ML20217B128

Forwards Matrix & Two Updated Attachments Re Critical Characteristics for Tier 2 Accident Mitigating Sys Discussed in Chapter 15 of Millstone Unit 3 FSAR

ML20217B128
Person / Time
Site:	Millstone
Issue date:	09/18/1997
From:	Schopfer D SARGENT & LUNDY, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
9583-100, NUDOCS 9709230238
Download: ML20217B128 (166)

Text

.

gl s

W, Sa r ge rirt"&h t_u ncly ' ' c

/

Don K. Schopter Vce President 312-269-6078 September 18,1997 Project No. 9583-100 Docket No. 50-423 Northeast Nuclear Energy Company Millstone Nuclear Power Station, Unit 3 Independent Corrective Action Verification Program United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 In accordance with the Sargent & Lundy (S&L) ICAVP Audit Plan, Project Instruction PI MP3-07 and the NRC ICAVP Oversight Inspection Plan, S&L submitted on June 30, 1997 the Critical Characteristics for the Tier 2 Accident Mitigating Systems discussed in Chapter 15 of the Millstone Unit 3 (MP3) FSAR. These Critical characteristics were approved by the NRC in letter dated August 21,1997. The NRC also identified addition 2 safety parameters to be included in the database. We have resiewed the additional critiest characteristics and provide the attached Exhibit I matrix in response to the NRC letter.

During the verification of the Critical Cnaracteristics, we revised some of the Critical Characteristics based on the Millstone Unit 3 design. We are enclosing,wo updated Attachments showing revision bars for specific Critical Characteristics which were e'.ther revised or added. Attachment I consists of a sort of the database by FSAR Chapter 15 accident. This report identifies the associated mitigating system, mitigating componem, critical parameter or characteristic, input assumption and the specific reference or source document (s) which utilize the assumptions. Attachment 2 is a sort of the database by the

(

mitigating system. This report identifies the mitigating system, the mitigating component, J

critical parameter or characteristic, input assumptions, the specific reference or source M

l documents which utilize the assumptions and the Chapter 15 associated accident. In some cases where the critical characteristic relates to the event itself and not a mitigating system it is identified as" Event Related"

ipagggg 7gggg j jigiijg g 3;;;g gj ij P

PDR 55 East Monroe Street. Chicago. IL 60603-5780 USA. 312-269-2000

/

^

c.

United States Nuclear Regulatory Commission September 18,1997 Document Control Desk Project No. 9583-100 Page 2 Please inform us if you need further information.

Yours very truly, 1sK e

D. K. Schopfer I

Vice President & ICAVP Manager DKS:spr Copies E. Imbro - (1/4) NRC T. Concannon (1/2) NEAC J. Fougere - (1/3) NU B. A. Erler -(1/0) S&L

- R. D. Raheja -(1/l) S&L mWavpbwn97w0918 a duc

j 9/18/97

. Exhibit 1 Additional Safety Parameters important to Safety Analysis for Millstone Unit 3 (Response to NRC Letter August 21,1997)

NRC item Description Category' Comments No.

I 11 fli 1

System Parameters Applicable to All Accident Analyses This figure applies to normal plant operational surveillance's which are covered by 2

1.a Rod Cluster Control x

Assembly (RCCA) position Tech. Spec. requirements 4.1.3.1.1 through 6. Rod Drop Events have been versus drop time (FSAR considued in critical characteristics (CC) 26.27,60,89,97.107.144,145.176, and 196.

Figure 15.0-3)

These figures apply to normal plant operational surveillance's which are covered by 2

1.b RCCA reactivity worth as a x

function of RCCA insertion Tech. Spec. requirements 4.1.3.1.1 through 6. Rod Drop Events have becri (FSAR Figures 15.0-4 and 5) considered in critical characteristics (CC) 26.27.60,89.97.107.144,145.176, and 196.

RCP coastdown start time is critical for a loss of offsite power (LOOP) since credit is 3

1.c Reactor Cooiant Pump (RCP) x coastdown flow rate taken for heat removal via natural circulation. Since LOOP is bounded by the analysis for Loss of Normal FW, CC 230 has been added to that accident.

2 Decrease in Feedwater Temperature Event This parameter is an output from the W analysis. The analysis proves that other 2.a The maximum decrease in x

the feedwater temperature transients are more significant; these assumptions and parameters are verified (35 deg F) assumed forloss through CC 9 and 10.

of one train of feedwater heater

' Category 1: The origined sutettal contains CC which address these sems. These CC are identiRed h the

• Comments" columrt Category 11: Additenal CC have been added to our June 30,1997 sutmital to address these llems. Details are provided in the"Commerts' colur9n.

Category 111: These items do not require any addihonal CC and justlRcebon is provided in the " Comments" columrt Tier 2 idenbfies CC for accident rndhgehng systems. The Rod Drop events have been considered in the identlhed CC. Actual drop trnes subsequent to the merHants have been vertRed.

2 8 RCP coastdown flow rate is W input assumpeon and is not a reedsy vennoble parameter Page 1 of 3

9/18/97 Exhibit 1 Additional Safety Parameters important to Safety Analysis for Millstone Unit 3 (Response to NRC Letter August 21,199'/)

NRC item Description Category' Comments No.

I il ll1 3

Steam Line Break 3.a RCP coastdown flow x

Steam Line Break without a LOOP is more severe than with a LOOP (Ref. NEU 623). See item 1.c for discussion of RCP coastdown flow and LOOP. CC 230 has been added for Loss of Normal FW.

3.b Reador Protedion System x

CC 218,219,224, 225, and 226 for overpower and SIS related reador trips have (RPS): Reador tiip setpoints been added for this accident.

and trip delay times 3.c High Pressure Safety x

x CC 227 and 228 were added 6 address setpoints. CC 211 was added to address injedion (HPSI) aduation:

flow. Existing CC 16 (Accident 15.1.4) addresses accumulator boron concentration.

Setpoints, delay time, safety injedion (SI) flow with and without a loss of offsite power, and Si boron concentrations 3.d Main Steam System:

x CC 207 added to address Main Steam isolation dosure time. CC 227,228, and 229 lsolation valve aduation added to add ess aduation setpoint.

setpoint, delay time, and isolation valve dosure time 3.e Main Feedwater System:

x CC 227 and 228 added to address Main Feedwaterisolation aduation setpoint. No Isolation valve aduation specific closure time identified in the analysis.

setpoint, delay time, and isolation valve closure time Page 2 of 3

9/18/97 Exhibit 1 Additional Safety Parameters important to Safety Analysis for Millstone Unit 3 (Response to NRC Letter August 21,1997)

NRC item Description Category' Comments No.

I il 111 t

3 Steam Line Break (continued) 3.f Auxiliary Feedwater System x

CC 227 and 228 added to address AFW actuation setpoints. CC 208 added to (AFWS): Aduation time of address delay time. Existing CC 138 and 139 (Accident 15.1.4) address AFW flow AFWS with and without a loss parameters. These accidents have been determined to be the limiting of offsite power, flow rate, characteristics with regard to AFW performance. Thus, the verification of these CCs temperature, isolation verifies AFW performance for the MSLB.

~

actuation setpoints, delay time, and isolation valve closure time 4

Turt>ine Trip t

b 4.a Pressurizer spray:

x This accident has been analyzed with arw1 wdhout the effects of pressunzer spray.

[

Operability, actuation The conclusion is that credit is not taken for pressurizer spray as a mitigating i

setpoint, flow rate, and fluid component for this event. (Ref. NEU-96-623) l temperature 5

A Loss of Normal Feedwater 5.a RCP coastdown flow x

CC 230 has been v$ded under the Loss of Normal Feedwater Accident. See item following the RCP trip thal 1.c.

results from a loss of offsite power i

i I

I Page 3 of 3 i

1

i Project No. 9583-100 File MP3-7.0-001 4

9/18/97 Millstone Unit 3 Accident Analysis Critical Characteristics for Chapter 15 Accident Mitigating Systems (Sorted by Chapter 15 Accidents) i (78 Pages)

Chapter 15 Accident Mitigating Systems sai>Ioa of Milstone t.'ait 3 pjg7 ICAVP (Affected Accidents)

AFFECTEDACCIDEVTS:

FeedwaterSystem Malfunctions that Result in a Decrease in FeedwaterTemperature (FSAR 15.1.1) i COMPONENT DESCRIPTION PARAMETER DESCRitilON INPUT ASSUMi'flON SAFIITY ANALYSIS REFERENCES

~

.. t t,. T.7

• TEi l't 7 7 ' "' ~ ~ ~ "

. 4.........

s?.

o. e.

=.:

~ ~ ~ ~

'~ ' " ~. ~ "....," " ' ^ ~ " ~~".' a

: t~

• . - 7. : "*e y.

p;.

4 Instrumentation - Steam Delay Time

1. 8: 2.5 seconds (The definition of delay t'ONE*-G-0075 Table 5.1.1-4 Generator Ifigh-liigh Level time is given in FSAR Table 15.0-4).

Sensor Verificarle fishod: Verification of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Survei!!ance and Calibration Test re.sults.

l Instrumentation - Steam Steam Generator Water Level

1. 7: Assumed trip setpoint is 100% of 90NE*-G-0075 Table 5.1.1-4 Generator fligh-Iligh Level narrow range span.

Sensor VenJIcarlos Method: Venfication of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Survei!!ance Test Procedures and Surveillance and Calibration Test results.

...., m. :. T-N....,- C..

2

..-;.g.q.

..vrqf-g...,.s.,1, y,. 5 n. - m..

q 77.. mf.

7 : w e p o m m 7r m. w m m m yr:r

~~-

.g Instrumentation - Neutron Flux Delay Time

1. 2: A delay time of 0.5 seconds for high 90NE*-G-0075 Table 5.1.1-4 Sensor 3 and low setpoint trip Ven]icarlon Method: Verification of the instrument performance will be accomplished by review of Technical Specifications.

Calibration and Surves!!ance Test Procedures and Surveillance and Calibration Test results

I Page i of78

Chapter 15. Accident Mitigating Systems l,",',,,"y']s""

Raision: of Cfl&97 ICAVP(Affected Accidents)

AFFECTEDACCIDENTS:

Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (FSAR 15.1.1)

COMPONENT DESCRil'IlON PARAMETER DESCRIPTION INPUT ASSUMl" TION SAFETY ANALYSIS REFERENCES lastrumentation - Neutron Flux Neutron Flux

1. I: Assumed high trip setpoint of 89% and 90NE*-G-0075 Tat le 5.1.1-4 Sensors 118% ofrared thermal power for 3 loop and 4 loop operation, respectively. Assumed low trip setpoint of 35% of rated thermal power for 3 and 4 loop operation.

Venylcation AIcthod: Venfication of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results.

Instrumentation - Neutron Flux Neutron Flux

1. 186 For power level greater than P6 but 90NE*-G-0075 Table 5.1.1-4 Sensors less than Pio, accident parameter exceeds setpoint.

VenJIcation AIcthod: Verification of the instrument performance will be accomplished by review of Technical Specification requirements. Calibration and SurveiIlance Test Procedures, and Calibration and Surveillance Test results.

bstrumentation - Overpower dT Delay time

1. 4:

7.0 seconds (The definition ofdelay 90NE*-G.0075 Table 5.1.1-4

, Sensors time is given in Note (a) of FSAR l

Table 15.0-4.)

Verifration AIcthod: Verification of the instrument performance will be accomplished by review of Technical Specifications, _

Calibration and Survei!!ance Test Procedures and Surveillance and Calibration Test results.

Page2 of 78

Chapter 15 Accident Mitigating Systems

[f",",ff""

Kaisioa or 9/1897 ICAVP(Affected Accidents)

AFFECTEDACCIDENTS:

Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (FSAR 15.1.1)

COMPONENT DESCRIPTION PARAMETER DESCR!VIION INPUT ASSUMVilON SAFETY ANALYSIS REFERENCFS Instrumentation - Overpower dT Differential Temperature

1. 3: Assumed trip setpoints cited as 90NE*-G-0075 Table 5.1.1-4 Sensors variable in the eference analysis 90NE*-G-0075 Figure 5.1.1-6 (Figures 5.1.1-6 and 5.1.1-7) 90NE*-G-0075 Figure 5.1.1-7

~

Verifrarion Method: Venf; cation of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Survei!!ance and Calibration Test results.

Instrumcotation -

Differential Temperature

1. 5: Assumed trip serpoints cited as 90NE*-G-0075 Table 5.1.1-4 i Overtemperature dT Sensors variable in the reference analysis 90NE*-G-0075 Figure 5.I.1-6 (Figures 5.I.1-6 and 5.1.1-7) 90NE*-G-0075 Figure 5.1.1-7 Verr 7xation Method: Verification of the instrument performance will be accomplished by review of Technical Specifications, Cahbration and Surveillance Test Procedures and Surveil lance and Calibration Test results.

I lastrumentation -

Differential Temperature

1. 6:

7.0 seconds (The definition ofdelay 90NE*-G-0075 Figure 5.1.1-4 Overtemperature dT Sensors time is given in Note (a) of FSAR 90NE*-G-0075 Figure 5.1.1-6 Table 15.0-4).

90NE*-G-0075 Figure 5.I.1-7 Veri ration Method: Venfication of the instrument performance will be accomplished by review of Technical Specifications, f

Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results.

I Page 3 of 78

~

scrision: of Chanter 15 Accident Mitioatine Systems B

o Milgone l'ais 3 a

CflL97 ICAVP(Affected Accidents)

AFFECTED ACCIDENTS:

Feedwater System Malfunctions that Result in an inercase in Feedwater Flow (FSAR 15.1.2)

COMPONENT DESCRIPTION PARAMETER DESCRIPI' ION INPUT ASSUMFflON SAFETY ANALYSIS REFERENCES

{,. 3:T ~,;',, fXf ; ' : :_,T" ~~ J.7"'G[,y ~f,~ " ~' Qp,);^"".7':; ~~**~;W79m.y'v**yr?7. ;Wpr3933?p?g;g e

lastrumentation - Steam Steam Generator Water Level

1. 207 liigh steam generator level at 100*6 of NEU-96-623 i

Generator Level Sensors narrow range span for closure of FSAR Table 15.0-4 isolation valves (7.0 s delay) and turbine trip (2.5 s delay)

Verificasion Method: Venfication wi!! be accomplished by a review of Technical Specifications, surveillances, calibration proceduras, and test results.

Qy.m rhG.iRG-- _ :.;a'TE't disfiCTiL!#iti"* #""IITMP'"JfEMI@%EElSTf05M$$$$$33Q$~~5Cdi$g]

Feedwater Control Valves Feedwater Flow

1. 9: A failure of a feedwater control valve 90NE*-G-0075 (pg 5-26) at full power will result in a step NEU-96-623 increase in feedwater flow to one steam generator not to exceed 140% of nominal feedwater flow. (FSAR 15.I.2.2, Case 2, item I) j Veripcation Merkod: Venfication of the system flow performance will be accomplished by review of system and component calculations and performance test results.

Feedwater Control Valves Feedwater Flow

1. 10: A failure of a feedwater control valve 90NE*-G-0075 (pg 5-26) at zero power will result in a step NEU-96-623 mcrease in feedwater flow to one steam generator not to exceed 200% of nominal feedwater flow. (FSAR l

15.I.2.2, Case 2, Item 2)

Verificarlos Method: Verification of the system flow performance will be accomplished by review of system and component calculations and performance test results.

Page 4 of 78

... ~

Chapter 15 Accident Mitigating Systems Ra

• ion: of Mibsone Iinit 3 CflOf97 ICAVP(Affected Accidents)

AFFECTEDACCIDENTS:

Feedwater System Malfunctions that Rrsuit in as increase in Feedwater Flow (FSAR 15.1.2)

COMPONENT DESCRIlilON PARAMETER DESCRIPTION INPUT ASSUMi'IlON SAFETY ANALYSIS REFERENCES EGeiisiraissis7-essiisiKeresif aWKas'asiAwareweb:rm-ieweise own,anssww-=muusmmuam:1.e.i.: rem 2:r a:+.nh;we c.c.r..i Instnunentation - Neutron Flux Delay Time

1. 200 A delay time of 0.5 seccads for high 90NE*-G-0075 Table 5.1.1-4 i

Sensors and low setpoint trip Venficarlon Method: Verification of tne instrurnent perforrnance will be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Survei!!ance and Calibration Test results.

lastrumentation - Neutron Flux Neutron Flux

1. 199 Assumed high trip serpoint of 89*& and 90NE*-G-0075 Table 5.I.1-4 1

Sensors 1)8% of rated thermal power for 3 i

loop and 4 loop operation, respectively. Assumed low trip setpoint of 35% ofrated thermal powri for 3 and 4 loop operation.

henT: cation Method: Venficaton of the instrument performance will be accomplished by review of Technical Specifications.

Calibration and Surveillance Test Procedures and Survel!!ance and Calibration Test results.

Instrumentation - Turbine Trip Reactor Trip

1. 201 Turbine trip due to high steam NEU-96-623 1

generator level generates a reactor trip.

(FSAR Fig _15.0-8)

Verification Method: Venfcation wi:1 be accomplished by a review of Technical Specifications and surveillance test data Page5 of 78

Chapter 15 Accident Mitigating Systems s< vision: of Milstone Unit 3 9/1&D7 ICAVP (Affected Accidents)

AFFECTEDACCIDENTS:

Imadvertent Opening of a Steam Gescrator Relief or Safety Valve Causing a Depressurization of the Main Steam System (FSAR 15.1.4)

COMPONENT DESCRIMION PARAMETER DESCRIMION INPUT ASSUMMION SAFETY ANALYSIS REFERENCES c;, ryy..r -. - ;, q _ q p g--,

, x, j ;,,.

m g, 7 - -- - -

. ~ ;

~~e c.-

7- - wsy;yg p q y-g ggy7p-Auxiliary Feedwater Pumps System Flow

1. 139 Minimum flow is greater than or equal 90NE*-G-0075 Section 5.1.3A.2 to 900 gallons per minute for 3 loop operation Verification Method: Venfication of system perfomiance will be accomplished by review of system and component calculations and performance and survei!!ance teu results.

Auxiliary Feedwater Pumps System Flow

1. 138 Minimum flov is greater than or equal 90NE*-G-0075 Section 5.1.3.4.2 to 1200 gallons per minute for 4 loop operation Verification Meskod: Venfication of system performance will be accomplished by review system and component calculations and performance and surveillance test test resuits.

,, ; p y grv t-y. + m,. : :~ c a c, g _,. 7 wwwwpyg T.; y 7 7 7

..e-.=;

7 ;,; n, 7.,. g.

,r, i

..v.....,.g.

1 Accumulators Concentration

1. 16: Greater than or equal to 2500 ppm NEU-96-623 (pg 15.1-14)

Vers]Icasion Method: Verification of this parameter will be accomplished by review of water chemistry sampling procedures and the historical sampling records.

Page 6 of78

Chanter 15 Accident Mitieatine Systems sceloa: of a

o o

Milstone l'ait J 07807 ICAVP (Affected Accidents)

AFFECTED ACCIDENTS:

Inadvertent Opening of a Steam Generator Relief or Safety Valve Causing a Depressurization of the Main Steam System (FSAR 15.1.4)

COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMl" TION SAFETY ANALYSIS REFERENCES

~4 7.L V Q,

NGT iT T::[.:Mf.: 4. -?'4= {: = 7,,, ge ;w. (J' e s Q [' j f.} '

- : + ~ _ f - ~ ?-N

.m m.-

Pumps Pump Flow

1. 13: Minimum flow from a single high 90NE*-G-0075 (pg 5-32) head safety injection pump is less than the minimum flow from a single centrifugal charging pump to assure the cccident analysis are conservative.

Ven7, carton Methoh Venfication of the system ilow performance will be accomplished by review of Startup Test and Surveillance Test results, and Technical Specification. Also, review of calculations extrapolating mini-flow results to fu!I flow conditions will be required.

m

. On ],, :. : ;. ; o a. aY3= i= r ? ? 'G rW},

j. '

7{Q~~" 77 y "* P y"*"y}"Q[Dfi}Qg.y l i 1

lastrumentation - Low-Low Steam Genentor Water Level

1. 12: Delay time will not exceed 2.0 FSAR Tab!c 15.0-4 l Steam Generator Level Sensor seconds. (Delay time definition is 90NE*-G-0075 Table 5.1.1-4 contained in FSAR Table 15.0-4)

Verij7 cation Method: Venfication of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results. Instrumentation - Lowd ow Steam Generator Water Level

1. 11: Trip setpoint is 0% of narrow range 90NE*-G-0075 Table 5.1.1-4 Steam Generator Level Sensor span for feedwater line break,10% of FSAR Table 15.0-4 narrow range span for loss of normal feedwater/ loss ofoffsite power.

Verification Methok Venfication of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results. Page 7of78

Chapter 15 Accident Mitigating Systems [,",','",,",'j';l" scrofon: or 9/l&47 ICAVP(Affected Accidents) AFFECTEDACC1DENTSr Inadvertent Opening of a Steam Generator Relief or Safety Valse Causing a Depressurization of the Main Steam System (FSAR 15.I.4) COMPONENT DESCRIFFION PARAMETER DESCRIFIlON INPUT ASSUMFIlON SAFETY ANALYSIS HEFERENCES Instrumentation - Pressurizer Delay Time

1. 204 2.0 seconds to initiate SI signal 90NE*-G-0075 Table 5.1.I-4 i

Pressure Sensors Verifkation Method: Venfication of the instrument performance will be accomplished by review of Technical Specifications; Cahbration and Survei!Iance Test Procedures and Surveillance and Calibration Test results. 1 Instrumentation - Pressurizer Pressunzer Pressure

1. 203 Assumed trip setpoint,1845 psig, cited 90NE*-G-0075 Table 5.1.1-4 i

in the reference analyses. Pressure Sensors Veri /icarlon Method: Venfication of the instrument performance will be accomplished by review of Technical Specification, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results. l .._ -,,, - ~ ~%m% CAentc5cDMrmv;Md.- E - ' y,_ m. _1t'JI5..=e G ".'*.:. ;e: 1- ~W. ~1?. n Safety lojection Pumps Systera Flow

1. I49 Full flow is achieved within 42 90NE*-G-0075 Section 5.1.3.4.2 I

seconds after accident with and without loss of offsite power. Venfkallon Method: 1) Verification of the sequencer performance wil1 be accomplished by review of cahbration and survei!!ance test results.

2) Diesel performance will be venfied by a review of surveillance procedures and test resuits from diesel test and ESF load sequencing tests.

Page 8 of 73

r n *lon: of Chanter 15 Accident Mitioatine Svstems s-D o J Milstone l'ait 3 C/lL97 ICAVP(Affected Accidents) AFFECTEDACCIDENTV: laadwertent Opening of a Steam Generator Relief or Safety Valve Causing a Depressurization of the Main Steam System (F3AR 15.I.4) COMPONENT DESCRIPTION PARAMt.TER DESCRIFTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES n

..w. w., ~

,n-2,,i wy f4;gs,wE%%%,ig w -- .- c,- S A m.n. u.- - - w.- ' : th N.4%%*. F - a-h_ - t 43 hw Main Steam isolation Valves Closure Time

1. 15: Main Steam Isolation Valves close in NEU-96-623 (pg 15.1-9) 3 MSS-CTV27A, B, C, D less than 10 seconds-Verification Afethod: Verification of valve performance will be accomplished by review of surveillance test results for the valve, Ca!bration Procedures and test results for valve logic and review of the maintenance and modification records.

Steam Dump Valve, Steam Relief Mass Flow Rate

1. 14: Maximum flow rate of 277 poun is per 90NE*-G-0075 (pg 5-33)

Valve,and Steam Safety Valve second at a secondary system pressure of 1200 psia with offsite power. l Venfication Aferhod: Verification of the valve's performance will be accomplished by review of bench test results. specification, l manufacturer's data and a review of maintenance and modification records. l l W M."R Y R W $# A.$. $ $ ' $0? ? V'? ? ? $ Instrumentation - Neutron Flux Delay Time

1. 223 A delay time of 0.5 seconds for high 90NE*-G-0075 Table 5.1.1-4 i

Sensors and low setpoint trip 5'ert/lcation Method: Verification of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Survei!!ance Test Procedures and Survei!!ance and Calibration Test results. i l Page 9 of 78

Chapter 15 Accident Mitigating Systems 7,",'l"l',y'l" Raision: or Cfl007 ICAVP(Affected Accidents) AFFECTEDACCIDEA75: Inadvertent Opening of a Steam Generator Relief or Safety Valve Causing a Depressurization of the Main Steam System (FSAR 15.1.4) SAFETY ANALYSIS REFERENUES COMPONENT DESCRIPTION PARAMETER DESCRil'IION INPUT ASSUMPTION Instrumentation - Neutron Flux Neutror Flux

1. 222 Assumed high trip setpoint of 89% and 90NE*-G-0075 Table 5.1.1-4 1

Sensors 118% ofrated thermal power for 3 ~ loop and 4 loop operation, respectively. Assumed low trip setpoint of 35% of rated thermal power for 3 and 4 loop operation. Ven7scation Method: Venfication of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results: Lastrumentation - Overpower dT Delay time

1. 221 7.0 seconds (The definition of delay 90NE*-G-0075 Table 5.1.1-4 i

Sensors time is given in Note (a) of FSAF_ Table 15.0-4.) Venfication Method: Venfication of the instrument performance will be accomplished by review of Technical Specifications, Calibretion and Survei!!ance Test Procedures a - d Surveillance and Calibration Test results. Instrumentation - Overpower dT Differential Temperats

1. 220 Assumed trip serpoints cited as 90NE*-G-0075 Table 5.1.1-4 i

Sensors variabic in the reference analysis 90NE*-G-0075 Figure 5.1.1-6 (Figures 5.1.1-6 and 5.1.1-7) 90NE*-G-0075 Figure 5.1.I-7 Verification Method: Verification of the instrument performance will be accomplished by review of Technical Specifications, 4 Calibration and Survei!!ance Test Procedures and Survei!!ance and Calibration Test results. Page 10 of 78

s< *io : of Chanter 15 Accident Mitigating Systems r o a s Milstone Ifait 3 Cfl&47 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Stemas Systems Piping Failure (FSAR 15.1.5) COMPONENT DESCRIPTION PARAMETER DESCRIMION INPUT ASSUMMION SAFETY ANALYSIS REFERENCES _. j a..., Auxiliary Feedwater Pumps Time to Full Flow

1. 208 Water supplied to steam generators no NEU-96-623 I

later than 10 min after accident. Verification Meshod: Venfy by reviewing design requirements, surveillance procedures, and test results. y. Instrumentation - Containment Containment Pressure

1. 228 Safety injection signal at lii-1 setpoint; HEU-96-623 I

Pressure Sensors steam line isolation signal at Ili-2 setpoint. Safety injection actuation logic initiates AFW flow and FW isolation. Verificasion Method: Venfy by reviewing Technical Specification requirernents, Surveillance and Calibration Procedures, and test results. 11 acntstion - Pressurizer Delay Time

1. 25i 2.0 seconds tc, initiate SI signal 90NE*-G-0075 Table 5.I.1-4 i

Pressure Sensors Verifics Jon MerAod: Venfication of the instrument performance veill be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results. Instrumentation-Pressurizer Pressurizer Pressure

1. 24: Assumed trip serpoint, I845 psig, cited 90NE*-G-CC75 Table 5.I.1-4 I

Pressure Sensors in the reference analyses. Verification Meded.- Venfication of the instrument pedormark:e will be accomplished by review of Technical Specification, Calibation and Surveillance Test Procedures and Surveillance tr.d Calibration Test resuits. Page iiof78t 2 -w-

m. Chapter 15 Accident Mitigating Systems l,"','",",,jl" revision: of C11097 ICAVP(Affected Accidents) AFEECTEDACCIDENTS: Steams System Piping Failure (FSAR 15.I.5) COMPONENT DESCRil" TION PARAMETER DESCRil' TION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES Instrumentation - Steam Line Iligh Negative Steam Pressure Rate

1. 229 Steam line isolation occurs at sepoint.

NEU-96-623 1 Pressure Sensors (Below P-1I) ( Verification Meskod: Venfy by reviewing Technical Specification requirement;, Surveillance and Calibration Procedures, and test results. Instrumentation - Stesm Line Low Compensated Leam Line Pressure

1. 227 Safety injection and steam line NEU-96-623 i

Prer:;ure Sensors isolation signal at 658.6 psig. Safety injection actuation logic initiates AFW flow and FW isolation. l Venfication Method: Venfy by reviewing Technical Specification requirements Survei!!ance and Calibration Procedures, and test results. l ~ g DL1267AhaafMssrmiandesafinnhEEEE55kiett!C11. N - f MERIBTW I Safety injection Pumps System Flow

1. 211 Full flow is achieved within 42 NEU-96-623 i

seconds after accident with and without loss ofoffsite power. Verification Method: 1) Verification of the sequencer perfctmance will be accomplished by review of calibration and surveillance test results.

2) Diesel performance wdl be venfied by a review of surveillance procedures and test resuits from diesel iest and ESFload sequencing tests.

u Page 12 of 73 C

merision: 81 Chanter 15 Accident Mitieatine S,ystems r o o M;3,tvec i:ais 3 CflL97 ICAVP (Affected Accidents) 'AFFECTEDACCIDEN75: Steam System Piping Failure (FSAR 15.1.5) COMPONENT DESCRIPTION PARAMETER DESCRafilON INPUT ASSUMITION SAFETY ANALYSIS REFERENCES ^ SiEiEiFT4N5Ml.9E1 Main Steam isolation Valves Closure Time

1. 207 Main Steam Isolation Valves close in NEU-96-623 1

3 MSS-CTV27A, B, C, D tess 0an :: seconds. Ven]kation Method: Verification of valve performance will be accomplished by review of survei!!ance test results for the valve, Cahbration Procedures and test results for valve logic and review of the maintenance and modification records. , 3:, ' -,. ; -% y _. a: $7'"TIqa-yp. +G .y,,,. .r~7~ --~~!- l ~. ~ i ~~ ~;~,~ ~~ Instrumentation - Neutron Flux Delay Time

1. 225 A delay time of 0.5 seconds for high 90NE*-G-0075 Table 5.1.1-4 I

Sensors and low setpoint trip p Venykation Method: Venfication of the instrument performance will be accomplished by review of Technical Specifications, Cahbration and Surveillance Test Procedures and Surveillance and Calibration Test results. Instrumentatios - Neutron Flux Neutron Flux

1. 224 Assumed high trip serpoint of 89% and 90NE*-G-0075 Table 5.1.1-4 i

Sensors

18% of uted thermal power for 3 loop and 4 loop operation, respectively. Assumed low trip l

setpoint of 35% of rated thermal power for 3 and 4 loop operation. VerijkasJon Meshed: Venfication of the instrument performarce will be accomplished by review of Technical Specifications, Cahbration and Surveillance Test Procedures and Surveillance and Calibration Test results. Page 13 of78

Chapter 15 Accident Mitigating Systems ["*"",['l' "l R<raio : of C'i&97 ICAVP (Affected Accidents) AFFECTLDACCIDENTS: Steard System Piping Failure (FSAR 15.I.5) COMI'ONENT DESCRil"flON PARAMETER DESCRil'1 ION INPUT ASSUMirTION SAFETY ANALYSIS REFERENCES lastrumentation - Overpower dT Delay time

1. 219 7.0 seconds (The definition of delay 90NE*-G-0075 Table 5.1.1-4 1

Sensors time is given in Note (a) of FSAR Tab!c 15.0-4.) Versylcation Aferkod: Venfication of the instrument performance will be accomplished by rev'ew of Technical Specifications, Cahbrabon and Survei!!ance Test Procedures and Succeillance and Calibration Test results. Instrumentation - Overpower dT Differential Temperature

1. 218 Assumed trip serpoints cited as 90NE*-G-0075 Table 5.I.1-4 1

Sensors variable in the reference analysis 90NE'-G-0075 Figure 5.1.1-6 (Figures 5.1.1-6 rnd 5.1.1-7) 90NE >-G-0075 Figure 5.1.1-7 l Veri /lcarlon.lferhod: Venfication of the instrument performance will be accomplished by review of Technical Specifcations, Cahbration and Surveillance Test Procedures and Survei!!ance and Calibration Test results. Instrumentation - 51 Actuation ReactorTnp

1. 226 Safety injection signal generates a NEU-96-623 I

legic reactor trip. Versibilon Aferhod: Verification will be accomplished by a review of Technical Specifications and surveillance test data. Page 14 of73

.....a me elon: 81 Chanter 15 Accident Mitiestine S<ystems Misse e tr i 3 a a-o o Cfl&97 ICAVP (Affected Accidents) AFFECTEDACCIDENJ5: Stenen Systema Piping Failure (FSAR 15.I.5) COMPONENT DESCRIFflON PARAMETER DESCRIFTION INPUT ASSUMFflON SAFETY ANALYSIS REFERENCES -en w

• v5..- -

- 9 ' y yy. - - ' ~ - - - = ~ ~ - " z' w977 1 Vanous OtTsite Dose (E)

1. 198 FSARTable 15.1-3 FSAR Section 15.I.3 i

Venfication Method: Verify try review of the vendor parameters and associated accident dose analyses. Vanous Offsite Dose due to Main Steam Line Rupture

1. 11: FSAR Table 15.I-3 and 15.0-2 FSAR Table 15.1-3 Venficssien Meshod: Venfy by assuring each parameter is consistent with plant design and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.0-8.

7.. _ _, Steam Generators Feedwater Flow

1. 23: 5F9,000 lbs for N and N-1 loop FSAR Table 15.1-3 operation (0-2 hours after acJint):

1,019,000 lbs for N nd N-! loop operation (2-8 heen arter accident) 3 (Unaffected Steam Generator) Venylcation Method Ventic;;00n of this parameter will be accomplished by review of design documentation and calculations which establish the equipment volume and operating flows and calculations for die accident conditions. Page 15 of78

Chapter 15 Accident Mitigating Systems l,"'l~','j','f" Raision: of Cfl&97 ICAVP(Affected Accidents) i AFFECTEDACCI3ENTS: Steam System Piping Failure (FSAR 15.1.5) COMi'ONENT DESrRIFFION PARAMETER DESCRit' TION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES Steam Generators Flow

1. 19: 167,000 pounds (Initial Steam and FSAR Table 15.I-3 Water Release From the Affected Steam Generator Over the First 30 Minutes of Accident)

Verificasion Merhod: Venfication of this parameter will be accomplished by review of design documentation and calculations which establish the equipment volume and operating flows and calculations for accident conditions. Steam Generators Flow

1. 20: 1300 pounds (Long Term Steam FSAR Table 15.1-3 Release (0-8 hours) from affected l

steam generator Venylcarlon Method: Venfication of this parameter will be accomplished by review of design documentation and calculations which estabbsh the equipment volume and operating flows and calculations for the accident conditions. Steam Generators Flow

1. 21: 417,000 pocads for N loop operation, FSAR Table 15.I-3 433,000 pounds for N-1 loop operation (0-2 hours after accident); 912,000 pounds for N loop operation,912,000 pounds for N-1 loop operation (2-8 hours after accident)(steam release from unafTected steam generators)

Verificasion Method: Venficath.n of this parameter will be accomplished by review of design documentation and calculations which establish the equipment volume and operating flows and calculations for the accident conditions. Page 16 of 78

Chapter 15 Accident Mitigating Systems [,'*','j';"" Revision: of C11LO7 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Steam System Piping Failure (FSAR 15.1.5) COMPONENT DESCR!!"IlON PARAMETER DESCRil" TION INPUT ASSUMFTION SAFETY ANALYSIS REFERENCES Steam Generators initialInventory

1. 22 Steam (pounds per generator): 8,000 FSAR Table 15.1-3 for N loop operation,7,600 for N-I loop operation. Liquid (pounds per generator) 103,000 for N loop operation,104,000 for N-1 loop operation Verification Aferked: Venfication of this parameter will be accomplished by review of design documentation and calculations which establish the equiprnant volume and operating flows and calculations for the accident conditions.

! Steam Generators Leak Rate

1. 17: Primary to secondary leakage FSAR Section 15.1.5.4 I

assumptions for the affected Steam l l Generators is 0.347 gym and unaffected Steam Generators is 0.653 epm. Verification Aferkod: Verification will be accomplished by review of the accident dose calculations for this accident. Page 17 of78

= - - Chapter 15 Accident Mitigating Systems [,',','",",,j;"" scridea: 81 D'i&97 ICAVP(Affected Accidents) AffECTEDACCIDENTS: Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow (FSAR 15.2.1) COMPONENT DESCRIPTION PARAME1 ER DESCRIPIlON INPUT ASSUMFflON SAFETY ANALYSIS HEFERENCES ,_ _,,3-- , ;. 7 -- (, - ; ;, ;, : 7- _ ;, 3 ---.. - -- m .~ .v .. y e .. + z ;. m. waa m gq g a Pressure Regulators Number

1. 78: No pressure regulators in system FSAR Section 15.2.1, page 15.2-1 l'crtyication Method: Venfy by reviewing system desig.n to determine if the system contains steam pressure regulators.

l l t ' Page 13 of 73

Chapter 15 Accident Mitigating S ' stems sa*io=: of ' Mitstone l'ait 3 9/l&97 ICAVP(Affected Accidents) AFFECTEDACCIDEVTS: Loss of Enternal Electrical Lead (FSAR 15.2.2) COMPONENT DESCRil'IlON PARAMETER DESCRit'IION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES ..,,g. -my,y g,. ,m,.mm.,_n < y., iW; WEid@JM .,1 E a c 2 O 3 ; E ' ' ' / I~.'**' ?"'i 5 ..,,2.- Stop Valves 3 MSS-MSVI,2,3, Closure Time

1. 79; Closure time <= 0.1 sec on turbine trip FSAR Section 15.2.2.1, page 15.2-3 1

4 NEU-96-623 Veryication Method: Verify by reviewmg any Technical Specification requirements, Survei!!ance Procedures and test results. Turbine Control Valves,3 MSS-Closure Time

1. 80: Closure time > 0.3 second on loss of NEU-96-623 1

MCVI,2,3,4 electrical load (For the turbine trip to FSAR Section 15.2.2.1, page 15.2-3 be controlling esent) Verificarlos Method: Venfy by reviewing Technical Specification requirements, Surveillance Procedures and test results i Fm9itiruTi afriliiG4skiiriiiEKins6irderum nonhi9ersiae ove"mciiin!5iUiisiiNNOT'fHf2.M??"" T"TTP?C:S'E"EC'T: v l l losuumentation - Pressurizer Pressurizer Pressure

1. 205 2425 psia, delay time 2 seconds (Table 90NE'-G-0075 (pg 5-21) 1 Pressure Sensors 15.2-1)

Vr VIcationMethod: Verify by reviewing Technical Specification requirements, Surveillance and Calibration Procedures and test results. Instrumentation - Steam Steam Generator Water Level

1. 206 Low Low reactor trip at 10% of 90NE*-G-0075 (pg 5-21)

I Generator Level Sensors narrow range span, delay time 2 seconds Verification ?.'erkoi Verify ~3y reviewing Technical Speibtinn requirements, Surveillance and Calibration Procedures and test results. Page 19 of78

revnio : of Chanter 15 Accident Mitigatine Svstems a o o J Mildene Itait 3 CJiL97 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Turbine Trip (FSAR 15.2.3) COMPONENT DESCRIPTION PARAMETER DESCRil'IlON INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES hvnTE:aa rWEa_E== m m&sdf asadfRTFAastiin m.m'a'faaRfLi55AAl""lTf*'"

• id22PII31".f.%}ISOfind *.11SfA&3,

'2Js2. I ii Steam Generator Safety Valves Mass Flow Rate

1. 81: The total capacity of the relief valves FSAR Section 15.2.2.1, page 15.2-2 I

are sized to pass 10$% of steam flow NEU-96-623 at rated power without exceeding i10% of steam system design pressure (I320 psia) ..,o Venfication AferAod: Verify by reviewing design requirements, Technical Specification requirements, Survei!!ance Procedures I and test results (Permissive at P9) $$b$$$?5$lN 1 Pressurizer Safety Valves, Mass Flow Rate

1. 86: The total capacity of the safety valves FSAR Section 15.2.3 2, page 15.2-7 1

3RCS'SV8010A,B,C are sized to prevent exceeding i10% 90NE*-G-0075, Vant ige 5I1 Fuel of RCS design (2500 psia) ( Vcnyication Afethod: Verify by reviewing design requirements, Surveillance Procedures ar.1 test resuits. ,7 _7-- n,e.. n,_.-,. m, w w 6 - L - ./r ua = - - th 48J -...-,...w.- o 1:strumentation - Differential Temperature

1. 83: Response time 7.0 seconds. Trip 90NE*-G-0075 (pg 5-21)

Overtemperature dT Sensors setpoint Figure 15.0-1 and 15.0-1 A 90NE'-G-0075 (pg 5-127) Venyicasion AferAod: Verify by reviewing Technical Specification requirements Survei!!ance and Calibration Procedures and test results. I Page 20 of 73

Chapter 15 Accident Mitigating Systems [,"*',,'fi"" sevalon: 81 Cff&97 ICAVP (Affected Accidents) AFFECTEDACCIDENTS: Turbine Trip (FSAR 15.2.3) COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMMION SAFETY ANALYSIS REFERENCES Instrumentation - Pressurizer Pressurizer Pressure

1. 84: 2425 psia, delay time 2 seconds (Table 90NE*-G-0075 (pg 5-21)

Pressure Sensors 15.2-1) Veripcation Method: Verify by seviewing Technical Specircation requirements, Survei!!ance and Calibration Procedures and test results. Instrumentation-St Actuation Reactor Trip

1. 82: Reactor trip occurs within 0.6 seconds 90NE*-G-0075 (pg 5-21)

Logic of turbine trip at >= 50% power Venylcasion Merhod: Verify by reviewing Technical Specifcation requirernents, Surveillance and Calibration Procedures and test results. Instrumentation - Steam Steam Generator Water Level

1. 85: Low Low reactor trip at 10% of 90NE*-G-0075 (pg 5-21)

Generator Level Sensors narrow range span, delay time 2 seconds Ver/pcarlon Method: Verify by reviewing Technical Specification requirernents, Surveil:ance and Cal;bration Procedures and test results. Page 21 of73

Raisioa of Chanter 15 Accident Mitioatine Systems es a v Milstone ttai 3 s 9/l&97 ICAVP (Affected Accidents) AFFECTEDACCIDENTS: ' Inadvertent Closure of Main Steam Isolation Valves (FSAR 15.2.4) COMPONENT DESCRif I'lON PARAMETER DESCRil'flON INPUT ASSUMI'flON SAFETY ANALYSIS REFERENCES .,.. 7.; ~ 7 :,~;p ;.7=== = - '=n77, Q '..=r.,,_,. y. ~"~ w ye~~ m r --- ->~ m~r=: m yn w ~r1:;-ng</V3'5;*. -~ s . :. r < - lastrumentation-SI Actuation Turbine Trip 487: Main steam isolation valve closure FSAR Section 15.2A Logic generates a turbine trip Verification Method: Venfy by reviewing Technical Specification requirernents, Sur/eillance and Calibratbn Procedures and test results,. l Page 22 of 78

" " * ~ ' ' " ' " " Chapter 15 Accident Mitigatina Svstems zam-of 6 J sidssese I mit J 9/1L97 ICAVP (Affected Accidents) AFFECTEDACCIDENTS: Loss oflioranal Feedwater Flow (FSAR 15.2.7) CGMPONENT DESCRIPTION PAHAMETER DESCRIPTION ANPUT ASSUMPTION SAfrETY ANALYSIS REFERENCES bEP)ibblebbNNNb Auxiliary Feedaster I%mps Sprem Flow

1. 93: Minimum syst-m flow.nith single FSAR Section 10A9.I, page 10.4-82 active failure is 510 gpm to four steam FSAR Section 15.2.6.1, page 15.2-10 generators, N loop operation NEU-96-623 90NE*-G-0075, Vantage 511 Fuel Ver.71carios Mahod: Venfy by revewing Techrucal Sphnnn requirements. Survedlance Procedures and test results.

Auxiliary Feedwater Pumps Tirw to Full Flor a91: 60 seconds from low-low steam FSAR Section 10A91. page 10.4-47 generator icvel trip, including diesel FSAR Section 15.2.6.1, page 15.2-10 loading on loss ofotTsite power for 90NE*-G-0075 Vantage 511 Fuct motor dnven pump NEU-96-623 Verrficarias Mahod: Venfy by reviewing design requirements, Survedlance Procedures and test results. Venfcaton of DG performance will be accomphshed by a review of servedlance procedures and test results for tre DG and ESF load sequencers. Page23 of73 L

s a sio=: 81 Chanter 15 Accident Mitinatina Svstems a o 6 J Maheese li is 3 a CTIO97 ICAVP(Affected Accidents) ATFECTEDACCIDEATS: Less of Normal Feedwater Flow (FSAR 15.2.7) COMPONENT DEKRIPTION PARAMETER DESCR11"IION INPUT ASSUM1"IlON SAFETY ANALYSIS REFERENCES Auxiliary Feedwater Pumps Time to Full Flow

1. 92: 70 seconds from low-low steam FSAR Section 10.4.9.3, page 10.4-47 generator level trip for turbine driven NEU-96-623 Pump 90NE*-G-0075 Vantage 51i Fuel s'erryication Mer&od-Venfy by reviewing desgn requirements. Survedlance Procedures and test results.

.,y an,pg:H & & m; h a enn:m c:7,. n., n m

~ w

.= -'. g3 , 1. '- m N -- '. y ;-7

• e M.. ~.1 i A,,,. L,.. ; c8, '

,m-- -e..

• ? I.

1.*.. ! t l l Instramash - Steam Steam Generatar Water Level

1. 209 Low Low level at 10*& ofnarrow FSAR Table 15.2-1 1

Generator Level Sensors range span,60 s delay for AFW flow NEU-96-623 l to steam generator renficados MerAod: Venfy by reviewing Technical Specification requirements. Survei!!ance and Cafibration procedures and test resuIts. ,.7 .._.,--7_.. . y_ 4 ,., u.. _,~~~ - .u.. ... ~... ..c -.. -.. a -: - Rrnm C-w Safety Va!>es Mass Flow Rate

1. 88: The total capacity of the relief valves FSAR Section 15.2.2.1, page 15.2-2 I

are sized to pass 105% ofsteam flow NEU-96-623 at rated power without exceed 4ng 90NE*-G-0075, Vantage 511 Fuct l 110*6 of steam system design press.are (1320 psia) Verryicarian MerAod: Venfy by reviewing design requirements. Tectnca! Specification requirements, Surveillance Procedures and test resuits Page 24 of 73

Chapter 15 Accident Mitigating Systems Kahia=: 81 Sidstone t'ai 3 CflOD7 ICAVP(Affected Accidents) AffECTEDACCIDENTS: Imss of Normal Feedwater Flow (FSAR 15.2.7) COMPONENT DESCRIPTION PARAMETER DESCRIPIION INPUT ASSUMFTION SAFETY ANALYSIS REFERESCES --_.-,_y.,m___.._-.----___,m,_c~,--,~,., - =, xx-ewrc-mO.nh:1wn t s m_. _ m_. Pressurizer Safety Valves, Mass Flow Rate

1. 90. The total capacity of the safety vahes FSAR Section 1523.2, page 15.2-7 1

3RCS*SV8010A,B,C are sized to prevent exceeding i10*k 90NE*-G-0075, Vantage 511 Fuel ofreactor coolant system design pressure (2500 psia) Veryication AferAod: Venfy by reviewing design requirements, Survedlance Procedures and test results. R~ tar Coolant Pumps Flow Rate

1. 94: Manual termination of flow FSAR Section 15.2.7.2, page 15.2-14 NEU-96-623 VctrJIcstion Aferked: Venfy by reviewing operating procedures.

l l Reactor Coolant Pumps Time Pump Cmodawn Begins

1. 230 For accident concurrent with a LOOP, NEU-96-623 I

pump cn=udawn begins at < = 63 s. Veryication Aferked: Venfy by reviewing surved:ance procedures Ir$$((I 5$$id Instrumentatum - Steam Steam Generator Water Level

1. 89: Low Low reactor trip at 10*& of NEU-96-623 FSAR I Ale 15.0-4 Generator Level Sensors narrow range span,two seconds delay setpoint to rod drop VertjIcsrion AferAod: Venfy by reviewing Technical Specification requirements, Surveillance and Cal.bration Procedures and test results Page25 of 78

Chapter 15 Accident Mitigating Systems [,";[",'jl"

• < eiaa: 81 C11D97 ICAVP (Affected Accidents)

AffECTEDACCIDENT& Feedwater System Pipe Break (FSAR 15.2.8) COMPONENT DESCR11" TION PARA.YETER DESCRil" TION P.NPUT ASSUMI" TION SAFETY ANALYSIS REFERENCES E AIeIw M 5 EhE 5 d N 8 E 6 E M Auxihary feedwater Pumps System Flow

1. 102 Minimum system flow with single FSAR Section l'O.4.9.f,Vdg810:442 active failure is 470 gpm to three FSAR Section 15.2.8.2, page 15.2-18 steam generators, N loop operation, NEU-96-623 300 gpm to two steam generators, N-1 loop operation Verrfication MerAod: Venfy by reviewing design requirements. Techruca: Specificaton requirements, Survei!!ance Procedures and test results.

l l Auxiliary Feedwater Pumps Time to Full Flow

1. ICi 68.5 seconds for N-loop and 69.1 for FSAR Tab!c 15.2-1 N-1 loop from low-low steam NEU-96-623 generator level trip, including dicsci loading on loss ofoffsite power for motor driven pumps Verr]Icarica Metho.f Venfy by revewing design requirements Techrucal Specification requirements, Surveillance Procedurcs and test results. Venfcabon of DG performance will be accomplished by a review of survei!!ance procedures and test results for the DG and ESF load sequencer Cavitating Venturiin Feedline, Maximum Flow Rate to Faulted Line
2. 104 39lbm/sec FSAR Section 15.2.8.2, page 15.2-19 3FDA*CAV60A, B, C, D FSAR Table 6.2-59 Verification Merked: Venfy by reviewing design and testing requirements for the venturi Page26 of78

.=.- Chapter 15 Accident Mitigating Systems seeA>n: of Milssene Omit 3 maa7 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Feedwater Systema Pipe Break (FSAR 15.2.8) COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMI TION SA. ETY ANALYSIS REl ERENCES Demineralued Water Storage Available Capacity

1. 103 SutTicient capacity to maintain 10 FSAR Section 10.4.9.1, page 10.4-42 Tank,3FWA*TKI hours of hot standby and 6 hours to cool down to 350 F l'errycation MerAod: Venf/ by reviewing design calculations. Techrucal Specification and Survedlance Procedures for the tank, and operating procedures meted with auxiliary feedwater supply.

,v.,,~~,_,.__ n n _,. m,, n,, _., n.. . am _ Jf*W ~- ' ' ' ' ' ~ "1 1'. '"'J9,'h . ! ' ~. Y T ~ L ;, ? M_ '. 21 1 lastrumentarm - Low Low Pressuruer Pressure

1. 107 SI signa! generated @ I860 psia,2 90NE*-G-0075 Vantage 51II'uct Pressuruer Pressure senws second delay before rod drop 5"crrycation Merked: Venfy by reviewing Techrucal SpWtm requirernents, Surveillance and Cahbration Procedures, and test results.

i l l Instrn=-nnvina hm Steam Generator Water Level

1. 210 low Low level at 10*L ofnarrow FSAR Tab:e 15.2-1 I

l Generator Level Sensors range span,62 s delay for AFW flow NEU-96-623 to steam generator Verrycarian Method: Venfy by reviewing Techrucal Specificaton requirements, Survei!!ance and Cahbration procedures and test results. Instrumentation - Steam Line Low Compensated tre== Line Pressure

1. 106 Safety injectmn and steam line FSAR Table 15.2-1 I

Pressure Sensors isolation signal at 658.6 psFg,2 90NE*-G-0075, Vantage 511 Fuel seconds delay Verrfication Merked: Venfy by reviewing Techrucal Specification requirements, Survei!!ance and Cahbraton Procedures, and test results. Page 27of73

Red *=: of Chanter 15 Accident Mitioatine Svstems s-e o J Milssene Itais 3 Cf13,97 ICAVP(Affected Accidents) AFFECTEDACCIDEA75: Feedwater System Pipe Break (FSAR 15.2.8) COMPONENT DESCR11410N PARAME'l ER DESCRIl410N INPUT ASSUMl" TION SAFETY ANALYSIS REFERENCES h Ci d FrfinE M MeV issi *~"" '" W HlGsei4 MIM15Glh553 1 solation Valves. Closure Time

1. 185 7 seconds on receipt of steam line FSAR Table 15.2-1 3FWS*CTV41 A, B, C, D isolation signal 90NE*-G-0075, V2.ntage 511 Fuel l'errycarios AferAod: Venfy by reviewing techncal sfebtsq requirements, survet!!ance tests and test results.

~ ""U'"5."* 7 7_ "._"'77_T,."' %.Ii3 ? MJ g. R 3, ?,T6 - 6 { g f E "., 7a :,iUi G J 2 M... ,.--nI.' uil.... .. 9. _} Safety injection Pumps Flow Capacity

1. 109 Adequate borated water is provided to FSAR Section 15.2.8.2, page 15.2-20 Leep the core covered 90NE*-G-0075, Vantage 511 Fuel l'erificados Aferked-Venfy by reviewing design requirements, Techncal Specification requirements, Surves!!ance Procedures y,

and test results. .e > g bm Q %y sjecaon Pumps Pump Flow

1. I10 Manually shut down SI pump prior to FSAR Section 15.2.8.2, page 15.2-19 g,

h.- solid pressuruer condition mQ ~4 5'erificarica AferAod: Venfy that operat:ng procedures identify parameters to be monitored and steps required to manually secure St. Safety Injection Pumps Staning Time

1. 103 2 seconds from SI signal with offsite FSAR Tabic 15.2-1 Power 90NE*-G-0075, Vantage 511 Fuel 17 seconds from SI signal wath loss of offsite power 1

5'erljicados AferAod: Venfy by reviewing design requirements. Technical SPhtinn requirements. Survei!!ance F tocedures and test results Venficanon of DG performance wi!! be accomplisheo by a review of survet!!ance procedures and test results for the DG and ESF load sequencers. ? Page 23 of73

-s .._.--l?'"'" Chapter 15 Accident Mitigating Systems scua,=: 81 Matssene t fais 3 t/lG97 ICAVP(Affected Accidents) AFFECTEDACCIDEYTI: FeedwaterSystem Pipe Break (FSAR 15.2.8) COMPONENT DESCRil" TION PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES 'aversaa~rwecomTww-manias'evsans eversas sasee'maac Etain Sicam Iso!.uion V.dses Cknure Time

1. 96: 12 seconds from time setpoint is FSAR Tabic 15.2-1 3 MSS-CTV27A, B, C, D reached renficarian Meraad: Venty by revemng Techncal Spectfcation requirements, Survedlance Procedures and test results.

s 5: The total capacity of the relief valves FSAR Section 15.22.I. page 152-2 I hn C-w Safety Valves hicss Flow Rate a i are sized to pass 105% of steam flow 90NE*-G-0075 Vantage 511 Fuel l at rated power without exceedmg I 10* & of steam sysicm design pressure (1320 psia) VerJIcarios Method-Venfy by reviewing design requirements. Techrucal Spenhtinq regt:frements, Survei!!ance Procedures and test r*<ntk 7,17~,'p;., 4.71...; gi ; y,,._. q. ~, s',,. 7

+. w -

c .p; .,n x... I knfarum Valses isola: ion Time

1. 105 30 minut.-s to isolate reactor coolant FSAR Section 15.2.8.2, p.sge 15.2-19 1

loop to faulted steam generator FSAR Table 6.2-59 rerijkarian Method: Venfy by rev; ewing operating procedures to assure procedures are in place to isolate the faulted steam generator when a feedwater line break occurs Page29 of73

e Chapter 15 Accident Mitigating Systems 2-61a=- 81 Mahtome l'ait 3 g, g ICAVP(Affected Accidents) AFFECTEDACCJDENTS: Feedwater Syster.: Pipe Break (FSAR 15.2.8) COMPONENT DESCR!!"1 ION PARAMETER DESCRIPIION INPUT ASSUMYI!ON SAI ETY ANALYSIS REFERENCES PressurizerSafety Vahes, Mass Flow Rate

1. 98: The total capacity of the safety vahes FSAR Section 15.2.32, page 152-7 1

3~~CS*SV8010A,B,C are sized to prevent exemu g !10*6 90NE*-G-0075, Vantage 51i Fuel n ofreact= coolant system design pressure (2500 psia) Verifension Afa&d Venfy by reviewmg design requirements, Techrucal Sp*tm requirements, Survei:!ance and Cahoration Prw wes, and test rest.its-Reactor Coolant Pumps Flow Rate

1. 100 Af anual termination of flow FSAR Section 152.82, page 15.2-19 Versfication Aferaad: Venfy by rewewing operating procedures for rnanual shutdown of reactor coolant pumps.

Reactor Coolant Pumps flow Rate

1. 99-Sufficient natural circulation through FSAR Section 152.82, page 152-19 coolant loop to remove decay heat as NEU-96-623 analyzed in Table 152-2 rerrficario. Aferme Venfy by reviewing startup tests.

-m _ xx.v q q 3 3.,.yggyg g h - Am emm me,m Generator Water Level

1. 97: Low Low reactor trip at 0*& of narrow 90NE*-G-0075 (pg 5-2I)

Generater Ixve! Sc.sars range span, two seconds delay setpoint to rod drop Verr 7xation Afe:Aad-Venfy by reviewing Techrucal S,*tm iequirements, Catitxr:. tion and Surveit!ance Test Procedures and test rest!ts. Page 30 of 73

Chapter 15 Accident Mitigatina Systems "*",s*is 3 2e e a: 81 6 3fdssene I C112.97 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Deeresse in Remeter Cootaat System Flow Rate (FSAR 153) COMPOSENT DESCRIPIlON PARAMETER DESCHJPI'lON ISPtJT ASSUMl"IlON SAFETY ANALYSIS REFERENCES eweraminammeesmenvanaa, amaamiewsAaneweTEa& dante _tedAACET ".~M.EiU-.1.'_. L7."'IF.5" 7 E."E O 7 0 E~ ~I 1 I i. ' 'I -. I l Sacam C-r~ PORV 1sobtion Closure Time

1. 29: Operator action to close valves in < or FSAR Section !$33A 1

Vahes 3MSSMOV!8A, B, C, D '20 minutes s'enpcmden uened-venfy using plant operatog procedures and if possible by simutating the conddion on the plant simulator , z, c,. ; ; ;. -.. 3. : +,,, m, n ;,,=. c : O. 7.- ~ ,-~,-u *;- m : v~~m...- m.u mar;: =ca r~ ~ a. - + - - Insr-nrh - Flow Sensors Reactor Trip Set Point (RCS Flow)

1. 27: At permissive > P8 Reactor trip at 92*e 90NE*-G-0075 (pg 5-21) ofnominal flow in 0.6 sec. Gravity NEU-96-623 FSAR Table 15.0-8 insertion ofcontrol rods per FSAR l

Tabic 153-1. 5~er Acarlos M ed ed: Verrfimwvi of system perfEance wdl be accomphshed by a review of-i) Trio setpomt in Techncal S,*6m i) review of Surved!ance Test Procedures and Sursedlance Tests as@W with the logic for P8 flow transmater logic & rod drop logic. iii) correlation betwe m reactor cootant flow and trip setpoint iv) flow delivered by RCS during startup tests and/or Surveillasce's currently performed. Instrumennrion - Flow %ws Reactor Trip Set Point (RCS Flow)

1. 26: At permissive power >P7,35*& of RCS FSAR Table 153-1 loop flow in 1.0 sec. Gravity insertion 90NE*-G-0075 (pg 5-21) ofcontrol rods per FSAR Table 153-1.

rcrification MerAod: Venfcabon of system performance wdl be accomphshed by a review of. i) Trip setpoird in Techrucal Spec fcaton h)revew of Survedlance Test Procedures and Survedlance Tests mnerded with the logic for P7 flow transmater toge & control rod logic hi)correlaton between reactor coolant flow and tnp setpoint iv) flow dehvered by RCS dunng startup tests and/or Survei!!ance's currently performed Page3I of73

Chapter 15 Accident Mitigating Systems 72",.jl" Kaha>=: of C11D97 ICAVP(Affected Accidents) AffECTEDACCJDEVTS: Decrease is Reactor Coolant system Flow Rate (FSAR 15.3) COMPONENT DESCR11'IlON PARAMETER DESCR!!' TION INPUT ASSUMPTION SAFET Y ANALYSIS REFERENCES laste-mL= - Speed Sensors Pusup Speed

1. 23: Reactor Coolant Pump Undenpeed 90NE*-G-0075 (pg 5-59) trip ser point l'arfication Sterhod: Venfcaton of system performance wd! be accompftshed by A review of-1)Trp setpcant in Techracal Specificaten ll) Review of Survedlance Test Procedures and Survedlance Tests associated with the logic for >P7 &

underspeed sensorlogic. l l l l rage 31of78

^"*""'* " Chapter 15 Accident Mitigating Systems K a hlen: or Stsissene I ess 3 cuoor ICAVP(Affected Accidents) AFFECTEDACCIDEN75: Deercase in Reactor Coolant System Flow Rate (FSAR 153) COMl'ONENT DESCRIFI!ON PARAMETER DESCRIFTION INPUT ASSUMFilON SAFETY ANAL \\ SIS REFERENCES 3c 4 A;;;r 3S ~y,3..n. n.~ -num ,n-nu-, Various Offste Dose (E)

1. 34: FSAR Table 153-3 FSAR Table 153-3 Vers]icasion Method: Venfy by assunng each parameter is consistent with plant design and operat:on and by reviewing the supporting dose mL'ifation to assure consistency wah FSAR Table 15.0-8 kirmainh' main 3HfiEiiiiiiiiF

~LiiTE.Esema:imiiniiIesskb'TJiWAW-WF992 9mm C-r-Feedwater Flow

1. 33: (0-2 hr) 771.277 lbm; (2-8 hr)

NEU-97-537 1,505,487 lbm for N toop and N-1 IwP FSAR Table 153-3 Verification Merhod: Venfy through Survedlance Test Procedures and Survedlance Tests or system flow calculat.ons ' m *siEEriifris N IVT T1"MEM4llB""CO'S?Ji??WTO% :s 9mm Generators !nitia!Inven:ory

1. 31: Liquid - 97,660 lbmSG; Steam -

NEU-97-537 I 8,301 lbm/SG - N loop FSAR Table 153-3 LiquiJ - 99,717 !!mi/SG, Steam - 7,731 lbm/SG - N-1 loop Verification Method- *fenfy through review of the verxior parameters and associated accxient dose calculations. Qnm C-rus I rak Rate

1. 30: I gpm for N loop and N-1 loop NEU-97-537 (Pnmary to M=dwy)

FSAR Table 153-3 Ver:fL -- Medad: V% through Te3nscal Specification, Surveinance Test Procedures and Surveillance Tests Page 33 of73

mawa-81 Chanter 15 Accident Mitiaatina Svstems r 6 h s unsa, arai 3 911597 ICAVP(Affected Accidents) AffECTEDACCIDEVTS: Decrease in Reactor Coolant System Flow Este (FSAR 15.3) COMPONENT DESCR!YTION PARAMETER DESCRIFIION INPUT A550MFTION SAFEIY ANALYSIS HEFERENCES

,.. _ _ ~ _

7, yy ...-~3 .v. m,. N'A Primary Coolant Inventory

1. 32: 520,000 lbm for N loop and 350,000 NEU-97-537 I

Ibm for N-1 loop FSAR Table 15.3-3 Venycation Aferkod: Venfy through revew of the vendor parameters and associated accident dose Oh M uvis. I l l l Page 34 of 73

sernion-81 Chanter 15 Accident Mitigatina Svstems a o 6 J Whtoac Itait 3 010D7 ICAVP(Affected Accidents) AffECTEDACCIDENTS: Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Soberitical or Low Power Startup Condition (FSAR 15.4.1) COMPONENT Dt.SCR11"IlON PARAMETER DESCR11'110N INPUT ASSUMl"f!ON SAFETY ANALYSIS REFERENCES De N dacB.355 ~s h 'w.rn-+~ M M m W L-, s Reactor Coolant Pumps Number

1. 115 3 re::cta coolant pumps in mode 3 FSAR Section 15A.I.2.8, page 15.4-4 operation NEU-96-623 VerificationMe Aod: Venfcation of the number of pumps in operation in Mode 3 will be accomplished by reviewing operating procedures and Technical Spebrimis to assure consisterey with the assumption.

l Varicus Coolant Temperature

1. 114 Themaximumaveragecoolant FSAR Section 15A.I.2.3, page 15 A-3 temperature is $57*F NEU-96-623 Verrficcrion Merhod: Venfication of the average temperature can be accomphshed by revicw of the operating procedures for zero power operation and Techncal Spehtint f

MEMUQ4EUN Instn-nr6 - Neutron Flux Neutron Flux

1. I12 Low se: point for reactor trip at 35% of NEU-96-623 FSAR Table !$ 0-4 Senscrs full power 90NE*-G-0075 Table 5.1.1-4 Vers]icarian Meraad: Venficaton of the reactor tnp setpoint will be accomphshed by review of Techncal Specifications.

Calibration and Surveillance Test Procedures and Cahbration and Survei!!ance test results. i instrumentation - Neutron Flux Neutron Flux Rate

1. 113 0.5 second delay time for trip signal NEU-96-623 FSAR Tabic s 5.0-4 Sensors actuation and RCCA release 90NE*-G-0075 Table 5.1.1-4 Ver:JIcarion M eraad: Venfication of the reactor inp delay tirr.s will be accomplished by review of Technical Specircations, Cahbraton and Surve.!!ance Test Piocedures and Cahbration and Surveillance test results.

Page 35 of78

...w:. x - - - ',",,**"','],""" l Chapter 15 Accident Mitigating Systems 2 - hio=r of C11097 ICAVP(Affected Accidents) AFFECTEDACCIDENTSr Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (FSAR 15.4.2) COMPONENT DESCRil"I!ON PARAMETE!* DESCRIPTION INPUT ASSUMl"110N SAFETY ANALYSIS REFERENCES T U 2 7 2 O 7 0 0 '5 7 0 Instrumentation - Neutron Flux Delay Time

1. 117 0.5 second delay time for trip signal NEU-96-623 FSAR Table 15.0-4 Sensors actuation and RCCA release for high 90NE*-G-0075 Table 5.I.1-4 neutron flux (high setting)(N and N-1 loop operation) rcrryIcation Merkob Venfcation of the reactor trip delay time will be accomplished by review of Technx.at Specifications.

Cahbration and Survei!!ance Test Procedures and Survei!Iance and Cahbrat;on Test results. l Instrumentation - Neutron Flux Neutron Flax

1. 116 Reactor trip at 1I8*6 of full power -

NEU-96-623 FSAR Table 15.0-4 l Sensors four imp operation 90NE*-G-0075 Table 5.1.1-4 l Reactor trip at 89*& of full power - I three loop operation Verificarios Merked Venficaton o' the reac. tor trip setpoint performance wi!! be accomplished by review of Technical Spectfcat.ons, Cahbration and Survei!!ance Test Procedures and Survet!!ance and Cahbrata,n Test results. Instrum,nrwinn. Differentir' Temperature

1. 212 Trip serpoint given in Figures 15.0-1 NEU-96-623 I

Overtemperature JT Sensors and 15.0-I A FSAR Section 15A 6 rcrryIcation Merkod-Verify that the operating procedures list the appropriate monitors and that there are proper pWures for alarm response. l Page 36 of 78

E S Chapter 15 Accident Miti atinE,ystems serai =: of .ua e.u cawr ICAVP(Affected Accidents) AffECTEDACCIDENTS: Startup of as lanctive Reactor Cootaat Pump at sa Incorrect femperature and Boron Concentratica (FSAR 15.4.4) COMPONENT DESCR11' IRON PARAMETER DESCRIPTION INPUT AS5UMFTION SAFETY ANALYSIS REFERENCES L " L ~ A M M C X 3 7 B M 3 7 E i*:: W C T 1 Isolation Valves Concentration

1. 14i Boron concentration of the isolated NEU-96-623 Section 15.4.1, page 15.4-loop is greater than or equal to the 15 boron concentration of the operating loops, or greater :: an 2600 ppm whichever is less.

Ferrfication d'.. %d: Review Techrucal S,WWs, Surved:ance Procedures eM Survei!!ance Tests for isolated and operaung reactor coolantloops. l l Isolation Valves Temperature

1. 140 Temperature of cold feg ofinactive NEU-%-623 Section 15.4.4, page 15.4-loop is within 20*F of the highest cold 15 leg temperature of the operating loops.

Verifcarian Method: Review Techrucal S,*tinos, Surveillance Procedures and Surveillance Tests for isolated and operating reactor coolant loops. l l l Page37 of 73

Chapter 15 Accident Mitigating Systems

"*"",'j,""

sowon: of C11LO7 ICAVP (Affected Accidents) AffEs~TEDACCIDDTS: Chemical and Volume Control System Malfunction that Results in a Decrcza in Boron Cone:ntration in the Rescior Coolant (FSAR 15.4.6) COMPONENT DESCRIMION PARAMETER DESCRIMION INPUT ASSUMI" TION SAFETY ANALYalS HEFERENCES [ % % 7 N M P *" W *[ Q [ Q f y { Q T Q ] - { Q ]. ] Boric AcidStorageTank Concentration

1. 137 Boran concentration >= 6600 ppm NEU-96-623 FSAR Section 15A.6 rcrryc=rion Method: Venfy that operating procedures sha!! provide instruction within CVCS configuration fcr bora'sd water s "

source l l Flow Path Valves Concentration

1. 133 Termination ofdilution when NEU-96-623 detected. Requires operator action FSAR Section 15A.6 renycation Method: Venfy tr:at operating procedures shall provide diluton path isolation instruction w: thin CVCS (LCV-1128 ard C) and configuration for borated water source (LCV-112D and E)

Flow Path Valves Concentration

1. 134 Inject boron when dilution has been NEU-96-623 terminated. Requires operator action FSAR Section 15A.6 Veryication Merked: Venfy that operaticg hures shall provide dduton path isolation instruction within CVCS (LCV-1128 and C) and cocinguraten for horated water source (LCV-112D and E) i Ibw Rate Control Valves F.'ow
2. 136 Operator action to limit flow to <= 150 NEU-96-623 gpm using valve V305 FSAR Section 15A.6 Venfication Mded: Venfy that operating procedures sha!I provide flow lirnet of 150 gprn and that administrative control limit exists.

Page 33 of 73

e Chapter 15 Accident Mitigating Systems Z"'j;'" seezaa: of C11097 ~~ ICAVP (Affected Accidents) AffECTED ACCIDEVTI.- Chemical and Volume ControlSystem Malfunction that Results in a Decrease 'n Boron Concentration in the Reactor Coolant (FSAR 15.4.6) COMPONENT DESCRIFTION PARAMETER DESCR11' TION INPUT ASSUMFTION SAFETY ANALYSIS REFERENCES lastro=~nt'= - Blended Flow Desistion Alarm

1. 121 >10*ideviation requires NEU-96-623 1

Rate Operator action FSAR Section 15.4.6 VerrTration Merhod: Venfy that the operating procedures list the appropnate morators and that there are proper procedures for alarm response lastrumentation - Blended Flow Flow Indication

1. 120 Abnormal mdication requires NEU-96-623 Rate Operator actim FSAR Section 15.4.6 Verr 7xation MerAod: Venfy that the operating procedures list the appropriate morutors and that there are proper procedures for l

alarm response Instrumentation - Boric Acid Deviation Alarm

1. 119 >10*ideviation requires NEU-96-623 Flow R2:e Operator action FSAR Section !$.4 6 Verrficarios Mer&od: Venfy that the operahng procedures list the appropnate morutors an1 that there are proper procedures for alarm response Instrumentation - Boric Acid Flow Indacaticn
2. 118 Abnormalindication requires NEU-96-623 Flow Ra e Operator action FSAR Section 15.4.6 Vers 7earios McA%od: Venfy that the operahng procedures Itst the appropnate morutors and tilat there are proper p rocedures for alarm response lastri-nr'= - Status Boric Acid Pump Status
3. 123 R. quires operator action to monitor NEU-90-623 Indatmg 1.ights stamt lights to detect dilution events FSAR Section 15.4.6 Verifraaien MerAad: Venfy that the operating procedures list the appropnate morutors and that there are proper p.hures for alarm response Page39 of 73

/

5 - ' -, :"':"'--^---' Chapter 15 Accideat Mitigating Systems C",'j,"" x< w oa: of crc 7 ICAVP(Affected Accidents) AffECTEDACCISEVII: Chemical and Volume Control System Malfunction that Results in a Decrease la Boron Concentration in the Reactor Coolant (FSAR 1S.4.6) COMPONENT DESCRIPTION PARAMETER DESCRil" TION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES lastru:nentation - Status Charging Pump tr,ns

1. 122 Requires operator action to monitor NEU-96-623 lainring Lights

<rann lights to detect dilution events FSAR Section 15.4.6 Veri /U=.h Aferhod: Venfy that the operatmg procedures hst the appropnate rnorutors and that there are proper procedures for a: arm response Instru.nnrition Krina Pnmary Water Pump Status

1. 124 Requires aperator action to monitor NEU-96-623 l

tminring Lights crarm lights to detect dilution events FSAR Section 15.4.6 Verryctrim Jferked: Venfy that the operating procedures hst the appropriate morutors and that there are proper procedures for alarm response. Rcfwling Water S:orage Tank C - arr" 6

1. 135 Operator action required to term' ate FSAR Section 15.4.6 1

m dilution NEU-96-623 VerifIcados Aferked-Venfy that operatog procedures sha!! provide dilution path isolation instruction within CVCS (LCV-1128 and C) and configuration for borated water source (LCV-1120 and E) n-7 7, Ek. 4' - 49.'./ utLJ t 'S m Y - Inse=~rita - Shutdown Shutdown Atargm

1. 132 Alarm at setpoint 2.0 which requires NEU-96-623 Margin Atonitors operator action rSAR Section 15.4.6 Verficados Afaked: Venfy that the operating procedures list the appropriate monitors and that there are proper procedures for alarm response Page 40 of78

Cinapter 15 Accident Mitigating Systems seelon: si Sidssaa t mit 3 01007 ICAVP(Affected Accidents) AffECTEDACCIDENTS: Che:sical and Volume Come.rol System Malfmaction that Results in a Decrease ia Boros Concentration in the Rea-tor Cooiant (FSAR 15.4.6) COMPONENT DESCR11410N PARAMETER DESCRII"1 ION INPUT ASSUM1410N SAFETY ANALYSIS REFERENCES Imtrumentarinn -Source Range Neutron Flux

1. 125 Visual ian'r= requiring cperator NEU-96-623 Tha action FSAR Section 15A.6.

8' esp are= JiaAoJ: Venfy that the operatng procedures 1:st the appropriate mondors and that there are proper procedures for alarm resporse lastr=~r'r= - Source Range Neutron Flux

1. 126 Audible count rate of flux requiring NEU-96-623 Flux operator e FSAR Section 15A.6 ;

- m 5 r l'erryicarian Makod: Verty ' hat the operatng procedures hst the appropnate rnorntors and that there are proper procedures for al arm resporse l Instn-r-r= - Source Range Neutron Flux

1. 127 Alarra requmng operator action NEU-9MC!3 ELx FSAR Section 15A.6 Verr]Icarios Maked: Venfy that the operatng procedures list the appropriate morutors and that there are proper procedures for alarm response

- w < < ~ ~ 6%- 7 -a /^ ww-n - rms - ....m..;, 7.;.... v1 . T - ,, f.,% ...m I Refueling Water Storage Tank Canerntranon

1. 187 Baron concentranon > 2000 ppm FSAR Section 15A.6 l

rcrryicarios Mahad: Vanfy that operatng procedures shait provide instrucbon within CVCS confguration for borated water source Page 41 y 73

Chapter 15 Accident Mitigating Systems C"'j"" se>*i=: ei C71Ls7 ICAVP(Affected Accidents) Ches6 cal aul Vcluse Control S stem Malfunction that Results in a Decrease in Bos on Concentration in the Reactor Coolant (FSAR AffECTED ACCJ9ENTS.- 3 15A.6) COMPONENT DESCRIPTION PAnddMETER DE5CRiPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCE 5 G,,* 3 - _a fe[,. L* : ; e,jf,,'aj'"_n/sGsez 7 je.~ ~a C - ~ - QW e f,M; :: & ";. ; '. (,}}Y~;QgKyG Instr--h - Ne aron Fin Axiat Thi%fTerence

1. 130 Alarm requires operator action NEU-95-623 Sensors FSAR Section 15A 6 l'ers/he Medsd: Verdy that the operating procedures Est the appropnate monitors and that there are proper procedures for a! arm rescortse lastrument= ion - heutrun Fgur Wurron Flux
2. 129 !!igh setpoint i18*.of rated power NEU-96-623 3ensors for N loop; 89*& ofrsted power for N-FSAR Section 15A.6 IINP rersficades Mcded: Venfy that the operating procedures hst the appropnate morutors and that there are proper procedures for alarm response.

Instrumentadca - DatTerenual Temperature

1. *28 Trip serpoint gisen in Figures 15.0-1 NEU-96-623 and 15.0-I A FSAk Section 15A 6 Ovenemperature dT Sensors Verificarian Medad: Venfy that the operating procedures hst the appropnate morutors and that there are proper p w:edures for alarm response

\\ i Instr-re - DitTerential Temperature

1. 147 Operator action upon receipt of turbine NEU-96-023 Overtemperature dT vmars runback als FSAR Section 15A.6 Verificados Mamad-Verty that the operating preres kst the appropnate rnonttors and tnat there are proper p ocedures for alarm response Page 42 of73

Chapter 15 Accident Mitigating Systems 7;",.' "','""_ s a n a: as C11097 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Chemical and Volume Control System Malfunction that Results in a Deercase in Boran Concentration in the Reactor Coolant (FSAR 15.4.6) COMPONENT DESCRIFTION PARAMETER DESCR1FTION INPUT ASSUMFTION sal ETY ANALYSIS REFERENCES laste-~*'rL= - Differential Temperature

1. 195 Alarm requires operator acuon NEU-96-623 FSAR Section 15.4.6 Osertempera:ure dT Sensors Venfication MerAod: Venfy mat me operating procedures hst me appropriate rnorutors and that there are proper procedures for l

alarrn response l l 51EEN~53} instr-=~tL= - Control Rod Control Rod Position

1. 131 Operator instructions are adequate to NEU-96-623 Pmm Sensors respond to indication FSAR Section 15A.6 i

renficazion Merked: Venfy that the operating procedures list the appropriate rnotutors and that there are proper procedures for alarm response lastr-riv - Control Rod Control Rod Posstwn

1. 194 Low alarm requires operator action NEU-96-623 FSAR Section 15A.6 Position Sensors Vers)7cados Me:Aod: Venfy that the operating procedures list the appropriate rnonitors and that there are proper procedures for alarm response 1astrumennrim - Control Rod Control Rod Position
2. 193 low Low alarm requires operator NEU-96-623 Pmvem Sensors acte n FSAR Section 15A.6 Ven]Ication Methodt Venty that the operatng procedures hst me appropnate rnorutors arW that there are proper procedures for alarm response-I Page 43 of78 m

2 * - 81 Chapter 15 Accident Mitigating Systems Mitsteme Uni 3 g 7 ICAVP(Affected Accidents) AFFECTEDACC1 DENTS: laadsertent I eding and Operation of a FucI Assembly in sa improper Position (FSAR 15.4.7) COMPONENT DESCR1PIlON PARAMETER DESCRIPTION INPUT ASSUMPI'lON SAFETY ANALYSIS REFERENCES i

• seinaminaissrsmiiiirarsaTsEminanimms Kiru:n'cansifs=iisiis" h h M Tr m L ZETr77T'r m --

m Fuel Core I nang

1. 142 !mplementation of quality control NEU-96-623 Section 15.4.7, page 15.4-during core loading 23 1

5'erryicarian Method-Venf correct fuel loading by remw of procedures implernented during core loading. f i i i P.zge 44 of78

maisio=: 81 Chanter 15 Accident Mitioatine Svstems I' o o J Aidstaae t! ass 3 Cf!L97 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Spectruse of Rod Claster Control Aswa.bly Ejection Accidents (FSAR 15.4.8) COMPONEN I DESCRIVIION PARAMETER DESCR1PIlON INPUT ASSUMFI1ON SAFETY ANALYSIS REFERENCES 'is,'. A.-.,. J. J, f _,... ci ~..: 7.~,.^,. TUR 1, J~. 7 ^ ~ ' 1 ~ ~ U-~ ~ ' ~. ~ T '~7;;] T { WQ ~ 1as:rumentation - Low Low Pressunzer Pressure

1. 173 Provides SI signal within one minute FSAR Section 15.4.8 Pressunzer Pressure Sensors

@ 1600 psia rcrsficadas Medad: Venfy by reviewing operabng procedures. W ' : ?.*,.***J'~.*.r~~ T, C.. "T. i, '...y ..,:.. :. w.,,.M ?. ' c ' W. ^L., . G [Y.. pp '..,. s. ,= Safety lajection Pumps Staning Time

1. 174 Flow starts within one minute sRer the FSAR Section 15.4.8 breal renJication Medad: Verdy by reviewing operahng procedures and surveil!ance tests.

7,. -. ;....:s u rn... ReliefVahes Operating Pressure

1. 178 All vahes are fully open at 1240 psia FSAR Table 15.4-6

- ~ ~ rer#icarias Medad: Venfy code test catafresults l l Page45of78

] Chapter 15 Accident Miti ating Systems Acral a: of 3 uma. _ Cf1197 l ICAVP(Affected Accidents) l AFFECTEDACCIDENTS: Specteam of Rod Cluster Control Assembly Ejection Accidents (FSAR 15A.8) COMPOSENT DESCR11' TION PARAMETER DESCRIPIlON INPUT ASSUMPTION SAFETY ANA!> 61S REFERENCES 7_ .y - ~ ,,,=.._,y____.,.,.,, s., ; vce s Con:rol Rod Merhanism floming Weld Integrity

1. 177 Penodicinspecten FSAR Section 15 4.8 Vcnycance Method: Verdy by reviewmg of procedures and inspecten data.

I 1 l -.. r-;..n.}Y; ~ 2. T :. < i,

- ~

' x + *~~s w :.* - - -.+ +ww, .[_, I Missile Shield Shield lategrity

1. 172 lastalled post refueling FSAR Section 15.4.8 rcrificaden Meded: Venfy that there is a procedure to remove and insta!!.

.;,.,..,m__g_g,g,_ ppq.q ..g n - 7y; lastn-r- '6 - Neutron Flux Neutron Flux

1. 169 Trip setpoint at high neutron flux FSAR Section 15.4.8 Sensors

!18% power (high range setpoint) renfL -saw M esk e d: Venfy by reviewing technca! spec:ficaton requirements surveillance and cahbration results Instr-m6 - Neutron Flux Neutron Flux

1. 170 liigh neutron flax 35% power Oow FSAR Secten 15.4.8 Smsors range serpoint)

Feriff-arL= Medad: Venfy by reviewing technical spwi'M -i requirements surveillance and cahbration resu!ts Page 46 of 78 'e -u+-- w- -e-- as- -e --m.-- . m

se mio-: si Chanter 15 Accident Mitientine Svstems o a J %5dweee t'ad 3 e DIO97 ICAVP(Affected Accidents) AFFECTED ACCIDENTS: Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSAR 15.4.8) COMPONENT DESCR11'flON PARAMETER DESCR!!"IION INPlJI ASSUMl"IION SAFETY ANALYSIS REFERENCES lastnerim - Neutron Flux Neutron Flux Rate

1. 171 Trip setpoint at high rate of flux FSAR Section 15.4.8 Semens Verification Method: Venty by revewing technical sphW requirements surves!!ance and cahbration results.

yygy &.e.s:%n :F =.M Refucimg Water Storage Tank Concentration

1. 175 Boron concentration x 2000 ppm FSAR Section 15.4.8 l

Versjication Merked: Venfy by revewing operating procedures and survei!!ance test. k W I N S 3 f--- !" M @ l b TI h k M N N N Y N h N NkNbilN NYf.IbMNI2-Ed lastne'L= - Control Rod r=rrol Rod Positios

1. 167 Operating instructens are adequate to FSAR Section 15.4.8 1

Posamn Sensors respond to indication limit a! ann and proside boration Verif!casion Mer&od-Venfy that the operating procedures list the appropnate morutors and that there are propr procedures for alann response l1 lastn-rarm - Control Rod Control Rod Pmmm

1. 168 Gperatag instructions are adec aate to FSAR Section 15.4.8 r

Posh hrs respond to RCCA deviation alarm and provide boration Verificaden Merhod: Venfy that the operating procedures kst the appropnate monitors and that e,ere are proper procedures for alarm response. Page 47 oj73

Chapter 15 Accident Mitigating Systems Ka* 81 sim r.u g,g7 ICAVP(Affected Accidents) AFFECTEDACCIDEVTS: Spectrum of Rod Claster Control Assembly Ejection Accidents (FSAR 15.4.8) COMPONENT DESCRIPTION PARAMETER DESCR1FTION INPUT ASSUMPTION SAFETY ANALYSIS REFEREP CES lastr"=~r*tL= - Control Rod Delay Time

1. 176 Rods begin to fallin 0.5 second after FSAR Table 15A-1 Pmn= Sensors the trip point is reached FSAR Section 15A.8.2.2, page 15A-32 l'ertycation Method: Venfy by reviewing surveinance tests.

~ l Page 48of73

Secchcasa t'g,s ics Chanter 13 Acc dent M t ent ne Svstems i ii i a n sion: of s o o J Maintame I. it 3 m CflL97 ICAVP(Affected Accidents) AffECTEDACCIDENTS: Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSAR 15.4.8) COMPONEh T DESCRi!"FION PARANtETER DESCR11'IlON INPUT ASSU3ti"IlON SAFETY ANALYh15 REtERENCE5 e-nn. i .c e ..is-t,. a os. 4.- n._ 2 _ n -. - r wa n. .-., 7. g. q,., 3 _. m Fans Flow (E)

1. 179 A secondary containment negative FSAR Section 15.4.8 pressure less than or equal to 0.25" w g is achieved in 60 seconds 8'errJJcarlos 3ferked: Venfy survedlance and test data support assurnptons. Review valgt:ty of assumpbon based on MP3 secondary containment funct:onal design calculaton.

tihers Ellici ncy(E)

1. 180 Fiher removes greater than 95*4 iodine FSAR Section 15.4.8 l

Veri /Jcation Slethod: Venfy survedlance and test data support assumptons. l 1 r,_ , -(._-. _ , ;-- y .n 3 4:; ~ .c Various Leak Rate

1. 183 The contamment leak rate is less than FSAR Section 15.4.8 or equal to 0.65?a containment total volume per day VerJJcation Afethod: Venfy test data and surveitance support asst.rnpto.s

• c.,,~,., ; s y _. r~-

^? ~ i.~~, m ' ~q _ ;

~. ~ ~ ~

.g- -w=-* } hm Generator Vanous Offsite Dose (Ell.)

1. 148 FSARTable 15.4-4,15.C-10 FSAR Section 15.4.8 I

VenJJca:Jon Atcskod: Venfy that input assiu.@4 are consistent with plant design and operaton and that dose calculat:ons are consistent with resu!ts of Table 15.0-8. Page 49 of 73 )

y seeias: Si Chanter 15 Accident Mitipatina Svstems a e 6 4 51dssone t'ast 3 O'!O97 ICAVP;Affected Accidents) AffECTEDACCIDEVII: Spectrum of Rod Claster Control Assembly Ejection Accidents (FSAR 15.4.8) COMPONENT DESCRIPTION PARAMETER DESCRil'IlON INPUT ASSUMi' TION SAFETY ANALYSIS REtERENCES n, _,,,,n.,,,_ g n,,, ,-3 .- z v.

. :p. 3. ;-~n Fans Flow (E)

1. 181 A secenbry containment negative FSAR Section 15A.8 pressure less than or equal to 0.25" wg is achieved in 60 seconds renjicadsn Afer&od: Venfy survei!!ance and test data support assumpbons. Review validity of assumption based on MP3 secondary cish urent funcbonal design calculation.

Filters Filter Ef&M (E)

1. 182 Filter removes grerer than 95% iodine FSAR Section 15.4.3 ren]Icarian Afer&od: Venfy surveit:ance and test data support assumpbons.

I Page 50 of 75

x<visLn: of Chanter 15 Accident Mitieatine Systems I Missioac tinis 3 a o o 9/l&97 ICAVP (Affected Accidents) AFFECTEDACCIDE$7S: Inadvertent Operation of Emergency Core Cooling System During Power Operation (FSAR 15.5.1) COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES I -,m, m.. .7. n... m.. m., e.,,,c...s., m...:.w - - 4.-- - m.. -,- aman .....,..,,m.p %, 4 Pressurizer Spray Supply Operating Status

1. 75: Pressurizer spray is operable FSAR Section 15.5.1.2.D Verificarlos Method: Assumed available to mitigate the transient.

Venfied by review of operating procedure, control tuning procedure and RCS Pressure Control Calibration records. 4 ~b EkMd Various Spurious Actuation

1. 74: Manual termination ofspurious St NEU-96-623 (pg 15.5-4) injection. Table 15.5-1,600 sec.

Requiring operator action Ven]Ication Method: An Abnormal Operating Procedure shot.ld be in place to mitigate this spurious function. Procedure should identify plant conditions and pressurizer parameters to be monitored for manual termination. 11 - a. - wwe~sw~;w. To ~ 7 . - =w om gg,.,, t w s. m ,3 :nr,.,.,..,,.,., m.c,..p n.e n um m PORV Selector Switch Status Auto Control

1. 77: Auto Control Selected by operator FSAR Section 15.5.1.2.0 action Ven]ication Method: Operation of PORVs assumed to result in the pressurizer pressure not reaching the PSRV set pressure (2500 psia +/- 3%, Tech Spec 3.4.2.2). Verify by confirming an operating procedure is in place and spucifies auto control for normal operation.

E Page5I of 78

o In LI PI Po n r n P C A ( 9 R s o s e s r ir s u Ow O F / t g t o l u cu u Re M F l su E 9 l m m m Vr P C s r O O T n e e e e ) 7 a n n Sn p t t e N E t a a e a s' r i u nti ^ e N A t a E D O s o o o o f t n n r n d T C s V S P R D C I r e A s ei E I a l S D e r io c s f C E u t u s u r V R N i T a z a I t i e l P S v o r e I n s l O h N Ve M V R e P V P P V e e r r r a r e r e r e A i f n ji a jl s fi s R J c i I i s i s c u c t c u c u A v a a a o a r a r e l r t r M r r i l i l zi e l z r o T o T o e o e E e t r n n ri n r n r i T n 3 p A p A P A P t f f f r r r E t t t s c s R O e e e e e p h h h m o u D e k s o o o r r r d d d e d e E a S t i C o AV V V rp A R n j er C be e e ce s I o nn ri ri +9 I os s f u T E h of i ci 3 r s f f r ci c dum I ma a a srie O m a ai i ti k e t t l o o o P on n n -{ zd N r 4~ e g I t r a e C e p b b b r v n ey y y ea c r ) r ai A P nc c i +' c al y aT T r a hb ie e e 1 e p n ni w eel C V 5 t v o e gh h e

Y wo C

(P s r g A t e r i c c o ar 5 o a al f n o c A c 2 cl 0 t o eS S T eti o rgi l i f i f d p p e sa n f d 7 u e 7 e 7 c 7 g e hc O c ci 1 ot 7 I 8 r c 2 h l e 6 N S c 3 n i e eci i f f dt P y n si i ( 5 a A s e Pp O U t h s e t t o a p a F1 Srep T e t u J* e i Sa l c Y at l o r o Ac S u t r Rs e m d t M t r A t t .e a Vsua dn a n Ru e e t p a S D A n i t ia o a Fti ci s Vi e r a s s ioi S u n r t z r c i cl n n i o i t e e t e n U d a f i u c d g n c 3 t un c o r n p r o M g i r a t it dS i S eg i p dP s s T o e u e ei. rp f a f r e r P P d eu o t i t t s e h r n o r 1 n i p v t v r r us e 5 e ,n I w n n r e be r aei o ei 0a e uP O a y v r r e 3 S c all m l t e O N r t a al 2e u e t r s rnn n n 2s r o v. en nR O ) S a vt oV p ec u c ) a v r t e e l e r i i s s e Y a e t r t l T a l e r u e r r s e wb se y l l8 o c c s t a e a a c t t s E n t t u ot r S o c 6 f o hi l o e i i r 0 o m t d p T ,e f' cl ns n b t a p oe g ( et r ri C si nd n F s i e D o.a a p S t c A al a r T i ow A b t b l a s o A. R e r r t t r a ue v t 1 e oi nr 5 if d N N n N nl F S 5 o E E E g w S A a 1 an U p A F ) n U U t r ..Y' oh R E d 9 9 d 9 r c o S T r 6 6 s 6 n ,N e u e Y o 6 6 e 6 dg c t 2 2 p 2 t uh i A ni 3 3 t o 3 r o o ( ( i ( eP n N g.- r p p n p , ~ aS A t 1 e g g g nR 5 L d d A a g.' d Y 1 1 1 V 5 5 5 5 f S o r 5 5 .a o. S c s I t 1 5 v' ai t 4 4 4 h n 2 S , h n b .si ) ) ) r c D H u *"

3. ?

aa E n .a is F t E m* oe c ti pS E o n R n q' ' N

v. "

C u "" c - A E S -~b ll \\li

W sevalon: or Chanter 15 Accident Mitioatina Systems 6 6 s Milstone l'ait 3 r 9/lL97 ICAVP (Affected Accidents) AFFECTEDACCIDENTS: Inadvertent Opening of a Pressurizer Safety or Relief Valve (FSAR 15.6.1) COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES ~ ^ ="% ' "'"s"*:,*..*; *'.i Y sw: !3,""

• w '

e e. e : > v. ~,~.h*', a' ~'"~:~L ,e<,= -,e ~ 'h. - ~ Various Concentration

1. 54: Positive moderator temperature FSAR Section 15.6.1.2, page 15.6 't _

coeflicient NEU 96-623 , n o r.. Verij7carlos Merkod: Verify that positive moderator temperature coefficient assumed in FSAR Chapter 15.6.1 is conservative with regard to entire core life. Review Technical Specification requirements and surveillance tests procedures. 1:.. y ;. s :::. i, v.. ;. ; 3. p, 7 g ;47 t. -',, f: 5 Y C/ 1 P ] F 9 % j K.T~ y"?if!! Q ((Q y y h M K M 7 ;y p y1 i Instrumentation - Delay Time

1. 144 Rod drop will begin 1.5 seconds after FSAR Table 15.6-1 Overtemperature dT Sensors setpoint is reached (N-loop operation NEU-96-615 case)

Venfication Method: Review Technical Specification and Sumeillance Testing for these instruments. 1 1 Instrumentation - Pressurizer Pressurizer Pressure

1. 145 Rod drop will begin 2.0 seconds after FSAR Table 15.6-1 Pressure Sensors setpoint is reached (N-1 loop operation NEU-96-615 case)

Verification Method: Review Technical Specification and Surveillance Testing for these instruments. Page53 of78

a s naion: si Chanter 15 Accident Mitieatino Systems a o o s Milstone Ifait 3 9/l&97 ICAVP (Affected Accidents) AFFECTEDACCIDENTS: Failure of Small Lines Carrying Primary Coolant Outside Containment (FSAR 15.6.2) COMPONENT DESCRINION PARAMETER DESCRINION INPUT ASSUMMION SAFETY ANALYSIS REFERENCES kWei:ains~rs:hi nanrusa.. cucisiiWiiriman uni'laamminaivoniTaiM-~ "'ar. '....IITJ I Letdown Line Penetrating Reicase Rate From Broken CVCS Line (E)

1. 57: CVCS letdown line break maximum FSAR Section 15.6.2, page 15.6-3 Containment flow of 152 gpm Verification Aferhod: Venfy using design drawings and calculations.

1,., J.. i f,' ffy - r '* T R i-l < '.'..M. '...'"*f WV. i '..... 4e,., a4, Containment Penetrations Line Size (E)

1. 55: No instrument lines connect to the FSAR Section 15.6.2, page 15.6-3 i

RCS that penetrate containment FSAR Section 15.0, page 15.0-1 Verification Aleshod: Review P&lDs for RCS and verify that no instrument lines, that penetrate containment directly connect with the RCS. Pressuriser Liquid and Steam Containment Isolation (E)

1. 56: Provisions of GDC 55 fe containment FSAR Section 15.6.2, page 15.6-3 Space Sample Line isolation are met and ensure that a sample line break will not exceed the limiting letdown line break size of 152 gpm for 30 minutes.

Verification ArciAod: Verify that the potential for release from the pressurizer liquid and steam space sample lines is within the letdown line break assumed. Check calculations of release rate for this versus the bounding scenario of 152 gpm CVCS letdown line break. Verify isolation provisions meet GDC 55. Page 54 of78

Aw I Chapter 15 Accident Mitigating Systems [,*l;",,];"" Revision: 01 9/lL97 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Failure of Small Lines Carrying Primary Coolant Outside Containment (FSAR 15.6.2) COMPONENT DESCRililON PARAMETER DESCRIPTION INPUT ASSUMiilON SAFETY ANALYSIS REFERENCES Various CVCS Letdown Line Break Dose (E)

1. 59: Per FSAR Table 15.6-2 FSAR Table 15.0-8 FSAR Section 15.6.2, page 15.6-4 FSAR Table 15.6-2 Verificarlon Method: Verify tnat each of the assumptions in the FSAR text and Table 15.6-2 is met and incorporated into the design basis calculation. In the case of conservative system alignment assumption, verify that the calculation takes no credit for those systems. Venfy dose calculation results consistent with Table 15.0-8.

Various Event termination (E)

1. 146 <= 30 minutes FSAR Section 15.6.2, page 15.6-3 Operator action NRC original SER for MP3 (8/2/1984)

Section 15.6.2 Veryicarlon Method: Venfy that procedures exist to support the operator action to isolate the limiting letdown system break within 30 minutes using the area radiation monitors. EiWiiiHsiiriG5EF#iliiiiRrWiiEi:FmTnTN~-MMG~ entrh;vw T wi mwGi Instrumentation - Area Monitors Detection of CVCS Leak (E)

1. 58: Area radiation monitoring and leakage FSAR Section 15.6.2, page 15.6-3 l

detection available and support detection and isolation of the design basis CVCS 152 gpm leak within 30 minutes. Vervicasion Method Verify that there are area radiation monitors to detect and mitigate this leak. Venfy monitor response time and associated testing is consistent with the bounding CVCS leak scenario. Page 55 of 78 . ~, V

Kerision: 01 Chapter 15 Accident Mitieating Systems khistone Iinit 3 = o 9/19M7 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Steam Generator Tube Failure (FSAR 15.6.3) COMPONENT DESCHil"I'lON PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES ~ SY5 TEM DESCRIPT10NfENGINIERED SIFEdd RDS MCTU ti$i5 5%E5K'-if0fifTT'IF ~ ~. ' ' ~ Insuumentation - Pressurizer Delay Time

1. 2 :4: 2.0 seconds to initiate SI signal w hich FSAR Table 15.0-4 i

Pressure Sensors initiates SI and AFW injection l'erification Aferhod: Verification of the instrument performance will be accomplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results. Instrumentation - Pressurizer Pressurizer Pressure

1. 213: Assumed trip setpoint,1845 psig, cited FSAR Table 15.0-4 1

Pressure Sensors in the reference analyses. l'erification Afrikod: Verification of the instrument performance will be accomplished by review of Technical Specification, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results. 1 5 5Ths~DsSURIPiiONik EgRhdisbijRiilidi361cil56n~s'%Usiiciigassocli@isitiditl5sy@]{ "Tl Various Offsite Dose Due to Steam Generator Tube

1. 150: FSAR Table 15.6-5 FSAR Table 15.6-5 1

i Rupture (FJL) l'ersyication Afethod: Verify by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with Table 15.0-8. ~.5T5. N DE.5 drip'T.fC5FR. 5KCi Rl5.6t5Cti6.N__5 5t.EG_5]Nf.S.T. @~ --)?W.hMwn.uwe~M. PE. T:"".""2 ~ ^ 8TW ~PJ. a -~..~-~s -O 6W9 u: x, S -.a a Instrumentation - Differential Temperature

1. 60: Rod drop occurs 1.5 seconds after RTS FSAR Table 15.6-4 i

Overtemperature dT Sensors trip signal l'erification Aferhod: Verify Technical Specificat;on and surveillance tests. ___~__ Page $6 of 78

g .u,.- Chapter 15 Accident Mitigating Systems Raisica: os Mihtone limit 3 1CAVP(Affected Accidents) AFFECTEDACCIDENTS: Steam Generator Tube Failure (FSAR 15.6.3) COMPONENT DESCRIVilON PARAMETER DESCRIVfION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES Instrumentation - Prenarieer Preuuriecr Pressurc

1. 196: Rod drop occurs 1.5 seconds after RTS FSAR Table 15.6-4 i

Pressure Sensors trip signal Verification Method: Venfy Technical Specification and surveillance tests. ~.., .aa ~ 1 1 i I Page $7of78

Chapter 15 Accident Mitigating Systems [','l;",'],","" sevnion: of D7&97 ICAVP (Affected Accidents) AFFECTEDACCIDENTS: Steam Generator Tube Failure (FSAR 15.6.3) COMPONENT DESCRIMION PARAMETER DESCRIMION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES ~ issiGiiiiii2ECE :::iC::T= a.4i:::::CEWhuhE%%Mi';ET;r 20T:I ~ w Various Event termination (E)

1. 62: Operator m nually isolates the faulted FSAR Section 15.6.3

team generator (including SG NRC SER for MP3 Supplemerit 4 b.owdown isolation) and cools down to equalize RCS/SG pressure within 30 minutes to minimize release rates Venfication Method; i) in original MP3 SER Supp.4, the NRC indicated that since NU committed to meet a currently evolving WOG initiative, the SGTR event results were acceptable. FSAR Chapter 15.6.3.4 descobes the WCAP that resulted from that effort. Verification of NU activities to meet the WOG guideltnes assumptions in WCAP-11002 and 10698 is necessary. Verify completion of NRC review of WCAP-10698.

ii) Venfy operating procedures. Various Steam Generator Tube Failure Dose (E/L)

1. 61i Table 15.6-5 List of parameters FSAR T ble 15.6-5 I

l Verification Method: Table 15 6-5 and the sechon 15.6.3 text cite many parameters used in the estimation of the radiological dose consequences. Venfy that each of these assumptions is appropriate per the Technical Specif. cation i limits and normal plant operations preceding the event. Venfy dose calculation reeds consistent with Table 15.0-8. Page 58 of 78

~,.: i i Chapter 15 Accident Mitigating Systems d sasiaa: 81 Milstone Ifait 3 ICAVP(Affected Accidents) j AFFECTEDACCIDENTS: Loss-of-Coolant Accidents Resulting from a Spectnam of Postulated Piping Breal6s Within the Reactor Coolant Pressure Boundary ] (FSAR 15.6.5) COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES j .=,,- ,y,,,, -y ~;7,,, =.: . y ~q p 77.. r g Various initiation of AFW Flow

1. 188 laitiated by SI Signal FSAR Section 15.6, page 15.6-11 1

NEU-96-615 l Verificesion Meshed: Venfy Technical Specifications and surveillance. ) -.7-- -~. - 7, .. -, - 7_-

y,.7 7 7 y

i Various Containment Design Provisions for LOCA

1. 159 Containment structural design FSAR Section 15.6.52, page 15.6-12 l

provisions per Table 6.2-3 j j Verificades Medad: Verify that the post-LOCA mass and energy release analysis is consistent with the latest reicad analysis results. Verify that the containment strength is sufficient providing that the post-LOCA depressurization systems operate. E %.~ ._.....g., _g 2 ,.-. :..s. j Pumps Flow Delivery Time

1. 152 Timing for ECCS Flow Delivery per FSAR Table 15.6-1 I

i Table 15.6-1 Verifcasion Meded: Venfy the Technical Specification requirements and surveiliance testing for the credited ECCS systems in Table 15.6-1. i Page59 of 78 ( --m -,m---.-

Chapter 15 Accident Mitigating Systems [,","",',", [l" scrision: or 9/19;97 ICAVP (Affected Accidents) AffECTED ACCIDENTS: Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Houndary (FSAR 15.6.5) COMPONENT DESCRil'flON PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES Various System Design Parameters Used in LOCA

1. 154: Assumptions per Table 15.6-8 FSAR Table 15.6-8 Analyses NEU-96-615 8'cryication.ticthod: Table 15.6-8 specifes the htmting ECCS performance parameters used in the LOCA analyses' Verify the '

.~ technical specification, the design, cperating, and testing characteristics of the system support the assumptions in the table. Note that Figures 15.6-20 and 15 643 are the reference for assumed pump / system (CVCS, HHSI, RHRS) flows. AYSTEM DESC{g]IO$ Muuggggggggs,' ages ocTueTl9N ftYftTM E!MfM%w -mEdem o M T oz - 4 Instrumentation - Containment Containment Pressure

1. 65: Generates Si signal which starts NEU-96-614 Pretsure Sensors SLCRS and ABVS at Ili-3 setpoint (10.0 psig) l'ervication Method: Venfy using Technical Specification and surveillance tests.

Instrumentation - Containment Containment laressure

1. 67: Control room isolation on containment NEU-96-614 i

Pressure Sensors pressure Ili-l signal FSAR Section 15.6.5.4 Vers]Ication Method: Venfy by Technical Specification surveillance test requirements. tastrumentation - Containment Containment Pressure

1. 63: Generates SI signal which generates NEU-96-614 Pressure Sensors ECCS actuation at Ili-1 serpoint (5.0 FSAR Table 15.6-1 psig) l'crification Method: Verify using Techrucal Specification and surveillance tests.

I Page 60 of 78

Chapter 15 Accident Mitigating Systems [],"",'Z"" w ion: of Cfl397 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: less-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6S COMPONENT DESCRIPTION PAR?. METER DESCR!!" TION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES Instrumentation - Pressunzer Delay Time

1. 216 2.0 seconds to initiate SI signal FSAR Table 15.0-4 I

Pressure Sensors Verificarlon Merkod: Venfication of the instrument performance will be accornplished by review of Technical Specifications, Calibration and Surveillance Test Procedures and Surveillance and Calibration Test results. Instrumentation - Pressurizer Pressurizer Pressure

1. 215 Assumed trip serpoint,1845 psig, cited FSAR Table 15.0-4 I

Pressure Sensors in the reference analyses. VerifIcarton MerAod: Venfication of the instrument performance will be accomplished by review of Technical Specification, Calibration and Surveillance Test Procedures and Surveillance and Ca'dbration Test results. r-weiswerirs ah-mmm*~mr~:r::::~mreewmmm r-mm, Vanous llot Leg Recirculation Criteria

1. 64: Operator action to transfer to hot leg NEU-96-615 recirculation within 9 hours FSAR Section 15.6.5, page 15.6-12 Verification Method: Verify that procedural instructions exist to ensure that the transfer to hot leg recirculation occurs within the 9 hour time period cited in the FSAR (page 15.6-12). Verification should include entrance enteria as well as appropriate re-alignment and confirmation actions.

Various Operator Action

1. 156 Sufficient indication and procedures to FSAR Section 15.6.5.2, page 15.6-12 perform transfer to post-LOCA cold leg recirculation mode of ECCS operation l

Verificarlon Method: Verify procedures and supporting indication exists to support the operator action to transfer the ECCS to the post-LOCA cold leg recirculation mode of operation. Event timing must be considered. Page 6I of78

, m a..; Kerhion: 01 Chanter 15 Accident Mitieatine Systems a o o e Mdstone limit J 9/1987 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) COMPONENT DESCRIlilON PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES AYlM N R W iM W M E M W N 3M E 3 " PMITENdEl?O3EsM4 M MG M+W UH4 5s uisdi! Instrumentation - Containment Containment Pressure

1. 157: System features to support FSAR Section 15.6.5.2, page 15.6-12 ll Pressure contTinment spray for containment NEU-96-614 pressure actuation at serpoint of 10 psig (Ili-3)

Verification Meskod: 1) Verify containment spray is consistent with FSAR Section 6.2.

2) Verify Technical Specification and surveillance test procedures for QSS system _

l [ ygpEw_cIM EHQIEEK]NBfTEgJSTH3E-N ~~~~~*=#"= ' *== <W=#" hia i~.~..m#. 23 Instrumentation - Pressurizer Pressurizer Pressure

1. 217: Rod drop occurs 1.5 seconds after RTS FSAR Tabic 15.6-4 ll Pressure Sensors trip signal Verification Medad: Verify technical speikatinn and surveillance tests.

p;ain:rDrc 4M-?_r W-g g_g g y;g g g g. g g y g g y y Pumps NPSil

1. 158: Adequate NPSil at minimum FSAR Section 15.6.5.2, page 15.6-12 containment post-LOCA pressure C1 Verification Meded: Venfy the design assumptions regarding the initial conditions in a post-LOCA containment and associated '

analysis use an acceptable approach to determination of minimum available NPSH for post LOCA ECCS, recirculation. Page 62 of 78

Chapter 15 Accident Mitigating Systems R<*lon: of Mastone Unit 3 7 ICAVP(Affected Accidents) AFFECTEDACCIDEATS: Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) COMPONENT DESCRilTION PARAMETER DESCRililON INPUT ASSUMITION SAFETY ANAL,YSIS REFERENCES ,.,,,,7 _y, i

15-1 & : 4, ; %.'. l. '.. k :. :.1 ?, :;. > 0yD1 y U',.,' :.. > i.: ' 1 ' '

.... a ;,. : - m: u.n g it m ~ +Ar4 % F ? ? M e x '**: n, y_md 7_. 7-_.. r-RefuctSg Water Storage Tank Volcme

1. 155 Su!Ticient RWST volume to provide FSAR Section 15.6.5.2, page 15.6-12 for capability to transfer to post-LOCA cold leg recirculation mode of ECCS operation.

Verification Method: Venfy design calculations for water volume and containment sump features provide for adequate NPSH for this mode of operation. Check for system start-up procedure and test results to venfy this mode of operation. Page 63 of78

Chapter 15 Accident Mitigating Systems l,"*"",,'];"" Keraion: or 9/198 7 ICAVP(Affected Accidents) AFFECTED ACCIDENTS: Loss-of-Coolant Accidents Resulting from a Spectrum of PostulaLd Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) COMPONENT DESCRIPTION PARAMETER DESCRitilON INPUT ASSUMITION SAFETY ANALYSIS REFERENCES $NNU5.3Mggg!(g!gyE {T][3}ggggJ[gjgl1431MNU%w &Ms3pWE r b u... w. _ m._.1 j $Y$Ib5 Fans Flow (E/UC)

1. If,9. Fan flow rate consistent with Reg.

FSAR Section 15.6.5.4 i Guide I.52 Verification Afethod: Verify that the appropriate system performance parameters are considered in the post-LOCA dose calculations for ESF leakage filtered by the auxiliary building ventilation and filtration system. Review Technical Specification surveillance requirements to ensure compliance with calculation assumptions per Reg. Guide 1.52. Filters Efficiency (E/UC)

1. 66: Filter performa - parameters FSAR Section 15.6.5.4 l1 efficiency l

l Verification Afethod: Verify that the appropriate system performance parameters are considered in the post-LOCA dose calculations for ESF leakage filtered by the auxiliary building ventilation and filtration system Review Technical Specification surveillance requirements to ensure compliance with calculation assumptions per Reg. Guide 1.52. Various Release Pathways (E/ llc)

1. 160: Release pathways for post-LOCA FSAR Section 15.6.5.4, page 15.6-23 doses are assumed to be limited to only the containment and the auxiliaiy building Verification Afethod: Verify by use of ventilation system design documents that the only release pathways for post-LOCA doses to evolve from the MP3 containment and ESF leakage is through the containment or auxiliary 1

building ventilation systems. Page 64 of 78

^ * " ' * ' ' ' ' ' ' " ' " Chapter 15 Accident Mitieating Systems Rcrision: 01 Alzhtune l'ait J o 9/19/97 ~ ICAVP(Affected Accidents) AFFECTED ACCIDENTS: Loss-of-Coolant Accidents Resulting from a Spectrurn of Postulated Piping Ilreaks Within the Reactor Coolant Pressure lloundary (FSAR 15.6.5) COMPONENT DESCRililON PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES gJgMMMiggg[AuggYBUl!.DgGEnTIONSyggejegcXfiTBATMBBVEli'lEddizain..u I aus I' low (till/C)

1. 164:In leakage into ductwork 1:SAR Section 15.6.5A, page 15.6-24 lI Item 2 Vers 7scation Sterhod: Verify by review of design drawings that air flow from areas of potential air borne activity are filtered prior to release through the unit 1 stack or accounted for in design calculations and surveillance testing.

EYETEMRE WMH.E9HIeMM!ET;!IBMSEfE6Tr.!-EBIAEEEliIRIS&REHEIMIl2N JEMIS11Mllex+%2==.mm.=. Various Leak Rate

1. 162: Lower leak rate aller T" I hour ( A FSAR Section 15.6.5 A lower leak rate is used post-LOCA)

Verification Afethod: FSAR indcates that a.nendrnent justified use of lower leak rate at T=1 hour. Verify that the design, operational, and testing provisions of that analysis are being rnet. SI!I!!Eb$b0!$I!SU5.bbbIU9b._0995YbuNI!95Iebb3dli[N#M"NW ' cS Cu.i Filters Filter El'liciency/ Flow Rate

1. 190: Removal ofiodine/ system flow per FSAR Section 15.6.5, page 15.6-12 ll Table 15.6-12 VenTscation Sterhod: Verify by Technical Specification and surveillance tests.

l Page 65 of 78

R aision: of Chanter 15 Accident Mitigatine Svstems a D o J Aldquac I:u44 3 9/198 7 ICAVP(Affected Accidents) AFFECTEDACCfDENTS: Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6 5) COMPONENT DESCRIPTION PARAMETER DESCRIFFION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES EI!IES Sb$b0$I!9NIEb095U9Y b905999.90k!Y$I!5 i!Eb!b'dm - ni. ' N ~~ M t_L_1 A.A a..i_ ...... l l' umps Lc.:L Kate(E/t/C)

1. 153: 5000 cc's per hour (Table 15.6-9) 1 SAR Section 15.6.5A, page 15.6-24 Vcnyicuilon Aferhod: Venfy system leakage using design calculation. Technical Specification and surveillance. Verify that the supporting dose calculation is consistent with FSAR Table 15.0-8.

RYSTFM RE!E6FJ!95LEYEHTEsteTER{Bd!912g!ca{ cggsqggggggy,gg3gggglgggg glggggg syg}mphen %g9Nw c w: c4m f y Secondary Containment Bypass Leakage (E/UC)

1. 163: Secondary containment bypass leakage FSAR Section 15.6.5A, page 15.6-24 l1 l

is documented per Table 15.6-9 and FSAR Table 15.6-9 FSAR page 15.6-24 Verification Afethod: Venfy that the analysis of bypass leakage has identified all of the bypass leakage paths and models them appropriately in the post-LOCA dose analyses. Various Atmospheric Dispersion Data (FJUC)

1. 70: Estimated T/Q's per Table 15.0-11 FSAR Section 15.0 l1 Vcnyication Afshod: Verify that the atmospheric dispersion data presented in Table 15.0-11 is appropnate for MP3 and is used in the post-LOCA dose calculations.

Page 66 of78

Chapter 15 Accident Mitigating Systems [,,";' "','j'7" Keelon: of 9/1847 ICAVP(Affected Accidents) AFFECTED ACCIDENTS: less-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 116.5) COMPONENT DESCRililON PARAMETER DESCRifilON INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES Vanous Control Room Dose (C)

1. 68: The post-LOCA radionuclide FSAR Table 15.6-10 inventory is a percentage of the FSAR Table 15.6-14 inventory available ia the reactor at the FSAR Table 15.6-15 time of assumed tuel damage as stated FSAR Tabl: 15.6-16 FSAR Table 15.6-17 FSAR Table 15.6-18 FSAR Table 15.6-19 FSAR Table 15.6-20 VenTscarism Aferhod: Venfy the control rocm bbitabihty calculation uses the referenced tables and supports the doses identified in 15.6J3.

Vanous Control Roo:n Dose (C)

1. 69: Assumptions per Table 15.6-12 FSAR Table 15 6-12 Verification Afeskod: Verify that assumptions listed in Table 15.6-12 re!!ect the plant design, operating and testing characteristics, and are property incorporated into the post-LOCA control room dor,a assessment per Table 15.6-13.

Vanous Control Room Dose (C)

1. 191 Post-LOCA radiological dose FSAR Section 15.0.9.1 consquence calculation uses an FSAR Section 15.6.5.4 inventory consistent with the TID-14844 assumption and MP3 core size Verificasion Method: Venfy that a calculation uses the inventones in TID-14844 and the MP3 core size to estimate the initial core radioisotope inventones for the post-LOCA dose estimates.

Page 67of78

Chapter 15 Accident Mitigating Systems l,1'l;';"'],';"" K ahlon: of 9/1D97 ICAVP (Affected Accidents) AFFECTED ACCIDENTS: lass-of-Coolant Accidents Resulting from a Spectruan of Postulated Piping Breaks Within the Reactor Coolant Pressure Houndary (FSAR 15.6.5) COMPONENT DESCHil" TION PARAMETER DESCRil" TION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES Various Dose Radiological Parameters

1. 166: Assumptions per Table 15.6-21 FSAR Section 15.6.5.4 FSAR Table 15.6-21 rcr:Tecarton Alcikod: Verify that assumptions listed in Table 15.6-21 reflect the plant design, operating and testing l

characteristics, and are properly incorporated into the post-LOCA TSC dose assessment in Table 15.6-22. Various OtTsite Dose (E/L)

1. 197: Assumptions per Table 15.6-9 FSAR Table 15.6-9 rcr7scation Sterhod: Venfy that assumptions listed in Table 15.6-9 reflect the plant design, operating and testmg characteristics, and are properly incorporated into the supporting dose calculation to assure mnsistency with FSAR Table 15.0-8.

Various Offsite Dose Due to ESF Leakage (E/tJC)

1. 192: Per Table 15.6-9 FSAR Table 15.0-8 ll Veri 7rcation Afethod: Venfy that the supporting dose calculation is consistent with Table 15.0-8.

mMladio St.cgSJltlfMI!9tGBEY+11f8Lo eWNd~h ma ascham aw fYETEM DEERigJION: e Fans flow (E/LjC)

1. 151: ESF filter performance parame.ers anJ FSAR Section 15.6.5.4 lI flow rate VertTscation Afeshod: Venfy that the appropriate system perfonnance parameters are considered in the post-LOCA dose calculat ons for ESF leakage filtered by the auxiliary building ventitation and filtration system Review Technical Specircation surveillance requirements to ensure comp!iance with calculation assumptions per Reg Gude 1.52.

Page 68 of 78

CNpter 15 Accident Mitigating Systems [,",'l"",'jl" Re slan: di Cfl&97 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Loss-of-Coolant Accidents Result'og from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMPIlON SAFETY ANALYSIS REFERENCES None Release Pathways

1. 161 Release pathways far post-LOCA FSAR Section 15.6.5.4, page 15.6-23 doses are assumed to be limited to only the containment and the auxiliary building.

Verification Method: Venfy by use of structural drawings and walkdowns that the only release pathways for post-LOCA doses to evolve from the MP3 containment and ESF leakage is through the containment or auxiliary building ventilation systems. Various Flow (E/lJC)

1. 165 SLCRS flow assumed to be directed to FSAR Section 15.6.5.4, page 15.6-24 i

Unit I stack VenJZeation Method: Verify by review of ventilation system design documents that the SLCRS is released through the unit 1 stack. I l l .... - ~. - Page 69of78

7 Chapter 15 Accident Mitigating Systems Re ason: ci Mildone Unii 3 9/1387 ICAVP (Affected Accidents) AFFECTEDXCCIDENTS: Radioactive Gaseous Waste system Failure (FSAR 15.7.I) COMPONENT DESChiPTION PARAMETER DESCRIPTION 7NPUT ASSUMPTION SAFETY ANALYSIS REFERENCES ,7. e, y y L ~., :.. . ::-; '[.~"T " ~}*"'"*f'"* ; ' ' [ ; ' ' f * *: ' f ;;"'i T*[? *?'*"}"~ft f}"! _.i

• 9y}TQ}

Filters Efficiency (E)

1. 41: Operating without filtration FSAR Section 15.7.1, page 15.7-1 I

l'erification Method: Venfy by reviewing the supporting dose calculation to dete mine if credit for fi;tration was taken. Various Ever termination (E)

1. 43: Terminateo <sthin one hour by FSAR Section 15.7.1, page 15.7-I operator action l'erification Me:Aad: Verify by reviewing the operating procedures to determine if there is a logical procedure to identify the source of the release and terminate the release.

Varicus Offsite doses due to waste gas system failure (E)

1. 40: FSAR Tables 15.7-2 and 15.7-3.

FSAR Section 15.7.1, page 15.7-I 5'erification Method: Venfy by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.0-8 Various Waste Gas System Release (E)

1. 42: Bypass of the waste gas filter is the FSAR Section 15.7.1, page 15.7-1 worst possible release frora the waste gas system Verification Method: Verify by reviewing system description to determine if there is a larger potential source for release.

Page 70 of75

Chapter 15 Accident Mitigating Systems K a hloa: of Milstone if ais J 9!!&97 ICAVP(Affected Accidents) AFFECTED ACCIDENTS: Radioactive Liquid Waste System Leak or Failure (Atmospheric Release)(FSAR 15.7.2) COMPONENT DESCRIFTION PARAMETER DESCRIPTION INPUT ASSUMPTION SAFETY ANALYSIS REFERENCES MeeiismhMIEXEClYJPTMCl ~ ~ kMSMeeY "AInin4 Boron Recovery Tank, TotalGaseous Activity (E)

1. 184 Largest liquid tank inventory.

FSAR Section 15.7.2, page 15.7-2 3BRS*TKIA,B Verl/lcetion Method: Venfy by reviewing design operating conditions for other tanks containing radioactive liquids and located in the plant yard. Various Offsste dose due to liquid waste system leak or

1. 44: FSAR Tables 15.7-4 and 15.7-5 FSAR Section 15.7.2. page 15.7-2 failure (E) l Verification Method: Verify by assuring each parameter is consistent with plant design and operation and by reviewing the l

l supporting dose calculation to assure consistency with FSAR Table 15.0-3. i i 1 t Page 71 of 78

""""*" 'j'l' " ,,,,, l Chapter 15 Accident Mitigating Systems sc elon: or Ul&97 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Liquid Containing Tank Failure (FSAR 15.7.3) COMPONENT DESCRIPTION PARAMETER DESCRil" TION INPUT ASSUMPTION SAFETY ANALYSIS REl'ERENCES -a-w.ii.r--mawmnsr_.g::ra..n r_cammerumsammusw.m:nzr - iskuum.:rumcwsAVtv Boron Recovery Tank, Total Liquid Activity (E)

1. 111 Largest liquid tank inventory FSAR Section 15.7.2, page 15.7-2 3BRS*TKI A, B Verification Method: Verify by reviewing design / operating conditions for other tanks containing radioactive liquids and located in the plant yard.

Vanous OfTsite liquid co,centrations due to liquid waste

1. 45: FSAR Table 15.7-4 and 15.7-5 FSAR Section 15.7.3, page 15.7-3 tank failure (E)

Verification Method: Venfy by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calcu'.ation to assure consistency with FSAR Table 15.7-7 l a l Page 72 of78 w 9

scelon: of Chanter 15 Accident Mitivatine Svstems a o o < Milstoac Unit 3 C/l&97 ICAVP(Affected Accidents) AFFECTEDACCIDENTS: Design Basis FucI Handling Accidents (FSAR 15.7.4) COMPONENT DESCRililON PARAMETER DESCRililON INPUT ASSUMITION SAFETY ANALYSIS REFERENCES .t 'r. )t _

4.[,,Q,--r~a

-v 7.-. r x pp w ~1 , * + _. 1 ~*..E e ~ e._.,,9 .t Con ainment Purge Supply and Closure Time (E)

1. 49: The valves close within 3 seconds of FSAR Section 15.7A.2.2, page 15.7-5 I

Exhaust Valves, receipt ofisolation signal 311VU*CTV32A/B.33A/B Verification Afethode Venfy by reviewing Technical Specification requirements for closure time, Surveillance test and test ~~ results. ag: .,f _e.,.t+,( 9 :'., a j L.*. j:C g-1. d f.m W m W W K.;., W j j P ? ~88* P 7 ;- M. 7 ( $ g g y Q Various Offsite Dose Dtte to Fuct flandling Accident (E)

1. 46: FSAR Table 15.7-8 FSAR Section 15.7A 2.1, page 15.7-5 Verification Method: Venfy by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.0-8.

..;.:~W.,;. 7. m. y g7- ^ r ': v. m. m.: :.~ >~.y-ee : <.: ~ ~ -~ - m. ;.7 :.,..;... :7...:m. m :., 3 w 7, ; ; : q,,.;.. 3 ,.,r_. Charcoal Filters,311VR* FLT2A, Filter Efficiency (E)

1. 48: The efficiency for organic and FSAR Table 15.7-8 inorganic iodine is 95%.

B Verification Afethoil: Verify by reviewing Technical Specification requirements for filter testing, Surveillance Test Procedures and test results. Various Operating Status (E)

1. 47: The system is operating in filtered FSAR Section 15.7A.2.I, page 15.7-5 mode when fuelis being moved.

Verl/lcation Method: Verify by reviewing Technical Sgifv* inn requirements and implementing operating procedures for moving fuel. Page 73 of 78

... _.. i.. _...........,,.... R aisiaa: or Chanter 15 Accident Mitieatine Systems Milstone Unis 3 a o o a' CflO97 ICAVP(Affected Accidents) AFFECTEDACCIDEVTS: Design Basis Fuel Handling Accidents (FSAR 15.7.4) COMPONENT DESCRitTION PARAMETER DESCRil410N INPUT ASSUMPTION SAFETY ANALYSIS HEFERENCFS

- -n -

-- -n y.,, n-.--- 7 -. m Contauunent Purge Air Exhaust Response Tirr<c (E)

1. 51: < 2 seconds to send closure signal to FSAR Section 15.7.4.2.2, page 15.7-5 Monitors,3RMS*RE4I,42 containment purge Verification Merhod: Venfy by reviewing the logic diagrarns for the monitor signals, identifying the expected dose rate at the monitor, the monitor setpoint and the associated Surveillance Procedures, and the moni*or response time specifed in vendor documents.

..j,'~ g g s.. W M M E g g {,,3.; p ff (.- f.;j 7--;, 7 :,,, r e ]..;3v-{ig A,4--g

g g 7 m

Spent Fuel Pool and S*orage Spent Fuel Pool Water Level

1. 50: 23 feet ab>ve the top of the fuel racks.

FSAR Section 15.7.4.2.1, page 15.7-4 Racks j Verification Method: Venfy by reviewing design drawings, leiel instrumentation and Technical Specification requPements. Page 74 of 78

Chapter 15 Accident Mitigating Systems Revision: Of Mducae limit 3 ICAVP (Affected Accidents) AFFECTED ACCIDENTS: Spent Fuel Cask Drop Accidents (FSAR 15.7.5) COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMl"IlON SAFETY ANALYSIS REFERENCFS imaame-amWase;menemmeandarmeimmaFur~nsen awaanziarirezacmzA-m*_E=m-E.r. 2 risacrm4Er.?.try?a:rrincam Bria t-e;! Spent Fuel Shipping Cask Trolley Cask Integrity (E)

1. 52: Maximum height oflift for spent fuel FSAR Section 15.7.1, page 15.7-6 cask < 30 f1 above any hard surface Verifkation Merked: Venfy by reviewing the design and operating procedures for rnoving the cask to assure no lift above 30 ft.

Spent Fuel Shipping Cask Trolley Cask integrity (E)

1. 53: Qualified lift height >= 30 feet above a FSAR Section 15.7.I, page 15.7-6 hard surface Ver!)kation Method: Venfy by reviewing the cask certification.

I Page 73 of78 -.. _... ~

Chapter 15 Accident Mitigating Systems [','[;",,'j'l'l" Kevisi=: 81 Cfl&97 ICAVP(Affected Accidents) AFFECTEDACCIDEVIT: Anticipated Transients Without Scram (FSAR 15.8) COMPONENT DESCRIPTION PARAMETER DESCRIPTION INPUT ASSUMMION SAFETY ANALYSIS REFERENCES h:rt (AT9EEtBREFEVERFMaGEEllEisC3REKm*M*EL~CC= 2 Icstrumentation-AMSAC Steam Generator Level

1. 35: Turbine trip <= 30 seconds after loss FSAR Section 15.8,7.8.I.3 ofmain feedwater WCAP 8330 Section 4-44 Verijkstlos Mc4ed: Venfcation of system performance and design will be accomplished by a review ot _.

l i a) ATWS modification documentation, including those involving installation, procurement, and modification testing. b) Functonal Tests / Procedures related with AMSAC l c) instrument setpoint and calibration records 4 + 1 This review would venfy the turbine trip setpoint, the independence of AMSAC from RPS and ESFAS, i l and the reliability of AMSAC. I Instrumentation - AMSAC Steam Generator Level

1. 37: asolation of Steam Generator FSAR Section 15.8, 7.8.1.3 blowdown and sampling lines after WCAP 8330 Section 4-44 loss ofmain feedwater Verijkastan Meshed: Verification of system performance and design will be accomplished by a review ot a) ATWS modification documentation, including those involving installation, procurement, and modifcation testing.

b) Functonal Tests / Procedures related with AMSAC c) instrument setpoint and calibration records This review would venfy isolation of blowdown and samping lines, the independence of AMSAC from I RPS and ESFAS, and the reliability of AMSAC. Page 76 of 78 2

Chapter 15 Accident Mitigating Systems [,"'ll;",,'j's"" Rc

• ion: of C11&97 ICAVP(Affected Accidents)

AFFECTEDACCIDENTS: Anticipated Transients Without Scram (FSAR 15.8) COMPONENT DESCRIPTION PARAMETER DESCR!FilON INPUT ASSUMl" TION SAFETY ANALYSIS REFERENCES Instrumentation - AMSAC Steam Gcaerator Level

1. 36: Initiate AFW Flow <= 60 seconds after FSAR Section 15.8,7.8.1.3 toss of main feedwater WCAP 8330 Section 4-44 Ven]ication Method: Venfication of system performance and design will be accomplished by a review of a) ATWS modification documentation, including those involving insta!!ation, procurement, and mndifetinn testiag.

b) FunctionalTests/ Procedures related with AMSAC c) instrument setpoint and calibration records This review would verify starting of all AFW pumps, the independence of AMSAC from RPS and ESFAS, and the reliability of AMSAC. I Instrumentation - AMSAC (C-20 Turbine Impulse Pressure

1. 39: AMSAC must be disabled at turbine FSAR Section 15.8,'7.8.1.3 l

Permissive) Ioad below 40% nominal after a time WCA;' 10858 delay between 180 to 420 seconds. Vcnyication Method: Venfication of system performance; and design will be accomplished by a review ot a) ATWS modification documentation, including those ir volving installation, procurement, and modification testing. b) Functional Tests / Procedures related with AMSAC c) instrument setpoint and calibrction records This revew would venfy the appropriate blocking of the ATWS signal, the independence of AMSAC from RPS and CCCAS, and the reliability of AMSAC. Surveillance tests / procedures for the assoc.ated permissive would also be reviewed. Page 77of 78

. eb.r W-N WP g

• Chapter 15 Accident Mitigating Systems

[,;,,',";'",,j"7' Ranion: of Gfl&97 ICAVP (Affected Accidents) AFFECTED ACCIDENTS: Anticipated Transients Without Scram (FSAR 15.8) COMPONENT DESCRIPTION PA RAMETER DESCRIPTION INPUT ASSUMl" TION SAFETY ANALYSIS REFERENCES Instrumentation - Steam Steam Generator Level

1. 38: AMSAC must actuate on low steam FSAR Section 15.8,7.8.1 3 Generator Level Transmitters renerator level (below 5 percent of WCAP 10858 Narrow Range Span) after sufficient time delay to allow feedwater system transients to momentarily disrupt feedwater flow without initiating AMSAC.

Veri ication Merhod: Verification of system performance and design wi!! be accomplished by a review ot f a) A'iWS modification documentation, including those involving insta!!ation, procurement, and mM## inn testing. b) Functional Tests / Procedures related with AMSAC c)lnstrument setpoint and calibration records This review would verify the steam generator level setpoints, the independence of AMSAC from RPS and ESFAS, and the reliability of AMSAC. Cales/ analyses determining the associated time delay would also need to be reviewed. Auxiliary Feedwater Pumps System Flow

1. 143 1760 gpm FSAR Section 15.8,7.8.1.3 WCAP 8330 Section 4-44 Verification Merhob Venfication of system performance and design wdl be accomplished by a revsw of surveillance s: o test / procedures and design drawings to venfy AFW flow.

Page 78 of 78

1 Project No. 9583 100 File MP3 7.0-001 9/18/97 Millstone Unit 3 Accident Analysis Critical Characteristics l'or Chapter 15 Accident Mitigating Systems (Sorted by Mitigating System) (83 Pages)

""j 8' Chapter 15 Accident Mitigating Systems l"C"'j"" ICAVP (Systems) SYSTElf DESCR1rT10N: ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC) PARAMETER DESCRIPTION INPUT ASSUMl" TION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES ymdf& m.:-- e.&y. ['-.} .} .-;e- } : sw :. 3;.',% - =,- - ~ ~

• 5 T..

u n.s.+ y. :.. &, e - a Sacam C""'"" Lesci Turbine trip <- 30 seconds after loss of main Anticipated Transients Wihr Scram FSAR Section 15.8,7.8.I.3 feedwater (FSAR 15.8) WCAP 8330 Section 4-44 Ven]Icarion Alethed: Venficaton of system performance and desagn wi:1 be accomphshed by a revew of-a) ATWS rWhtm documentation, includmg those involving inct,Wm, procurement, and rMhtm testing b) Functonal Tests / Procedures related with AMSAC c) Instrument setpoint and calibrabon records This review would ven'y the turbine trip setpoint, the independence of AMSAC from RPS and ESFAS, and the rdnMty of AMSAC. FSAR Section 15.8,7.8.1.3 Initiate AFW Flow <= 60 seconds after loss of main Anticipated Transients Without Scram feedwater (FSAR 15.8) WCAP 8330 Section 4-44 Venfirariaa Aterked: Verptm of system performance and design wdt be accomphshed by a review of' a) ATWS rWhtino documentabon, including those involving insta!!abon, procurement, and rWhtm testing. b) Functonal Tests / Procedures related with N.tSAC c)!nstrument setpoint and cahbration reccrds This review would venfy starung of a!! AFW pumps, the independence of NASAC from RPS and ESFAS, and the reliab&ty af AMSAC. Page1 of33

Chapter 15 Accident Mitigating Systems "a *3"" 8' w-n.n ,,,9,7 ICAVP (Systems) SYSTEWDERRIPTION: ATWS MITICATION SYSTEM ACTUATION CIRCUITRY (AMSAC) PARAMETER DESCRIPTION INPUT ASSUMFTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERFliCES FSAR Section 15.8,7.8.1 3 hm Generator level Imbrum of srnm C-~ blowdown and Anticipated Transients Witbaa Scram samp!Mg imes after loss of main feedwater (FSAR 15.8) WCAP 8330 ktion 4-44 l'enficar&w Meshed: Ven5 cation of system performance and design wiH be acces.piished by a review of a) ATWS nvv6hW documentabon, including those invorving installabon, procurement, and nw hN testing b) FuncbonalTests/ Procedures related with AMSAC c)1nstrument setpoint and cabbracon records This revew would venfy isolation of blowdown and sampfing lines, the independence of AMSAC from RPS and ESFAS, and the reliabihty of AMSAC. 7, - -- 3, ; 7...- - ,--ya , y a,r g-~; - --- m7.gpp g g g g g- -w .-.y- -n FSAR Section 15.8,7.8.?3 Turbine Impulse Pressure AMSAC must be disabled at turbine load below 40% Anticipated Transients Without Scram nominal aner a time delay between 180 o 420 (FSAR 15.8) WCAP 10858 seenuuk I'enyicaden Meshed: Venficaben of system performance and design will be accomphshed try a review of a) AlWS rehW documentation, including those involving instauation, procurement, and moddication testng b) Funcbonal Tests / Procedures related with AMSAC Page 2 of83

Chapter 15 Accident Mitigating Systems C"'T "*i""' 8' ICAVP (Systems) SYSTDiDESCRIPTION: ATWS MITICATION SYSTEM ACTUATION CXRCUITRY (AMSAC) PA2AMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES ~.,,.,.. - +- ..w,, -- n.,eag y . x,.m ,;x.,;:,;p y; 4,4%, - u. w .m ex n - - -- < ~ s. n. .. + m,- .a.-- FSAR Section 15.8,7.8.I.3 Steam Generator lock A MS AC must actuate on low steam generator level Anticipated Transients Without Scram (below 5 percent et Nrrow Range Span) after (FSAR 15.8) WCAP 10858 sutlicient time delay to allow feedwater system transients to momentarily disrupt feedwater flow without initiating AMSAC. 3'errfication Merked: Venfcation of system performance and design will be accompisshed by a review ot a) ATWS rnoddication documentation, including those involving insta!!stm, procurement, and l mtviWm tesbng b) Funcbonal Tests /Procedaes related with AMSAC Page3 of33

t i { t i f I "d"' Chapter 15 Accident Mitigating Systems i. a,a m v.a s ICAVP (Systems) SnTEwDESCRIPTJON: AUXILIARY BtJILDING VENTILATION SYSTEM (RPV-3314A) PARAMETER DESCRIFFION INPUT ASSUMFTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES i ~~ ' ~ S ,g I Flow (E) A secondary corirainment negative pressure less than Spectrum of Rod Cluster Control Assembly FSAR Section 15.4.8 j or equal to 0.25" wg is achieved in 60 seconds Ejecuon Accidents (FSAR 15.4.8) i t t rerpicasime mes6.it: verdy survedlance and est data support assumptons Review validdy of assumpton based on MP3 secondary contamment funcbonal desagn rak'dahnet I I \\ } Flow (Ell >C) Fan flow rate cr=<Mene with Reg. Guide 1.52 1.ms.of-Coolant Accidents Remirme from a FSAR Section 15.6.5.4 1 { i Spectrum of Pel2rM Piping Breaks Within the Reactor Coolant Pressure B=ndary (FSAR 15.6.5) ) Ver#iicasiner Meshed: Venfy that the imppicpiusic system performance parameters are considered in the post-L OCA dose rak,dahans for ESF leakage filtered by the auxdiary buddog ventilabon and fdtrabon systent Review j Techncal Spardirahnq $Jrygdlance requirements 10 ensure comphance with F23en32fM assumptons per i Reg. Guide 1.52. i 1 = a ? 1 F W m (E) Fiher removes greater than 95% iodine Spectrum of Rod Cluster Control Assembly FSAR Section 15.4.8 Ejection Accht* (FSAR 15.4.2) 5 L P ' - N-Venfy survesRance and test data support assu spicits. .l i j ) 1 i Page 4of83

Chapter 15 Accident Mitigating Systems ""hi"" 8' ma-v.n g,, ICAVP (Systems) SYSTElf DESCRlf710.V: AUXILIARY BUILDING VENTILATION SYSTEM (RPV-3314A) PA'2AMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Section 15.6.5.4 I Etriciency (E'L/C) Fitter performance parameters efficiency Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) renfcadon Jiahod: Venfy that the appropnate system performance parameters are considered in the post-LOCA dose c*nbhnns for ESF leakage fdtered by the auxthary buildmg ventilation and fdtrat:on system. Review Techncal Specifcation survedlance requirements to ensure compliance with calculation assumptions per Reg. Guide 1.52. ~~ ~ 'T T~'~5-- ~ 7 ay- .._.'VaNo'Es F '7 7 5, ^ T. ~ ~7'2561 =. u EN_ TI)E._Sch, _-wm I m. Release Pathways (ELC) Release pathways for post-LOCA doses are mnmed Loss-of-Coolant Accidents Resulting from a FSAR Section 15.6.5.4, page 15.6-2 to be limited to only the containment and the Spectrum of Postulated Piping Breaks ausiliary building Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) renycodon 3ferhod: Venfy by use of ventdation system design documents that the only release pathways for post-LOCA doses to evolve from the MP3 containment and ESF leakage is through the containment or auxdiary budding ventdation systems. I Page 5 of83

Chapter 15 Accident Mitigating Systems ""^1"'" 8' SI:!ssone t'ais J gyg7 ICAVP (Systems) S)TTElf DESCRIPTIO.V: AUXILIARY BU;LDING VENTILATION SYSTEM (RPV-3314A)/SLCRS FILTRATION (RPV-33141) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCFS ? 4 9. -\\;..% 4 0.. N D ' ?Y l ~xw I - ~ ?.$$ ?. A " :f+ -cSNM$ SF'WNMW' e Flow (E/UC) In leakage irao ductwort Imf-Coolant Accidents Resulting from a FSAR Section 15.6.5.4, page 15.6-24 i item 2 Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR l'.6.5) l'errAcarion AferAod: Venfy by review of design drawuigs that air flow from areas of po'v.tial air borne actmty are filtered prior to release through the unit 1 stack or accounted for in desgn calculations and survetilance testing. l Page 6 of83

Chapter 15 Accident Mitigating Systems l,"[*",7l"' "'"d*"' 8' C11197 ICAVP (Systems) S)TTElf DESCRIPTION: AUXILIARY FEEDWATER (AFW-3322) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES _7:_ ;"yl?lf % Q p y n y y g M " ?.~' [~.~q [.'~ ( & " ~ 7" _~M. System Flow Minimum system Aw with single active failure is Loss of Normal Feedwater Flow (FSAR FSAR Section 10.4.9.I, page 10.4-42 510 gpm to four steam generators. N loop operation 15.2.7) FSAR Section 15.2.6.1, page 15.2-10 NEU-96-623 90NE*-G-0075, Vantage 511 Fuel l VenTrario= Meded: Venty by reviewing Techncal Specircabon requirements, Survesance Procedures and test resu!*a 90NE*-G-0075 Section 5.13.4.2 l Minimum flow is greater than or equal to 900 Inadvertent Opening of a Steam Generator l gallons per minute for 3 loop operanon Relief or Safety Valse Causing a Depressuruation of the Main Steam Sysem (FSAR 15.1.4) renTestion Method: Venfication of system performance will be accomplished by review of system and component calcinations and performance and surveillance test results. FSAR Section 15.8,7.8.1 3 1760 gpm Anticipated Transients Without Scram (FSAR 15.8) WCAP 8330 Section 4-44 VenTration Mes&od: Venficahon of system performance and design will be accomphshed by a review of survesitance test /ep and design drawings to venfy AFW flow. 90NE*-G-0075 Section 5.13.4.2 Minimum flow is greater than or equal to 1200 Inadvertent Opening of a Steam Generator gallons per minute for 4 loop operanon Reliefor Safety Valve Causing a Depressurtzation of the Main Steam System (FSAR 15.1.4) VerTrades Meded: Venfcabon of system penem-a will be accomplished by review system and component <-nhfatirvis and performance and survedlance test test results-Page 7 of33

'j" Chapter 15 Accident Mitigating Systems C"[l" 8' ICAVP (Systems) SIITEM DESCRIPTJON: AUXILIARY FEEDWATER (AFW-3322) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES System Flow ILiinimum system flow with single active failure is Feedwater System Pipe Break (FSAR FSAR Section 10A.9.I, page 10A-42 470 gpm to three steam generators, N loop operation, 1518) FSAR Section 1518.2, page 15.2-18 300 gpm to two steam generators, N-1 loop operasion NEU-96-623 Versficaties Method: Venfy by reviewog design requirements. Techrucal Spec:feation requirements, Suvet!!ance Procedures and test results. FSAR Table 15.2-1 Time to Full Flow 68.$ seconds for N-loop and 69.1 for N-1 loop from Feedwater System Pipe Break (FSAR low-low steam generator level trip, including diesel 1518) NEU-96-623 LMag on loss of offsite power for motor driven pumps rers]icamies Meded: Venfy by revewing design requirements. Techncal Spectfcation requirements. Survestlance Procedures and test results. Venf= taxi of DG performance will be accomphshed try a review of surveillance procedures and test results for the DG and ESF load sequencer 60 sernmk from low-low steam generator level trip, Loss of Normal Feedwater Flow (FSAR FSAR Section 10A.93, page 10A-47 including diese! Inadmg on loss of offsite power for 1517) FSAR Section 1516.1, page 15.2-10 motor driven pu:np 90NE*-G-0075, Vantage 511 Fuel NEU-96-623 Veryicaties Meded: Venfy by rewewing design requirements. Surveillance Procedures and test results Venfcation of DG performance will be accomplished by a review of survedlance procedures and test results for the DG and ESFload sequencerst 70 seconds frcun low-low steam gene:ator lesel trip Loss of Normal Feedwater Flow (FSAR FSAR Section 10A.93, page 10A-47 for turbine driven pump 1517) NEU-96-623 90NE*-G-0075 Vantage 511 Fuel Versy3cmales Medad: Venty by reviewmg design requirements, Survesitance Procedures and test results Page 8 of83

L i l r "'"da" 8' Chapter 15 Accident Mitigating Systems hiddene l'ait 3 9ffg7 ICAVP (Systems) } .SlTTEMDESCAlrTION: AUXILIARY FEEDWATER (AFW-3322) i t ] PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES Time to Full Flow Water supplied to steam generators no later than 10 Steam System Piping Failure (FSAR 15.1.5) NEU-9M23 I min after accident. T renTuar&= Meded-venty by reviemng design requirements, surveillance procedures, and test results i i 6 62_{:DW@nhwi.:t h W M{ t Maximum Flow Rate to Factred 39 lbm/sec Feedwater System Pipe Break (FSAR FSAR Section 15.2.82, page 15.2-19 I_ine 15 2.8) FSAR Table 6.2-59 i Venfkar&= Meded.- Venfy by reviewing design and testing requirements for the venturi. j Anail^!e Capacity Sufficient capacity to maintam 10 hours of hot Feedwater System Pipe Break (FSAR FSAR Section 10.4.9.1, page 10.4-42 crmahy and 6 hours to cool down to 350*F 15.2.8) l renTrados Medad: Verdy by reviewing design cabil*ns. Technical Specification and Surveillance Procedures for the tank, ] and operating procedures xdW with auxiliary feedwater suppt/. t l I t 4 Page 9 of83

C -e'-"N" '"""-'""?. Chapter 15 Accident Mitigating Systems "* 8' us-vm g,, ICAVP (Systems) SYSTDIDESCRIPTION: AUXILIARY FEEDWATER (AFW-3322) PARAMETER DESCRIPTION INPUT ASSUM71' ION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 7 i i :."'+ i'i 2 ". O ' ote 5C MS.M i ? 7'W[} Y~.C ~ J.,1 (i.[.41.? M'47 # ' u h; : ' .. n...- Initiation of AFW Flow laitiated by SI Signal Loss-of-Coobat Accidents Resulting from a FSAR Section 15.6, page 15.6-1I I Spectrum of Postulated Piping Breaks NEU-96-615 Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) renTestian Merhost: Venfy Technical Speedications and surve 11ance. l l l Page 10 of33

Chapter 15 Accident Mitigating Systems

• • 8' m

im g,, ICAVP (Systems) SYSTEWDETCXIFTION: CHEMICAL AND VOLUME CONTROL SYSTEM (CVC-3364) PARAMETER DESCRIPTION INPUT ASSUMITION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES Q ;.~ D., 7,,. _ ;, W"~. ^^ " -+-

' ~ ~,

t - G ; - i \\ # ~k % f Concentration Greater than or equal to 2500 ppm laadvertent Opening of a Steam Ger.crator NEU-96-623 (pg 15.1-14)

RelieforSafety Vahe Causing a Depressurization of the Main Steam System (FSAR 15.1.4) Vers]ication Meded: Venficabon of this parameter wiR be accomplished by review of water chemistry sampling procedures and the insiicizi sampling records. -~T . ' ~ '. 'Qa q R T. - ~ L c% L, MR="B?&WMP) ~

• ^

~ . 4. #,..a. ..J.G..,.A u* W ~ ~. " 3 T.. i....A -M ..~ NEU-96-623 Concentranon Boron concentranon >= 6600 ppm Chemical and Volume Control System l Malfuncuon that Results in a Decrease in FSAR Section 15.4.6 l Boron Concennation in t!. Reactor Coolant (FSAR 15.4.6) Verificaden Meded: Venfy that operating procedures shah provide instruchon wittun CVCS configuration for borated water source NEU-96-623 Cencentrauon Termination ofdilution when M-c'-4 Requires Chemicaland Volume Control System operator acuan Malfuncuan that Results in a Decrease in FSAR Sectica 15.4.6 Baron Concentration in the Reactor Coolant (FSAR 15.4.6) Verficades Meded-Verify that operahng procedures shas provxie ddubon path Mahnr3 anStruchon within CVCS (LCV-1128 and C) and configurabon for borated water source (LCV-112D and E) Page ll of83

= _. ""p 8' Chapter 15 Accident Mitigating Systems C"y'jl" ICAVP (Systems) - SBTEMDESCRIPTION: CHEMICAL AND VOLUME CONTROL SYSTEM (CVC-3304) PADAMETER DESCRifTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-96-623 Concentration laject boron when dilution has been terminated. Chemical and Volume Control System Requires operator acten Malfuncton that Results in a Decrease in FSAR Section 15 4.6 Boro a Concentration in the Reaaor Coolant (FSAR 15.4.6) renykasime Meded: Venfy that operatng pixedures shall provide dilution path isolation instruction within CVCS (LCV-1128 and C) and configurabon for borated water source (LCV-112D and E)

4,,

~T~~ . 17 -{ ; 3Agy&W[ ' ~ ~ - ' ~ 7:, j ;;, .1-.., R Q 1 2 I Flow Operator action to limit flow to <= 150 gpm using Che=61 and Volume Control Sprem NEU-96-623 valve V305 Malfimamn that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) Verfkaden Meded: Venfy that operating procedures shaB provide flow limit of 150 gpm and that administratue control limit exists. g .,,,. g _y g a, .:. 3 ;... NEU-96-623 i Deviation Alarm >IO*a deviation requires che=61 and Volume Control Sptem Operatoracten Malfunction that Results in a Decrease in FSAR Section 15.4.6 i Boron CweaM_= in the Reactor Coolant (FSAR 15.4.6) renyicaden Meded: Verdy that the operating procedures list the appropnate monitars and that there are proper procedures for alarm response Page12of83

i """"' 8 ' Chapter 15 Accident Mitigating Systems 4 w.. - w a

gin, L

] ICAVP (Systems) S)$TEMDESCA1PT10.V: CHEMICAL AND VOLUME CONTROL SYSTEM (CVC-3304) l PADAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFL.tENCES Flow !~4W'k a Abnormalindication requires Chemical and Volume Control System NEU-96-623 ] Operator actia, Malfanction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) l'er:7scation Meshed: Venfy that tne operating procedures list the appropriate rnonitors and that there are proper procedures for ] alarm response i l l T. 7.K.,,, q 7.... -. y.... , - v5 - ).,..Y. r - . ;<m- ~7 .:: ; ;.. 7y- ' 9. . ;% -- A y.. ? ~ Deviation Alarm >10% deviation requires Chemical and Voluute Control System NEU-96-623 Operator action Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration 'm the Reactor Coolant (FSAR 15.4.6) 5'erJIcanes Meded: Venfy that the operating proceds.tres list the appropnate monitors and that there ase proper procedures for alarm response i Flow Imirarian Abnormal kiiearvwn requires rh-WI and Volume Control System NEU-96-623 Operator action Malfunctim that Results Si a Decrease in FSAR Section 15.4.6 Baron C4 ncentration in the Reactor Coolant (FSAR 15.4.6) l'enTscarses Meded: Venfy that the operating procedures list the appropriate mondors and that there are proper procedures for alarm response Page 13 of83

Chapter 15 Accident Mitigating Systems

• * ' "' 8' um tw g,,,7 ICAVP (Systems)

SnTEM DESCRffTIO.V: CHEMICAL AND VOLUME CONTROL SYSTEM (CVC-3394) PARAMETER DESCRIPTION INPUTASSUMPTION AFFECFED ACCIDENTS SAFETY ANALYSIS REFERENCFS .g .. _y,j ...._j.,.. c, w.r. g.,. 5 - = "-"'e+ . _ - - +. a 4 s_-. NEU-96-623 l Boric Acid Pump Status Requires operator action to monitor status lights to Chemical and Volume Control System ) detect dilution events Malfunction that Results in a Decrease in FSAR Section 15.4.6 Borca Concentration in the Reactor Coolant i l (FSAR 15.4.6) l Verificarian Meded: Venfy that the operating procedures list the appropriate morutors and that tnero are proper procedures for alarm response NEU-96-623 Charging Pump Status Requires operator actum to monitor status lights to n-=W and Velume Control System dete:t dilution events Malfunction b Results in a Decrease in FSAR Scction 15.4.6 Baron C-earnrion in the Reactor Coolant (FSAR 15.4.6) Venficados Meded: Venfy that the operating procedures list the appropnate morutors and that there are proper procedures for alarm response. NEU-96-623 Primary Water Pump Erm Requires operator action to monitor status lights to nemical and Volume Control System detect dilution events Malhanction that Results in a Decrease in FSAR Section 15.4.6 Boron C-eatntb in the Reactor Ccolant (FSAR 15.4.6) Verifkation Meded: Venfy that the operating procedures kst the appropnate morutors and that there are proper procedures for alarm response Page 14 of83

Chapter 15 Accident Mitigating Systems "'***8' u-twa ICAVP (Systems) SYSTDIDESCRIPTIO.V: CHEMICAL AND VOLUME CONTROL SYSTEM (CVC-3304) PARAMETER DESCRIPTION INPtJT ASSUMl" TION AFFECTED ACCIDENTS SAFETY ANALYS85 REFERENCES A. 3 5 A 4t*.. WYE %Wf3M ..?.i ' W ' iE. 5A M +'.M NOwM%*NlMME4 FSAR Section 15.5.1.2.D Opeming Smus Pressurizer spray is operable Inadvertent Operation of Emergency Core Cooling System During Power Operation (FSAR 15.5.1) Ver#icarios Medd Assumed avadable to rningate the trasisent. Venfied by review of operating procedure, contred turung procedure and RCS Pressure Control Cahbrabon records. m m: p p. y m y -z. FSAR Section 15.4.6 I Concentration Operator acten required to lerminate dilution Chemical and Volume Control System hhtfimw-that Results in a Decrease in NSU-%-623 Boron Contration in the Reactor Coolant (FSAR 15.4.6) Verificarlos Medd Venfy that operating procedures sha5 provide dilution path isolation instructon wdhin CVCS (LCV-1128 and C) and configurabon for borated water source (LCV-112D and E) j Page15 of33

mai+-: 3' Chapter 15 Accident Mitigating Systems C"l7 CflL 97 ICAVP (Systems) S3 ITEM DESCRIPTION: CalEMICAL AND VOLUME CONTROL SYSTEM (CVC-3304) PARAMETER DESCRIPTION INPtJT ASSUMITION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES b-- $$$ik3[b Reicase R.ue From !kolen CVCS letdown line break nusimum ikru of 152 gpm Failure ofSmall Lines Carrying % FSAR Section 15.6.2, page 15.6-3 CVCS Line(E) Coolant Or=<ide Cnania-t (FSAR 15.6.2) Var]Ication Merkost: Venfy using design drawngs and mb deinas. l l l Page 16 sf83

Chapter 15 Accident Mitigating Systems Z"O 8' ICAVP (Systems) SYSTEM DESCRIPTION: CONTAINMENT PURCE (HVIUHVU-3J13E) PARAMETER DESCR11" TION INPUT ASSUMPTION AFFECTL) ACCIDENTS SAFETY ANALYSIS REFERENCES l l NSiNE3U.INEE2I FSAR Section 15.7.4.2.2, paga 15.7-1 Ckuure T~ane (E) The vahrs close u ithin 3 seconds of receipt of Design Basis Fuelllandling Accidents imhtm signal (FSAR 15.7.4) l renTendon Merkat: Venfy by reviewing Techrucal Specificatxm requirements for clostwe trne Survesuance test and test I resuus. Page 17of33 ms

Chapter 15 Accident Mitigating Systems Z",'j"" 8' ICAVP (Systems) SISTEM DESCK1PTION: CONTAINMENT STRUCTURE (ILRT, LLRT & ELECTRICAL PENETRATION (CMT-3312A)) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES ,-., = u n n, _

• • .- wr > z e o t am-9 *.9^

-T".T'*'FCC 3 %'. ' '. w ! M L ; d. t b7 4 s2

rM * :

Om e=~' Design Provisions Cwr= ment structural design prowisi.xts per Tab!c Loss-of-Coolant Accidents Resulting from a FSAR Section 15.6.5.2, page 15.6-12 for LOCA 6.2-3 Spectrum of Pmmt"ad Piping Breaks Within the Reactar Coolant Pressure Boundary (FSAR 15.6.5) Ved/Jcation Method: Venfy that the post-LOCA rnass anu erergy release analysis is consistent with the latest reload ana!ysis resutts Venfy that the containment strength is suffeient providmg that the post-LOCA depressunzation l systems operate l l l Page 13 of33

Chapter 15 Accident Mitigating Systems C"i"",""

• • 8' ICAVP (Systems)

SUTEM DESCRIPTIO.Y: CONTAINMENT STRUCTURE (ILRT, LLRT & ELECTRICAL PENETRATION (CMT-3312A)) PARAMETEF. DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 7E" 5. 25$3 FSAR Section 15.6.5.4 Leak Rare Lower leak rate aAer T=1 hour (A lower leak rate is Loss-of-Coolar.1 Accidents Resuhing from a used post-LOCA) Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) VenJIcation Meskod: FSAR ineb% that amencment justfied use of lower leak rate at T=1 hour. Venfy that the design, operational, and testing provtssons of that a.Wsis are besng met. FSAR Section 15.4.8 The rmt-ment leak rate is less than or equal to Specaum of Rod Cluster ConLol Assembly l C.65% contamment total volume per day Ejecten Accidents (FSAR 15.4.8) l'enylcarian Meskod: Venfy test data and survei!!a.v,e support assumptms.

""6'"" 3' Chapter 15 Accident Mitigating Systems Staldoor t sis 3 gg7 ICAVP (Systems) SITTE.lf OESCRIPTION: CONTROL ROOM VENTILATION (ACC-3314F) PARAMETER DESCRil' TION INPUT ASSUMf" TION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES y -,; ::3 7 g ~ y p q g - -. g7_3 97 ;;, - r:(ye; po - - .,--~; .._, a Fiher Efficiency,Tiow Rate Remosal ofi.*bne' system flow per Table 15.6-12 less-of-Coolant Accidents Resatting from a FSAR Section 15.6.5, page 15.6-12 I l Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) Vaification 3fakoh Venfy by Techrucal Sp*tm and survAance te sts. l Page20 of83

!%.rtheast I taisaws Chapter 15 Accident Mitigating Systems Revan: ei st _ im,3 mwr ICAVP (Systems) SISTElf DESCRIPTiO.V: EMERGENCY CORE COOLING SYSTEM (ECCS) PARAMETER DESCRIPTION INPUT ASSUM.* TION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES ..,, s.L... ' x 7 T.y 1.~ OQ&% [O a. [G "fTMfhEdQ ' ? - [ E 5 1-C 'Q' Q C 3 j e.'

3*,-

FSAR Table 15.6-1 1 flow Deliscry Time Timing for ECCS Flow Delivery p r Table 15.6-1 Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) I Vm74carian Meded: Venfy the Technical Sehnq requirements and survestlance testing for the credited ECCS systems in Table 15.6-1. Pump Flow Minimum flow from a single high head safety inadvertent Opening of a Steam Generat 90NE*-G-0075 (pg 5-32) injection permp is less than the minimum flow from a Relief or S.ifety Valve Causing a single centrifugal charging pump to ama e the Depressuruation of the Main Stevn Sprem accident analysis are cordervative. (FSAR 15.I A) Ver:Tscation Medod: Ver6hno of the system flow performance wi!! be accomplished by review of Startup Test and Surves!!ance Test resu!!s, and Techrucal Spehhnn. Also, revew of calculations extrapolahng mini-flow results to full fkww corvthnns will be required. E5*hM1[$I [bYb5 FSAR Table 15.6-8 System Design Parameters Used Mmmptions per Table 15.6-8 Imss-of-Coolant Accidents Resulting from a in LOCA Analyses Spectrum of Postulated Piping Breaks NEU-96-615 Within the Reactor Coolant Pressu<e Boundary (FSAR 15.6.5) M he limiting ECCS per.imance parameters used in the LOCA analyses. Venfy the VerrTacarios Meded: Table 15.6-8 P t techrucal 4*hno, the design, operating, and testng characteristics of the system supM the asstmytic.w in the table. Note that Figures 15.6-20 and 15.6-43 are the reference for assumed pump / system (CVCS HHSI,RHRS) flows. fage 21 of83

Chapter 15 Accident Mitigating Systems " *" ".o"" 8' muo-i g,g, ICAVP (Systems) SITTElfDESCRIPTION: EMERGENCY CORE COOLING SYSTEM (ECCS) PARAMETER DESCRIFTs 3N INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES

• e-f.c.4 U,7 6 "l': k J <4 *

-, 5 6 sf T'~,.- I U~ ' S==.~* + " G4/WM,"[ - OfEP : - 33f479"':9 ~ ~ 2 .

• 7. 2 i Isal Rate (LIJC) 5000 cc's per hour (Table 15.6-9)

Loss-of-Coolant Accidents Resulting from a FSAR Section 15 614, page 15.6-24 Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) Venfication Merkat: Venfy system leakage using design calculation. Tectuucal Sybbnn and survedlance Venfy that the j supporting dose # 'fnW is consistent *.anth FSAR Table 15.0-8. Page22 of83

Chrpter 15 Accident Mitigating Systems "

• 8' uaa r.u g,,,

ICAVP (Systems) SITTElfDESCRIPTION: ENGINEERED SAFECUARDS ACTUATION SYSTEM (ESA-3467A) ~ PARAMETER DESCRIPTION INPUT ASSUMl" TION AFFELTED ACCIDENTS SAFETY ANALYSIS REFERENCES CIOf.rO8ENT DESCWTION: liestnsaneetation # M S M +6[M Mf3d_ @ c$[E P E 7 1-]_f,$;f~' ~ j NEU-96-614 CmrUnment Pressure C-w SI signal w hich generates ECCS actuation Loss-of-Coolant Accidents Resuhing from a at fli-I epoint (5.0 psig) Spectrum of Postulded Piping Breaks FSAR Table 15.6-1 Within the Reactor Coolant Pressure j Boundary (FSAR 15.6.5) s'enTrasaw Medod: Venfy using Techncal Whtm and survedlance tests. NEU-96-614 I Control room isotation en containment pressure !!i-1 Lef-Coolant Accidents Resulting from a signal Spectrum of Postulate. Piping I!::2Ls FSAR. Section 15.6 5.4 Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) l'enTrasics Meded: Venfy by Techncal Sp*tm surved!ance test requirements. NEU-96-614 Generates SI signal which stans SLCRS and ABVS Loss-of-Coolant Accidents Resulting from a at fli-3 setpoint (10.0 psig) Spectrum of Postulated Piping Breaks Within the Reactor Caolant Pressure Boundary (FSAR 15.0.5) s'enTraries Meded: Venfy using Techncal SPhtion and survedlance tests. Page23 of33

Chapter 15 Accident Mitigating Systems C J"' 2* 8' ICAVP (Systems) SITTElfDESCRIPT10.V: ENGINEERED SAFEGUARDS ACTUATION SYSTEM (ESA-3407A) PARAMETER DESCRIPTION INPUTASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-96-623 I c - n w r Pressure Safety injecuon signa! si Ifi-1 setpoint; < tam line Steam System Piping Failure (FS.* R 15.1.5) isolation sigm.1 at Ili-2 serpoint. Safety injecuan actuartion logic inirisre< AFW flow and FW isolarum Veryication Medo.fr Venfy by revewing Techrucal Wtion requirements. Survedlance and Calibration Procedures, and test results 1 .1 ,.T" ' 'Y c,.. s 6. ;.72DMI ' ~ ' ~ " ' " ' " ~ ~ ~~ "~ .. (i V gR T #: .~-2 ~ .~ ~ ~ - ~ ~ ^ F ~' !. tem Pressuruer Pressure Si signal generated @ 1860 psia 2 second delay Feedwater System Pipe Break (FSAR 90NE*-G-0075. Vantage 511 Fuel before rod drop 15.1 8) l'er]ication Meded: Venfy by reviewing Techrucal Wfic=6m requirements, Survedlance and Cahbration Procedures, and test results. FSAR Section 15.4.8 Provides SI signal within one minute @ 1600 psia Spectrum of Rod Cluster Control Assembly Ejecuon Accidents (FSAR 114.8) l'ers]icaties Meded: Venfy by revieweg operating procedures. g .... ~ _. u-c . ~ FSAR Table 110-4 Szcam Genera or Water level Delay time will not exceed 10 seemd< (Dela) time Inadvertent Opening of a Steam Geacrator definition is e-nu in FSAR Table 110-4) Reliefor Safety Valve Causing a 90NE*-G-0075 Table 17.1-4 Depressuruation of the Main Steam System (FSAR 111.4) VerP-W Meded: Verdicahno of the instrument performance wWI be scicingsshed by review of Technical Smfie=Hms, (whration and Survedlance Test PrWnres and Survedlance and Cahbration Test results Page24 of83

Chapter 15 Accident Mitigating Systems C"7 8' ICAVP (Systems) S)TTER DESCRIPTION: ENGINEERED SAFECUARDS ACTUATION SYSTEM (ESA-3407A) PARAMETER DESCRIPTION INPUT ASSUMf" TION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 90NE*-G-0075 Table 5.1.1-4 Steam C-~ Water l_evel Trip setpomt is 0*k ofnarrow rauge span for inadsertent Opening ofa Steam Generator feedwater !ine breal,10*& of narrow range span for Reliefor Saf.-ty Valve Causing a FSAR Table 15.0-4 loss of normal feedwatertloss of offsite power. Depressurization of the Main Steam System (FSAR 15.1.4) Ver#Icarien Medad: Venficabon of the instrument performance ws!! be accomphshed by review of Techncal Sphtms, Cahbraton and Survedtance Test Procedures and SurveAance and Cahbration Test resuits. L'_,*..,;it' [ C 5-[ 7.~ i: T I C D N" ~ ~ D2 ENM f 90NE*-G-0075 Table 5.I.I-4 I Delay Tune 2.0 seconds to initiate SI signal hm System Piping Failure (FSAR 15.1.5) Ter#Ication Medad: Versfrahm of the instrument performance will be accomplished by review of Te.chrucal Sphfims, t'aahraton and Surveillance Test Procedures and Surveillaxe and Cahbration Test results. FSAR Table 15.0-4 I 2.0 seemk to inante SI signal Loss-of-Coolant Acadents Resulting from a Spectrum of Postulated Piping Breaks Within the Rearts Coolant Pressure i Boundary (FSAR 15 6.5) l v s-rat ri of the instrument performance will be accomphshed by review of Techncal Specifications, l rergicamien MM-e f Mhraton and Surved!ance Test Procedures and Survediance and Cahbration Test results I rage 25 of83

Chapter 15 Accident Mitigating Systems Z*"Cl" "a* 8' Cf!O97 ICAVP (Systems) l SITTER DESCRfPTIO.V: ENCINEERED SAFECUARDS ACTUATION SYSTEM (ESA-3407A) i PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 90NE*-G-0075 Table 5.1.1-4 i 2.0 sec=h o intenre SI signal Inadvertent Opening ofa Steam h Delay Time t Reliefor Safety Valve Causing a Depressurization of thc Main Steam System (FSAR 15.1.4) renK wien Meded-Verihtuvi of the instrument perfor.aance wdl be accompashed by review of Technical Sphtuvis, Neation and Servc21ance Test Procedures and Survediance and Cahbration Test results 2.0 seconds so initiate SI signal which initiates Si and Steam Generator Tube Failure (FSAR FSAR Table 15.0-4 1 AFW injection 15.6.3) renfication Meded: Venficaton of the instrument performance will be accomphshed by revew of Technical SpebWs, Cahbrabon and Survei2ance Test Procedures and Suned:ance and Cahbration Test results %NE*-G-0075 Table 5.1.1-4 I I essunzer Pressure A unmed trip setpoint, I845 psig, cited in the Steam System Piping Failure (FSAR 15.1.5) reference analyses. rer#icados Meded: Ver*atrvi of the instrument performance wi t be accomphshed by review of Techncal Specificabon, Cahbrabon and Survedlance Test Procedures and Survedlance and Cahbration Test results 90NE*-G-0075 Table 5.1.1-4 i A<urmed rip setpomt, !845 psig cited in the inadvertent Opening of a Steam Generator t reference analyses. Reliefor Safety Valve Causmg a Depressurization of the Main Steam Systern (FSAR 15.1.4) I I rerificaales Meded: Ver* arrvi of the instrument perkirniac will be accomplished by review of Technical SWtrui. Cahbrabon and Surveinance Test Prne** ues and Survedance and Cahbration Test results. Page 76 ofSi 1

Chapter 15 Accident Mitigating Systems Z",'7

• • 8' ICAVP (Systems)

SFSTEMDESCAfrTION: ENCINEERED SAFECUARDS ACT UATION SYSTEM (ESA-3407A) PARAMETER DESCRIPTION INPUT ASSUMl"FION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Table 15.04 I Pressurtzer Pressure Assumed trip serpoint,1845 psig, cited in tiv: Lossa>f-Coolant Accidents Resulting from a reference analyses. Spectrum of P="'2r-d Piping Breaks Within the Revmr Coobnt Pressure Bound.ary (FSAR 15.65) 5'er.ficarisa Meded: Veri-4-- of the instrument perbnv=.ca wiB be accomphshed by review of Techrucal Specification, caLhration and Surve: Hance Test Procedures and Survei8:ance and Cahbration Test results l FSAR Table 15.04 I A <urmed trip serpomt,1845 psig, cited in the Steam Generator Tube failure (FSAR reference analyses. 15.63) ren]icaden Medad: Venfcation of the instrument performance win be accomplished by review of Tt chrucal Specification, Cahbraton and Surve Bance Test Procedures and Survedtance and cahhration Test results k@EEM*kd5M[$ %NE*-G-0075 Table 5.1.1-4 Deby Time 2.5 seconds (The definhion ofdelay time is given in Feedwater System Malfunctions that Result FSAR Table 15.04). in a Decrease in Feedwater Ten.perature (FSAR 15.1.1) renyIcaden Meded: Venfication of the instrument performance wiR be accomp'ished by review of Techncal WWoons, cahhration and Survedlance Test Procedares and Survedlance and Cahbrabon Test results %NE*-G-0075 Table 5.1.1-4 Sacam C-= Water Level A summed trip serpoint is 100% of narrow range span. Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (FSAR 15.1.1) renylcade Nemed: ver#atuvt of the instrument performance win be accomphshed by review of Technical Specifications, cahhrabon and Survedance Test Procedures and Survedlance and Cahbration Test resu:ts, l' age 27of83

Chapter 15 Accident Mitigating Systems [',","'j,7"

1. di"" 8' 9/l&47 ICAVP (Systems)

SFSTE3fLESCRIPTION: ENGINEERED SAFEGUARDS ACFUATION SYSTEM (ESA-3407A) PARAMETER DESCRIPTION INPUY ASSUMPFION AFFF.CTED ACCIDENTS SAFETY ANALYSIS REFERENCES kbhi&M 5'.'.W? ?CI.ETgfh~ h sevarr?? ?????$!$?$Nb?hl'b NEU-96-623 I Steam Generator Water Level liigh steam generator level at 100% of rurrow range ' v.fwater System Malfunctio..s that Result span for closure ofisolation valves (7.0 s delay) and .a.r, Increase in Feedwater Flow (FSAR FSAR Table 15.0-4 turbine trip (2.5 s delay) !$. I.2) l'errylcation Jfethod: Verification will be accornplished by a review of Technical Specifications, survei. lances, calibration procedures, and test retufts. FSAR Table 15.2-1 1 Low Low level at 10% of narrow ruge span,60 s Loss of Normal Feedwater Flow (FSAR delay for AFW flow to steam generator 15.2.7) NEU-96-623 l'ersylcation Sterhod: Venfy by reviewing Technical Specification regulements, Surveillance and Cahbration procedures and test results. FSAR Table 15.2-1 1 Low Low level at 10% of narrow range span,62 s Feedwater System Pipe Break (FSAR delay for AFW flow to steam generator 15.2.8) NEU-96-623 l'crification Aferhod: Venfy by reviewing Technical Specification requirements. Survei!!ance and C'libration procedures and test results. I E D E $d5 NEU-96-623 I pressure Steam line isolation occurs at sepoint. (Below P-II) Steam Systen. Piping Failure (FSAR 15.1.5) Iligh begativ Rate l'erificarlon AIctkod: Venty by reviewing Technical Specification requirernents. Surveillance and Cahbration P '"res, and test resu'*s. ~ Page 23 of83 e

y. Chapter 15 Accident Mitigating Systems [,'l,',','lll',']lf ""^'""'8' 9/1597 ICAVP (Systems) SFSTElfDESCRIPTION: ENGINEERED SAFEGUARDS ACTUATION SYSTEM (ESA-3407A) PARAMETER DESCRIFTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Table 15.2-1 i Low Compensated Steam Line Safety injection and steam line isolation signal at Feedwater System Pipe Break (FSAR Pressure 658.6 psig,2 seconds delay 15.2.8) 90NE*-G-0075,. Vantage 511 Fuci l'enyicarisn Aferhod: Venfy by reviewing Technical Specification requirements, Surveillance and Calibration Procedures, and test results. NEU-96-623 1 Safety injection and steam line isolation signal at Steam System Piping Failure (FSAR 15.l.5) 658.6 psig. Safety irejection actuation logic initiates AFW flow and FW isolation. renyication Method: Verify by res.gechnical Specification requirements, Surveillance and Calibration Procedures, and test results r=y==-=%_ ~ggg =lW@yes _ _==_ -- - - wpm.. Aw-

==mn-ms;2,+ggr:.nnwTainm:rO3,,=.9M5g cre=q-- _igJ .r u _al --__ m ---m =mm mz m au onu_ am. - i Spurious Actuation Manual termination of spurious SI injection. Table Inadvertent Operation of Emergency Core NEU-96-623 (pg 15.5-4) 15.5-1,600 sec. Requiring operator action Cooling System During Power Operation (FSAR 15.5.1) l Ver!/icarlon Method: An Abnormal Operating Procedure should be in place to mitigate this spurious function. Procedure should identify plant conditions and pressurizer parameters to be monitored for manual termination. i l S Page 29 of83

J. ' "a *i""' 8' Chapter 15 Accident Mitigating Systems Mihtunc !!nis 3 9ffg ICAVP (Systems) SYSTElfDESCRIPTIO.V: EVENT RELATED (Radiologicel consequences - no associated mitigating system) PARAMETER DESCRIFTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 'Mgfmf ggt' T yr"?'"'77*wT?iRYr"!"="'"mW"fEyLTpmrmcem+~~~eWwmm:m T*W7b ?"yf (** i T" ' W- . u w _ _tittus M ~ w =; m * @ *=' + M+W ~H- " %d.ar" wAu k m "= 'w=1s w 2Li Offsite Dose Due to Steam FSAR Table 15.6-5 Steam Generator Tube Failure (FSAR FSAR Table 15.6-5 i Generator Tube Rupture (Ell.) 15.6.3) Verification Method: Verify by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with Table 15.0-8. l l l Page10 of83

puure.. i--...,. -.. '2'----- -t--- Chapter 15 Accident Mitigating Systems

',"l;"y'],';""

"a**" 8' O'IL97 ICAVP (Systems) SYSTElf DESCRIPTION: EVENT RELATED (Radiological consequences - no associated mitigating system) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES ~ ?. ; Af "" L Q!9'**QMyQ:l v. ') ~ M J : su Tr y.. ~,~ 's M G *% ~. n .,xe. e. TotalGaseous Activity (E) Largest liquid tank inventory. Radioactive Liquid Waste System Leak or FSAR Section 15.7.2, page 15.7-2 Failure (Atmospheric Release)(FSAR 15.7 2) Versykstbn Method: Verify by reviewing design operating conditions for other tanks containing radioactive liquids and located in the plant yard. Total Liquid Activity (E) Largest liquid tank inventory Liquid Containing Tank Failure (FSAR FSAR Section 15.7.2, page 15.7-2 15.7 3) Vers]ication Method: Verify by reviewing design / operating conditions for other tanks containing radioactive liquids and located in the plant yart --?revwla'tsennummisarrce:::mw=mMwnew??wm=7m*coNom!rSemi Line Size (E) No instrument lines connect to the RCS that Failure of Small Lines Carrying Primary FSAR Section 15.6.2, page 15.6-3 i penetrate containment Coolant Outside Containment (FSAR 15.6.2) FSAR Section 15.0, page 15.0-l Versfication Method: Review P&lDs for RCS and verify that no instrurnent lines, that penetrate containrnent directly connect with the RCS. i

• 9

. ;., _ 3 y, ....-.7 .p. ,qr Efficiency (E) Operating without filtration Radioactive Gaseous Waste system Failure FSAR Section 15.7.1, page 15.7-1 I (FSAR 15.7.1) Verificasion Method: Venfy by reviewing the supporting dose calculation to deternune if credit for filtration was taken. Page 31 of83

Chapter 15 Accident Mitigating Systems 8'dd*" 8' Milstone t mit J ICAVP (Systems) SYSTElfDESCRIPTION: EVENT RELATED (Radiological consequences - no associated mitigating system) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES

a e ; e,;.;, an.w:,,. f. _;; : :4:m esmanymy.9

.},,,

;; q ;- m.,. n

. m y, ; w.;:..; m.p,,

.., - 2, ....ub Containment Isolation (E) Provisions of GDC 55 for containment isolation are Failure of Small Lines Carrying Primary FSAR Section 15.6.2, page 15.6-3 met and ensure that a sample line break will not Coolant Outside Containment (FSAR 15.6.2) exceed the limiting letdown line break size of 152 gpm for 30 minutes. l'erification Afethod: Venfy that the potential for release from the pressurizer liquid and steam space sample lines is within the letdown liae break assumed. Check calculations of release rate for this versus the bounding scenario of 152 gpm CVCS letdown line break. Venfy isolation provisions meet GDC 55.

• " Y -e t ' v W,n**~

"*e. m g g m-- ~ + A _=;~+- ~m p . _~_.. %.:..... ~

,. m

-c w+md

?... W. -i. I m M M_ Lk ...: : a Bypass Leakage (E/L/C) Secondary containment bypass leakage is Loss-of-Coolant Accidents Resulting from a FSAR Section 15.6.5.4, page 15.6-24 I l documented per Table 15.6-9 and FSAR page 15.6-Spectrum of Postulated Piping Breaks FSAR Table 15.6-9 24 Within the Reactor Coo' ant Pressure Boundary (FSAR I5.6.5) Venfication Afethod: Verify that the analysis of bypass leakage has identified all of the bypass leakage paths and models them appropriately in the post-LOCA dose analyses. / ^ 2-k.we %s+4c:q ~ [. =. i ' :,,} ?..}~~j f Q ;- ~~.(7-T 7 ' %. f r 7 Cask Integrity (E) Maximum height oflift for spent fuel cask < 30 ft Spent Fuel Cask Drop Accidents (FSAR FSAR Section 15.7.1, page 15.7-6 above any hard surface 15.7.5) Ven7scation Afethod: Venfy by reviewing the design and operating procedures for moving the cask to assure no Inft above 30 ft. Page32 of83

"*i""'" Chapter 15 Acc: dent Mitigating Systems Milstone t'nis J g ICAVP (Systems) SYSTElfDESCRIPTION: EVENT RELATED (Radiological consequences - no associated mitigating system) PARAMETER DESCRII" TION INPUT ASSUMl" TION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES Cask Integrity (E) Qualified lift height > 30 feet above a hard surface Spent FucI Cask Drop Accidents (FSAR FSAR Section 15.7.1, page 15.7-f I5.7.5) Venylcation Method: Verify by rcviewing the cask certification. 'N._ . 7'6"My~a"r6uTs*~"W,Y5,35"Ef@l"f97hT~~_0EE737@n"3E3. _j? T P 7 I"ST_'73 E ~ n wa a FSAR Section 15.4.8 I OtTsite Dose (E/L) FSAR Table 15.4-4, 15.0-10 Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSAR 15.4.8) Venylcation Method: Verify that input assurnptions are consistent with plant design and operation and that dose calculations are consistent with results of Table 15.0-8. -.__._ ~ . - _ _ _ _ _ _. s y e < # 3_g. y4wu.? -g@.g_past;; .._.,.m, . _y ,,,,,e_' x.,,,_Om,g - c 3 s ; %c % _gg _CORFONENT DESCRIPTION;Tari_ou 4;W FSAR Section 15.0 I Atmospheric Dispersion Data Estimated X/Q's per Table 15.0-1 i Loss-of-Coolant Accidents Resulting from a (E/L/C) Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR I5.6.5) Venylcation Method: Venfy that the atmospheric dispersion data presented in Table 15.0-11 is appropriate for MP3 and is used in the post-LOCA dose calculations. Page 33 of83

,... /.,,., - --..___ _-- - _ Chapter 15 Accident Mitigating Systems "",','"",,],'f" "ad*" 8' Cf1597 ICAVP (Systems) SYSTEM DESCRIPTION: EVENT RELATED (Radiological consequences - no associated mitigating system) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Table 15.6-10 Control Room Dose (C) The post-LOCA radionuclide inventory is a Loss-of-Coolant Accidents Resulting from a percentage of the inventory available in the reactor at Spectrum of Postulated Piping Breaks FSAR Table 15.6-14 the time of assumed fuel damage as stated in the Within the Reactor Coolant Pressure FSAR Table 15.6-15 references Boundary (FSAR 15.6.5) FSAR Table 15.6-16 FSAR Table 15.6-17 FSAR Table 15.6-18 FSAR Table 15.6-19 k' FSAR Table 15.6-20 Verificasten Meshed: Verify the control room habitability calculation uses the referenced tables a id supports the doses l identified in 15.6-13. FSAR Table 15.6-12 Assumptions per Table 15.6-12 Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR I5.6.5) Verificarian Meskod: Verify that assumptions listed in Table 15.6-12 reflect the plant design, operating and testing characteristics, and are properly incorporated into the post-LOCA control room dose assessment per Table 15.6-13. FSAR Section 15.0.9.1 Post-LOCA radiological dose ccnsequence Loss-of-Coolant Accidents Resulting from a calculation uses an inventory consistent with the TID-Spectrum of Postulated Piping Breaks FSAR Section 15.6.5.4 14844 assumption and MP3 core size Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) Verificasien Methed: Verify that a calculation uses the inventories in TID-14844 and the MP3 core size to estimate the initial core radioisotope inventories for the post-LOCA dose estimates. Page 34 of83

Chapter 15 Accident Mitigating Systems "dd*"' 8' Mdssone tr it 3 a gj g ICAVP (Systems) SYS'IElf DE7C#trTION: EVENT RELATED (Radiological consequences - no associated neitigating system) PARAMETER DESCRitTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Table 15.0-8 CVCS Letdown Line Break Dose Per FSAR Table 15.6-2 Failure of Small Lines Carrying Primary (E) Coolant Outside Containment (FSAR 15.6.2) FSAR Section 15.6.2, page 15.6-4 FSAR Tabic 15.6-2 Vertpcstles Merked: Venfy that each of the assumptions in the FSAR text and Table 15.6-2 is met and incorporated into the design basis calculation. In the case of conservative system alignment assumption, verify that the calculation takes no credit for those systems. Verify dose calculation resuits consistent with Table 15.0-8. FSAR Section 15.6.5.4 I Dose Radiological Parameters Assumptions per Table 15.6-21 Loss-of-Coolant Accidents Resulting from a f Spectrum of Postulated Piping Breaks FSAR Table 15.6-2I Within the Reactor Coolant Pressure l l Boundary (FSAR 15.6.5) Verspcasion Merhod: Verify that assumptions in (ed in Table l's.6-21 reflect the plant design, operating and testing characteristics, and aie poperty incorporated into the post-LOCA TSC dose assessment in Table 15.6-22. Event termination (E) <=- 30 minutes Failure of Small Lines Carrying Prima y FSAR Section 15.6.2, page 15.6-3 Operator action Coolant Outside Containment (FSAR 15.6.2) NRC original SER for MP3 (8011984), Section 15.6.2 Vers ication Method: Venfy that procedures exist to support the operator action to isolate the limiting letdown system break f within 30 minutes using the area radiation monitors. l Terminated within one hour by operator action Radioactive Gaseous Waste system Failure FSAR Section 15.7.I, page 15.7-1 (FSAR 15.7.1) Verificasion Method: Verify by reviewing the operating procedures to determine if there is a logical procedure to identify the source of tha release and terminate the release. Page35 of83

Kerwo r or Chapter 15 Accident Mitigating Systems Northeast Utilities Cf1L97 sti,,,,,, c.,,3 ICAVP (Systems) STSTElfDESCRIPTION: EVENT RELATED (Radiological consequences - no associated mitigating system) PARAMETER DESCRll' TION INPlTF ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES Event termination (E) Operator manually isolates the faulted steam Steam Generator Tube Failure (FSAR FSAR Section 15.63 generator (including SG blowdown isolation) and 15.6 3) NRC SER for MP3 Supplement 4 cools down to equalize RCS/SG pressure within 30 minutes to minimize release rates Veripcasion Method: i) in original MP3 SER Supp.4, the NRC indicated that since NU committed to meet a currently evolving WOG initiative, the SGTR event results were acceptable. FSAR Chapter 15.6.3.4 describes the WCAP that resulted from that effort. Verification of NU activities to meet the WOG guidelines assumptions in WCAP-11002 and 10698 is necessary. Verify completion of NRC review of WCAP-10698. Offsite Dose (E) FSAR Table 153-3 Decrease in Reactor Coolant System Flow FSAR Table 15.3-3 Rate (FSAR 153) Venfication Meshod: Verify by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.0-8 FSAR Table 15.1-3 Steam System Piping Failure (FSAR 15.1.5) FSAR Section 15.13 i Venycasion Meshed: Verify by review of the vendor parameters and associated accident dose analyses. Offsite Dose (E/L) Assumptions per Table 15.6-9 Loss-of-Coolant Accidents Resulting from a FSAR Tabic 15.6-9 Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure I Boundary (FSAR 15.6.5) Verification Method: Verify that assumptions listed in Table 15.6-9 reflect the plant design, operating and testing characteristics, and are property incorporated into the supporting dose calculation to assure consistency with FSAR Table 15.0-8. Page 36 of83

""6 '"" 8' Chapter 15 Accident Mitigating Systems [,",",*,"I'fl" 9/19/97 ICAVP (Systems) SYSTElf DESCKIPTION: EVENT RELATED (Radiological consequences - no associated mitigating system) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES Offsite Dose Due to ESF Per Table 15.6-9 Loss-of-Coolant Accidents Resulting from a FSAR Table 15.0-8 I Leakage (E/L/C) Spectrum of Postulated Piping Breaks i j Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) Vcnyication Method: Verify that the supporting dose calculation is consistent with Table 15.0-8. I Offsite Dose Due to FucI FSAR Table 15.7-8 Design Basis Fuel flandling Accidents FSAR Section 15.7.4.2.1, page 15.7-C llandling Accident (E) (FSAR 15.7.4) i Venylcation Aferhod: Verify by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.0-8. Offsite dose due to liquid waste FSAR Tables 15.7-4 and 15.7-5 Radioactive Liquid Waste System Leak or FSAR Section 15.7.2, page 15.7-2 system leak or failure (E) F,ilure(Atmospheric Release)(FSAR 15.7.2) Ven]Ication Method: Venfy by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.0-8. Offsite Dose due to Main Steam FSAR Table 15.1-3 and 15.0-2 Steam System Piping Failure (FSAR 13.1.5) FSAR Table 15.1-3 Line Rupture Vcnyication Method: Venfy by assuring each parameter is consistent with plant design and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.0-8. l Page37 of83

xevision: or Chapter 15 Accident Mitigating Systems Northeast Utilitics 9/l&97 3,,,,,,,, c i, 3 ICAVP (Systems) SFSTElf DETCRIPTION: EVENT RELATED (Radiological consequences - no associated mitigating system) PARAMETER DESCRII" TION INPUT ASSUMI' TION AFFECTED ACCIDENTS SAFETY ANnLYSIS REFERENCES Offsite doses due to waste gas FSAR Tables 15.7-2 and 15.7-3. Radioactive Gaseous Waste system Failure FSAR Section 15.7.1, page 15.7-1 system failure (E) (FSAR I5.7.1) VersTscation Afethod: Venfy by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.0-8 Offsite liquid concentrations due FSAR Table 15.7-4 and 15.7-5 Liquid Containing Tank Failure (FSAR FSAR Section 15.7.3, page 15.7-3 to liquid waste tank failure (E) 15.73) Verification Afethod: Verify by assuring each parameter is consistent with plant design and operation and by reviewing the supporting dose calculation to assure consistency with FSAR Table 15.7-7 1 Steam Generator Tube Failure Ts.51e 15.6-5 List of parameters Steam Generator Tube Failure (FSAR FSAR Table 15.6-5 Dose (E/L) 15.6 3) Verification Aferho.t: Table 15.6-5 and the section 15.6.3 text cite many parameters used in the estimation of the radiological dose consequences. Verify that each of these assumptions is appropnate per the Technical Specification limits and normal plant operations preceding the event. Venfy dose calculation results consistent with Table 15.0-8. Waste Gas System Release (E) Bypass of the waste gas filter is the worst possible Radioactive Gaseous Waste system Failure FSAR Section 15.7.1, page 15.7-1 release from the waste gas system (FSAR 15.7.I) Veripcation Afethod: Verify by reviewing system description to determine if there is a larger potentia! source for release. Page 38 of83

Chapter 15 Accident Mitigating Systems [',','","',[ ,l "*i"" 8' 9/l&97 ICAVP (Systems) SYSTElf DESCRIPTION: EVENT RELATED-REACTOR CORE PARAMETER DESCRifTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES .. ;; ; y,.wg e s l.g 3 7....

5~..,.,+

p._.

c. - ( y 7;.

Core Loading implementation of quality control during core loading Inadvertent Leading and Operation of a NEU-96-623 Section 15.4.7, page 15.4-23 Fuel Assembly in an Improper Position (FSAR 15.4.7) Verificarlos Method: Verify correct fuel loading by review of procedures implemented during core loading.

---

~.--m.m-s.

m__.._________..._ ...-.-..m. NEU-96-615 llot Leg Recirculation Criteria Operator action to transfer to hot leg recirculation Loss-of-Coolant Accidents Resulting from a within 9 hours Spectrum of Postulated Piping Breaks FSAR Section 15.6.5, page 15.6-12 Widiin the Reactor Coolant Pressure Boundary (FSAR 15.6.5) Venfication Method: Venfy that procedural instructions exist to ensure that the transfer to hot leg recirculation occurs within the 9 hour time period cited in the FSAR (page 15.6-12). Verification should include entrance cr teria as well as appropriate re-alignment and confirmation actions Operator Action Sufficient indication and procedures to perform Loss-of-Coolant Accidents Resulting from a FSAR Section 15.6.5.2, page 15.6-12 transfer to post-LOCA cold leg recirculation node of Spectrum of Postulated Piping Breaks ECCS operation Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) Venficasion Meskod: Verify procedures and supoorting indication exists to support the operator action to transfer the ECCS to the post-LOCA cold leg recirculabon mode of operation. Event timing must be considered. Page 39 of83

""h'""' 8' Chapter 15 Accident Mitigating Systems Milstoac Unis 3 gg ICAVP (Systems) SYSTElfDESCRIPTION: FEEDWATER (FWS-332I A) PARAMETER DESCRif710N INPUT ASSUMITION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES

m. y. :pw

,77 y,.., s rpz z 7.. n, 3 r. - q.. '...p'.- 'A L' ., *

• q,.: * ' L.

E' '. s ! '. *?*s.- & _4, feedwater Flow A failure of a feedwater control valve at zero power Feedwater System Malfunctions that Result 90NE*-G-0075 (pg 5-26) will result in a step increase in feedwater flow to one in an increase in Feedwater Flow (FSAR NEU-96-623 steam generator not to exceed 200% of nominal 15.1.2) feedwater flow. (FSAR 15.I.2.2, Case 2, item 2) renAcation Jferhod: Verification of the system flow performance will be accomplished by review of system and component calculations and performance test results. A failure of a feufwater control valve at full power Feedwater System Malfunctions that Result 90NE*-G-0075 (pg 5-26) will result in a step increase in feedwater flow to one in an increase in Feedwater Flow (FSAR NEU-96-623 I stcain generator not to exceed 140% of nominal 15.1.2) l feedwater tL w. (FSAR 15.1.2.2, Case 2, item 1) VenJIcation 3feskod: Verification of the system flow performance will be accomplished by review of system and component calculations and performance test results. Ed '.lliR .: C M,'i?. Z.'~~ T T.1 -T ~ MQllC i ~ .I ~ ~ ' ~. ' ' ~ N ' Y *i"M FSAR Table 15.2-1 Closure Time 7 seconds on receipt of steam line isolation signal Feedwater System Pipe Break (FSAR 15.2.3) 90NE*-G-0075, Vantage 511 Fuel Verification Afethod: Venfy by reviewing technical specification requirements, survei!!ance tests and test results. t Page 40 of83

Chapter 15 Accident Mitigating Systems [*,7',",",,'];'7" ""6'""' 8' 9/lLQ7 ICAVP (Systems) SYSTElf DESCRIPTION: FEEDWATER (FWS-3321 A) PARAMETER DESCRIFFION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES ~

' q 2K

%, 7 *. h b G r((R,.

• ' e +? ~ '.

NEU-97-537 Feedwater Flow (0-2 hr) 771,277 lbm; (2-8 hr) I,505,487 lbm for N Decicase in Reactor Coolant System Flow loop and N-1 loop Rate (FSAR 15.3) FSAR Table 15.3-3 l'enylcation Method: Venfy through Surveillance Test Procedures and Surveillance Tests or system flow calculations l l Page 41 of83

l 1 i ""6'*"' 8' Chapter 15 Accident Mitigating Systems Manoac Uni: 3 yyg ICAVP (Systems) SYSTElf DESCRIPTION: FUEL BUILDING VENTILATION SYSTEM (RPV-3314C) i PARAMETER DESCRIITION INPUT ASSUMFTION AFr'ECTED ACCIDENTS SAFETY ANALYSIS REFERENCES e p _, . ( (. FSAR Table 15.7-8 Filter Efficiency (E) 'Ihe efficiency for organic and inorganic iodine is Design Basis 1 uct flandling Accidents 95%. (FSAR 15.7.4) i j Ven]icarton Method: Venfy by reviewing Technical Specification requirements for filter testing, Surveillance Test Procedures and test results. I i i -~- ~ ~ w.~ v ~.~ -" ~ ~ m:~ r......x 2' "5 .:, r.. ' ' W rr. y;.T=.~, ;. :-,,m..v.- ..........s..%. ~. m-1 FSAR Section 15.7.4.2.1, page 15.7-f j Operating Status (E) The system is operating in filtered mode when fuel is Design Basis Fuel llandling Accidents j being moved. (FSAR 15.7.4) i j VenJIcstion Method: Verify by reviewing Technical Specification requirements and implementing operating procedures for i moving fuel. i i 1 l 5 i i A Page 42 of83 1 -n,- ,-es, ..w-.- -w-r --w., y 2.., w

---

w--

i !s Chapter 15 Accident Mitigating Systems ""6'*"' 8' m v.n g,w ICAVP (Systems) SYSTEMDESCRIPTION: HICH AND LOW PRESSURE SAFETY INJECflON SYSTEM (HPI-3308, LPI-3307A) PADAMETER DESCRIPTION INPUT ASSUMFflON AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES . 1:: Flow Capacity Adequate borated water is provided to keep the core Feedwater System Pipe Break (FSAR FSAR Section 15.2.8.2, page 15.2-20 covered 15.2.8) 90NE*-G-0075, Vantage 511 Fuel ) Venfication Method: Venfy by reviewing design requirements, Technical Specificadon requirernents, Surveillance Procedures and test results. k. f Pump Flow Manually shut down Si pump prior to solid Feedwater System Pipe Break (FSAR FSAR Section 15.2.8.2, page 15.2-19 pressurizer condition 15.2.8) Ven]Icstion Method: Venfy that operatirtg procedures identify parameters to be monitored and steps required to manually secure St. FSAR Table 15.2-1 Startmg Time 2 seconds from Si signal with offsite power Feedwater System Pipe Break (FSAR i 17 seconds from Si signal with loss of ofTsite power 15.2.8) 90NE'-G-0075, Vantage 511 Fuel 1 Verification Method: Verify by reviewing design requirernents, Technical Specification requirements, Surveillance Procedures i and test results. Venficabon of DG performance will be accomplished by a review of surveillance procedures and test results for the DG and ESF load sequencers. FSAR Section 15.4.8 Flow stans within one minute after the break Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSr.R 15.4.8) VerJ/lcation Method: Venfy by reviewing operating procedures and surveillance tests. D Page 43 of83 L ~_

Chapter 15 Accident Mitigating Systems l,"**,","j','!"" 8*'*a: 8' D"lL97 ICAVP (Systems) SYSTElfDESCRIPTION: HICH AND LOW PRESSURE SAFETY INJE' *lON SYSTEM (IIPI-3308, LPI-3307A) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 90NE*-G-0075 Section 5.1.3.4.2 I System Flow Full flow is achieved within 42 seconds after inadvertent Opening of a Steam Generator accident with and without loss ofoffsite power. Relief or Safety Velve Causing a Depressurization of the Main Steam System (FSAR 15.1.4) Verification Method: 1) Venfication of the sequencer performance will be accomplished by review of calibration and survei'iance test results.

2) Diesel performance will be verified by a review of surveillance procedures and test results from diesel test and ESF load sequencing tests.

NEU-96-623 1 Full flow is achieved within 42 seconds after Steam System Piping Failure (FSAR 15.1.5) accident with and without loss of offsite power. Vers

• cation Method: 1) Venfication of the sequencer performance will be accomplished by review of calibration and surveillance test results.

2) Diesel perfom1ance will be venfied by a review of surveillance procedures and test results from diesel test and ESFload sequencing tests.

Page 44 of83

Chapter 15 Accident Mitigating Systems [",','ll,,j'l "*'*"8' CflOO7 ICAVP (Systems) SYSTElf DESCRIMION: MAIN STEAM SYSTEM (MSS-3316A) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES u l. ~ . 'fr 16 E W ' A +TK ..MWm 5 a'i -, s p 9ii ris.s ~ : J si.. t. i - . : ; r.1 FSAR Table 15.2-1 Closure Time 12 seconds from time sr. point is reached Feedwater System Pipe Break (FSAR 15.2.8) Veri /kasion Meshod: Venfy by reviewing Technical Specification requirements, Surveillance Procedures and test results. l NEU-96-623 (pg 15.1-9) Main Steam Isolation Valves close in less than 10 Inadvenent Opening of a Steam Generator I seconds. Relief or Safety Valve Causing a I Depressurization of the Main Steam System (FSAR 15.1.4) Verification Meskod: Venficabon of vaive performance will be accomplished by review of surveillance test results for the valve, Calsbration Procedures and test results for valve logic and review of the maintenance and modificatbn records. NEU-96-623 I Main Steam Isolation Valves close in I.:ss than 13 Steam System Piping Failure (FSAR 15.1.5) seconds. Veri /ication Method: Verification of valve performance will be accomplished by review of surveillance test results for the valve, Cahbration Procedures and test results for valve logic and review of the maintenance and modification records. a y7 .-.g3, _ e.,. : ..~m- .n- -r ; FSAR Section 15.2.1, page 15.2-1 Number No pressure regulators in system Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow (FSAR 15.2.1) Verifkaden Mested: Venfy by reviewmg system design to determine if the system contains steam pressure regulators. Page 45 of83

. _.. = =. Chapter 15 Accident Mitigating Systems l;,,*",",,'f,""" ""'5'**: 8' ICAVP (Systems) SYSTElfDETCRIPTiON: MAIN STEAM SVSTEM (MSS-33I6A) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 6 wr:Fw.i.e W/. . nx e 1 .e FSAR Table 15A-6 Operating Pressure All valves are fully open at 1240 psia Spectrum of Red Cbster Control Assembly Ejection Accidents (FSAR 15.4.8) Venfication Method: Venty code test datatresults SS?$55 M~. ~, _] l 90NE*-G-0075 (pg 5-33) Mass Flow Rate Maximum flow rate of 277 pounds per secoed at a Inadvertent Opening of a Steam Generator secondary system pressure of 1200 psia with offsite Reliefor Safety Wive Causing a Depressurization of the Main Steam System power. (FSAR 15.I.4) Verification Method: Venfication of the valve's performance will be accomplished by review of bench test results, specification, manufacturer's data and a review of maintenance and modificaticn records. g, g : g, ;,, 7 -,,, - .. ~ _... 7 73. y, qm. :.;y; - y ;, a,7 m..:. j ; y; :7,. mg.qsy, j.;g y 4.e-g %cp 3 FSAR Section 1533A I Closure Time Operator action to close valves in < or = 20 minutes Decrease in Reactor Coolant System Flow Rate (FSAR 153) Verification Method: Venfy using plant operating procedures and if possible by simulating the condition on the plant simulator Page 46of83

i i Chapter 15 Accident Mitigating Systems ""'d*"' 8' Milstone tf ait J j gyy7 ICAVP (Systems) SYSTEMDESCRIPTION: MAIN STEAM SYSTEM (MSS.3316A) PACAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES a.

.. w.:,y _,,

y ,,....x_..... v~. .~. Mass Flow Rate De total capacity of the relief valves are sized to Loss of Nonnal Feedwater Flow (FSAR FSAR Section 15.2.2.1, page 15.2-2 1 1 pass 105% of steam flow at rated power without 15.2.7) NEU-96-623 exceeding i 10% of steam system design pressure 90NE*-G-0075, Vantage 511 Fuel (1320 psia) Ven]icesian Meskod: Verify by reviewing design requirements Technical Specification requirements Surveillance Procedures and test results. The total capacity of the relief valves are sized to TurbineTrip(FSAR 15.2.3) FSAR Section 15.2.2.1, page 15.2-2 I NEU-96-623 pass 105% of steam llow at rated power without exceeding i 10% of steam system design pressure (1320 psia) I Verification Meskod: Venfy by revieanng design requirements, Technical Specification requirements Surveillance Procedures and test results(Permissive at P9) He total capacity of the relief valves are sized to Feedwater System Pipe Break (FSAR FSAR Section 15.2.2.1, page 15.2-2 1 j pass 105% of steam flow at rated power without 15.2.8) 90NE*-G-0075, Vantage 511 Fuel exceeding 110% of steam system design pressure (1320 psia) i Verificasion Meskod: Venfy by reviewing design requirements, Technical Specification requirements, Surveillance Procedures and test results. Page 47 of83

Chapter 15 Accident Mitigating Systems l,",','""'],"7' "' * '*" 8' 9/18M7 ICAVP (Systems) SISTEMDESCK1PTION: MAIN STEAM SYSTEM (MSS-3316A) PARAMETER DESCRIPTION INPUTASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES _..- ~ ' ;. i 7,' -.. ~; L,.z ? _ }. :3y TLt4 f g I

^

FSAR Section 15.2.2.I, page 15.2-3 I Closure Thne Closure Une <= 0.1 sec on turbine trip Loss of External Electrical Load (FSAR NEU-96-623 15.2.2) Penyicana.vMedad: venfy by reviewing any Technical Specification requirements, Surveillance Procedures and test results. 1 . L J 'N ..-'W.' l,:,vS 73fg gi4p NEU-96-623 I Closure Time Closure time > 0.3 second on less ofelectrica ioad Loss of External Electrical Load (FSAR (For the turbine trip to be controlling event) 15.2.2) FSAR Section 15.2.2.1, page 15.2-3 VenJJcaden Medad: Venfy by reviewing Technical Specifcation requirernents, Surveillance Procedures and test results. Page 48 oj83

Chapter 15 Accident Mitigating Systems [,["",'jf" "d'*"' 8' Cfl&97 ' ICAVP (Systems) SFSTEMDESCRIPTION: MAIN STEAM SYSTEM (MSS-3316A) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES -,7r,, _.,., 7 ;; 3 FSAR Table 15.I-3 Feedwater Flow 589,000 lbs for N and N-I loop operation (0-2 hours Steam System Piping Failure (FSAR 15.1.5) after accident): 1,019,000 lbs for N and N-1 loop operation (2-8 hours after accident) (Unaffected Steam Generator) I Ver!/ication Method: VerificaSon of this parameter will be accornplished by review of design documentation and calculations which establish the equipment v.)lume a;xi operating flows and calculations for the accident conditions. I FSAR Table 15.1-3 Flow 167,000 pounds (Initial Steam and Water Release Steam System Piping Failure (FSAR 15.I.5) From the Affected Steam Generator Over the First 30 Minutes of Accident) Verification MerAod: Verification of this parameter will be accomplished by review of design documentation and calculations wtuch establish the equipment volume and operating flows and cak.ulations for accident conditions. FSAR Table 15.1-3 417,000 pounds for N loop operation,433,000 Steam System Piping Failure (FSAR 15.1.5) pounds for N-I loop operation (0-2 hours after accident); 912,000 pounds for N loop operation, 912,000 poumis for N-1 loop operation (2-8 hours after accident) (steam release from unaffected srcam generators) Veri /icarlos Meskod: Verification of this parameter will be accomplished by review of design documentation and calculations which establish the equipment volums and operating flows and calculations fcr the accident conditions. Page 49 of83 o

Northeast Utilities Chapter 15 Accident Mitigating Systems Kevisio : ei 3,31,,,,a.,, 3 01397 ICAVP (Systems) SYSTEWDESCRIPTION: MAIN STEAM SYSTEM (MSS-3316A) PAGA."JETER DESCRII" TION INPUT ASSUMl" TION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Table 15.1-3 Flow 1300 pounds (lang Tenn Steam Release (0-8 hours) Steam System Piping Failure (FSAR 15.1.5) from effected steaar generator l'erification Method: Venfication of this pararreter will be accomplished by review of design documentation and cal'tlations which establish the equipment volume and operating flows and calculations for the accident conditions. NEU-97-537 I InitialInventory Liquid - 97,660 lbm/SG; Steam - 8,301 lbm/SG - N Cecrease in Reactor Coolant System flow l loop Rate (FSAR 153) FSAR Table 15.3-3 l Liquid - 99,717 lbm/SG; Steam - 7,73 I lbm/SG - N-Il*P VerJication Method: Venfy through review of the vendor parameters and associated accident dose calculations. l l FSAR Table 15.I-3 Steam (pounds per generator): 8,000 for N loop Steam System Piping Failure (FSAR 15.1.5) operation,7,600 for N-I loop operation. Liquid (pounds per generator) 103,000 for N loop operation, 1M,000 for N-1 loop operation Verification Method: Verification of this parameter will be accomplished by review of design documentation and calculations wtuch establish the equipment volume and operating flows and calculations for the accident conditions. FSAR Section 15.1.5.4 I Leak Rate Pnmary to secondary leakage assumptions for the Steam System Piping Fa;ture (FSAR 15.1.5) affected Steam Generators is 0347 gpm and unaffected Steam Generators is 0.653 gpm. Verification Meded: Verification will be accomplished by review of the accident dose calculations for this accident. Page 50 of33 x

Chapter 15 Accident Mitigating Systems l,""*,"',,'],'f" "*'*"'8' 9/1597 ICAVP (Systems) SYSTElfDESCRIPTION: MAIN STEAM SYSTEM (MSS-33I6A) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-97-537 Leak Rare I gpm for N loop and N-1 loop (Primary to Decrease in Reactor Coolant Syst m Flow Secondary) Rate (FSAR 153) FSAR Table 153-3 l l'enfication Method: Verify through Technical Specification, Survei!!ance Test Procedures and Surveillance Tests l l l l ) l Page 51 of83

Chapter 15 Accident Mitigating Systems ""*"" I"it 3

1. a'h18"' 8' Milstone t m g

ICAVP (Systems) SFSTE3fDESCRIPTION: NUCLEAR INSTRUMENTATION SYSTEMS (NME/N MI!NMP/NMS-340I) PARAMETER DESCRIPTION INPUT ASSUMFrlON AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES nW+.%%K} j. -QpfQ ~-' ' ; L N^W. '"?:. Y : -n Q,'7."e[Q T-[;[.f y_/ M." %p'y:;"f};;) L.- NEU-96-623 Shutdown Margin Alarm at setpoint 2.0 w hich requires operator action Chanical and Volume Control System Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) Verification Merkad: Verify that the operating procedures list the appropriate monitors and that there are proper procedures for alarm response. 1 K'~,. Ws. v. - : T ST'7 Pt :'K 1. ' j 'O;:. M 7 C' ~~ T ~~~ ~ M P: W W ~~~f~ W AWEf MWffM.9 NEU-96-623 Neutron Flux Visual indication requiring operator action Chemical and Volume Control System Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) Verification Iferhod: Venfy that the operating procedures list the appropriate monitors and that there are proper procedures for alarm response. 1 NEU-96-623 Audible count rate of flux requirit.g operator action Chemical and Volume Control System Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) Per7 cation Meskod: Verify that the operating procedures list the appropriate monitors and that there are proper procedures for alarm response. Page$2 of33

""^'""'8' Chapter 15 Accident Mitigating Systems Milstone Unit 3 gfy7 ICAVP (Systems) SYSTDIDFJCRIPTION: NUCLEAR INSTRUMENTATION SYSTEMS (NME/NMUNMP/NMS-3401) PARAMETER DESCRII" TION INPUT ASSUMFTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-96-623 Neutron Flux Alarm requiring operator action Chemical and Vo!ume Control System Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) Venficarlos Method: Verify that the operating procedures list the appropriate rnonitors and that there are proper procedures for alarm response. j Page $3 of83

nown: or Northeast Utilitin Chapter 15 Accident Mitigating Systems si,t, . u,,,3 D'ID97 ICAVP (Systems) SYSTDIDESCRIPTION: QUENCH SPRAY INCLUDES RWST (QSS-3309) PAR.OtETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES [ [DE M M M T h " M i!?95?A?i:l @ bfG Y Q P f? E N Ffi.M T ? E 7 C E ] ] Containment Pressure System features to support containment spray for Loss-of-Coolant Accidents Resulting from a FSAR Section 15.6.5.2, page 15.6-11 I containment pressure actuation at setpoint of 10 psig Spectrum of Postulated Piping Breaks NEU-96-614 (Ili-3) Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) V< 'ication Method: 1) Verify containtnent spray is consistent with FSAR Section 6.2.

2) Venfy Technical Specification and surveillance test procedures for OSS system.

l [MENIDESCRIPTiDNi?RifMikig WMS@ TEE"jN7'MinMHFS@EfRjfNMM[R35%!IE?f"W35J; S EKij ~ ~ 2 Concentration Boron concentration > 2000 ppm Chemical and Volume Centrol System FSAR Section 15.4.6 Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant (FSAR 15.4.6) Ven71 cation Method: Venfy that operating procedures shall provide instruction within CVCS configuration for borated water source. 4 Page $4 of83

Chapter 15 Accident Mitigating Systems ""#*"' 8' Milstonc Unit 3 gyp 7 ICAVP (Systems) SYSTEMDESCRIPTION: RADIATION MONITORS (RMS-3404) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES w3 3 y., py ; ;nq ygws ( s.. 7 g ;,,. 7 3 _, y g. . x.;., ; ;g,..,., y. ; ;.,g. g_ FSAR Section 15.7.4.2.2, page 15.7 Response Time (E) < 2 seconds to send closure signal to containment Design Basis Fuct llandling Accidents purge (FSAR 15.7.4) Verification Merkod: Verify by reviewing the logic diagrams for the monitor signals, identifying the expected dose rate at the monitor, the monitor setpoint and the associated Surveillance Procedures, and the monitor response time specified in vendor documents. ?1t@*~W'IfMit Detection of CVCS Leak (E) Area radiation monitoring and leakage detection Failure of Small Lines Carrying Primary FSAR Section 15.6.2, page 15.6-3 i available and support detection and isolation of the Cociant Outside Containment (FSAR 15.6.2) design basis CVCS 152 gpm leak within 30 minutes. Verification Method: Venfy that thwe are area radiation monitors to detect and mitigate this leak Verify monitor response time and associated.Msting is cmsistent with the bounding CVCS leak scenario. 1 i Page55 of83

...--__-.? i. __.___.m'_...._ --- - --'--'-.,.. _.. "'i'-.-'"---------'------- Chapter 25 Accident Mitigating Systems >Unis 3 l l gy g l ICAVP (Systems) SYSTEMDESCRIPTJON: REACf0R COOLAPtTSYSTEM (RCS-33GI) FARAMETER DESCRirTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 'Ch[Ds5CRFTIOSETMM ~ l Z7 ,;g4 m ' ; 31.M FSAR ocr-15.4.3 Spectnan of Rod Chater Control Assembly Welllategrity Periodic P'- Eject = Aerh(FSAR 15.4.3) rer#Lai.s mesh.e verity by reweang or procedwes and 'e data. _._ u t M.r d L 1 .DESCHrTIOsi 'N6'M@@ ^ 4 __m. Concentration Boron concentration of the isoi u : 'oop is greaser Startup of an inactive Reactor Cooiant NEU-96-623 Section 15.4.4, page 15.4-15 than or equal to the boroa eaare..arion of the Pump at an Ir arrect T+-- and W=3 oops,or greaser than 2600 ppan whichever Baron Cn= centration (FSAR 15.4.4) l is less. Ver#icasies.w-Rewew Techncal W m, Survedlance Procedures and Survedlance Tests for isolated and operatmg reactorcoolantloops. Isalation Tune 30 miname< 80 isolase rertar coolant loop to f=ned Feedwater Systma Pipe Break (FSAR FSAk Sectaan !5.2.82.,mge 152-N I ce= generator 15 2.3) FSAR Table C 659 rer#icades W VerWy by rn ;is operaeng procedures to assure pramtres are in place to h the faulled steam generator when a hIme break occurs. ragenop3 L

Chapter 15 Accident Mitigating Systems Z" 'j,"" "" 8' Cy1197 ICAVP (Systems) SYSTEWDESCRIPTJON: REACTOR COOLANT SYSTEM (RCS-33el) FARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-96-623 Section 15.4.4, page 15,v-Temperature Temperature of cokt leg of invtive loop is within Startup of an inactive Reactor Coolant 15 20*F of the highest cold Icg temperature of the Pump at an lacorrect Temperature and operating loops. Boron C-entration (FSAR 15A.4) l'ervicarios Mdd-Review Tectuucal Whm, Survedance Procedures and Survedlance Tests for isolated and operating teenr coolantloops 1 3e ., & ug s ,, _ ~ _, _, FSAR Section 15.5.l.2.D Owm AutoControl Auto Control Selected by operator acten inadvenent Operation of Emergency Core Cooling Sysicm Du ing Power Operation (FSAR 15.5.1) 8'er#icanes M ed ed: Operabon of PORVs assumed to result in the pressunzer pressure not reaching the PSRV set pressure (2500 psia +/- 3%. Tech Spec 3A22). Venfy by confirrrang an operahng procedure is in place and W auto controlfor normaioperabon. -...y; 3 FSAR Section 15.5.1.2.D I Pressurizer Pressure Operanon of the PORVs tr<nh< in pressurtzer inadsenent Operation of Emergency Core pressure not reaching PSRV set pn:ssure Cooling System During Power Operanon (FSAR 15.5.1) Ferfcasies Wda Assumed avstnN to nwyana the transient. Pravent subcooled water flow through PSRVs in case pressunzer reaches water-solid state. Venfied by review of control tuning procedure and caleration records. Page57of33

Chapter 15 Accident Mitigating Systems Z*",.'7 8" ~ 8' asaar ICAVP (Systems) SYSTElf DESCXfrTJO.V: REACFOR COOULNT SYSTEM (RCS-33el) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES - _ - rm..:_;.; .; a qw xan

y..,

l Mass Flow Rate The total capacity of the safety vahes are sized to Feedwater Sprem Pipe Break (FSAR FSAR Section 15.233, page 15.2-7 I J prevent exceeding i 10% of reactor coolant system 15.2.3) 90NE*-G-0075, Vantage 511 Fuel l design pressure (2500 psia) s'e@smien Meded: Venfy by reveweg desagn requirements. Technca! When requirements. Surveillance and Cabbrabon Procedures, and test results. FSAR Secuon 15.23.2, page 15.2-7 I He total capacity of the safety vahes are sized to Turbine Trip (FSAR 15.23) prevent exceed =g i 10% of RCS design (2500 psia) 90NE*-G-0075, Vantage 511 Fuel rergicamine MM Venfy by reviewing de".ign requirements. Surveillance Procedures and test results. The total capacity of the safety vahes are sized to Loss of Normal Feedwater Flow (FSAR FSAR Section 15.23.2, page 15.2-7 I ) prevent exceeding i 10% of rnetar coolant system 15.2.7) 90NE*-G-0075, Vantage 511 Fuel l design pressure (2500 psia) r Veryicmaios Meded: Venfy by reviewing design requirements. Survesitance Procedures and test resul:s I Flow Rate Sufracient natural cirr'daian through coolant loop to Feedwater Sptem Pipe Break (FSAR FSAR Section 15.2.8.2, page 152-19 remove decay heat as analyzed in Table 15.2-2 15.2.8) NEU-96-623 recifice Men.d: Venry by reviewing stattup mmes. Page58 of33

Chapter 15 Accidern Mitigating Systems ""- 8 ' u.so a 3 g,g,7 ICAVP (Systems) SYSTEW DESCRIPTION: REACFOR COOLANT SYSTEM (RCS-330I) FARAMETER DESCRil" TION INPtJT ASSUMf" TION AFFECTED ACCIDENTS ' SAFETY ANALYSIS REFERENCES Flow Rate Maru.a! termmanon of flow Loss ofNormal Feedwater Flow (FSAR FSAR Section 15.2.7.2, page 15.2-14 15.2.7) NEU-96-623 Ver#kcles Meded: Verdy by reveanng operating procedures. Man m1 terminir= of flow Feedwater System Pipe Break (FSAR FSAR Section 15.2.8.2, page 15.2-19 1518) Ver#icaden Medad Verdy by reviewing operabng procedures for manual shutdown of rmetnr ::colant pumps FSAR Secuon 15.4.128, page 15.4-Number 3 reactor coolant pumps in mode 3 operation Uncontrolled Rod Cluster Control Asembly Bank Withdrawal From a Subcritical or NEU4-623 law PowerStanup Condition (FSAR 15.4.1) Veryicaden Meded: Venfcahon of the number of pumps in operation in Mode 3 will be accomplished by reviewing operating procedures and Techrucal Wtions to assure consistency wth the assumption. l Time Pump r=thwn Begins For accident concurrent with a LOOP, pump less ofNormal Feedwater Flow (FSAR . NEU-96-623 I raastAmin begins at < = 63 s. 15.2.7) rergic.de.Meded: venty by reveeng survemance procedures Page39of83

Chapter 15 Accident Mitigating Systems ""

• 8'

- om c7,,w ICAVP (Systems) SYSTEM DETCAIPTION: REACTOR COOLANTSYSTEM (RCS-330I) PARAMETER DFSCRIPTION INP1TT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES gmeerus1f,~ g j -f y . - (p4.tw v r.,p.. r. ', ;, 3.j sg.~~;>E ' - N ~ iRR

e. r.

Concentrataxi Postive moderator temperanse ci efficient Inadvertent Opening of a Pressurizer Safety FSAR S ction 15.6.1.2, page 15.6-2 or Relief Valve (FSAR 15.6.1) NEU-96-623 VerfrodenMedad: Venfy that poscve moderator temperature coeffcent assumed in FSAR Chapter 15.6.1 is conservatrve with regard to entre core fife. Review Technical Specification requirements and survedlance tests procedures. FSAR Section 15.4.1.2.3, page 15.4-3 Coolant Temperature The mn 3-= average coolant temperature is $57*F Uncontrolled Rod Nter Control Assembly Bank Withdrawal From a Subcrincal or NEU-96-623 Low Power Startup Condition (FSAR 15.4.1) Verfradon Merked: Venfication of the average temperature can be accomplished by review of the operating procedures for zero power operabon and Techncal S,*tM i Page 60cf83

Chapter 15 Accident Mitigating Systems ["U,'j"" Ka* 8' C11097 ICAVP (Systems) SYSTEMDESCKirTJON: RE/MCA COOLANTSYSTEM(RCS-3301) PARAMETER DESCRIPTION 1NIy! UMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES =- EEU-97-537 I Pnmary Cool.ue Imentory $20,000 lbm for N loop and 350,000 !bm for N-1 Decrease in Reactor Coolant System Flow loop Rate (FSAR 153) FSAR Table 153-3 5'erificarios Merhod: Venfy through review of the vendor parameters and assoczated accident dose calculations. i Page 61 of83 w

Chapter 15 Accident Mitigating Systems ""*" 8' Mddene i!ais 3 gyg ICAVP (Systems) SYSTEM DESCRIPTION: REACTOR HEAD-MISSILE SHIELD PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES e .m.. :9.. -,.. q y:,4,,g> ,gg .+ .,;.3,,r FSAR Section 15.4.8 Shield lategrity lasta!!cd post refueling Spectrum of Rod Cluster Centrol Assembly Ejecuon Accidents (FSAR 15.4.8) 3"crr]ication Merlsod: Venfy that there is a procedure to remove and insta!L l l [ 1 l l Page 62 of83

Chapter 15 Accident Mitigating Systems l,"j, *",';"'"" 8' ICAVP (Systems) SYSTEWDESCRIPTION: REACTOR PROTECTION SYSTEMS, RTS (RPS-3466) PARAMETER DESCRIPTION INPLTT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES . s c o r s,. r w - ~ + ,p - FSAR Table 153-1 ReactorTrip Set Point (RCS At pennissive power >P7,85% of RCS toop flow in Decrease in Reactor Coolant System Flow Flow) 1.0 sec. Gravity insemon of control rods per FSAR Rate (FSAR 153) 90NE*-G-0075 (pg 5-2I) Table 153-1. l'enykarisa Meu Venfication of system performance wiB be accomplished by a review ot i) Trip setpoint in Techncal Sm'--4i ii) review of Survedlance Test Procedures and Survedlance Tests associated with the logic ".or P7, flow transmitter logic & control rod logic At permissive > P8 Reactor trip at 92% of nominal Decrease in Reactor Coolant System Flow 90NE*-G-0075 (pg 5-21) flow in 0.6 sec. Gravity insertion ofcontrol rods per Rate (FSAR 153) NEU-96-623 FSAR Table 15.0-1 FSAR Table 153-1. l'enykardes Meded-Venfication of system performance wiB be accomphshed by a review of i) Trip setpoint in Technical Whno ii) review of Survenilance Test Procedures and Surveillance Tests xW with the logic for P8, flow transmeter logic & rod drop logic. iii) correlation between reactor coolant flow and trip setpoint iv) flow dehvered by RCS during startup tests and/or Surveillance's currently g3,3.__. x;_ _ y;, NEU-96-623 Axial Flux Difference Alarm requires operator acuan Chemical and Volume Control System Malfunction that Results in a Decrease in FSAR Section 15A.6 Boron Concentrauon in the Reactor Coolant (FSAR 15A.6) 5'erificades Wad-Venfy that the operahng procedures list the appropnate morutors and that there are proper procedures for alarm response. Page 63 of83

Northcaw U staties Chapter 15 Accident Mitigating Systems A n a s: ei ,,,,,,,, u, 3 9/lL97 ICAVP (Systems) SYSTEMDESCAfrTIO.V: REACTOR PROTECTION SYSTEMS, RTS (RPS-3406) 5/24 METER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 90NE*-G-0075 Table 5.1.14 Delay Time A delay time of 0.5 seconds for high and low Feedwater System Malfunctions that Result serpoint trip in a Decrease in Feedwater Temperature (FSAR 15.1.1) Ver#ication Meded: Venficabon of the instrument performance will be accomphshed by review of Techncal Specif. cations. Calabration and SurveBance Test Procedures and Survedlance and Cahbration Test results. l NEU-96-623 FSAR Table 15.04 0.5 second delay time for trip signal actuanon and Uncontrolled Rod Cluster Control Assembly RCCA release for high neutron flux (high setting)(N Bank Withdrawal at Power (FSAR 15A.2) 90NE*-G-0075 Table 5.1.14 and N-I loop operation) Ver#3 canon Meded: Venfcation of the reactor inp delay time wel be accomplisned by review of Techncal Specircations. Cahbration and SurveJiance Test Procedures and Survedlance and Cahbration Test resu!ts 90NE*-G-0075 Table 5.1.14 I A delay time of 0.5 seconds for high and low Steam System Piping Failure (FSAR 15.1.5) serpoint trip Verification Meded: Venficanon of the instrument performance win be accomplished by review of Technical Specifications. Cahbrabon and Surveinance Test Procedures and Surveillance and Calibration Test results 90NE*-G-0075 Table 5.1.14 I A delay time of 0.5 <eennA for high and low Inadvenent Opening of a Steam Generator setPoint trip Reliefor Safety Valve Causing a Depressurization of the Main Steam Sym (FSAR 15.l A) Ver#3 cades Meded: Venfrahnq of the instrument performance will be accomplished by review of Technical Specifcations, Cahbrabon and Survedlance Test Procedures and Survedlance and Cahbrabon Test results. Page 64 of83

Chapter 15 Accident Mitigating Systems l,*J" "",. """

• • 8' t'1L97 ICAVP (Systems)

SnTEMDESCRIPTION: REACIOR PROTECrlON SYSTEMS, RTS (RPS-3406) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDEN'IS SAFETY ANALYSIS REFERENCES 90NE*-G-0075 Table 5.1.1-4 I l Delay Time A delay time of 0.5 seconds for high and low Feedwater System Malfunctions that Result setPoint trip in an increase in Feedwater Flow (FSAR 15.1.2) l rergicasim Siedsd: Verfrahnn of the instrument performance will be accomplished by review of Technical Specifications, cahhraton and Surveitance Test Procedures and Surveillance and Cahbration Test resuP.s. NEU-96-623 Neutron Flux liigh setpoint i18*& of rated power for N loop; 895& n~nW1 and Volume Control System ofrated power for N-I loop Malfunction that Results in a Decrease in FSAR Section 15.4.6 Baron Concentration in the Reactor Coolant (FSAR 15.4.6) Versficerica Meded: Venfy that the operating procedures fast the appropriate monitors and that there are proper procedures for alarm response NEU-96423 FSAR Table 15 0-4 R ~sar trip at 1 I 8?& of full power - four loop Uncontro!!ed Rod Cluster Control Assembly operation Bank Withdrawal at Power (FSAR 15.4.2) 90NE*-G-0075 Table 5.1.1-4 R-tar trip at 89*& of full power-three loop operation Ver#icamies Meded: Venficaten of the reactor trip setpoint performance will be accomphshed by review of Techrucal Sm? 2-:-w, Cahbrabon and Survedlance Test Procedures and Surveillance and Cahbration Test re Page 65 of83

Chapter 15 Accident Mitigating Systems [*"" 'f"" "*'*" 8' CfiO97 ICAVP (Systems) SYSTElfDESCRIPTION: REACTOR PROTECTION SYSTEMS, RTS (RPS-3486) PARAMETER DESCRIFTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-96-623 FSAR Table 15.0-4 Neutron Flux Low serpoint for reww trip at 35% of full power Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Subcritical or 90NE*-G-0075 Table 5.1.1-4 Low Power Startup Condition (FSAR 15.4.1) Verification Meded: Verrtv atvvi of the reenr trip setpomt will be accomphshed by review of Technical Specifcations, r'ahhraton and Survemance Test Procedures and Cahbration and Surveillance test results 90NE*-G-0075 Table 5.1.1-4 Assumed high trip setpoint of 89*& and 1I8*& of Feedwater System htalfunctions that Result rated thermal power for 3 loop and 4 loop operation, in a Decrease in Feedwater Temperature respectively. A unmed low trip setpoint of 35'E of (FSAR 15.1.I) rated thermal power for 3 and 4 loop operation. Ver#icark m Med e.fr Venficaton of the instrument performance will be accomphshed by review of Technical Speofa.ations, Cahbration and SurveiBance Test Procedures and Surveillance and Cahbration Test results. FSAR Section 15.4.8 l Trip serpoint at higi. neutron flux ! 18% power (high Spectrun of Rod Cluster Control Assembly range serpoint) Ejection Accidents (FSAR 15.4.8) Ver#ication Ecded: Venfy by reviewing lechncal specification requirements surveillance and cabbration results ~0NE*-G-0075 Table 5.1.I-4 I Aunmed high trip serpoint of 89th and iI5% of inadvertent Opening of a Steam Generator rated thermal power for 3 loop and 4 loop operarmn. RelieforSafety Valve ('ammg a respectively. Assumed low trip serpoint of 35% of Depressuruation of the Main Steam System rated thermal power for 3 and 4 loop operarma (FSAR 15.1.4) Ver#icsales Mede fr Verrrratuvi of the instrument peihmance will be as.cinpEshed by review of Technca! Specifications. Cahbraton and Survesitance Test Procedures and Survesance and Cahbration Test results. Page 66 of83 i

w. v uw.

Chapter 15 Accident Mitigating Systems me.a.: #1 ,i,,_ tm,3 9/13D7 ICAVP (Systems) SYSTEWDESCz1FTJON: REACTOR PROTECTION SYSTEMS, RTS (RPS-3406) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Secuan 15.4.8 Neutron Flux liigh neutron flux 35?& power (Iow range serpoint) Spectrum of Rod Chister Control Assembly Ejectxxa Acci.ht< (FSAR 15.4.8) l'er#icasies Medad: Venfy by re. viewing techrucal WM requirements survesitance and cahbration results. 90NE*-G-0075 Table 5.1.1-4 I Asummed high trip setpoint of 89*& and i18*& of Steam System Piping failure (FSAR 15.1.5) rated thermal power for 3 loop and 4 loop operauon, respectively. Assumed low trip serpoint of 35*& of rated thermal power for 3 and 4 loop operation. l l'er#Icasion Medad: Venfcaton of the instrument pedum=>ce wiB be vervnplished by review cf Technical Specircations, Cahbraton and Survedance Test Precedures and Survedance and Cahbrabon Test results 90NE*-G-0075 Table 5.1.1-4 I A <uimM high trip setpoint of 89*& and 1I8*& of Feedwater System Malfunctions that Result rated thermal power for 3 loop and 4 loop operation, in an increase in Feedwater Flow (FSAR respectively. Asummed low trip wtpoint of 35?& of 15.1.2) rated thermal pow er for 3 and 4 loop prh I'cryiceries Meded: Venficahnn of the instrument performance will be accomphshed by review of Techrucal Spefientians, Cahbraton and Survedance Test Procedures and Survedance and Cahbration Test results. 90NE*-G-0075 Table 5.1.1-4 i For power level greater than P6 but less than P10 Feedwater System Matrueh< that Result remhr f+== exceeds serpoint. in a Decrease in Feedwater Temperature (FSAR 15.1.1) Ver#Icasies Medad: Ven6cabon of the irstrument performance we be accomphshed by review of Techrucal Specifcation requirements. Cahbraton and Survedance Test Procedures, and Cabbrabon and Surverwce Test recdk Page 67 of83

Chapter 15 Accident Mitigating Systems 7"",7 8' ICAVP (Systems) SYSTEMDESCAlrT10N: REACTOR PROTECFION SYSTEMS, RTS (RPS-3406) PARAMETER DESCRIPTION INPUT ASSUMPTION AI'FECFED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-96-623 FSAR Table 15.0-4 Neutron Flux Rate 0.5 secoad delay time for trip signal actn= rum and UncontroIIed Rod Cluster Control Assembly RCCA release Bank Withdrawal From a Subcritical or 90NE*-G.0075 Table 5.1.1-4 Low Power Startup Condition (FSAR 15.4.1) Verfication Meded: Venfication of the reactor trip delay time w2 be accomphshed by review of Technical Speficatms, Cdhrabon and Survedance Test Procedures and Cahbraton and Surveillance test results. FSAR Section 15.4.8 Trip setpoint at high rat.of flux Spectrum of Rod Cluster Control Assembly Ejecuan Accidents (FSAR 15.4.8) Veryication Meded: Venfy tre reveanng technical specifcation requirements survenilance ard calibration results. 90NE*-G-0075 Table 5.1.1-4 Delay time 7.0 seconds (The definition of delay time is given in Feedwater System Malfunctions that Result Note (a) of FSAR Table 15.0-4.) in a Decrease in Feedwwer Temperature (FSAR.15.1.1) Ver#Ic=da, Meded: Venfw-www of the instrutnent performance wm be accomplished by review of Techncal Specifications. Cahbrabon and Survedance Test Procedures and Surveillance and Cahbrabon Test results. 90NE*-G-0075 Table 5.1.1-4 1 j 7.0 seccads (The definition of delay time is given in inadvenent Opening of a Steam Gener.stor Note (a) of FSAR Table 15.0-4.) RelieforSafety Valve Faming a Dsp ization of the Main Steam S;mem (FSAR 15.1.4) Veryicades Meded: Venfr-Men of the instrument penb uusrica wi: be accomplished by review of Techncal Specifications. Cahbration and Survedance Test Procedures and Survemance and Cabbration Test results. Page 68 of33

= - - Chapter 15 Accident Mitigating Systems C 'j,"" 8' ICAVP (Systems) SITTEMDESCRtrTION: REACTOR PROTECTION SYSTEMS, RTS (RPS-3466) PARAMETER DESCRIFTION INPUT ASSUMPTION AFFECFED ACCIDENTS SAFET'.' ANALYSIS REFERENCES 90fIE*-G-0075 Table 5.1.1-4 I Delay time 7.0 seconds (The definition of delay time is given in Steam System Piping Failure (FSA1.15.I.5) Note (a) of FSAR Table 15.0-4.) Veryicarime Merbf: Ver# *4 of the instrument performance wiR be accomplished by review of Tecincal Specifications, Cahbrabon and Survedance Test Procedures and Sursediance and Cahbration Test results. 90NE*-G-0075 Table 5.1.!-4 DifferennalTemperature Assumed trip serpoints cited as variable in the Feedwater System Malfunctions that._sult i reference analysis (Figures 5.1.I-6 and 11.1-7) in a Decrease in Feedwa:er Temperature 90NE*-G-0075 Figure 5.1.14 (FSAR 15.1.1) 90NE*-G-0075 Figure 5.1.I-7 l 1 reryicaties Meded: Venficabon of the instrument performance will be accomphshed by review of Technscal Specif. cations, Cais.toi and Survedance Test Procedures and Survedlance and Cahbration Test resulta 90NE*-G-0075 Table 5.1.1-4 I A tmmat trip serpoints cited as variable in the Inadvertent Opening of a Steam Generator reference analysts (Figures 5.1.1-6 and 5.1.1-7) Reliefor Safety Valse Causing a 90NE*-G-0075 Figure 5.3.1-6 Depressurization of the Main Steam System 90NE*-G-0075 Figure 5.1.I-7 (FSAR 15.I.4) Vergicaties Meded: Venficabon of the instrument performance will be accomphshed by review of Technical Specifications. Cahbrabon ar.d Survedance Test Procedures and Survedlance and Cahbration Test results. 90NE*-G-0075 Table 5.1.I-4 I A sm-f trip -*=sats cised as variabk in the Steam System Piping Failure (FSAR 15.I.5) reference analysis (Figures 5.1.14 and 5.1.1-7) 90NE*-G-0075 Figure 5.I.1-6 90NE*-G-0075 Figure 5.1.1-7 Ver#L

• N - Venfica6on of the instrument peih==s wiu be accomplished by review of Technical Specifications, Cahhracon and Survedance Test Procedures and Survedance and Cahbrabon Test results.

Page 69of83

Chapter 15 Accident Mitigating Systems C,.'jj"' ""6'*" 8' D10D7 ICAVP (Systems) SYSTEMDESCRIPTJON: REACFOR PROTECFION SYSTEMS, RTS (RPS-3406) PARAMETER DESCRIPTION INPUT ASSUMPThON AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 3 gy;q FSAR Table 15.6-I DelayTime Rod drop will begin 1.5 seconds after setpoint is inadvertent Opening of a Pressurizer Safety t reached (N-loop operatgm case) or Relief Valve (FSAR 15.6.1) NEU-96-615 rerification Meded: Review Techncal WW and Survedlance Testing for these instruments. 90NE*-G-0075 Table 5.1.1-4 j DifferentsalTemperature Assuced trip serpoints cited as variable in the Feedwater System Malfunctions that Result reference analysis (Figures 5.1.1-6 and 5.1.1-7) in a Decrease in Feedw ater Temperature 90NE*-G-0075 Figure 5.i.1-6 (FSAR 15.1.1) 90NE*-G-0075 Figure 5.I.I-7 l'er#icarlos Meded: Venfcation of the instrutnent performance will be accornplished by review of Technical Sperrems, Cahhration and Surveillance Test Procedures and Survedlance and Cahbration Test results 7.0 seconds (The definition of delay time is given in Feedwater System Malfunctions that Result 90NE*-G-0075 Figure 5.1.1-4 Note (a) of FSAR Tahle 15.0-4). in a Decrease ie Feedwater Temperature 90NE*-G-0075 Figure 5.I.1-6 (FSAR 15.1.1) 90NE*-G-0075 Figure 5.1.1-7 5'er#ication Meded: Venfratm of the instrutnent perfortnance will be accornplished by review of Technical Sprentims, Cahbration and Surveillance Test Procedures and Surveillance and Cahbration Test results NEU-96-623 Trip serpoint given in Figures 15.0-I and 15.0-I A Ch snicaland Volume Control System Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concencration in the Reactor Coolant (FSAR 15.4.6) Ver#icadse Meded: Venfy that the operahng procedures list the appropnate monitors and that there are proper procedures for alarm response Page 70 of33

Chapter 15 Accident Mitigating Systems C"'j,"" 8' ICAVP (Systems) S13TEW DESCRIPTJO.V: REACTOR PROTECTION SYSTEMS, RTS (RPS-3406) l l PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 1 NEU-96-623 DifferentialTemperstare Operator actam upon receipt of turbine runback Chemical and Volume Control System alarm Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) l renykation Afeded: Venfy that the operating procedures list the appropnate morutors and that there are proper procedures for alarm response I FSAR Table 15.6-4 1 l Rod drop occurs 1.5 seconds after RTS trip signal Steam Generator Tube Failure (FSAR 15.6.3) renfication Afeded: Venfy Techncal Specification and surveillance tests. Re>poc: e time 7.0 seconds. Trip setpoint Figure 15.0-Turbine Tr p (FSAR 15.2.3) 90NE*-G-0075 (pg 5-21) I and 15.0-1 A 90NE*-G-0075 (pg 5-127) renfication 3feded: Venfy by revewing Techncal Speci5 cation requirements. Surveillance and Cahbration Procedures and test results NEU-96-623 Alarm requires operator action Chemical and Volume Control Systern Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) renykarier %ded: Venfy that the operating procedures list the appropriate morutors and that there are proper procedures for I alarm response Page 71 of33

Chapter 15 Accident Mitigating Systems 7"',"l"

• • 8' ICAVP (Systems)

SYSTEM DESCRIPTION: REACTOR ?ROTECTION SYSTEMS, R'I5 (RPS-3406) PARAMETER DESCRIPTION INPtJTASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-96-623 i DifferenualTemperature Trip setpoint given in Figures 15.0-1 and 15.0-! A Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (FSAR 15.42) FSAR Section 15.4.6 Venfy that the operating procedures list the appro.3te morutors and that there are proper procedures for Verification Mated: r alarm response. .T' ' ~ -Q'

y Pressurizer Pressure Actuates to trip rear @ l860 psia Inadvertent Operation of Emergency Core NEU-96-623 (pg 15.5-1)

Cooling System During Power Operation (FSAR 15.5.1) Verpcation Meded: Verdication by revew of Techncal WhW, Surves!!ance, Calbration and setpoint data. FSAR Tabic 15 6-1 Rod drop will begin 2.0 wanA< after setpoint is Inadvertent Opening of a PressurM-Safety rewhed (N-1 loop operanon case) or Felief Valve (FSAR 15.6.1) NEU-9M15 rergication Meded.- Re. view Techncal W and Surve Rance Testing for these ristruments. 90NE*-G-0075 (pg 5-2!) 2425 psia, delay time 2 seconds (Table 152-1) Turbine Trip (FSAR 15.2.3) Veryicanen Meded: Venfy by reviewing Techncal S-htion requirements. Surveillance and Caleraten Procedures and test results Page 72 of83

Chapter 15 Accident Mitigating Systems " *". it J "'*'*"8' Massene i m C119,97 ICAVP (Systems) l l SITTEMDESCRIPTION: REACTOR PROTECTION SYSTEMS, RTS (RPS-3406) PARAMETER DESCRIPTION INPtTT ASSUMPTION AFFECFED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Table 15.6-4 1 Pressurizer Pressure Rod drop occurs 1.5 <-d< after RTS trip signal Loss-of-Coolant Accidents Resuhing from a Spectrum of Postulated Piping Breaks l Within the Reactor Coolant Pressure LL=hry (FSAR 15.6.5) Venf techrucal vanfraw and surved!ance tests. l'erfcados M M - f 2425 psia, delay time 2 seconds (Table 15.2-1) Loss of External Electrical Load (FSAR 90NE*-G-0075 (pg 5-21) I 15.2.2: l'<rificarios Meded: Venfy by revewing Techncal Specification requirerrents, Surveillance and Cahbranon Procedures and test rxntk l SAR Table 15.6-4 1 Rod drop occurs 1.5 seconds after RTS trip signal Steam Generator Tube Failure (FSAR 15.6.3) I'crificarian Meded: Venfy Techncal S* tion and surveJ!ance tests. 7m jigijij,,WW7eeC liiiiW4MWJNRMWWITfRF@iEW7WITWWE'c M79 NEU-96-623 (pg 153-4) Fe~ Trip SI erb generates a reactor trip (FSAR Figure inadsenent Operation of Emergency Core 15.0-22) Cooling System During Power Operation (FSAR 15.5.1) l'ervication Meded: Venfcabon by Techrucal SWhtimi and Surveillance TEST DATA. Page 73 of33

Chapter 15 Accident Mitigating Systems ^",,,',",",,jj"' 8' ICAVP (Systems) SISTEM DESCRIPTION: REACTOR PROTECTION SYSTEMS, RTS (RPS-3406) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES l Reactor Trip Reactor trip occurs within 0.6 seconds of turbine trip Turbine Trip (FSAR 15.23) 90NE*-G-0075 (pg 5-21) l at x 50% power l rcrificasien Meded: Venfy by revewog Tectncal Specification requirements. Surveillance and Cahbration Procedures and l test results. NEU-96-623 i Safety injection signal generates a reactor trip. Steam System Piping Failure (FSAR 15.1.5) Veryicaskw Meshed: Venfcation will be accomphshed by a review of Tectncal Spenhtinos and surveillance test data. FSAR Section 15.2.4 TurbineTrip Main steam isolation valve closure generates a laadvertent Closure of Main Steam Isolation t irbine trip Valves (FSAR 15.2.4) rcryicesion Meded: Venfy by revieweg Technical SpehtM requirements. Survedlance and Calibration Procedures and test results = ~,- 3 Pump Speed Reactor Coolant Pump Underspeed trip set point Decrease in Rnetar Coolant System Flow 90NE*40075 (pg 5-59) Rate (FSAR 153) Veryicadw Meded: Verdication of system performance wiB be accomphshed by A reven of I) Trip setpomt in Techncal W.'--A -7 II)Revew of Survedlance Test Procedures an.1 Survedlance Tests WW with tre loge for >P7 & underspeed sensorlogic. Page 74 of83

)% Chapter 15 Accident Mitigating Systems Z"7 8' ICAVP (Systems) SiTTEMDESCRIPTION: REACTOR PROTECTION SYSTEMS, RTS (RPS-3406) l PARAMETER DESCRIPTION INPLTT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES

3. =

1--.. p ~ x-y,,-~,.. 77 - --- n Steam C-= Water Level Low Law rem = trip at f% of narrow range span, Feedwater System Pipe Break (FSAR 90NE*-G-0075 (pg 5-21) l two seconds delay setpoint to rod drop 15.2 8) l'crification Meded: Venfy by reviewing Techrucal Specification requirements, Cahbration and Surved!ance Test Procedures j 1 I and test results l NEU-96-623 FSAR Table 15.0-4 Low Low reactor trip at 10*L of narrow range span, Less of Normal Feedwater Flow (FSAR two seconds delay setpoint to rod drop 15.2.7) 8'erJIcarion M ed ed: Venfy by reviewing Techrucal Specification requirements, Survedlance and Cahbration Procedures and test resatts Low Low reactor trip at 10*a of narrow range span, Turbine Trip (FSAR 15.2.3) 90NE*-G-0075 (pg 5-21) delay time 2 seconds l'eryication Mesmod: Venfy by reviewing Techrucal Specification requirements Survedlance and Calibration Procedures and test results. Low I.ow reactor trip at 10*L of narrow range span, Loss of External Electrical Load (FSAR 90NE*-G-0075 (pg 5-21) I delay time 2 seconds 15.2.2) l'ersfication Meded-Venfy by reviewing Techrucal SghW requirements. Surveillance and Cahbration Procedures and test resultst Page 75 of33

Chapter 15 Accident Mitigating Systems " " ". *n " "'"d""' 8' mm g,g ICAVP (Systems) S237E3fDESCAJPTION: REACTOR PROTECTION SYSTEMS, RTS (RPS-J406) 1*ARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES ,,-;.7- ~;- -,- ,.._ _y. y_ . 4.y;p.gengq NEU-96-623 I Revtne Trip Turbine trip due to high < tram generator level Feedwater System Milfunctmas that Result generates a reactor trip. (FSAR Fig.15.0-8) in an increase in Feedwater Flow (FSAR 15r) Verifi daa Alcskod: Venfw ::rinri ws!! be accornplished by a review of Techrucal Specife atx>ns and su vci!!arx:e test data. 3 ....,,.;,. w w,

.,y _.-.

...-...-~.-r. n. Muual Trip Operator actwo to mena!!y trip reactor inadvertent Operation of Emergency Core NEU-96-623 (pg 15.5-1) Cooling System During Power Operation (FSAR 15.5.1) reirficarianMe Ao* Ven5caten by Techrucal Specrfcation and Survedlance test data. Abnormal operatrg procedure should include parameters to be observed and monitored for thd.cuon l Page 76 of83

Chapter 15 Accident Mitigating Systems [*" '7,""

• • " ~ 8' ICAVP (Systems)

SISTE3f DESCRIPTION: RECIRCULATION SPRAY (CRS-3306) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS S,_FETY ANALYSIS REFERENCES wm o-. ..: _ 4..g/ 4 -~y. _. p.... ,..; ~. as - t ; :_ r <r_ s.-.: 1. 7 _ -. _ FSAR Section 116.5.2, page 15.6-12 WPe.JI Adequate NPSH at minimum containment post-Loss-of-Coolant Accidents Resulting from a LOCA pressure Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure l Boundary (FSAR 15.6.5) Venycadon Aferked: Venfy the de. sign assurnptions regarding the irubal conditons in a post-LOCA containment and associated analysis use an acceptable approach to determinabon of minirnum availt.bie NPSH for post-LOCA ECCS reorculaten. Page 77cf83

Chapter 15 Accident Mitigating Systems C','j"" n'*" 8' ICAVP (Systems) SYSTEM DESCR1PTION: REFUELINC WATER STORACE TANK (QSS-3307C) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SA t t. :Y ANALYSIS REFERENCES ...yy. 7 _._ q, . 7 y. .m. g-2 ;_. FSAR Section 15.4.3 Concentration Boron concentrationx 2000 ppm Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSAR 15.4.8) i Versfication Me.kad: Venfy by renewing operating procedures and survedlance test. FSAR Section 15.6.5.2, page 15.6-12 Volume SutTeient RWST volume to provide for capability to Imf-Coolant Accidents Resulting from a transfer to post-LOCA cold leg recirculation mode of Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure ECCS operation. Boundary (FSAR 15.6.5) Ver!/ication Medad: Venfy design n-ab deirvm br water volume and containtnent sump features prov:de for adequate NPSH for this rnode of operation. Check for system scart-up pocedure and test resu!ts to venfy this rnode of operabort. Page 78 of83

i. Chapter 15 Accident Mitigating Systems

"2","""

"d*" 8'

• !!L97 ICAVP (Systems)

SISTEMDESCMPTION: ROD POSITION INDICATION (RDI-3409) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 3 NEU-96-623 Control Rod Position Operator instructions are adequate to respond to Chemical and Volume Control System wi.corL= Malfunction that Results in a Decrease in FSAR Section 15.4.6 Baron Coacentration in the Reactor Coolant (FSAR 15.4.6) Ver#icaden Meded: Venfy that the operahng procedures Inst the appropnate morutors and that there are proper procedures for alarm response ( l NEU-96-623 Low alarm requires operator action ch W and Volume Control Sysicm Malfunction that Results in a Decrease in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) Ver#icados Mened: Venfy that the operatog procedures list the appropnate morutors and that there are proper procedures for alarm response. FSAR Section 15A.8 1 Operatmg instructions are Ay* to respond to Spectrum of Rod Cluster Control Assembly 6iaL= limit alarm and provide boration Ejection Accidents (FSAR 15.4.8) Ver#icadse Meded: Venfy that the operahng procedures list the appropnate morutors and that there are proper procedures for alarm response Page 79of83

Chapter 15 Accident Mitigating Systems ["j"],7 " a'** 8' olm7 ICAVP (Systems) SITTElf DESCRIPTION: ROD POSITION INDICATION (RDI-3409) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES NEU-9-i-623 Control Rod Position Low Low alarm req' ires operator action Chemical and Volume Control System Malfunction that Results in a Decreax in FSAR Section 15.4.6 Boron Concentration in the Reactor Coolant (FSAR 15.4.6) Verifor%n Afer&od: Venfy that the operating procedures hst the appropriate rnonstors and that there are proper procedures for alarm response FSAR Section 15.4.8 I Opersang instructions are adequate to respond to Spectrum of Rod Cluster Control Assembly RCCA deviation alarm and provide boration Ejection Accidents (FSAR 15.4.8) l VerrJication 3Icskod: Venfy that the operatmg procedures list the appropriate rnorutors and that there are proper procedures for alarm response l FSAR Tabic 15.4-1 Delay Tune Rods begin to fall in 0.5 second after the trip point is Spectrum of Rod Cluster Control Assembly rwhi Ejection Accidents (FSAR 15.4.8) FSAR Section 15.4.8.2.2, page 15.4-3A Verificadom Jferaad: Venfy by reviewing survei!!ance tests. Page 30 of83

. _ = 1 Chapter 15 Accident Mitigating Systems 8"* 8' m-im l m, l l ICAVP (Systems) SIITElf DESCK1PTJO.V: SLCRS FILTRATION (RPV-33141) PARAMETER DESCRIt' TION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES m 17n3 _ " y. _ ~ _ a _ _ m _ a ~gy; - 3;y 3- - - ,g 3

333m7,

.. :( ~ FSAR Section 15.4.8 FL* (E) A sen=Ary o_=ra-negatisc pressure less than Spectrum of Rod Cluster Control Assembly or equal to 0.25" wg is achieved in 60 seconds Ejecien Accidents (FSAR 15.4.8) s'enfication Weshod: Venfy surveillance and test data support assumptons. Review validty of assumption based on MP3 secondary containnent funcional design Wtenn. FSAR Section 15.6.5.4 i Flow (E/1JC) ESF filter performance parameters and flow rate Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR I5.6.5) ren]ication Method: Venfy that !!'e app,yiate system performance parameters are considered in the post-LOCA dose calmitenns for ESF leakage fdtered by the auxiliary buildsng ventdation and fdtraton system Review Techruca! SWMnn survesitance requirements to ensure compliance with calculation assumptions per Reg Guide 1.52. M 6 ~ @ j Q W @ ** ~ ' K Q % 4 % 7 @ ? Q V 7 E T y^," T' Q T "- T V M ] } QI W p' f " ] 1 FSAR Section 15.4.8 Fiher Efficiency (E) Fiher removes greater than 99e iodme Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSAR 15.4.8) renyication Meshod: Venfy surver.ance and test data support assumptions. Page 81 of33

Northeast l'talstaes insi - as Chapter 15 Accident Mitigating Systems 3,m,,,, v it 3 01197 ICAVP (Systems) S137EW DETC11PTIO.V: SLCRS FILTRATION (RPV-33I41) PARAMETER DESCRII" TION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES 3 3 _- y.wn y - * ;. ,,;" ^ Y ;.. <... - - .= ~

. ~.

Release Pathways Release pathways for post-LOCA doses are aunmed Loss-of-Coolant Accidents Resulting from a FSAR Section 15.6.5.4, page 15.6-23 to be limited to only the containment and the Spectrum of Postulated Piping Breaks auxiliaiy building. Within the Reactor Coolant Pressure Boundary (FSAR 15.6.5) Ven]Icarion Merhod-Venfy by use of structural drawings and walkdowns that the only release pathways for post-LOCA doses to evolve from the MP3 containment and ESF leakage is through the containtnent or auxiliary building ventilation systems. oms +r+d Fluw (E/lJC) SLCRS flow assumed to be directed to Unit I stack Loss-of-Coolant Accidents Resulting from a FSAR Section 15.6.5.4, page 15.6-24 1 i Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure l Boundary (FSAR 15.6.5) l renficadon Mer&od: Venfy by review of ventilation system design documents that the SLCRS is released through the unit i stack. Page 82 cf83

Chapter 15 Accident Mitigating Systems 72"'j""

• 8' C11&97 ICAVP (Systems)

S13TEMDESCRIPTION: SPENT FUEL POOL COOLING & PURIFICATION (SFC-3305) PARAMETER DESCRIPTION INPUT ASSUMPTION AFFECTED ACCIDENTS SAFETY ANALYSIS REFERENCES FSAR Section 15.7.4.2.1, page 15.7-4 Spent Fuct Pool Water Level 23 feet above the top of the fuel racks. Design Basis Fuelliandling Accidents (FSAR 15.7.4) l'enficariam Mesked: Verdy by reviewing design drawings, level instrurnentaten and Techrucal SWhtM requirements. ( i ( Page 83 of83}}