolatim. Ibis safety acticn is expected to cccur, but is not a requireeent fcr this event since there is no fuel failure.

                                                                                                                         ~ t
                                                                                                                      ~  ~

Li

I/ kpyemlix 14 Page 1 of. H6~ Prepared: Date. ~%~//g e Lose Flow ot hll Peedeater

 -3FNWSG3&48 (Rev 1)

Checked: C. Eckert T. L. Garg

                                     ~~~g~~gg~~gggg~~+g>gggtQSjZBCXCXZRCt>>RCQQC 8/a/g p ROD'OOCQCRR'OSCXSCCO 1+0  DESCRIPTION OF THE EVENT The Feedwater System is a part of the feedwater/condensate                        network which returns primary coolant to the reactor vessel to replace coolant removed by steam flow. This feedwater supplied makeup f low is sub" cooled, which allows the operation of the reactor recirculation pumps in a non-cavitating region of their operating envelope.

During operating states C, D, E, axd F, feedwater control system failures or feedwater pump trips, causing partial or complete loss of feedwater flow, can lead to an unplanned core coolant decrease. Until corrective actions can be impleme..ed,, loss of feedwater flow causes the mass of steam leaving the reactor vessel to exceed the mass of, water entering the vessel, resulting in a net decrease in the coolant inventory and consequential drop in reactor water level (Reference 6b, figure 14.5-10) ~ This appendix is based on the previous issue (reference 6i) plus new inputs and comments subsequently received. The FSAR, Appendix G (reference 6c) also addresses the loss of all feedwater flow. 0 The reduction in subcooling corresponding to the reduction in feed-water flow causes a rise in the average temperature of the coolant entering the core, so that neutron flux decreases due to increased void fraction.. Level. will fall at a rate dependent on the initial operating condition of the reactor. High power operation ('state F) ard long core exposure history initial conditions tend to decrease level at the fastest rate due to the higher heat energy produced which mus't be carried away by steam.' loss of feedwater event, from low power operat tate D will be a much milder event. f Trip settings at. r evel(s) initiate co te 8asures nclu ding scram (states azd F), and RCIC- vent co inued loss of coolan (a re d d tb fuel fa lure. Initiation of a fo ccurs after approximately s e s s r IC, HPCI, and recircul tion pump trip occur a longer time (FSAR 14.5. rence 6b). Pressure contro is provided throu h n turbine control a+i/or bypass valves a steam to the main condenser. The current P lant has change e low water level setpoint of the MSIV closure from level 2 to level 1'(reference 6k), so that no isolation is expected now even though the original FSAR analysis included that action and the need for subsequent isolated shutdown (reference 6f).

tl of kypeadia 14 Prepared: 6 Page 2 Date. / Tr'c kO gA g~ Lose of k11 Fee@water E. C. Eckert Flow BFN-OSG3&48 (Rev 1) Checked:

                                                                      ~  Garg Date:  ~~~

aa~aaaa aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa Either RCIC or HPCI is capable of avoiding the functions which are initiated at low level Ll (reference 61). cannot be restored, the reactor should be depreesurized using If normal feedwater appropriate procedures (e.g. reference 6j).

  ~
    . It may be necessary to depressurize the vessel under certain possib1e conditions. 1) Loss of feedwater occurs axd neither HPCI or RCIC are available for vessel makeup. 2) Pressure control using the relief system, or operation of HPCI or RCIC occurs resulting in suppression poo1 heating such that continued pressure or level control will raise suppression pool temperature above the maximum limit for its capability to safely absorb the mergy in the vessel coolant. Since vessel isolation is moved to Level l ani should not occur, the suppression gcc                pool heating should not be a significant problem. PCS                           <<oi~'~g ~

the isolation (atd subsequent SRV opening) also

                              ~                                                 ~

reduced+'he likelihood of the loss of feedwater plus stuck open SRV case discussed in reference 61. In'ither case, depreesurizing the vessel will permit the low pressure coolant injection systems to be used to maintain level (reference 6a, 6.5.2) ~ -0 2.0 EVENT CATEGORY The BFNP FSAR addresses Transient. this event as an Abnormal Operational anal 3.0 PLANT PARAMETERS AND SAFETY CONCERNS

3. 1 R tor Water'.,".Level The Los of Feedwater Event threa cladding fission t

pr oduc b b ing the primary high pressure essel invento r prxate safe actions, the rea s e e i )f for fission leadin to possxb e sdFo produc s into the primary coolant sys ea $ e ia 2-1 ani 2-2).

                                  , ~ >'I ','" "

This event also challenges primary system s cceptable rp 1 2-3) during

                       ~handle this concern.

system cooldown, but normal procedures should

Page 3 of k6P cC Date: ~/JSI'~ Loss oE kll Feedeater E. C. Eckert 7 Flem Checked: c74 9 Date: BPNWSG3W48 <Rev 1) T. L. Garg ~WMP'. ammu'maaaamrcaaamaaama=ma~~=~-~-~~~ 0 EVENT MITIGATION 4.1 Safet Actions A licable 4.1.1 Avoid Release of Radioactive Material (2-1) In this event with dropping water level, early, anticipatory isolation steps (safety actions 20, 26, 27, 36) are 4-4(Shaken to provide assurance against any radiological release to the outside or to the plant operators &of'eP> "<<~g~q/f~ 4.1.2 Avoid Puel Pailure (2-2) Low level scram (Safety Action 17) is needeR to rapidly reduce core power to decrease the steam flow/feed flow mismatch caused by loss of the Feedwater System (reference 6c, G.S.3.5.1, event 27) axd thereby moderate the water level reduction which threatens to uncover the fuel. Action to reduce power (safety action 24) through core flow reduction are also involved in this event, although scram actually occurs ahead of the low level (L2> functions.. .0 Hi h pressure water supply systems must be provid pr mar s st e ~to to Core make up Co fe t E s ceded if ca e the reactor be P bf igh pressure t rhine-dr iven) sy tems. Bringing the system(s) on line to repla e the lost fe water assures that the d and the fu equate y cooled. I Loss of feedwater from power and the resulting low level scram is shown in the PSAR to result in vessel isolation on low level (although current setpoints should avoid this consequence). This safety action would terminate steam flow to the main condenser (reference 6c, G.5.3.5.1, event 27), but it is not included here because of the new setpoints. 4.2 S stems Re uired Table 14-1 summarizes the system functions required to achieve the safety actions identified in section 5.1 ~ It was collated from event~nique actions axd applicable standard sequences fran section 7.6 of the main body of the report (Tables 7.6W, -17~ <8>

                -18,    and   -20).
                                                                                                ~,f)p Rl
 ~ '
                                                  ~
                                                    ~

V

Page 4 af Aft)CC Date: Loss of'11 Fecdeater E. C. Eckert Floe BFNWSG3&48 (Rev 1) Checked: arg Dat e. ~~~~

  ~aaaoaaeassaaaacaaaaaaaoaaaaaaaaaaaaacsaaa>a>+~++~+

4.2.1 Scram (Safety Action 17) When reactor water level reachee level 3, the Reactor Protection System (RPS, 99) is required to be activated ard to send a scram signal to the Control Rod Drive System. This

• signal originates in the Feedwater System (03) (reference 6d).

The Control Rod Drive System (CRD, 85) in turn receives a scram signal from RPS to cause all control rods to be rapidly inserted into the reactor for prompt shutdown- Only the CRD scram function .is required for mitigation of this transient. The above discussion appli s to operating states D and F onlv', states C axd E do not require scram (reference 6c, figures 35 a and b). 4.2.2 Core Coolin (19) The low vessel water level detection circuitry (part of the Feedwater System 03) is required to initiate a signal for operation of RCIC axd HPCI (L2) (reference 6e, note 1). The RCIC System (71) azd HPCI System (73), are activated by the low level.(L2) signal ard must function to maintain sufficient water in the RPV to cool. the core and,maint ep boiler in standby condit ' water can be restored. They act o sat'is fy the single failu ~iterion. Some of,. e systems which eu Ioperat ion are" also . ho 4.2.3 Initial 'S Secondary' e Protectio'n (I6) f 'n 1 (27) and

                                                                      ) Rima     Containment (26)

Control Ba Enviro When react water lev'el drops to Level 3, (sensed by System 03, Feedwater), several anticipatory steps are taken toward RPV isolation, axd establishment, of piimary and secondary

                                       .2.4 containment. The primary facets are included in Table 14-1 for systems 03, 31, 64, 65, 69, 74, 94, and 99.(Wry~~ b>>T~<~~M).

Cc< V~/h Power (Safet Action 24) 'educe When reactor water level drops to Level 2, (sensed by System 03, F e water)<tIxp recirculation pumps are tripped. This backs up the< ptredirculation flow control runback features that function with the water level controls to avoid low level (L3) if the event was only a partial loss of feedwater. -0 scram

                                                                                              /I ShPK SHDXDt%%              hMLMXS                                                             Bf~g Loca oE h11 Fcedeater
                                              ~pemdxx 14 Prepared:       gCF  C. Eckert Page Date:

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                                                                                             //g E.

Plod BFNWSG3&48 (Rev 1) Checked: T. L. Garg Date: Q~P~ gggggzgggg gggggg~~gg ~~~ggssssags~gsaaaaaaemsasaa33%%>>+ 4~2~5 Control Ba Environmental Control (36) Steps toward Control Bay isolation and the supply of emergency, pressurized air are initiated from the low water level signal (L3, Feedwater system, 03) via the RPS (99) atd System 64 to system 31 the control bay HVAC system. Table 13-1 shows the functions.

5. 0 LONG-TERM SHUTDOWN 5.1 Normal'solated Shutdown The steps provided above achieve stable, hot shutdown conditions. Should it be impossible to restore normal feedwater (or lower pressure level control using the condensate system),

manual depressurization and cooldown may be initiated. The sequence of equipment required is given in section 7.6 of the main body of the report for this standard path if needed (Table 7.6-2). 5.2 own ilizes all systems 1 y all) normal t 4 he unavailable the reactor can t e s tdown by utilizing only safety equ pment as given in the Safety Shutdown sequen won ~ 6 of the main report x.s event from an already isolated o ation (Table 7.6-3) ~ 6~0 REFERENCE S

a. Brown's Ferry Nuclear Plant FSAR, Chapter 6; Core Standby Cooling Systems.

b. Brown's Ferry Nuclear Plant FSAR, Chapter 14; System Safety Ana lys is.

c. Brown's Ferry Nuclear Plant FSAR, Appendix G; Safety Operations Ana lys is.

d. 730E915, Revision 13, Sheet 9, RPS Elementary Diagram.

e. 47W610-73-1, Revision 28, Note 1, HPCI Mechanical Control Diagram.

f. BFNP Technical Speci.fication, Table 3.2.a, Revision 9"19-84.

g. BFNWSG3-037, MCEL Design Basis.

h. Browns Ferry Nuclear Plant FSAR, Chapter 5; Containment.

i. BFN Safe Shutdown Analysis, Appendix 14, (Revision 0), V. G.

Blanchette, Jr. atd L. E. Pohl, BFNWSG3-048, June 28, 1986.

ad SHVmma aaumIS hyped~ prepared: 14 6+ Q& Page 6 Date: of ktf ~

                                                                                 /4/ /'7 e Loss Flo>>

of 411 Feehrater BFN-OSG3-048 (Rev 1) Checked: csaasscraaaaaaaassaassaeaaaasaaaaasaassssameeaa+++I E. C. Eckert arg

                                                           ~~~~~~~~~~~~~~

Date.'M>~F PSAR Volume 7, Question $ 13.11, List of BFNP Emergency Operating Instructions (including feedwater/cond'ensate system failure) ~

k. NEB RAC 1306 (B45 860624 068), TVA letter dated July 9~ 1984, fran L. M. Mills to H. R. Denton (NRC), Change MSIV Closure Low Level Setpoint from L2 to Ll (470 to 378 inches).

l. NEB RAC 1725 (B45 860701 037), TVA letter dated April 24, 1979, from J. E. Gilleland to J. P. O'Reilly (NRC), Enclosure pages 3&,

Description of loss of feedwater event and core cooling system performance. 7.0 NON-APPLICABLE INPUTS Sections 4.2 axd 4.3 of the main report show results of an assessment of licensing references and PSAR questions. The items which have been identified as likely SSA (Rev. 1) inputs for this event are given in Tables 4.2-1 and 4 .3-1). These references were either used above in this appendix or they have been evaluated and not utilized as described below. 7.1 MEB WWA 1125 (B44 860731 024). TVA letter dated December 3, 1980 from L. M. Mills to J. P. O'Reilly (NRC). This reference discusses details of diesel generator, load shedding, axd RHR Service Water actions wh'ich do not apply to this event (see Appendix 15A, 25A). 7.2 NEB RAC 1080 (B45 860619 040) ard NEB RAC 1083 (B45 860619 043), both referring to a TVA letter dated May 23, 1978 fran J. E. Gillelard to G. Lear '(NRC). They concern plant performance for a 20-inch change in the Level 2 setpoint before the shift of isolation to Level 1. No signficant impact on performance is reported, and there is no impact on the SSA.

SAFE QUINN AN.ySIS APPE)OII N Page 7 of 11 Loss of Feedwater Rm Sate. ~H/

                                                                                   .E.C. Eckert ONcked)                   4   M         0ate:  < +<~+ 7 IRH8~8       Nev 1)                                                         T.L  Sarg .

TARE 11 -1 REACT INACCEPTARE IVA IRR. 'E91IS SAFETT ACTISI STS STStE)I STATE Code Title Code Title NK 0F 2-2 Fuel Failure 17 Scran SF Provide low water level IL3) trip signal to reactcr protectim systen I99) failmfe logic. 0F 2-2 Fuel Failure 17 Scrae Scrae signal frm reactor prutectim systee I99) will activate the cmtrol rod drive systen to insert rods. Scran functim mly. 0F 2-2 Fuel Failure 17 Scran 99 REACTIR PR0IEIN SF Provide scran signal to the cmtrol rod drive IDS) systea I85) m feedwater systen I03) low water level IL3) trip signal. C0EF 2-2 Fuel Failure 19 Ccrc Cooling I IIAIN STENl Provide stean fcr RCIC 171) turbine. C0EF 2-2 Fuel Failure 19 hre Cooling I IIAIN SEN Provide stean fa )PCI I73) turbine in appcrt of IfCI initiatim m low water level IL2). IXEF 2-2 Fuel Failure 19 Care Cooling 2 NiENSAIE Provide ncrnally open water supply fcr RCIC systen 01). (

\

G6'-2 Fuel Failure19 Core Cooling 2 CMENSATE Provide ncraally <yen water supply fm .fFCI systen I73) initiaticn m low water evel IL2). C0EF 2-2 Fuel Failure 19 e 7ing 3 FEE0NTER SF Provide low reactor water level si I IL2) to tta I/a t t' I71) ni i IXEF 2-2 Fuel Failure 19 'd p fl'CIC 171) flow to through the feedwater spar gers. C0EF 2-2 Fuel Failure 19' Provide low reactcr water level signal IL2) to IPCI systen I73). CtEF 2-2 Fuel Failure 19 Ccrc'Cooling 3 ffEENTER Provide path for IPCI I73) flow to tbe vessel through the feedsater spargers for fPCI initiaticn m low water level (L2). C0EF 2-2 Fuel Failure 19 Core Cooling 6I PRI CCNTAIMSIT Provide alternate sarce water supply fl RCIC systen t71) through Core Spray systen f75) piping frow sqspressim pool. Provide suppressim pool level indicaticn. Accept RCIC turbine eshaust Steaae

SAFE S)llXQN AN.YSIS APPERIZ 14 Page 8 of 11 Loss of Feedwater Flow F C. Ectert Q>>chdi NHKQ~ (Rev I) T.L Sarg I TAHE 14 -I KD MXPMLE SA KR. RERLTS SAFETY ACTIOI SYS

  ~

STATE Code Title Code Title -

                                                  )L GP      2-2 Fuel Failure 19 Core Cooling      Q   PRI CITAINBIT        Provide alternate sarce waar supply fcr )PCI systen 03) free srppressim pool fcr )fCI initiaticn m low water level IL2) signal.

Accept SCI turbine exhaust stean. GIEF 2 Fuel Faille 19 Ccrc Cooling 71 RCIC SF Provide HXS ATU power to Rf systen (03) low water level Lm instrueentaticn.fN'CIC systen iritiabcn. GIEF 2-2 Fuel Failure 19 Ccrc holing 71 RCIC SFS RCIC initiatim cn lor reactcr water level (L2) signaI frow M systen (74). GXF 2-2 Fuel Failure 19 Ccrc Cooling 71 RCIC SF Provide HXS ATI) power to Rf systee lm) low water level IL2) instrueentatim fa IPCI systen

                                                                          ~

03) initiatim, G IEF 2-2 Fuel Failure 19 Core Cooling 73 )fCI SFS )fCI initiatim cn Rf systen 103) Im reactcr water level ILD signal.

CCEF 2-2 Fuel Failxe 19 Ccrc Cooling 74 IR SF Send feedwater systen NQ Ia reactcr water level signal IQ) ). IXEF 2-2 Fuel Failure 19 Ccrc ling 75 ORE SPRAY Provide flow path fcr alt te e NEF 2-1 RadRelease 20 RPUIsol 9 signal frow prieary cmtainant systen ). This safety actim is epected to ocnr is even since there is no fuel failure. GEF 2-1 Rad Release 20 RP/Isol 74 M SF M Isolatim signaI tripped m Im water level IL3) signai fron prinary cmtainant systen f64). This safety actim is Npected to occN'g hut is not a requlfMRt event since there is oo fuel faille. fl'his lF 2-2 Fuel Failure24 PowReduce 3 FEHIATER Provide Iow water level (L2) signal to open feclrculatim Il/8 drive enter tjreaters 1200) electrical systen) fl trip of recirculatim puaps. This safety acticn is expected to ocor, but is not a requirement fl this event. (4

SAFE QUESN AI4)LYSIS APPEM)IT 14 Page 9 of 11 6'an loss of Feedwater Flw TRE 14 -1 Owksli a1 T.L c Sarg

                                                                                        ~

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REACT tMNXHARE TN SER. RERLIS SAFETY ACTI% SYS SYSIEII STATE Code Title Code Title ML - ~ N% 2-2 Fuel Failure 24 Pw Reduce 66 RECIN1ATIN Recirculatim I/6 drive aotor breaters f200, electrical systeo) tripped m lw water level IL2) signal. This safety actim is epected to occur, but is not a requirmnt fcr this event. If 2-2 Fuel Failure 24 Pw Reduce 99 REACim PROIEI)I Provide RPS ATU paar to feedwater systen INl low water level K2) instruaentatim fcr trip of reclrculatim pumps. Ibis safety acticn is expected to ocnr, but is not a requireaat fcr this'event. If 2-2 Fuel Failure 24 Pw Reduce 200 BEIRICAL (pen recirculation IN drive outcr breahrs f200, systea) m feedwater systeo IN) lw 'lectrical water level (L2) signal. This safety actim is Npected to occur) but Is not a feqLLlr Mlt fcr this event. 2-1 Rad Release 26 Est Pri Cent 3 FEHNTER SF Provide Iw water level (U) signal to reacts protectim systen f99) fcr initiatim of Q actim is e ed to occur, but is not a requireeent for this t

                                                                 'nce                 s no   fuel failure.

IXEF 2-1 Rad Release 26 Est Pr 0j. i m) ti Ir ei <<g Prioary Ccntaineent systeo f64). This saf ty actim is espected to occur, but is not a requlreeen or as failure. IXEF 2-1 Rad Release 26 Est Pri Cmt 64 PRI CDITAIQEXT  %'jpm lM. (O) signal frc~ RPS 199), initiate Q isolatim actims b send priaary/secmdary ccntainant isolaticn signals to systems 2, 65, 69, 74, 75, 76) 77, 64, 90, and 94. This arum is erpected to occur,but is not a requircwot for this event. CKF 2-1 Rad Release 26 Est Pri Cont 7S CNE SPRAT Perforn isolatim actimts) upon receiving lw water level (LZ isolatim signai froa the Prioary Cmtainaent systeo (64). This safety action is epected to occur, but is not a requireaent fa this event since there is no. fuel failure.

58)TIKI AIIALYS1S NRR)1 ll

'SAFE sa     Nlrb Loss of Feedwater IRHKN~ IRev fltnt 1)

TSE )I -1 l7ecked: F C. Eckert T.l. Sarg htll ~P REACT IACCEPTAKE TYA SAFE IRK REM.TS SAFETY ACTlll Sm SYSTf)I FSC STATE Cade Title Cade Title )L Nf GEE IXEF 2-1 Rad Release 26 Est Pri Cent 76 GKA)IISIT MT SF Perfara isolatim actimls) upm receiving law water level IQ) isolatim signal fran the Priaary Cmtainaent systea N). Ms safety ac4m is expected to occN't but is aot a requireaat fcr this event since there is no fuel failure. GKF 2-1 Rad Release 26 Est Pri Cmt 77 SNSIE SF Pertcra isolatim actiaals) upan receiving law water level IQ) isolatim signal frm the Priaary Cmtaintwtt system N). This safety actlm is Npect8I to accN't but is nat a requireant fN this event since there is no fuel failure. GEF 2-1 Rad Release 26 Est Pri Cmt N I SF Perfara water level isolatim actimls) upm receiving IQ) isolatim signal froa the low Priaary Cmtainant systea N). This safety ac'tim is expected to Kit but is liat a requirenent fN this event since there is no fuel failure. lf 2-1 Rad Release 26 Est Pri Cant SS Im SF, 'er,far acnls) upm rec vlng scraa sa'gnal fron the Reactar Protecti tea I99).

   ,  GEF    2-1   Rad Release     26       t Pri Cant 90 RAMATIN        1                                                     i 1                                aa systea  N). This actim is espected to occur, but is require+at far                      ~      e is oo  fuel GEF    2-1   Rad Release     26     Est           9$ TlP                       initiate
                                                                                                       .(W r'f'~'jtdg~lt7 TlP wiMrawaIlAm law water level IQ) signal froa priary cmtainant systea           IN.

This safety actim is espected to ocnrt but is aat a requirmnt for this event since there is no fuel 'failure. 2-1 Lul Re)ease 26 Est Pri Cmt 99 REACHR PITECCl)l SF Provide law water level (Q) signal froa feedwater systea INl to priaary cmtainaent 0'EF systea IM) for initiatim of Q isolatims. This safety actim is espected to murt but is not a requireaent fm this event since there is ao fuel failure. l -~ "' 4. P l ~ ~

SAFE BUIQN A)NL'61S APPEM)lI 1l Page 11 of f,evgi O tN Flm +crt Loss of Feedwater jossss: ~iPW~ E.C. htsI ~4w 7 fRHjSS3~ tRev 1) T,L Sarg TARE 1l "1 ND INXPTARE M tPER. RESIUS SAFETT ACT)% StS NSTE)t SIATE Code Title Code Title )L N% IXEF 2-1 RadRelease 27 Est Sec Cmt 6l PR1 GNAINSIT SF Perfora isolaticn actimts) cpm receiving low water level tQ) isolatim signal free the Reactor protectim Systaa N). This safety actim is expected to occur, but is not a requirmnt for this event since there is no fuel failure. tXEF 2-1 Rad Release 27 Est Sec Cmt 65 SST SF initiate SOT plant start m low water level tQ) signal froa priaary cmtainent systea N). This safety actlm is eapected to occur, but is sot a requireant fa'his event since there is no fuel failure. IXEF 3-5 Pers Overep 36 Cont t)ay Env 3 fEENIER SF Provide low water level tQ) signal to RPS systea tgg) fl initiatim of ccntrol bay isolatim. This safety actim is espected to occur, hut 1s ant a requireant fcr this event since there is no fuel failure. c CKF 3-5 Pers Overep 36 Cent t)ay Env 31 AIR CtM)ITlMIS SF Air cond. 0 0 supply ducts isolate I Eaerg. Pres. Systea tCmtrol Rona Eaerg. Yet. Systea) supplies pres. filtered air to KR cn LLL IQ) signal froa systea 6l, This safety actim is epected to occur, hut is not a requireaent for this event. GEF 3-5 Pas Orere1p 36 Cent Say Env 6l SF tjpm low water level tQ)'ignal 1 ti C 1 e 'e f i failure. ev his t s nce GEF 3-5 Pers Overep 36 Cmt Say Env signal froa ter systea tO3) to priaary cmt. systea N) for initiatim pf control bay isolatim. This safety acticn is apected to anr, hut is not a requireaent fcr this event since there is ao fuel failure. 7)Mao< tz r+~ i SoQ) A4 jrC~ys4~ os)s r"@3

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