ML20100G011

Rev 0 to PNPP EALs Bases Document

ML20100G011
Person / Time
Site:	Perry
Issue date:	02/01/1996
From:	James Anderson CENTERIOR ENERGY
To:
Shared Package
ML20100G010	List: ML20100G011 ML20100G016 PY-CEI-NRR-2016, Forwards Rev 0 to PNPP EALs Bases Document & Draft Rev 0 to PNPP NUMARC/NESP-007 Plant-Specific EAL Guidelines
References
NUDOCS 9602220302
Download: ML20100G011 (153)

Text

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i i

4 PERRY NUCLEAR POWER PLANT EMERGENCY ACTION LEVELS (EALs)

BASES DOCUMENT Based on Conversion to NUMARC/NESP-007 Methodology l

1 l

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Revision 0 f

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Prepared By:

/mo nhr nA.~~._

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s

[

Emergency Planning Unit

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~~

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l Reviewed By:

• Perry Operations Section 1

PORC Meeting No.:

96-013 Meeting Date:

1-25-96 2[I 76 Submitted:

[lanagdr, Regulatory Affairs Section PbR DO O

O 40 F

PDR

. EMERGENCY ACTION LEVELS (EALs) B ASES DOCUMENT Table of Contents Section Tide Eage 1.0 PURPOSE 1

2.0 SCOPE 1

3.0 REFERENCES

1 4.0 DEFINITIONS 2

5.0 RESPONSIBILITIES 2

5.1 Supervisor, Emergency Planning Unit 2

5.2 Manager, Regulatory Affairs Section 3-6.0 DETAILS 3

6.1 Methodology for Assessment and Classification of an Event 3

6.2 EAL Decision Table Logic Examples 4

i ATTACHMENTS - Initiating Condition Index 7 - EAL Entry Criteria and Bases 9

i

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

l Page1 EMERGENCY ACTION LEVRI.R EAT 4 B ASES DOCUMENT t

1.0 PURPOSE 1

To provide the applicable initiating conditions, entry criteria, and technical bases to justify the classification of an event per Section 4.0 of the Emergency Plan in accordance with NUM. ARC /NESP-007," Methodology for Development of Emergency Action Levels".

l 2.0 SCOPE This document outlines the bases used for the development of the EALs in accordance with NUMARC/NESP-007. As such, the intent is to provide a historical record for the l

l training of Emergency Response Organization (ERO) personnel and for the maintenance of the EALs themselves. A detailed comparison of this document against the criteria outlined in NUMARC/NESP-007 is provided in the Plant-Specific EAL Guidelines (PSEG) document.

While this document contains the EAL initiating conditions and entry criteria, Emergency i

l Plan Implementing Instruction (EPI) A1," Emergency Action Levels", shall be used at the l

time of an incident or transient for the classification of an event per the Emergency Plan.

l l

3.0 REFERENCES

l

'~

3.1 Source References 1.

Regulatory Guide 1.101," Emergency Planning and Preparedness for Nuclear Power Reactors (Revision 3) 2.

NUREG-0654," Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants (Revision 1) l l

\\

l 3.2 Use References l

1.

Emergency Plan for the Perry Nuclear Power Plant, Docket Nos. 50-440/50-441 2.

Emergency Plan Implementing Instruction (EPI) A1," Emergency Action 12vels" 3.

Plant-Specific EAL Guidelines (PSEG) i l

l l

Page 2 4.

Code of Federal Regulations (CFR), Title 10, Part 50.47 4.0 DEFINITIONS 4.1 Initiating Condition (IC)

One of a predetermined subset of nuclear power plant conditions defined by NUMARC/NESP-007, where either the potential exists for a radiological emergency or such an emergency has occurred. Initiating conditions are established based on the four emergency classes established by 10CFR50.47, e.g. Unusual Event, Alert, Site Area Emergency, and General Emergency.

4.2 Event Category A collection of similar initiating conditions grouped to allow for the prompt recognition of the transient or event and assessment of severity based on the four emergency classes.

4.3 Emergency Action Level AL)

A pre-determined, site-specific, observable threshold or entry criteria for a given initiating condition that places the plant in a given emergency class. An EAL entry criteria can be:

an instrument reading; an equipment status indicator; a measurable parameter (onsite or offsite); a discrete, observable event; results of analysis; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency class.

4.4 Ooerating Mode There are six applicable operating modes associated with the initiating conditions used in this document: numbers 1 through 5, and the letter "D". Numbers 1 - 5 correspond to the Modes 1 through 5 defined by Technical Specification Table 1.2; the letter "D" stands for the reactor DEFUEL condition.

5.0 RESPONSIBILITIES 5.1 Supervisor, Emergency Planning Unit:

1.

Shall ensure that the EAL and Bases Document and Plant-Specific EAL Guidelines (PSEG) are revised accordingly to reflect changes to the EALs being evaluated as part of Section 4.0 of the Emergency Plan and implemented under EPI-A1.

1 Pcge 3 i

2.

Ensure that a thorough interdisciplinary review of proposed changes is conducted by sections affected by the change or knowledgeable in the subject matter.

5.2 Manager, Regulatory Affairs Section 1.

Shall approve any changes to this document and the PSEO following a thorough interdisciplinary review as part of a proposed change or revision to the EALs contained in Section 4.0 of the Perry Plan and EPI-A1.

6.0 DETAILS 6.1 Methodolorv for the Assessment and Placcification of an Event 6.1.1 Using the Initiating Condition Index (PNPP Form No. 8851 A, Attachment 1), identify the emergency by event category and determine the most appropriate initiating condition (IC) or ICs based on the operating mode at the time of event initiation, plant conditions and j

event severity, j

6.1.2 Refer to the applicable sections of the EAL Entry Criteria and Bases (Attachment 2) for the ICs identified to determine, using the following guidelines, if the entry criteria list are met:

1.

If the values or setpoints established as entry criteria are reached as an anticipated part of a planned and approved maintenance or testing evolution, declaration of an 1

emergency class is HQIrequired.

2.

For those EAL entry criteria with a permitted out of service time or duration, e.g.

loss greater than 15 minutes, the following shall apply:

j a.

Start the time clock at the time of discovery, unless there is firm evidence to believe otherwise, in which case the clock start time is retroactive.

b.

Declare the emergency class as soon as it is determined that the transient will last longer than the specified duration, and do HQI postpone until the allotted time has expired.

6.1.3 Declare the stated emergency class for the applicable IC when the EAL entry criteria listed are met.

1.

Declare the most severe emergency class when multiple ICs are applicable.

Page 4 6.1.4 Based on the emergency class declared, implement the appropriate EPI to initiate emergency response actions required by the Emergency Plan.

l 6.2 EAL Decision Table Logic Examples 6.2.1 EXAMPLE #1:

Condition a AND Condition b AND Condition c... is replaced by:

Moving down &

Condition a Condition b Condition c 6.2.2 EXAMPLE #2:

Condition 1 OR Condition 2 OR Condition 3... replaced by:

Moving down 3 Condition 1 Condition 2 Condition 3 l

I i

Page 5 6.2.3 EXAMPLE #3:

Condition a AND l

Condition b AND l

Either:

Condition le l

OR 1

Condition 2c OR Condition 3c... replaced with:

Moving down 8 i

Condition a Condition b Condition Ic Condition 2c Condition 3e I

l l

r

Page 6 6.2.4 EXAMPLE #4:

Either:

Condition la AND Condition Ib AND Condition Ic OR Condition 2a

'AND Condition 2b AND Condition 4 OR Condition 3a AND Condition 3b AND Condition 4... replaced with:

Moving down 8 1

Condition la Condition 2a Condition 3a l

Condition Ib Condition 2b Condition 3b Condition Ic Condition 4

4 i

f I

PNPP No. 8851 A EVENT CATEGORY UNUSUAL EVENT I.

ALERT

8. ; SITE AREA EMERCENCY

!. l CENERAL EMERGENCY p Fuel clad degradation p Any loss or challenge to the Fuet

[l uncover fuei.

Loss of RPV water tevet that has or wtit p Loss of two barriert AND a loss or ciad barrier i cnaisense to the tnard barrier G

Al s

3 s

s Page 9 - AU1 Page 15 iTable A15 - AA1 f

Page is - AS1 I

Page 15 (Table A1) AC1 h

1 1

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A: FISSION PRODUCT p Reacw cooiant system seawage p Any loss or chanenge to the Reactor Either a chattenge or loss of bo,th the

=

r si cooiant system barner.

y; Fuel Ctad barrier AND Reactor Coolant BARRIER q

q j system barrier DEGRADATION 1

Page 11 - AU2 Page 15 (Table A1)- AA2

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Page 15 (Table A1)- A$2

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31 Chattenge to either the Fuel Ctad barr6er

'4 Any loss or challenge to the b Containment barrier

[ OR Reactor Coolant System barrier, AND v

the loss of any additionat Darrier i

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i Page 15 (Table All-AUs Page 15 (Table All-A$s i

j inabiitty to maintain piant in p Complete loss of functions needed to

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B: LOSS OF DECAY NOT APPLICABLE

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H NOT APPLICABLE HEAT REMOVAL page ss - eAi Page ss, - est FUNCTIONS j[' ilmtts p; Fatture to inittate or complete an Inabiuty to reacn reauired snutdown U Failure to intilate or complete an 3] Failure to initiate or complete a C: LOSS OF within Technical Specification automatic Reactor Scram once an i automatic Reactor Scram once an RPS successful shutdown, AND indtcation j Res function is reautred 1 function is reautred. AND a manual Q of an extreme chaNenge to the aoulty CM SHUTDOWN FUNCTIONS E,

OR FAILURE TO i

3 d SC'** **5 E 5"CC'"'"'-

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Page 37 - m Page 39 - cA1 page 41 : C$1 Page 44 - CC1 k

SHUTDOWN O

9 Loss of att offstte power to Division i Power capability to olvtston 1 and 2 j; 1 and 2 EH Busses for greater than j;

lj]; Loss of all offsite power and onsite i

T1 protonged loss of alt offsite power

, EH Busses reduced to a single power to Olvtslon 1 and 2 EH Busses and onstte power to Division 1 and 2 h) d y is minutei j power source for greater tnan is EH Busses. AND continuing a for greater than 15 minutes, t

y minutes, such enat any additionai a

7 degradation of core cooiing o!

_J single fatture would result in a J

capabliity 5'

" 8 "C"

• D: A.C. POWEP40SS Page 47 - DU1 Page 50 - DA1 Page $4 - D$1 Page 56 - DC1 l Loss of all offstte power AND onsite

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p for greater than 15 minutei s.

D' Page $2 - DA2 g Degradation of Dtvtsson 1 and 2 p Degracation of Divtsson 1 and 2

'jN'm"Nutes NOT APPLICABLE NOT APPLICABLE ER AT ON s

ute S

1 Page $9 - EU1 9

Page 61 - ES1 r

p Fire witnen a sate Snutdown Buriding p Hre OR erososion af fecting tne e

2

. j NOT extinguished within 15 minutes g

s operaoility of plant safety systems j.

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reoutred to establisn or maintain F: FIRE OR v

D safe shutdown' TMMKME MTMPMMW EXPLOSION p Explosion within a Safe Shutdown Page 68 - FA1

'l' Building.

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1 p Unexpected increase in plant p Major damage to trradiated fuet.

3 radiation levels 3:

Page 73

• CLF1 Page 77 - CA1 D-D:

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DrdR N

. M M M N N DWM R i

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l2/lOLAAleMLL4 LASE j! cecrease witn irraatated fues

j. safe snutdown sundings that N

L outside tne RPV remain!ng 4j' impede operation of systems L

s reoutrea to maintain safe operations C "d R

• on to estabitsn or maintain Coto SHUTDOWN 6

l Page 75 - CU2 Page 79 - CA2 sj Any unplanneo release of gaseous 4 Any unptanneo release of gaseous ii Site Bounoary cose resuitsng f rom an t

(y $lte Boundary oOse resuttrng from an h radioactivity to the enviroment p racioactivity to tne enviroment f actual or imminent reiease of actual or imminent release of 4"

tnat exceeds two times tne ooCu mi tnat exceeos 200 times tne ODCM 41 gaseous raoloactivity tnat exceeos aj gaseous raatoactivity tnat exceeos s Control timet for 60 minutes si Control limit for 15 mtnutes s' 100 mR TEDE dose OR 500 mW CDE Cnt:o

( 1000 mR TEDE oR 5000 mR COE Cntid 4 invroid oose for the actuai oR projected e invroto cose for the actuai or projecteo E

or greater.

H: INCREASED or greater ouration of the retease curation of tne release RADIATION Page 83 - HU1 Page 88 - HA1 Page 93 - HS1 Page 96 - HC1 RELEASE TO THE y[ Any unplanneo release of HQuid AnV Unplanned release of HQuia i

radioactivity to the environment

.! raatoact vity to tne environment ENVIRONMENT that exceeds two times tne ooCu

.l tnat exceeos 200 times tne sj Control limit for 60 minutes s ODCM Control limit for 15 minutes D! or greater.

D' or greater Page 85 - HU2 Page 90 - HA2 h Control Room Evacuation has oeen 4 Control Room evacuation nas oeen 1: CONTROL ROOM NOT APPLICABLE lj intitateo.

h initiateo, AND plant Control CANNOT EVACUATION t

41 ee estaoitsneo witnin is minutes NOT APPLICABLE s;:

Page es -iA1 o

Page 100 -IS1 s '-

n si Loss of most annunciators or u Los of most annunciators or 3: inaositty to mumtor a segnaticant f nansient in omgrm h enai n in ene Conte Room with inoican n in tne C ntr i R mfr J: LOSS OF i greater than is minutes i ettner m a significant transient in j

ANNUNCIATORS OR

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j progre;ss. OR <2) compensatory 4

NOT APPLICABLE 2

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INDICATIONS inoications are N_OT avaltable.

Page 103 - JU1 Page 106 - JA1 Page 109 - JS1 Loss of onsite UR instant g{S y

p communications capaomtiei Page 113 - KU1

)g NOT APPLICABLE NOT APPLICABLE] O g gg APPLICABLE y[ significant oegradation of offsite

~ o K: LOSS OF ag A,y COMMUNICATIONS (g communications capablittiet

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g av Page 115 - KU2

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}N Naturai oR destructtve pnenomenaf affecting Safe Snutdown Bullofngi kg P Naturat OR destructive pnenomena M..m o

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DESTRUCTIVE

'ai councarv.

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NOT APPLICABLE

[TNOT APPLICABLE 3

PHENOMENA lj Page 117 - Lui Page 120 - LA1 5

p Reiease of toxic oR fiammaoie p Reiease of toxic oR fiammaose v gases affecting tne Protected Area if gases witnan a sate Snutoown M: RELEASE OF 4! ooundarv e-meo oetnmentae

suocing which jeoparoizes operation 4

TOXIC OR

_k to sa*a operatio. of tne piant.

[ of systems reautreo to maintain NOT APPLICABLE NOT APPLICABLE

- safe operations OR to estaotisn FLAMM ABLE GAS or maintain Coto SHUTDOWN Page 125

.MU1 Page 127 - MA1

.t ; Confirmeo security went whacn

.h, Security event en the plant

.h h of aoility to reacn ano maintam Jl Security event in a Plant Vital Area 5 Security ewnt resulting in loss

[ inoicates a potenttaa oegracation Protecteo Area.

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N: SECURITY EVENTS

. in tne seves of safety of tne piant.

4;

Coto snuroowN a

h Page 131 - Nyt Page 133 - NA1

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Page 136 - NC1 o

p otner conoittons existing, wnien in p otner conoitions existing. wnicn in p/ the judgement of tne Emergency otner concittons existing wnicn in y otner conoitions entsting wnicn in 3 the juogement of the Emergency s'

tne juogement of tne Emergency N

0: EMERGENCY g tne juogement of tne Emergency 4;

t Cooroinator. warrant dectaration of

. Cooromator. warrant ceciaration of C Cooroinator. warrant oectaration of C Cooroinator. warrant ceciarat:on o*

COORDINATOR *S h an unusual Event h an Alert

,p a Site Area Emergency h a Generai Emergency JUDGEMENT page 139 - ou1 Page 141 - oA1 Page 143 - osi Page 1as - oc1 t

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Initiating Condidon Index (Cont.)

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IC AU1 is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.

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'Ihe Off-Gas Pretreatment process radiation monitor reflects the steamjet air ejector effluent and would be one j

of the first indicators of degrading fuel conditions in Modes 1,2 and 3. Therefore, elevated offgas radiation e

activity represents a potential degradation in the level of safety of the plant and a potential precursor of more t

serious problems.

6.1.1 AUl (Cont.)

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s c.u Coolant activity in excess of allowable Technical Specifications 3.4.5 limits also reflects a degraded or degrading core condition and a potential pwmor of more serious problems. 'Ihis condition is elevated to an Alert should coolant activity exceed 300 microcuries/ gram dose equivalent I-131, a value which represents approximately 2%

clad damage.

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2. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Section 3/4.4.5 g.

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3. Offsite Dose Calculation Manual (ODCM), Appendix C - Control 3/4.11.2.7, Rev. 5 5.

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4. FCR 16986, Calculating Radiation Monitor Readings

5. Perry SP-810-07, Drywell Radiation Plots and Technical Bases, dated 5/10/83 F3 O

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l 6.1 Catenorv A: Fission Prnduct Barrier Deeradation (Cant )

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o Initiatina ConditTens Entry Criteria G

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Reactor Coolant System Unidentified leakage in Drywell greater Identified leakage in Drywell greater than S

leakage than or equal to 10 gpm.

or equal to 30 gpm averaged over 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> U

gi period.

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Both the unidentified and identified Drywellleakage rates corresponds to 5 gpm greater than the value in

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Technical Specification 3.4.3.2 and are observable on normal Control Room indication. His allows for a shutdown to commence per the action statement of Tech. Spec. 3.4.3.2 without declaring an Emergency Plan event unless the leakage is significantly greater and has the potential to degrade. His will also permit actions to isolate non-reactor coolant boundary systems in order to identify the source of leakage.

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l 6.1.2 AU2 (Cont.)

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Technical Specifications, Perry Nuclear Power Plant, Unit 1, Section 3.4.3.2 m

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AS1 RPV water level CANNOI be rnaintained greater than 0".

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that has or will uncover Reactor is " shutdown under all conditions without boron"

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This IC and its associated EAL address: (1) a loss of the Reactor Coolant System, defined here as the inability to G

l maintain level above the top of active fuel; and (2) a challenge to the fuel clad when the core becomes uncovered.

This could ultimately result in a release to the environment.

ASI is applicable only to non-ATWS situations in which RPV level was NOT intentionally lowered per PEI-B13 l

(A'IWS) as a means of power control. Refer to Event Category "C" for classification under ATWS conditions in which RPV water level is intentionally lowered.

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b The fission product barrier loss and challenge thresholds defined in the Fission Product Barrier Matrix only apply g

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under Modes 1,2 or 3. This separate IC, established outside the Fission Product Barrier Matrix, is based on J

j application during Modes 4 and 5 in addition to Modes 1,2 and 3. Refer to the Fission Product Barrier Matrix g-y for possible event escalation to a General Emergency in Modes 1,2 or 3 based upon this condition in g

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combination with a loss or challenge to the Containment barrier.

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2. Technical Specifications, Perry Nuclear Power Plant, Unit 1, Sections 3.4.9.1 and 3.4.9.2 l

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3. Plant Emergency Instruction (PEI) B13, RPV Control (Non-ATWS), Rev. 2 l

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4. Plant Emergency Instruction (PEI) B13, RPV Control (A'IWS), Rev. A i

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s Sheet 7 of 138 TABLE A-t FISSION PRODUC1 NF929A i

I i

l REACTOR PRESSURE VESSEL LEVEL, DRYWELL RADIATION REACTOR COOLANT SYSTEM ACTIVITY EMERCEN(

mh Entry into PEl T23. Containment Drywell radiation monitor Sample activity is equal to or greater Any conc mw Flooding reading greater than than 300 UCl/gm dose eaulvalent judgmen 3h 4,000 R/hr.

iodine-131.

Coording d

v Fuel Clad g

RPV level is less RPV level CANNOT Any conc 4

g5 than 0".

be determined.

Judgeme l

"j E NOT APPLICABLE NOT APPLICABLE Coordina gh to the F' 4

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t REACTOR PRESSURE ORYWELL RADIATION REACTOR PRESSI'"E CONTROL DRYWELL PRESSURE R

VESSEL LEVEL, RPV level less Drywell radiation SRVstuck An SRV is beit a ' Emergency Drywell pressure ME than 0".

monitor reading open.

cycled to Depressurization greater than 168 01 greater than control RPv is required.

psig.

j 135 R/hr.

pressure.

Co' i

ow on-8 b Sample activity is equal to indication of RCS b

or greater than 300 UCi/gm leakage inside the I*I dose equivalent lodine 131.

Drywell.

pe, K-i.w g

HE CP zs "l E NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE Co; ifE 6"

Im' Rc>

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REACTOR PRESSURE CONTAINMENT RADIATION CONTAINMENT CONTAINMENT VESSEL LEVEL, HYDROGEN PRESSURE Entry into PEl-T23, Intentional venting intentional venting of Contat Containment Flooding of Containment Containment per PEl-T23.

doesg per PEl-M51/MS6.

Closur; Immec 4

mE the CO:

yy NOT APPLICABLE succes; penetr 8 g PathWk exists 8

--ummuum Containment radiation in the UNSAFE Containment

$h monitor reading greater than region on the pressure is ag NOT APPLICABLE than 20.000 R/hr.

NOT APPLICABLE HCL figure.

greater than yg 15 psig and v

increasing z

FOOTNOTES

1. A LOSS or CHALLENGE. to any barrier based on RPV water levelis applicable only to Non-ATWS conditions-2 Those thresholds for which a LOSS or CHALLENCE is determined to be IMMINENT (ie, within the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />), classify as

1 j'

Page 15 BARRIER MATRIX INITIATING CONDITIONS INSTRUCTIONS UNUSUAL EVENT F

f C0ORDINATOR JUDGEMENT

. For each of the three barriers' AU3 Any loss or challenge to the I

determine if any LOSS or Containment barrier.

b tion that, in the or CHALLENCE criteria have of the Emergency been met.

Modes:[Nk[3]T]

or indicates loss of the g

3arrier. *

2. Compare the barrier LOSS (es)

ALERT and CHALLENCE(s) to the 0

c AA1 Any loss or challenge to Fuel Clad tion that, in the initiating conditions listed, and N

't of the Emergency make the appropriate event bard n or, Indicates a challenge declaration.

Modes: Q"IK-3 I-[- "I]--

el Clad barrier. 8 P

AA2 Any loss or challenge to the Reactor R

Coolant System barrier.

O Modes:[1] 3)) ~]

ACTOR COOLANT SYSTEM BYPASS EMERCENCY COORDINATOR JUDOMENT U

SITE AREA EMERGENCY

. break outside Containment exceeding one Any condition that, in the judgment O

nore MSiv Vech. Spec. Isolation setpoints.

of the Emergency Coordinator.

AS2 Elther a challenge or loss of ItgLh T

indicates loss of the RCS barrier.

the Fuel Clad barrier AND Reactor itainment penetration does NElsolate Coolant System barrier.

a valid closure signal.

Modes:[1[2]37Ti~)

s A

1ediate Operator actions in the Control An Cnauenge to eithM the Fuel Qad bardM m are! Lor successfui in isolating affected etration.

OR Reactor Coolant System barrier, AND R the loss of any additional barrier.

R r or more of the Maximum Safe Operating Any condition that, h the judgment M0d'5: b

.f _U I

3 iditions per PEl-N11 has been exceeded.

of the Emergency Coordinator, E

Indicates a challenge to the RCS itainment penetration does!Loiisolate barrier. >

GENERAL EMERGENCY R

a valid closure signal.

AC1 Loss of two barriers, AND a loss or nediate Operator actions in the Control challenge to the third barrier.

M r

successful in isolating affected A

T R

CONTAINMENT ISOLATION EMERGENCY COORDINATOR JUDGEMENT I

iment penetration Primary system is discharging outside Containment.

Any Condition that,in the judgement E isolate on a valid of the Emergency Coordinator, indicates signal.

loss of the Containment barrier. 2 (Loss of the Containment barrier may ate Operator actions in include a rapid unexplained decrease ltrol Room are NOT in Contalnment pressure following an fulin isolating aYected One or more of the Maximum Safe MSL break in the Turbine Buildin9 initial increase.)

ition.

Values per PEl-N11 is exceeded due indicated by either:

to Reactor Coolant System leakage

• elevated TB radiation levels outside Contalnment.

Table 3 3 2 2 isolation setpoint.

• TB area temperatures greater than a De n

Tech. Spet Table 3.3.2-2 isolation setpoint.

Any condition that. In the judgement of the Emergency Coordinator, Indicates a challenge to the Containment barrier 8 NOT APPLICABLE hough the thresholdts) has been exceeded.

Sheet 8 of 138 Page lo EAL Entry Criteriit and Ilases (Coni)

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6.1 Cateeory A: Fiuinn Product Barrier Deeradatian (Cont.)

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1. The term" imminent" as used in the EMERGENCY COORDINA'IUR JUDGMENT category for all three categories is intended to infer that a barrier loss will occur within the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

2. Although the logic used for these initiating conditions (ICs) appears overly complex, it is necessary to reflect the following considerations:

The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Containment barrier.

Unusual Event ICs associated with RCS and Fuel Clad barriers are addressed under separate Fission f

e Product Barrier ICs (AUI, AU2),

At the Site Area Emergency level, there must be some ability to dynamically assess how far present G

r conditions are from General Emergency. For example, if Fuel Clad barrier and RCS barrier LOSS EALs exist, this would indicate to the Emergency Coordinator that, in addition to offsite dose assessments, continual assessments of radioactive inventory and Containment integrity must be focused on. If, on the other hand, both Fuel Clad barrier and RCS barrier CHALLENGE EALs exist, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.

The ability to escalate to higher emergency classes as an event gets worse must be maintained. For example, RCS leakage steadily increasing would represent a greater risk to public health and safety.

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f NOTE: A LOSS or CHALLENGE to the Fuel Clad barrier based on RPV waterlevelis applicable only to non-ATWS conditions. Refer to Event Category 'C' for event 7

classification based on the intentionally lowering of RPV water level below the top of 4

active fuel (TAF) as means of power control for ATWS scenarios per PEI-B13 g!

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A CHALLENGE to the Fuel Clad barrier has been established as either 0" (TAF) or RPV level o,

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Per the PEI Bases Document, adequate core cooling can be assured if RPV level is maintained g

greater than 0". At TAF, an emergency depressurization is required per PEI-B13, RPV Control p

l (Non-ATWS) to allow for the injection of low pressure make-up systems, if not aheady initiated, 5

I to restore and maintain RPV water level.

The inability to determine RPV water level has also been established as a conservative threshold for a CHALLENGE to the Fuel Clad barrier. If RPV water level CANNOT be determined per

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PEI-B13, RPV Control (Non-ATWS), the operator is directed to emergency depressurize to l

ensure core submergence. If RPV level can still NOT be restored, the operator is directed to enter PEI-B13, Containment Flooding, thus creating the sequence described below for a LOSS.

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s >g: v ..' ^' c e i s <, itw His value is higher than that specified for RCS barrier LOSS. Rus, this EAL indicates a LOSS of both Fuel Clad barrier and RCS barrMr. NOTE: It is important to recognize that in the event the radiation monitor is sensitive to shine from the reactor vessel or piping, spurious readings will be present and another indicator of fuel clad damage is necessary to t b A.3 Reactor Coolant System Activity ,a (1) Reactor coolant activity is NOT used to indicate a CHALLENGE to the Fuel Clad barrier. um 2 [ (2) A LOSS of the Fuel Clad barrier is indicated by a coohnt activity of 300 pCi/gm dose [. O i equivalent I-131. His amount of activity is well above that expected for iodine spikes and l e corresponds to approximately 2%-5% fuel clad damage. This amount of clad damage indicates [ W significant clad heating and thus the Fuel Clad barrier is considered lost. g 5 ac I 1 it oO s + 6.1.4 (Cont.) .m s i [ ^ ' ~ % ~ ) : ip ' '['" . DISCUSSION,(Cont.)]ig, Dip?gu < ' ' '[ ^ MI,b '..... ', ~ ~ ', l~ J [_ ( .m.. ~ *bs gdt,,; ^ 1 E- ? l: m , a m ~ ~ U B. REACTOR COOLANT SYSTEM l B.1 Reactor Pressure VesselIzvel m>r NOTE: A LOSS or CHALLENGE to the RCS barrier based on RPV water level is applicable only to m non-ATWS conditions. Refer to Event Category 'C' for event classification based on the 3 l@ intentional lowering RPV water level below TAP per PElr.B13 (ATWS) as a means of power g control for ATWS scenarios. g B-CI (1) RPV water levelis NOT used to indicate a CHALLENGE to the RCS barrier. { R b (2) A LOSS of RCS barrier integrity has been defined as water level less than 0" (TAF). Per the NUMARC/NESP-007, a RCS LOSS is considered the same as a CHALLENGE to the Fuel = Clad barrier for a known LOCA resulting in RPV water level dropping below TAF. Thus, this y [ condition appropriately escalates tne emergency classification to a Site Area Emergency. g B.2 Drvwell Radiation (1) Drywell activity, as indicated on the D19 radiation monitors, is NOT used to indicate a CHALLENGE to the RCS barrier. (2) A 135 R/hr reading on 1D19-R100A or ID19-R100B is used to indicate a LOSS of the RCS barrier. The threshold of 138.5 R/hr was determined under FCR 16986 and was conservatively

p rounded off to 135 R/hr for readability. This value assumes an instantaneous release and dispersal li of the reactor coolant noble gas and iodine inventory associated with Technical

!d Specification 3.4.5 limit concentrations into the Drywell atmosphere at shutdown. t 6.1.4 (Cont.) ,. vug. ~m augg g t; w ,er m gg g mg gg t a www s -g. a,

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f hh"b..Yd b[N,% $k!!h I hy i$$ 5DkSCUSSIONNh*?'; N ' ' f k wh> ek hv%C 9%M % ~ $ *? A 135 R/hr reading is less than that specified for Fuel Clad barrier LOSS. Thus, this EAL would G j be indicative of a RCS leak only. If the radiation monitor reading increased to that value specified by Fuel Clad barrier, fuel damage would also be indicated. l r B.3 Reactor Pres-c Control t n (1) RPV pressure is NOT used to indicate a CHALLENGE to the RCS barrier. (2) A LOSS of the RCS under this category focuses on the intentional bypassing of the RCS barrier y via the SRVs as a means of pressure control. Each of the EAL indicators listed below y compromise RCS integrity and creates a direct release path for fission products to the y Suppression Pool. g g 3 SRV stuck onen OR being cycled to control RPV oressure. /MD fuel clad failure 5 o indicated by RCS samole activity canal to or ereater than 300 uCi/em dose canivalent E. O i Iodine-131. f M i ~ ~ ~ R Credit should be given for the actions taken under ONI-B21-1 to successfully close a B stuck open SRV. [ L Emereency Deoressurization required per PEls. Specific plant conditions requiring j cmergency RPV depressurization are given in the individual PEls and are listed in the PEI l Bases Document under PEI-B13, Emergency Depressurization. B.4 Drvwell Pressure j i (1) Drywell pressure is not used to indicate a CHALLENGE to the RCS barrier. RCS leakage in y excess of Technical Specifications is covered under Initiating Condition AU2. A O t M N r I 6.1.4 (Cont.) g- ' y,;.::' , y:,;,~;; g' 3,,. g g 4 ' y ^7 % c c,,, 7 ' : ~ r, 7 y ;7 4 y i c y, ; f ' ; +o ' ^c' ' O ' '; ^l' ' N ? 1 ;',,, [ { 1i C :' LDISCUSSION(Cont.) ' ' O' ^Jdd'S V - ~ I, q'

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(2) A Drywell pressure of 1.68 psig is used per PEI-T23, Containment Control (Pressure), to indicate C a LOSS of RCS barrier integrity. 'Ihis threshold addresses a loss of coolant accident (LOCA) due to breaks inside the Drywell and is an easily identifiable reactor scram setpoint. 'Ihe quahfier of" indication of RCS leakage inside Drywell" (e.g., an increase in Drywell Floor i Sump Fill Rate) is included as an indicator of RCS boundary degradation and climinates requirement to classify based on a Drywell pressure increase due to a loss of Drywell l ventilation / cooling. B.5 Reactor Coolant System Bvnass 3 j v1 (1) A CHALLENGE to the RCS barrier is indicated by an unisolated RCS leakage outside l Containment from reactor support systems. MSL breaks are considered under RCS LOSS. B-G Entry into PEI-N11, Containment Izakage Control, is listed to qualify affected systems, plant { 9., areas and establish severity threshold. m g i Per the PEI Bases Document, the purpose of PEI-N11 is to protect equipment in the Annulus and 1 i surrounding Containment, limit radioactive releases to the Annulus and surrounding Containment 9 [ integrity or limit radioactivity release from the Annulus and surrounding Containment. 6 l PEI-N11 Maximum Safe Operating Condition values are used to quantify the magnitude of the 1 Reactor Coolant System leak and provide site-specific indications. Per the PEI Bases Document under PEI-N11, these conditions are the highest parameter value at which either: (1) equipment necessary for safe shutdown of the plant will fail; or (2) personnel access necessary for the safe operation of the plant will be precluded. i 90it 1 ts i m. ..m. l 6.1.4 (Cont.) i . m__. D O T M se u w ^M ' ' ^ ~ ~*6> > c M '" ' "u ' a. (i;M 'L !?:' g

DISCUSSION 1, Cont.)3 ~ =F'& in p * ' ' ',,'

^ a s-g- 4g gy y .~ -, m, + , ( i M,. The failure to isolate the affected system is defined by 2 of the following criteria: G Containment penetrati~ does NOT close on a valid isolarian sienal. 'Ihis criteria refers to the i successful automatic.,sure of at least one isolation valve in an affected system. Redundant closure of both the inboard and outboarti isolation valves, if applicable, is NOT required. Operators should assess whether isolation is successful based on Control Room indication. [ m Immedinne Operator actions in the Control Room are NOT successful in isolating the affeggd y penetration: This criteria is limited to actions taken to remotely isolate the penetration from the m Control Room panels within the first 5 minutes after the failure to isolate is identified. Actions j f g! taken to dispatch personnel in-plant to attempt to manually close a valve / damper to isolate the g 2, penetration are NOT considered. g B-E (2) A RCS LOSS is limited to RCS inventory loss due to a MSL break outside Containment. The { 9,, magnitude of the break is quantified by requiring that plant conditions result in an automatic g MSIV isolation signal per Technical Specification Table 33.2-2. g = "Ihe integrity of other reactor support systems, applicable under PEI-N11, are evaluated as a RCS h j barrier CHALLENGE. E i p An RCS LOSS due to an RCS bypass scenario is focusing on the unsuccessfulisolation of a MSL [ break. 'Ihe successful isolation of a MSL break would be classified under Event Categories 'O f and H', based on its radiological in-plant and offsite effluent significance. The failure of the affected MSL to isolate is defined by d of the following criteria: f Containment nenetration does NOT close on a valid isolation sienal. This criteria refers to the ,2 successful automatic closure of at least one Main Steam Isolation Valve (MSIV) in an affected [ MSL. Redundant closure of both the inboard and outboard MSIVs is NOT required. l Operators should assess whether isolation is successful based on available Control Room indication. I .m._ 6.1.4 (Cont.) ll,~> i ,e,.~~,, ,, 'f s,,,, h . - s +, nn > -

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i w ; ( l ' '. ' 't ~s j, s ,. v s ^ ' ' v._g N s B Immadiate C-iar actions in the Cmtrol Room are NOT amdulin innIntine the affacead G l MSL. 'Ihis criteria is limited to actions taken to remotely isolate the penetration from the Control Room panels within the first 5 minutes after the failure to isolate is identified. i Actions taken to dispatch personnelin-plant to attempt to manually close a valve to isolate the penetration are NOTconsidered. l r 1 t C. CONTAINMENT g1 l 1 v> 1 p( 3. l 4 =r C.1 Reactor Pressure VesselIzvel 3-G f NOTE: A LOSS to the Containment barrier based on RPV water level is applicable only to non-A'IWS [ 2, i conditions. Refer to Event Category 'C' for event classification based on the intentional te C i lowering of RPV water level below TAF per PEI-B 13 (ATWS) as a means of power control for f ATWS scenarios. 1O O i (1) RPV water level is NOT used to indicate a CHALLENGE to the Containment barrier. f l (2) A LOSS to the Containment barrier is defined as the inability to provide adequate core cooling to an extent requiring Primary Containment Flooding. 'Ihis logic is consistent with the Containment LOSS criteria under Containment Hydrogen and Containment Pressure categories due to the intentional venting of Containment per PEI-T23, Containment Flooding, to restore adequate core cooling via submersion. au w Un 6.1.4 (Cont.) i jhMif# i , 2o ^5sjii.1,163 f$g" "pRj$j?l, NC 'de "ge*ylEspr jhg@%g ^ ,) :

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, ~jlgg{}:. 4 n f 58 ] gn 'j p; e g g ] Q asg g g g% ( D,ISCUSSION(ConL) ~ > x -a + -, ~ ~ E C.2 Containment Radiation 9 (1) A 20,000 R/nr reading on ID19-R200A or ID19-R200B is used to indicate CHALLENGE to l i the Containment barrier. This level of activity is indicative of approximately 20% clad damage. Per FCR 16986,20,000 R/hr is based on the Containment radiation monitor reading associated I with 10% failed fuel from NUS Ixteer SP-810-07, Attachment 1 (Curve 3), which was doubled to [ account for 20% fuel failure. It is a vahe that indicates significant fuel damage well in excess of [ that associated with the LOSS of both Fuel Clad and RCS barriers. l i A radioactive release requiring offsite protective actions is NOT possible unless a major fuel 3 cladding failure allows radioactive material to be released from the core into the reactor coolant. g d 'Ihis amount of activity in Containment, if released, would have severe consequences justifying a g CHALLENGE to Containment. As such a General Emergency declaration is warranted. B-i;; 18 O E. (2) Containment activity, as indicated on the D19 radiation monitors, is NOT used to indicate a m g LOSS of the Containment barrier per NUMARC/NESP-007 bases. g" [ i O C.3 Containment Hydroeen t (1) Hydrogen concentration is NOT used to indicate a CHALLENGE of Containment barrier j integrity. Per the PEI Bases Document under PEI-M51/M56, Hydrogen Control, RPV water [ level must drop below TAF for a minimum of 10 minutes for significant amounts of hydrogen to j be generated. Table A-1 criteria therefore focuses on ability to maintain RPV water level as a CHALLENGE to Containment. l i t

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(2) PEI-M51/M56, Hydmgen Control, provides Operator actions to mitigate the buildup of hydrogen G concentrations in the Drywell and Containment and prevent the Hydrogen Deflagration Overpressure Limit (HDOL) from being exceeded. Per the PEI Bases Document under PEI-M51/M56, Hydrogen Control, the HDOL curve assures that the postulated combustion of hydrogen and oxygen will NOT result in sufficiently high i pressure that will cause the structural failure of Containment or adversely affect Drywell integrity. m b A barrier LOSS shall therefore be implied due to the intentional venting of Containment per PEI-p M51/M56 which purposely bypasses the Containment barrier. J r.n h. d C.4 Containment Pressure g B-3 (1) A CHALLENGE to the Containment barrier is based on exceeding any of the following criteria: { P, to G A Containment pws-of 15 psig and increasing was selected since at this point preparations g oo e are taken per PEI-T23, Containment Control (Pressure), to vent Containment prior to 4 exceeding the Primary Containment Limit (PCL). This criteria therefore indicates a clear p CHALLENGE to Containment. Refer to C.4(2) for a Containment LOSS based on E intentional venting to prevent exceeding the PCL. j In the UNSAFE region on the Heat Canacity Limit (HCL) figure, due to either: (a) Suppression Pool temperature above the Heat Capacity Temperature Limit (HCIL). Per the PEI Basis Document PEI-T23, Containment Control (Suppression Pool i Temperature), the HCIL is defined to be the highest Suppression Pool temperature at l which initiation of RPV depressurization will not result in eweeding- (1) the suppression lP chamber design temperature, OR (2) the PCL before the rate of energy transfer from the W RPV to the Containment is within the capacity of the Containment vent.. This t$ temperature is a function of RPV pressure, and the limit is used to prevent failure of the Containment or equipment necessary for the safe shutdown of the plant. l l t .E' 6.1.4 (Cont.) r i - 2, '1,,c w ~ e ',: ;

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~. _ ~-.. ,; p 9, E u (b) Suppression Pool level below the Heat Capacity level Limit (HCLL). Per the PEI bases Document PEI-T23, Containment Control (Suppression Pool Temperature), the HCLL is defined to be the higher of either- (1) two feet above the elevation of horizontal vents (14.25 feet); or (2) the lowest water level at which initiation of RPV depressurization t will NOT result in exceeding the HCFL. De HCLL is used in conjunction with the HCFL to prevent failure of the Containment or failure of equipment necessary for the m safe shutdown of the plant, and to prevent loss of the pressure suppression function of p the Containment. m l t E (2) A Containment barrier LOSS is based on the intentional venting of Containment being required a per PEI-T23 to prevent exceeding the Primary Containment Pressure Limit (PCPL). The PCPL [ ensures that pressure is maintained below the most limiting factor based on the Perry Plant design,

3.

G which is the pressure capability of Containment. De limiting component for the PCL curve is the l l Containment Equipment Hatch. E. 2 ? Y + A Containment LOSS based on a rapid unexplained decrease of Containment pressure (i.e., NOT i attributed to Containment Spray or condensation effects) following an initial increase is addressed 3 [ under the Emergency Coordinator Judgment category. Containment pressure willincrease as a i result of the mass and energy release into the Drywell from a IDCA. Dus, Containment pressure R not increasing under these circumstances might indicate a LOSS of Containment integrity. i Per the PEI Bases Document under PEI-T23, Containment Control (Pressure), the PCL is a i combination of the PCPL and the Maximum Containment Water Level Limit (MCWLL). PEI-l T23, Containment Control, directs Operator actions to mitigate increases in Contaisment pressure to prevent exceeding the PCL. { m so i M h i 6.1.4 (Cont.) - dyl n " j'C'?c ~ ' ,~ ~' j.: ^,, '^ ^ ~

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yy, mg ' ' ':': m a'- Rw 'Ihe MCWLL ensures that Containment water level is kept below the level which will result in exceeding the pressure capability of Containment. With Containment water level above PCL, as a result of Containment Flooding efforts, integrity can no longer be assured. Therefore, irrespective of whether adequate core cooling is assured, injection into Containment from systems 4 which can only take suction from outside Containment, must be terminated. f C.5 Containment Isolation m (1) The failure of a Containment penetration to isolate successfully is NOT used to indicate a j CHALLENGE to the Containment barrier. m h-k G (2) A LOSS of Containment barrier integrity is defined by either of the following conditions: = bo[ Failure of both isolation valves in any one line to close AND downstream pathway to the e W g environment exists, as indicated by all of the following: (a). Containment nenetration does NOT close on a valid isolation sienal. 'Ihis criteria refers F3 to the successful automatic closure of at least one isolation valve in an affected system. 8C Redundant closure of both the inboard and outboard isolation valves, if applicable, is NOT required. (b) Imnwiinte Ooerator actions in the Control Room are NOT successful in isolating the affected gm.ireilon. 'Ihis criteria is limited to actions taken to remotely isolate the penetration from the Control Room panels within the first 5 minutes after the failure to isolate is identified. Actions taken to dispatch personnel in-plant to attempt to manually close a valve / damper to isolate the penetration are NOT considered. y li i 6.1.4 (Cont.) l i $g8 < ?d$f72%gy J ~ qqfch ? 1._ +mg %$ l@[$k $$ [ d kh I k Q' ^ ~ ^, M Ddih! Y N ^^ W5@ l ^ yML._. x .gg k Is M h "NM h e g I o (c) Pathway to the environnzei exists via the nenetration. The intent of this criteria is NOT to consider a LOSS in Containment simply due to the failure of a penetration to isolate when coeuwended, which is covered under Technical Specifications. 'Ihe concem is that a pathway exists through a break or system penetration which would in effect bypass Containment creating a pathway to the cuivironment outside the normal process stream or with the normal filtration path NOT intact. For example, a failure of the MSIVs on a given MSL where a pathway exists to the Main Condenser, in which a vacuum is maintained via the SJAEs through Off-gas, is NOT considered a pathway to the environment. p 4 Normal system leakage is NOT considered. g i Unisolable primary system leakage outside Containment, as indicated by: B-13 e o I g, m i i (a) Primary system dischareing outside Containment. Criteria refers to a break or failure to e g isolate which results in a LOSS of RCS inventory, greater than normal system leakage, y o outside Containment. 1 n i o The magnitude of the break is quantified by either of the following criteria being met for a i pipe break outside Containment: (b) One or more of the Maximum Safe Oneratine Conditions listed under PEI-N11 has been exceeded. Per the PEI Bases Document under PEI-N11, these conditions are the highest { parameter values at which either: (1) equipment necessary for safe shutdown of the plant will fail; OR (2) personnel access necessary for the safe operation of the plant will be precluded. m i g t (c) Indication of a MSL break in the Turbine BniMiw Since a MSL break in the Turbine a Building is outside the scope of PEI-N11, the magnitude of the MSL break is based on d either a corresponding MSL low pressure and elevated Turbine Building temperatures per i Technical Specifications Table 33.2-2, or elevated Turbine Building radiation levels. i t i 6.1.4 (Cont.) , > $hn ~ ~ ~, .~ f 'l,, l:~,' ~'A ' ~k, ^ ~ ~,Y<, ~ ~ 4,f' y - ,9-( y'~-p ;,,~., 1 " $y,, ' ') g, 'f' ' ~,] n% 9 ;': j :( 3 g i REFERENCES + J ' " y ' ,o',, n v z~,, y

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a~;+ n> ~' e:~ ~n g 3 1. NUMARC/NESP-007 (Rev.2) Table 3 G 2. Plant Emergency Instruction (PEI) B13, RPV Control (Non-ATWS), Rev. A 3. Plant Emergency Instruction (PEI) B13, Emergency Depressurize, Rev. B 4. Plant Emergency Instruction (PEI) B13, RPV Flooding, Rev. A 5. Plant Emergency Instruction (PEI) M51/56, Hydrogen Control, Rev. A r 6. Plant Emergency Instruction (PEI) N11, Containment 12akage Control, Rev. B 4 i p, E 7. Plant Emergency Instruction (PEI) T23, Containment Flooding, Rev. A g B-0 8. Plant Emergency Instruction (PEI) T23, Containment Control, Rev. A { 9. Plant Emergency Instruction (PEI) Bases Document, (Rev. 2 / Update 1) { E$ = r 10. Off-Normal Instruction (ONI) N11, Pipe Break Outside Containment, (Rev. 6) y a 11. Off-Normal Instruction (ONI) B21-1, SRV Inadvertent Opening / Stuck Open, Rev. 3/IEN-3 C 12. Technical Specifications, Perry Nuclear Po'wer Plant, Unit 1, Section 2.1 and Table 3.3.2-2 13. FCR 16986, Calculating Radiation Monitor Readings i 14. PERRY SP-810-07, Drywell Radiation Plots and Technical Bases, dated 5/10/83 15. NUMARC/NESP-007 (Rev. 2)" Question and Answers", dated June 1993

P li 16.

FCR 17163, Table 1 - MAAP Run Results for 'IRANIEAK E! 17. System Design Manual (SDM) G50, Liquid Radioactive Waste Systems, Rev. 3 I 6.1.4 (Cont.) ',s' ))[{ ,,' ] g esi:lA j^ g f ' g hgg g-lss Y....hy wm ~ sa ' e s jQg3Fi ~ 'x, (. ~ R_ 4r<- y.. e. .. m,.,s w,m;y wa .m m s m ll n 18. System Design Manual (SDM) G61, Liquid RadWaste Sumps Systems, Rev. 3 0 19. System Design Manual (SDM) G50, Floor Drain Collector System, Rev. 4 m m n = w* 3" l3. M m A N O k M m oo = o pr* s i "U l m OQ O ( W N 5 t ( l i l I 6.2 Category B: Loss of Decav Heat Removal Functions ' : ~, ; ' ' _ '- ' ',; ; = c ' : ' ', ' ' L> f u,'::o;;.

~ 'Y,

3 , M ' M ' ;'s'^ ' > 7, z ' ~, ~ 3l c c' l ? ~ ' ^: ' :'^ '. ; ' ': :' E x 5 6.2.1 INITIITING CONDITION'BAf s, } f ' : ' ~ ~ :, 'l ' j ,;<;,,,; ' y;;, ' l',',_ x ' ~ ' ' Q. l Bo E Initiating Conditions Entry Criteria w Inability to a maintain RCS temperature less than 200 F. BA1 BA1 A Inability to maintain L m plant in COLD E SHUTDOWN R g1 T m Q if if a Applicable Modes: 3-N E 2, 4 5 $c a iMs 3 x ^ 'DISCdSSION3 wa 4 g e m, This IC and its EAL address a complete loss of functions required for core cooling during REFUELING and COLD SHUTDOWN Modes. He IC remains applicable for situations in which an uncontrolled increase in RCS temperature greater than 200 F results in a change to Mode 3. The criteria," inability to maintain a reactor temperature of less than 200 F," is met as soon as it becomes known that sufficient cooling CANNOT be restored to maintain temperature below 200 F regardless of the current temperature. An Alert is declared in the event RCS temperature exceeds 200 F unless the required systems are a functionally available to restore the temperature below 200 F. He intent of IC BA1 is NOT to classify based on C 'd a momentary unplanned excursion above 200 F when heat removal capability is available. L .t 6.2.1 BA1 (Cont.) w2 - su - m.. m m#c~. w s s y . ;, d;' ,,[q, [.k 7.....m.y. b. y, [khh .jpflj ( . Yi., I}NhN f[{:, 'L; ,*$;.r:: ,,. f ;" h,j, ^ W4 ..,.f g [ } 4 ^. pl$., v #,i i y,._ x ^ q.;.. p. s y ^ @f ws*::D?M.S$3% hi 'a.aW,'WMin,es' gp %(g!! L, Waham ' i>:: REFERE.NCES% ~ ~ 1,s qyse g e t x -. 4 ~ J ~ %.Ae % %p%se74"irg;;p M!

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l. NUMARC/NESP-007 (Rev. 2) Alert SA3 w

2. Off-Normal Instruction (ONI) E12-2, Loss of Decay Heat Removal, Rev. 4

3. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Table 1.2 m>

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O i R 3. u W Ch O b m oo t sn O M i O t 3 r f* 1 l I e l-t u OQ O 143 A-r 6.2 Category B: Loss of Decay Heat Removal Functions (ContJ c , '..:. ~,, z 's a ~ n :n ':i a +';l, . ~, '.' ~ :~,;: : :,,

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~, a,:: ~ >:, ~ c' s . ~,,. n. 8 u .^: " ' ^' 'c ':,' = 3 : :,w2:r:,':, -v. '#:L~::n ,s n &,, , ~ - - ~ ~ <,, ;~. < - ~ nw- ~. ,m, ~: - ~, w:;, N,,;~es' ', ",i'M4!'0,,2 j, 'd,':? ' ; y.; y:.. ,v: N d '0' E m.,. 'INITIA, TD4G C.O,v :,DFI1,DN,BSi ["i$,e'la'i^ir;eJb,'l:*^'^:'z:; ^ l b';513 'tl 2' : c ^ }[{ > 9 3"': < :<, a

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.~; ;..- ' ~, ,,. ~, '\\ '' z':ll, ', f &,':i_l l : ",';'. ' ' / 7, '~ i \\ : n *? ~~_^*3'Sn- [ ?; ',* lUS 1, ~ }, '* l W ,U '. '_'} ' g ?;"Z{, ;, j' '. \\ , 'z ; >' ~ ~ ;^' { ?, ", ' t ". ' + \\ '. 's, u.* .oD Initiating Conditions Entry Criteria G E.S1 S I m RHR Loops A and B are NOT capable oflowering RPV temperature. T BS1 E m

s N

g-y Completeloss of Suppr:ssion Pool temperature is above the HCL. A functions needed to R g a achieve COLD E B-tj SHUTDOWN A g" o to w E "Mi = M = E O O R S G E N C Y Applicable Modes: 1 2 3 l? oa W t.#t 5 6.2.2 BS1 (Cont.)

DISCUSSIONiddE)"U

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O - [ ';; %^ f 7- [ <s y 'V y s " A' : -.'j g + 'L, ' D" fM ^ l 'O n v' a ;s ' f', ig! C' u." ~ s r = This IC and its associated EALs address the loss of systems needed to reach COLD SHUTDOWN from Modes G 1,2, or 3. l L The normal method for rejecting heat in Modes 1,2, and 3 is via the Main Condenser. If the Main Condenser is i NOT available, heat may be rejected directly to the Suppression Pool via the SRVs. i The Suppression Pool will act as a limited heat sink until the ability to remove heat to the ultimate heat sink is m restored. If Suppression Pool temperature is greater than the HCL, it is assumed that the Main Condenser is unavailable AND heat capacity of the pool is severely degraded due to Suppression Pool low level or high m temperature. Per PEI Bases Document under PEI-T-23, Containment Control (Suppression Pool Temperature), 4 so long as the plant is maintained below the HCL, the Suppression Pool temperature will not exceed the design g limit of I85 F following RPV depressurization. g

2.

w w w Losing both divisions of the Shutdown Cooling and Suppression Pool Cooling modes of RHR inhibits the ability { P,, to reduce reactor coolant temperatures to less than 200 F. Loss of RHR Shutdown Cooling means all RHR m g modes as defined in SOI-E12 or ONI-E12-2. B: 3 i e

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REFERENCES' J, < '-

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l. NUMARC/NESP-007 (Rev. 2), Site Area Emergency SS4

2. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Table 1.2

3. Plant Emergency Instruction (PEI) Bases Document, Rev. 2 / Update 1 c

4. Off-Normal Instruction (ONI) E12-2, Loss of Decay Heat Removal, Rev. 4 i

w*

5. System Operating Instruction (SOI) E12, Residual Heat Removal System, Rev. 8 i

l 6.3 Category C: Loss of Shutdown Functions or Failure to Shutdown y cq3 j: ;, a.:

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': ' ~, ' : ^ , 3 M' y ~, ;, ; J ^ 1; ~(' ';e .a E ' ' ' ' ', ', l ' ! /, ': ^ ^r 2 s 'M ' ' ' n C, n '.,, i g. 63.1 INITIATING CONDITION CU1h d 2', ' y' 4' , :.;:~, 7 -^

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-( B 8 Initiating Conditions Entry Criteria G .C_U.1 U CU1 Plant is NOT brought to the required operating mode within the Technical Specification N action statement time following entry into an LCO. U m Inability to reach S required shutdown U m a within Technical A 3 Specification limits L 2. y o Fi 2 E g-g V = o 3. E e N Ti Applicable Modes: T {n o" 1 2 3 DISCUSSION) ~ d Limiting Conditions of Operation (LCOs) require the plant to be brought to a shutdown condition when the Technical Specification required configuration CANNOT be restored. The plant is within its safety envelope when being shutdown within the allowable action statement time in the Technical Specification. An immediate [ Unusual Event is required when the plant is NOT brought to the required operating conditions within the [ 4 allowable action statement time. Declaration of the Unusual Event is based on the time at which the LCO-specified action statement time period elapses due to failure in equipment needed to meet the action statement or it becomes obvious that the action statement will NOT be met. 6.3.1 CU1 (Cont.) .s.... ~-. s e s ,,s .s s e-e<s ~ <~, 4 ~ m ,<w. ,s. 3 ,,.,,,s 'c, ::(s5),.,,[ Cm ' M' ~ ' n $ : ' 'J<~'h~ ;i : ' .W>, <bN,.- ~ g s ,s , s b N ' M O ,' J[ [ : ^' ?'i; N: DISCUSSION}a at.)(' ' 1 (. ' 1; _ ~,,s, e :~ : . x:,.,+s ,) s

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< s $D Declaration should NOT be made because of an administrative oversight that results in an LCO action statement being exceeded and a controlled shutdown started. (i.e., unaware that the plant was in an active LCO.) Declaration should be made because of equipment failures that prevent the performance of an ordered shutdown or failure to meet the shutdown action statement from the time discovered and an active LCO entered. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement period elapses and is NOT related to how long a condition may have existed. s , ^.. ~ ~ ~ ~ . ~ s 's \\ h.'... f'. h t n +We '.-. .~.".v. v si[v/4r , i p ^ 6 .:g";y <, a :' J -

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1. NUMARC/NESP-007 (Rev. 2), Unusual Event SU2 g

a. w n o

2. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Sections 3.0.3 and 3.0.4

{ 2, 1 W m ao u o3Fs e m 00 O tu t 00 i 63 Category C: Loss of Shutdown Functions or Failure to Shutdown (Cont.) . ~ r -s h o- ~: y. w d:' w. ,s-c- v

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.u. [' ,,. q z q & ',, >~ z '>s,, ^~ ~._ ; ~ g \\ c - g%, ~ "' o ?., n~,;, i.,,:w<i, : a g f 3 J'^: a ; ' '_~.c : ' '^"r'c 5 A [ 63.2 ', INITIATING CONDITION CA1 '.' 4'5 J' ' y '" ' o \\. + - g., . _,-, ;, ;; p p< - e.- ~ y' ;;g,,., p ;;,; <,+;- ,a ~ 3 .;, - e + s ,,/ ,,=,e-o Initiating Conditions Entry Critena w l CA1 CA1 Following automatic actuations of either of the following," shutdown under all conditions without boron" has NOT been obtained: Failure to initiate or A

• RPS L

m complete an automatic Reactor Scram once an

• RRCS E

RPS function is R m a required T 3 v> k s Applicable Modes: 3-Sf E R 1 2 $g 5 = r ^ y ^ y ,Ihfh ^ - .g ' 9 ^ ~ DISCUSSIONS m.. g ne l CA1 is applicable if either Mode 1 or 2 existed when the transient started and NOT the mode which exists at the time of classification. This condition indicates a failure of the automatic protection system to fully scram the reactor. It is more than a potential degradation of a safety system in that a front line automatic protection system did NOT function in response to a plant transient OR an inadequate number of control rods inserted. Redundant Reactivity Control y System (RRCS) is included since it is an automatic system designed to back up the Reactor Protection System (RPS) for low level /high pressure situations. $e 6.3.2 CA1 (Cont.) ~ .. ~ = - <.~. v ., - - -~,,, sw,, s+ y . V'.'i't O '~.'7%^ ' .c, e. J'

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iDISCUSSION(Cont.) 6 .') L :: '- ' :: 'm' i ~ : ; ;,, o 1 s r' s, s s ,,s. ; n.' :Y.,.:'.m '>, s s~,, a., < ~. 4 , ~ x. .s ,c s s i e The RPS/RRCS failure could be due to 1) electronics such that all control rods did NOT receive the scram G signal; or 2) hydraulics such that all control rods did NOT full insert to complete the scram. Thus plant safety has been compromised, and design limits of the fuel may have been exceeded. " Shutdown under all conditions without boron" is defined by the PEI Bases Document under PEI-B13, RPV Control (ATWS) - ENTRY, as l either 1) all control rods are determined to be full-in, except one control rod may be at any position; or 2) as determined by a Reactor Engineer. m This EAL may be terminated if manual Operator actions achieve shutdown conditions OR when a Reactor p j Engineer determines that the reactor is shut down. Failure of the manual scram in Mode I would escalate this m i event to Site Area Emergency CSI. j l vs 9. J TD s S b'N ~ ' ? ' ' ' y -w 'x. g g wy, ,y' x y y ^ '; ,;c. ...y: REFERENCES 'ygg pp - w ~ 4 ~-"2 M..Ev '

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1. NUMARC/NESP-007 (Rev. 2), Alert SA2 ga

2. Plant Emergency Instruction (PEI) Bases Document, Rev. 2 / Update 1 F3 o

E R

3. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Table 2.2.1-1 h

T co OQ O 4O 4 h ..m m 63 Category C: Loss of Shutdown Functions or Failure to Shutdown (Cont.) 'g V V -, o / f's,. '. e \\ f ,/ jf ,<,Am ~ o ;' + ~:: g s ~ ' g, k-O ,\\ , ~ >,-, j,m~ /, ~ - Y, 3+ ; +,,

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y g f g,- ~ ~~ ,N : ~ ~,... 's: 'i: l E d IP'5',7 C;E :Ji!&l<;? ? i '6.3.3 ' INITIATING CONDITIONCSE J ,-l'E:J:' ': d '! s: r ;' ;'^ ;'I O f ': ,+~*s,y' y, ~,$~' ,'G ,~ - l tr ;,, yv.' g,: 2 : y,,, :,a,'a ;, :< e;,,, y ' ':' + l :c, f,, , a.+- :, ;, -,. ca ;; z z ;y ;:~ ~ ~. ,.m : c ~'.',', s,; y,y ,,s.,, ' ' +. g a Initiatina Conditions Entry Criteria ~ 9 CSI Following automatic actuations of either of the following, " shutdown under all conditions without boron" has NOT been obtained: S RPS I nc T CS1 F E m Failure toinitiate or j complete an automatic Manual operator actions taken at 1H13-Reactor power CANNOT be determined. A y Reactor Scram once an P680 were NOT successfulin lowering R 2 Fi RPS function is Reactor Power to less than 4%. E 3.- ww required, AND a A { o,, manual Scram was to w N_DIsuccessful E g = M = E O O R G E N C Y Applicable Modes: c 1 so OQ O b w 6.3.3 CS1 (Cont.) - - ~ n - p ,7 - ~ c , c- -,, ,:: - u:- n : :z ~' ~ $' L:f l'? ! 'f':' 'i' D 'D 6['? 3::b' ' yww ./~-- ;;,' ':'l'^Y;3- ~ DISCUSSION' ': l!~' J'

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CSI is applicable if Mode 1 existed when the transient started and NOT the mode which exists at the time of G } t classification. Refer to CA1 for Mode 2 applicability. This condition indicates a failure of both the automatic protection system and manual efforts at Control Room .i panel IH13-P680 to scram the reactor. Four percent reactor power was selected to identify a successful manual scram. This power level is consistent m i with the decision process used in PEI-B13, RPV Control (ATWS) - ENTRY. Power levels above the average y power range monitor (APRM) downscale trip setpoint of 4% may challenge the ability to limit Containment m heatup and may require actions to deliberately lower RPV water level per PEI-B13 (ATWS) to reduce reactor j-i power. This threshold should NOT be confused with the definition of" shutdown under all conditions without g! boron," defined in the PEI Bases Document under PEI-B13, RPV Control (ATWS) - ENTRY as: 1) all control rods are determined to be full-in, except one control rod may be at any position; or 2) as determined by a Reactor g. y Engineer. gg 6 A manual scram is any set of actions by the Reactor Operator at IH13-P680 which results in a scram signal. I h These actions include placing the Reactor Mode Switch in the SHUTDOWN position, arming and depressing the 5 l RPS Manual Scram push buttons, and arming and depressing the RRCS Manual ARI push buttons. Injection of y boron is NOT considered in reducing reactor power below 4%. g s I If Reactor power is unknown and the Reactor is NOT " shutdown under all conditions without boron," then it CANNOT be verified that power is less than 4%. A concurrent challenge to the ability to cool the core would escalate this event to General Emergency CGl. [ t %e7. i N i I t l i L 6.3.3 CSI (Cont.) ~ ~, - ~ ~ W. ve$ ? ~ \\ ~,?+ i s, ~ h '~ r '< > ' ' ' ', ' r '< > c .' ; E ^ A y$k: i' l~ , f, c' ' l REFERENCES 'c ' c :-. - t.,' U2,'i 'M 00 O b t.*) 63 ('ategory C: Loss of Shntanwn Functions or Failure to Shutdown (Cont 3 w s:, y :+,, ;7

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s Initiatina Conditions Entry Criteria

~p Following automatic actuations of either of the following," shutdown under all Cfd conditions without boron" has NOT be obtained: G

• RPS E

CGI

• RRCS N

E Failure toinitiate or R complete a successful Manual operator actions taken at 1H13-Reactor power CANNOTbe determined. A y shutdown, AND P680 were NOT successfulin lowering L g-E indication of an extreme Reactorpower to less than 4%. g challenge to the ability E 2-g to cool the core M = o 8. E to w R g = Either of the followin8 conditions exist: G =, E O O Entry into PEI-T23, Containment Flooding. N In the UNSAFE region on the HCL figure C Applicable Modes: Y 1

• u ao 00 0

~.... _. -... - -. - 6.3.4 CGI (Cont.) n. . a ~ sm ma gy ^' E h ~yg'SSIdN - , -[',,' d em%g a= a ,r gyrn a sm ~ n= = ~ ~ ~ ~- 5 w I CGI is applicable if Mode 1 existed when the transient started and NOT the mode which exists at the time of classification. Refer to CA1 for Mode 2 applicability. This condition indicates a failure of both the automatic protection system and manual efforts to scram the reactor concurrent with a challenge to the ability to cool the core. Four percent was selected to identify a successful manual scram. This power level is consistent with the decision m process used in the PEI-B13, RPV Control (ATWS)-ENTRY. Power levels above the average power range y monitor (APRM) downscale trip setpoint of 4% may challenge the ability to limit Containment heat-up and may gs require actions to deliberately lower RPV water level per PEI-B13 (ATWS) to reduce reactor power. It should j NOT be confused with the definition of" shutdown under all conditions without boron." E! Fir-3 A manual scram is any set of actions by the Reactor Operator (s) which results in a scram as defined above. 'Ihese 2-actions include placing the Reactor Mode Switch in the SHUTDOWN position, arming and depressing the RPS g P, Manual Scram push buttons, and arming and depressing the RRCS Manual ARI push buttons, injection of $g j boron, and PEI-SPI actions. If control rod insertion actions are still being implemented when a core limit is = = reached, a General Emergency shall be declared. { 9 If Reactor power is unknown and the Reactor is NOT " shutdown under all conditions without boron," then it { CANNOT be verified that power is less than 4%. For event classification purposes,"an extreme challenge to the ability to cool the core"is defmed as either-Entry into PEI-T23. Containment Flooding, based on an inability to adequately cool the core. During an ATWS condition, PEI-B13 RPV Control (ATWS), directs the Operator to deliberately lower RPV water level below the TAF (0") to reduce reactor power. Assurance of adequate core cooling is achieved when RPV level can be maintained at or above the Minimum Steam Cooling RPV Water Izvel (-30"). Under this l? I 4 ATWS condition, steam flow through the core is sufficient to preclude the peak clad temperature of the E hottest fuel rod from exceeding 1500 degrees F. If level CANNOT be maintained at or above the Minimum 8 Steam Cooling RPV Water level, Operators are directed to initiate Containment Flooding per PEI-T23 in an attempt to re-establish adequate core cooling. i i i 6.3.4 CG1 (Cont.) ', w.

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p, ; z 3 ~;;;::;; ~ ; z a. ~ 5 t n r In the UNSAFE region on the Heat Canacity Limit (HCL) figure. PEI-T23 directs the Operator to initiate an G emergency depressurization per PEI-B13 in support of Containment Flooding. Per the PEI Bases Document under PEI-T23, Containment Control (Suppression Pool), sufficient Suppression Pool heat capacity will be available to ensure that the initiation of RPV depressurization will NOT result in exceedmg the PCL before the rate of energy transfer from the RPV to the Containment is within the capacity of the Containment vent, so long as Suppression Pool parameters are maintained outside the UNSAFE region on the HCL figure. i Therefore, availability of the Suppression Pool is critical in support of restoring adequate core cooling m through Containment Flooding and ensuring a heat sink is available for heat removal via the SRVs. 1 m ' : (;, }'N f ~,[^,[.^j~ f :: , ' 5'U i'( Q^' ' : ~ ~ ;,,l ^ l J. ; . ':'

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u ; ' ',, +

g - Z., w - :,..

,L ? ,,5, - g g w = oo

1. NUMARC/NESP-007 (Rev. 2), General Emergency SG2

= o B. i ~

2. Plant Emergency Instruction (PEI) B13, RPV Control (ATWS), Rev. A g3 g

l i

3. Plant Emergency Instruction (PEI) T23, Containment Flooding, Rev. A l

O i o e t

4. Plant Emergency Instruction (PEI) Basis Document, Rev. 2 / Update 1

?.

5. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Table 2.2.1-1 I

l t 't2 23 0"2 l O A-O f l r f l 6.4 Category D: A. C. Power Loss ^ : ' ' c- :' l ; L

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', ' 7':

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y c ,p.7' y,,~,; I5M11) INITIATING CONDITION DUI /': E ;, ' ~' j X'5'F"<l6:'"; J * (, #-if d:if Oi:';;4;;;: [ if ^' ~' , ' : 7, M y' 8- '~

m, y

'e', n . in 8 Initiating Conditions Entry Criteria G Dill U N U m DU1 ONI-RIO entered for a Loss of Off-site Power (LOOP). S y U m a Loss of all offsite power A 3 to Division 1 and 2 EH Either of the following power sources Either of the following power sources L g E Busses for greater than CANNOTbe made available within CANNOT be made availabic within g 15 minutes 15 minutes for energizing Bus EH11: 15 minutes for energizing Bus EH12: E S-M V g 2, Normal Preferred E $g NormalPreferred e Altemate Preferred e Altemate Preferred N 5 T S Applicable Modes: p av 1 2 3 4 5 D .m .DISCUSSIONii ^ ^ '^ mme %% ~ w; i Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a loss of off-site power (LOOP) as defined by ONI-RIO. [ o A 4 6.4.1 DUI (Cont.) g v:., ~, ,; m - ~., 2 ~ ~ y,;y (. 3;;, up .g - ag,, y s, w, n v, ', l',.. g, a': '7'E ~ n s e _w, DISCUSSION (Cent.)L'i. - -~ p g ay y 3 ;,;;',:73 ,; 3,

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s Technical Specification 3.8.1.1 for Modes 1,2 and 3 requires two physically independent circuits between :he G

offsite transmission network and the onsite Class 1E distribution system. 'Ihe Normal Preferred power source from the Perry Plant Transmission yard for hatit usses EH11 and EH12 is from the Unit 1 Start-up Transformer, b supplied through Interbus Transformer LH-1-A to bus THl. The Alternate Preferred power source to either EH11 or EH12 is from the Unit 2 Start-up Transformer, supplied through Interbus Transformer LH-2-A to bus TH21. Per ONI-RIO, a loss of offsite power (LOOP) is defined as ".a loss of offsite power has occurred and at least f one of the Division 1 and 2 Diesel Generators has supplied its respective bus". IC DU) addresses the ability to y re-energize either bus EH11 or EH12 from the Normal Preferred or Alternate Preferred power source within 4 15 minutes. g! j d E Fifteen minutes was selected as a threshcid to exclude transient or momentary power losses. However, g. g classification should be made as soon as it is known that the availability of offsite power will NOT be { p, re-established within 15 minutes. The intent of DUI is the availability of either the normal preferred or alternate i preferred sources to energize their respective bus (es), and NOT the physical connection of either of these fE [ supplies to the bus. g 9 Credit is NOT taken in this EAL for the Division 3 Diesel Generator because it only supplies power to the High 5 l Pressure Core Spray (HPCS) pump and associated loads, NOT for any long term decay heat removal systems. l In particular, Suppression Pool cooling mechanisms would be essential subsequent to a station blackout. l f Failure of either bus EH11 or EH12 to be supplied from its respective diesel generator is evaluated for escalation [ l to an Alert under IC DA1 for Modes 1,2 and 3. Failure of both busses EH11 and EH12 to be supplied from their respective diesel generators (Station Black Out) is evaluated for escalation to an Alert under IC DA2 for Modes 4 and 5 and to a Site Area Emergency under IC 7 DS1 for Modes I,2 and 3. t i 6.4.1 DU1 (Cont.) ~ y car - w .,.c ~ N ^ s n ,N _- '[. /, i As 18c ^y:$Q'>. . <............. g)$( ~' ~ ~ 'l'q;. q*g ^ '^ ^ x s m, g g v REFERENCES 2

1. 19 J

gi: Whs-59... ; hin. ", ^. ' s D* n

1. NUMARC/NESP-007 (Rev. 2), Unusual Event SUI G

2. Technical Specifications, Perry Nuclear Power Plant, Unit 1

3. Off-Normal Instruction (ONI) R10, Loss of AC Power, Rev. 4/TCN-3

4. System Design Manual (SDM) RIO, Plant Electrical AC System, Rev. 8

5. Updated Safety Analysis Report (USAR), Chapters 15.2.6 and 8.2.1

%r m s m? s. a n. u m 88 O 3 m A w CU W

=

oo N = o U.v N as OQ O AC >w 6.4 Category D: A. C. Power Loss (Cont.) v ^ ', ; :; i S;; L 'x' ' } 3 33 :;\\, u:,,; ;[ x :? e 2 = :, :', w ' 9 ' :n: c'm 3 ; M. g:: 'q" 3 72;f:; y a' 6A2 : INITIATING CONDITION DA1 L w'4' 'i'$y';:': j 'i'y+ 'f  ; ', : q ;cp '"'; p "ivt ? j~ ' %', g;^y'gyhy$1 g

43.'F:l{'5' !:'I '"'l4 n'f ME' f $3M $ g 2 ' i:, /~

4 ' '; p s y l p ' f: p/ g. s: 1 j ; ~' ; O ' y: ' '(p g;; s a Initiating Conditions Entry Criteria ~ Essential AC power reduced to only onc of the following power sources for greater DAl DA1 than 15 minutes: Power capability to Division 1 and 2 EH + NormalPreferred m Busses reduced to a . Altemate Preferred A p single power source for

• Division 1 DieselGenerator L

m j greater than

• Division 2 DieselGenerator E

15 minutes, such that R g any additional single T Fi failure would result in Loss of the single remaining power source will result in a loss of AC power to both ?=~ iD Station Blackout busses EH11 and EH12. [ 2 to G Applicable Modes: } 1 2 3 9 aC v :.: . e * + r.f:q s

w:

DISCUSSION?

~ 4 m M*g 'M,zy*q ' lh(. - v + y This IC and its associated EAL provide an escalation from IC DUI," Loss of all offsite power to Division 1 and 2 EH Buses for greater than 15 minutes." 'Ihe condition indicated by this EAL is the degradation of the offsite and onsite power system such that any additional single failure would result in a Station Blackout (SBO). m = oon IC DA1 is only applicable to Modes 1,2 and 3 and is concemed with the degradation of offsite and onsite AC g power such that the loss of any single source would result in a SBO, as defined in ONI-RIO. Credit is NOT taken for Bus EH13 which only supplies power to the HPCS pump and associated loads, but does NOT provide power to any decay heat removal systems that would be critical in a SBO scenario. 6.4.2 DA1 (Cont.) 2,, au m, m ~>, w ,+ "m;& % '~ - N~ Efr. 4 1 ^ " " /"% W a4!i_ @ w %'? d if

DISCUSSION l(Coni.)! * *?

~ ~ g g y,, g. = ,3 ; p g B B Fifteen (15) minutes is allowed prior to classification to either: G Restore redundant AC power source to EH11 or EH12 e Provide separate independent sources of AC power to EH11 and EH12 i The loss of the normal preferred and alternate preferred sources with the respective diesel generators powering hath EH11 and EH12,is classified as an ALERT under DU1. m>r Escaladon to a Site Area Emergency is evaluated under IC DS1, for Operating Modes 1,2 and 3, based on a g1 totalloss of AC power to hath busses EH11 and EH12. 3 m.,. A total loss of AC power to busses EH11 and EH12 while in Operating Modes 4 and 5 is classified as an Alert I under IC DA2. No escalation path exists to a Site Area Emergency for Operating Modes 4 and 5. g-8 E 2 a

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m .y... +~~ -y ( g 9 4

m.

~.. " 5 . REFERENCES! ~ ~ m<_ w~ - a sys ~e v

e::

.q

g;:j s._.-

- ;gg;g - - - - ~ ra %o O

1. NUMARC/NESP-007 (Rev. 2), Alert SA5

{

2. Off-Normal Instruction (ONI-R10), Loss of AC Power (Rev. 4)

3. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Section 3/4.8.1

4. Updated Safety Analysis Report (USAR), Chapter 15.2.6 2?

4: S 6.4 Category D: A. C. Power Loss (Cont.) n . ^: 7,', 't,m c ; '; -;, ' ':a; %3;,x ~s.' :, M.,,y ;. 6.4.3 ; INITIATING CONDmON PA2 M~, i ~';,x :+.,;;4 :,,',,}:M'; e ::_ 'T 'T*f p',~y ,e : -. < < ~ +, =..., ~ ,- U

., z

<4 MM. ui M c':: ,;,~;c z: ~ y .y B :(: J6 ;;%:'# g ~ - ;,c'3 r,.; p ' 3; % ;<,,;; p y ;;;; ,. ~, ; : : g 3 :, : y ~ -:, : y q, g 3, a-

', ' ;c,

'y,. '3-c, ~ Bae Initiating Conditions Entry Criteria G DA2 DA1 Loss of all offsite power Both busses EH11 and EH12 CANNOT be energized from any source within AND onsite power to 15 minutes. A Division 1 and 2 EH L m Busses for greater than E y 15 minutes. R m = T 4 Applicable Modes: n y if. o Fs 2 4 5 D B-g = .o, o

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^#@ ' F ~ y a s sw- 'f a E = 0o IC DA2 deals with loss of the minimum required offsite circuits and failure of the diesel generators to restore power to the emergency busses resulting in a loss of AC power to all plant safety systems requiring AC power including: RHR, ECCS, Containment cooling systems, spent fuel heat removal systems, and Suppression Pool cooling systems. IC DA2, is only applicable to Modes 4 and 5, and is concerned with a total loss of AC power to both busses EH11 and EH12. Credit is NOT taken for Bus EH13 which only supplies power to the HPCS pump and associated loads, but does NOT provide power to any decay heat removal systems that would be critical in a Station Blackout (SBO) scenario. 2 =0 t/t tJ 6.4.3 DA2 (Cont.) , - as tie: m.;m en ^ c~ ' .', c ' r,, ~, c, ;~ y 2 e M A: p h.~-.~~.> ^.,., > ~ m 4 3 ns ,w.s s s s MP.c as:. s .m .,a8.

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^ m we'*s*.T *.W'"I*J - t . - f~ ' :' E if '~ lDISCUSSION (Ca z s a ww - y , ^ e o z s- ^ f- <s. sf s s s 3" pg W^"6 w

s+d G

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j.

m Fifteen (15) minutes is allowed prior to classification to wstore a single source of redundant AC power source O to either EH11 and EH12. Fifteen minutes was selected as a conservative lower threshold that retains the anticipatory nature of the EALs while excluding transient or momentary power losses. However, classification should be made as soon as it is known that power to either EH11 or EH12 will NOT be able to be restored within 15 minutes. No escalation path exists to a Site Area Emergency for Modes 4 and 5 in Event Category 'D'. Criteria m established in IC DA2 would be considered a Site Area Emergency in Modes 1,2 and 3. m s. s,, s. w _... ;O o yge. y :.:.m. w: ., f;::va TiiMMx g:g.a r ^^?M

• 2 x

MA T:p:; ' x ^ gg:"s Y s .e KIN!! '.. g 'A ^' e 5 '$N:$3s n sm:. - - zwffp:;MLpr wwmm ' V. g ~>.s a ~ s, ~ ,.o e. - ~ .:.mp-Qg seg m s. me g c. A m u

1. NUMARC/NESP-007 (Rev. 2), Alert SA1

{ 2, to w

2. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Section 3/4.8.1 l

m t o

3. Off-Normal Instruction (ONI) RIO, Loss of AC Power, Rev. 4 g

r-s

4. Updated Safety Analysis Report (USAR), Chapter 15.2.6 em 90 4

O t.A W i 6.4 Category D: A. C. Power Loss (Cont.) s s . m ;-, a.e 2 1, >s, ,2 <,x %;- c 7.,, .,4-y. ,,;~ ..4,,

• p-g ; ;q:, 3;;

y ;,:p %y. Ei ^ T c '~e,;, 'R M' ~ ':# 77: 'c; A: 6.4.4INflTATING CONDITION DS1

2

&, c- :' 'c s- ~ :~?. ' +' - e 's s .s" ,s :gy ' - ' ', ~ <:- n-c e, ': ';; / o ~. a.m, ',- ~.- -, ~ cc ', s.- - m,- p, ,, :~ e ~.e ~, w, .a '~ Boa Initiating Conditions Entry Criteria G M S I T m E c~ DS1 m t u A 3 Loss of all offsite AND R y o onsde power to E g a Division 1 and 2 EH Both busses EH11 and EH12 CANNOT be energized from any source within A g-g Busses for greater than 15 minutes. g o,, 15 minutes E to w M = o E = R O O G ?. E N C Applicable Modes: Y 1 2 3 7 .,.s gy ~ s [f.33,. g 1

DISCUSSION =:t

~ ~ ^ ^ ~ k: ^ ' a

m y,g g, ;

y y g IC DSI deals with a loss of all AC power compromising plant safety systems requiring electric power, including: RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. l 6.4.4 DS1 (Cont.)

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s IC DS1, is only applicable to Modes 1,2 and 3 and is concerned with a total loss of AC power to both busses G

EH11 and EH12. Credit is NOT taken for bus EH13 which only supplies power to the HPCS pump and associated loads, but does NOT provide power to any decay heat removal systems that would be critical in a Station Blackout (SBO) scenario. I Fifteen (15) minutes is allowed prior to classification to restore a single AC power source to EH11 or EH12. I Fifteen minutes was selected as a conservative lower threshold that retains the anticipatory nature of the EALs m while excluding transient or momer.ury losses. However, classification should be made as soon as it is known y that power to either EH11 or EH12 will NOT be able to be restored within 15 minutes. m =1 Escalation to a General Emergency is evaluated under IC DG1, for Modes 1,2 and 3, based on a continued g-g! degradation of core cooling capability. g Q

2.

S = w .p:-Jf ^ MMP + ugs.9 M:~/gpqg- ?;e.> yer m9 y O s: ^ 7 ~J - b:jipgggg0i ):{. .,i:ge M4+ mms x ~. % ~p~

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<s 3 m 3 gg $y.s,gppi;g[ ggg,g ,......r.. mg: 7.m3x, x 4 iN$d J agg g" g l yey<. ',, nd T.y 4, - iREFERENCES, ?!? m; ~ m.. - : m m 9;. my gy g o i ra

1. NUMARC/NESP-007 (Rev. 2), Site Area Emergency SSI B

O3

2. PNPP Engineering Calculation CEI-03, MAAP Run 10_00. i',, Station Blackout (no injection)

C r

3. Off-Normal Instruction (ONI) R10, Loss of AC Power, Rev. 4

4. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Section 3/4.8.1

5. Updated Safety Analysis Report (USAR), Chapter 15.2.6 m

so 00 O t.#t t.on 6.4 Category D: A. C. Power Los c (Cont.) s, .,.s , '- ~ s.~ _.m....<m.s s, -:< (^ ) s t 3 (< 's y e,I .s/< >^ 'm. S,, s ~ ^ '^ '6A.5 - INITIATING CONDITION DG1 lN - 1;:'J M'<c C : e ^1' ' ' ' DC': " ' 2 f - is ' t' J- : %'.? '.:'a ' #"I' E -.. M. ',,

" '

. <.> ^,',:.

E, n'.e< l, l 7- ~s. > '

' ^ + +

:

n-o B '_ r. 2:, >,, : ;, - :,, ' '; ^ ~ v s. s s, - ;. as- ,s Oa Initiating Conditions Entry Criteria G Dfd G DG1 E N m Pmionged loss of all Both of the following busses CANNOT be Entry into PEI-T23, Containment E y offsite AND onsite energized from any somce in less than Flooding, based on inadequate com R m j power to 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s: cooling due to a loss of ECCS capability A Division 1 and 2 EH L p . EH11 Fii-3 Busses, AND continuing

• EH12 E

L A

degradation of core = m cooling capability M = o o m E en <a R g G 2 E o O N E C Applicable Modes: Y 1 2 3 - ~ um,. . m

• ~

DISCUSSION! 1 ^ s 't '^ ^

w n W

D3 m O u Loss of all AC power compromises d plant safety systems requiring electric power including RHR, ECCS, m Containment Heat Removal and the Ultimate Heat Sink. 'Ihis event is escalated to a General Emergency based on a prolonged loss of d AC power leading eventually to a loss of fuel clad, RCS, and Containment. 6.4.5 DG1 (Cont.) y 'p: n ,,, ;, ~ :y,";- ;;; ;,~ ; c, 'y,~, z ~, : ~; m. m~. n - :.. w.

~

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. ' - > - 7^, j,m; : ' l,, ". O, ji ':: - cs.,,+, ' '~ '.,' q~T >n<, y ;', y > ~;> ,;Q j, ,, ?: ,,~ , s,'~y s' 'c;>,4,,,':; C M':: ,:'L : ' ' ' :~ M if ^ DISCUSSION (Co,at.) >. - s. s c-o

s

:,.

5 <,,tc ,~-,;'t , ~' 2, : ' ' ',, ~. 4 . :.- ^ *, y, ,, ~ ~, ',',~,'<:, Q* <s n ~ s,. n A four hour restoration time is allocated to re-energize either bus EH11 or EH12 from any AC power source. G This restoration time is based on the Station Blackout (SBO) Coping Analysis described in USAR Chapter 15, Appendix H. For event classification purposes, the " continued degradation of core cooling capability" is defined as entry into PEI-T23, Containment Flooding, to re-establish adequate core cooling for ATWS and non-A'IWS conditions. l To ensure continuity with the Fission Product Matrix (Table A-1) and Initiating Condition CGI, the criteria m established by the PEI Bases Document for adequate core cooling under ATWS and non-ATWS conditions is p i used. m = ' 4 u., 3,;,,, p .o ,x oy. - ~. ,,g,.,, , e .,,w, m

. ~. -,

- 3 ', ;. ; '; [~ ' v z (/ <~ ~ ~' _~, ~ , y 1 :: y~ [, n', 4

REFERENCES, _ ; J,,

', ',:f:f ',. L ' a: ;~ -;;' L'] :':i:'. x '.;: ' :: ':n, y,,,': y )y ~, g ( m c w o

1. NUMARC/NESP-007 (Rev. 2), General Emergency SGI E.

[ tn w w oo

2. Plant Emergency Instruction (PEI) B13, RPV Control (Non-ATWS), Rev. A 19 i

3. Plant Emergency Instruction (PEI) B13, RPV Flooding, Rev. A h

l

4. Plant Emergency Instruction (PEI)'I23, Containment Flooding, Rev. A i

t

5. Plant Emergency Instruction (PEI) Basis Document, Rev. 2 / Update 1 i

6. Off-NormalInstruction (ONI), RIO, Imss of AC Power, Rev. 4

7. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Section 3/4.8.1

8. Updated safety Analysis Report (USAR), Chapter 15.2.6 and Appendix 15H

.o 00 O t>t 4 +- r Sheet 50 of 138 Page 58 EAL Entry Criteria ml Bases (Cont 3 l I THIS PAGE INTENTIONALLY BLANK 6.5 Categorv E: D. C. Power Degradation ~ 3' ' ' ~

, p

' :

. x ;; p - ~' .c .;~.. ~ ':, ~ 6.5.1 INITIATING CONDITION lEU1' ' ^> < J ' ' 'f' ': ;, ' ' cs ~ ' O ~ ^' '; [ ' ; l ' ' ' [; ' ' l'!'$ '^ "; ' ~' g. 5 c - : - - .s ' ' - ^ ^ B 8 Initiating Conditions Entry Criteria G Elli EU1 Voltage on both of the following busses is less than 105 VDC for greater than U 15 minutes: N Degradation of U m Division 1 and 2

• ED-1-A S

>r essential DC power for

• ED-1-B U

m j greater than 15 minutes A L m S. I a a E 1 u m V = o a m E a to w N g = Applicable Modes: T 3 ,o o 4 5 Ev S.^

f:.?

DISC. U.S.SIONs. ~ Yfs i@dE ~ ~ ll.@3;4 ~ 's ^ 41 v.,, ~ s; n / [^ IkN ~ h. y. 4 ,g; This IC and its EAL recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during COLD SHUTDOWN or REFUELING. This EAL is intended to be anticipatory in that the Operators may NOT have the necessary indication and control of equipment needed to respond to the loss. 2?% Credit is NOT taken in this EAL for the Division 3 DC bus because this power source does NOT affect any E decay heat removal systems. 6.5.1 EU1 (Cont.) u -v'm'

:= ~:+- ' r :'- '

^~~ zuw ma :ssg' G; k:yM ' ~ ::: ' i'N' 'l 'b' cL' U 6'I: mn, as~ s '., : >-'~ 'O ' : ' > ' ' b ' 5' h. I ' N? ' ' [$ $N:h'h lbbk'U5iSIUlIlbNdI/N ~' '. :~- va ,c ~ y,z, V 'n o ~1 t

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v' ' ' ' - ' w . 'c > ',':'. ~ > ' '. - Q* 'l .~ ^ Lyy $n One hundred five (105) volts DC is the minimum design voltage of the 125 VDC,60 cell Division I and II G batteries. Thus 105 volts DC has been selected as the threshold for a loss DC power. The same set of conditions as described in this EAL would be classified as Site Area Emergency ES2 if they occurred during Modes 1,2, or 3. s's, aan'perg:p+:egd$h,. e w usv : en e w g~~*<~4 g6,.,; % *

• py y # F-na -

% w me gm mm ww =g:a w + w w.ch.,..~ -we; - n tr3 Az 7- ..s m::e 7jjg' y g m jyg, ,g f 'Te d d 5

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~ za! REFERENC.gS?*5 T-, s an.m /,**@ ew.,,,,... ",, E .. ~ g^ g .e'a r gg. s a ag +~ gar < li[kE:$s@7 "gg, gg,. ^ $y[7 ' ~hikppy ${$ Mit;$. ^ 4 s t.5 ff y "e~ "' 45s 24@j$ N:R s m f s y4[ ERO, $$sW:f's?:!g:;:" N 08W ~5 3

1. NUMARC/NESP-007 (Rev. 2), Unusual Event SU7 lp a

Fi R

2. Updated Safety Analysis Report (USAR), Chapter 83.2 p,.

g [ 'i o 8

3. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Section 3/4.8.2.1 to w

n oo

4. FCR #021930," Emergency Action levels, EAL DC Voltage, R42" O5r*

? h i L

• t3 cm 09 L

os i o i L v i 6.5 Categorv E: D. C. Power Decradadon (Cont.). s >. ; ;,' i ;'m, :^ E,,, .'~-s, }7] y;' :'b';;;Q ~ '>

, q x v.x, ;.,

x :u,,',x,;, x, ~,s'::-' ? ~~ x A; ' r,. z,, s ?,,.,,y,,,,, ; b. '. a-

, ~' c~.,. ;,..

,. 7,

'85 ' INITIATING CONDITION EST,:-e'cO /Q;::QU

u 'H'6' V: c'a,,'.2 7 f ?'sf'; / x ;:'!4 #*,'6!s:i'y

~ h.~ ' if s;': - : a f3' ; w:.,.: 3. ; ~; < ~,:7 :i~'q; a ' e ' '. ~:6.,, ':l 'y' + ' ' ','.M '., ~l ,' ; : :~-' ' t'; y,!: :::;~ ~ ~:xBx e; %_, 8 i gan Initiating Conditions Entry Criteria G s E.Sl S I T trs E t~ m n ESI A y R m 9-g" Degradation of Voltage on both of the following busses is less than 105 VDC for greater than E g Division 1 and 2 15 minutes: A g-g essential DC power for E o greater than 15 minutes . ED-1-A E gg . ED-1-B M oo g E = R Q O E E N G C I Y Applicable Modes: 1 2 3 m to OQ O Ch. e-m 6.5.2 ESI (Cont.) ^w m u. w, m

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,,'x .,3 s ss 'mn ~,- gh,~}5 r? ' ', < ;, ^ n fg ;3nsygb >. n' <s s -, < A~, s

DISCUS 5IOl@Sny*DQP 8[' Ui30$[, Q '"' d', 'Q 5 l'i / [ ;.1, { 'e !

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E m" ,, 3 i nema $%ny \\M%Wn u + M? Y ' ' ' ' > ~' J' ~ ~

p' : ;:

?, 'r ;' ': ' 8-Bo:s This IC and its EAL recognize a loss of DC power compromising the ability to monitor and control the removal G i of decay heat during POWER OPERATIONS, START-UP, and HOT SHUTDOWN conditions. It is intended to be anticipatory in that the operating crew may NOT have the necessary indication and contml of equipment needed to respond to the loss. This EAL represents a more serious condition than that described in Unusual Event EUl in that the initial temperatures, pressures, and available decay heat may be substantially higher than in l Unusual Event EUI, resulting in significantly less time available before failure of systems needed to protect the public. m>r Loss of M DC power compromises the ability to monitor and control plant safety functions. Prolonged loss of m I d DC power may result in core uncovery and loss of Containment integrity when there is significant decay heat j and residual heatin the Reactor. m I 9 Y f i a Credit is NOT taken in this EAL for the Division 3 DC bus because this power source does NOT affect any g. g decay heat removal systems. m o 8, i r One hundred five (105) volts DC is the minimum design voltage of the 125 VDC,60 cell Division I and II E M l batteries. Thus 105 volts DC has been selected as the threshold for a loss DC power. ,o g i ~. ..m . 4 ~ ~. _ : g.., e 1 ggggggycgg  ; ~ ,,L OL; L, fc 'p ' : ~ ~ ' ' ;s ... c: s - ", ' ; 3 ':; ; ; .~' i '?,

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1. NUMARC/NESP-007 (Rev. 2), Site Area Emergency SS3 I

t

2. Updated Safety Analysis Report (USAR), Chapter 83.2

3. Technical Specifications, Perry Nuclear Power Plant (Unit 1), Section 3/4.8.2.1

.o =%

4. FCR #021930," Emergency Action Izvels, EAL DC Voltage, R42" m

o i. F i 6.6 Category F: Fire or Explosion ~ ;>,,Q.;:' #y ';!,;'g;', :jj"jQ:, ;~ [LQr'R]:l 611 ? INITIATING CONDITION FU1 r-:', ' ?;;,{:g: iQ;=':g[;;}f' M :'; '^ic -;}~r c^

1. ' ^[;N:r ;]-Qj,

[^ L 4'lLll< Ql,;; ';l ^ ' Q s) :' L' ~.e,8 7:H <: c'le' - '4%A [

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, y. ~:3';yy,,:'p y,; :y,;y' 'n ; ;,, +4 5n Initiating Conditions Entry Criteria G Fue within any Safe Shutdown Building. FUI U Fire CANNOT be extinguished within 15 minutes of either of the following: N FUI U m e Verification of alarm. S Notification received in the Control Room from plant personnel that a fire exists. U g1 Fire within a Safe Shutdown Building A g y NQT extinguished L within 15 minutes g a E

s.

u = u V N O o Q. E to u N 5 = 1 c T = .,o o 1 2 3 4 5 D E s " 4gd: v Q-idf hif:;?'^ 4^ fu f

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,.p y x m. J-M,'y- ~ DISCUSSION 5-s.. ^ ~ yp;g,.34g.. 3ffjs/ [t s...., + -W.

.3

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A fire is as defined in PAP-1911," Fire Emergency." m This IC and its associated EAL address fires that are of sufficient magnitude that they may be potentially g significant precursors to damage to safety systems. This excludes items such as fires within administrative w buildings or other structures NOT contiguous with a Safe Shutdown Building, and other fires of no safety consequence or threat to a Safe Shutdown Building. 6.6.1 FUI (Cont.) i f

' ;'[

Q ly'd ' ; ;' lf'2~' y a, ' x"',' ( 'y [Q' o{'4'(w' 1 :d%' '".:a: 7 ' ^',lL;': c, ';';, Q' ':;x y:' ', [' y) :.; u, ' ~ ':' ' c'. ;. < ',' - ~~ <, + . u ,L ' : 1~ i' M ' g '. '? '! S v : DISCUSSION (Cont.) < < ~ ',x n cy ; g e .o 'p', -,, ;, e,~,, e,c :;;~ ' c ; e, 7~,,- . q c' - "; : y e,,s 1 ?~v s e,~, ze 3*. m:,,.n w u, 'v ~ o .~ s 5 l = Verification in this context means those actions taken in the Secondary Alarm Station (SAS) to determine that G the alarm is NOT spurious. Verification includes the receipt of multiple or independent alarms or confirmation of a single detector by visualinspection of the affected area by a first responder. If an inspection of the area is completed within 15 minutes with no evidence ofit fire, i.e., spurious alarm, no declaration need be made. De 15 minute time frame has been established to exclude small fires that can be controlled by Fire Brigade resources and have no impact on the performance of required safe shutdown systems m or components. p m Notification includes all verbal means oflearning of a fire, j v1 For the purposes of this IC, Safe Shutdown Buildings / areas are considered to be the following locations: ( a. u ControlComplex(allelevations) l 1 2 Auxiliary Building (all elevations) Intermediate Building (all elevations) = Fuel Handling Building (all elevations) 2 ReactorBuilding(allelevations) 8 Emergency Service Water Pump House (all elevations) Electrical Duct Chase Ixading to ESW Building Diesel Generator Build 4 (all areas except the Unit 2 Division 1,2, and 3 DG Rooms) SteamTunnel(allelevations) Diesel Generator Fuel Oil Storage Area Condensate Storage Tank Intake / Discharge Structure 2? 'E t 6.6.1 FU1 (Cont.) as a 'W ed-a /,a 3 m - g >' @jf^ @g + 'i ' %.3,$.A% ?M ',

m...

., ~ ~. ' ' t, ,s, e r / s-s s' s 't E$f g: ;: ?! MN 4 s s a

1. NUMARC/NESP-007 (Rev. 2), Unusual Event HU2 w

2. Updated Safety Analysis Report (USAR), Chapter 9 and Appendix 9A

3. Fire Protection Evaluation Report, Sections 3 and 4 - Table 3-1, Rev. 4

4. Appendix R - Evaluation, Safe Shutdown Capability Report, Section 4, Rev. 5 m>

F

5. Plant Adannistrative Procedure (PAP) 1911, Fire Emergency, Rev. 4 m

s m

o n-a

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te so -4 D8 O m ti 5 = n ra oRv ca 90 O Ch tA 6.6 Category F: Fire or Explosion (Cont.) ,, ::. :..Gm ' :r:7,11?:. e..... # i3%. ~..:,

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-x g. Bea Initiating Conditions Entry Criteria G EU2 U N FU2 Report by plant personnel confirming the occurrence of an explosion within a Safe U m Shutdown Building. S >r Explosion within a Safe U mj' Shutdown Building A L v' 9. T y a E 1 u m m V = o o E to w N = m Applicable Modes: T {Q o" 1 2 3 4 5 D t + - X?:::( s ~ ^ ^ ^ W ~~ DISd. USSIO.. N...3. w/. ~ An explosion is considered a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potendally imparts significant energy to near-by structures and materials. m == Consideration is given to unanticipated explosions which exhibit sufficient force to cause minor damage to plant structures or equipment within a Safe Shutdown Building. Unanticipated is referred here since an anticipated m occurrence would indicate a security event where Security and/or plant personnel are aware of the possibility of an explosion taking place. 6.6.2 FU2 (Cont.) ~: -o ^: mn :an- ~ - -

ng >

,3... e ~ >\\ 1 we w 3 lc5 . '^~ s et f ^ >;^. if o e ._......g...... ....s en y ~ wg w y;g. c ~>

DISCUSSION (Cont.)iLf u :

a , gp g g n gqy ~p, ny g = No attempt is made to assess the magnitude of the damage. The occurrence of the explosion with reports of G damage (deformation / scorching) is sufficient for declaration. Any security aspects of this event should be considered under Event Category N," Security Events." If structural or equipment damage occurs within areas housing safe shutdown equipment and functions, the event may be escalated to Alert, LA1. tn For the purposes of this IC, Safe Shutdown Buildings / areas are considered to be the following locations: rn

s ControlComplex(allelevations) y Auxiliary Building (all elevations) g g2 Intermediate Building (all elevations) g y

Fuel Handling Building (all elevations) g-g a' Reactor Building (allelevations) g, o. ~ Emergency Service Water Pump House (all elevations) g Electrical Duct Chase Leading to ESW Building g I Diesel Generator Building (all areas except the Unit 2 Division 1,2, and 3 DG Rooms) = SteamTunnel(allelevations) p DieselGenerator FuelOilStorage Area f' Condensate Storage Tank Intake / Discharge Structure ~~ w - - ,;,, n.,; mm n m: f4 sm ,qw. g4 @_ '..4E f ' e e 2;22 : Jsn asti. ~ ~ iy!.. # $ 2,a;1. E~ 5-iMENME ~ i ~

.x.;..

~;, x g, y v ,sn g. v

1. NUMARC/NESP-007 (Rev. 2), Unusual Event HUl g

a

2. Updated Safety Analysis Report (USAR), Chapter 9 and Appendix 9A

3. Fire Protection Evaluation Report Sections 3 and 4, Table 3-1, Rev. 4

4. Appendix R - Evaluation, Safe Shutdown Capability Report, Section 4, Rev. 5 s

6.6 Category F: Fire or Explosion (Cont.) , ~flw%' ;^'(,'>, g,,, % ' ~ 'I ; -', C-n.: '} ', ' ^ 'a

~.

n... .a, e a - '+ .s n ~ xe^ -,.^~'.,,~,.:,, - + ' ~3 < 2 ,'s. :' + ', ~:.' s. ' I ' l h 6.6.3fINITIATING CONDITION FAI d's y.4^W} ~R 's.M;:7'? A . ,$;,. ; ';,. '.':'u./'M^x'i ;? 1; p <J' ' ^.t e a;, ~*, dm :' ' i;E ~,' s, v' s + ~ <, ~ ~, * ~, s ~ f' . r: .: a,- - a -s - x,. - s. s t:s. - s c -: - . z ~;s s,. s ~ m. e s c::: . n,.~;z, .x.- r ..- s s< s, y e .s-u <a a, s. - >, y p* o-y, c-Baw Initiating Conditions Entry Criteria G Either of the following has been confirmed: Eal Fire in a Safe Shutdown Building. FA1 Explosion in a Safe Shutdown Building. Fire OR explosion m affecting the operability of plant safety systems m o required to establish or A q !? maintain safe shutdown Plant personnel at the scene report visible Affected safe shutdown system indicates L g damage to safe shutdown equipment or degraded performance. E g E components. R g-T = o. ~ Q. tz w Affected safe shutdown system or component is required to be operable per Technical y Specifications for the present plant operating mode.

o o

Applicable Modes: { 1 2 3 4 5 D M [.,' EM% M < 'l. ' g' ~ ~~ .....m..; ~ W e ps.s.;

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+ i l A fire is as defined in PAP-1911," Fire Emergency." g a a t An explosion is considered a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized = equipment, that potentially imparts significant energy to near-by structures and materials. 6.6.3 FAI (Cont.) y y[9', ;';'yn cQ^- ~. . ? 'S ^:

a' s ;,: '

L,; ; %p

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^= y [ DISCUSSION (Cont.) ' ,ll3?.s:e:,3i:',."1i~, ' ' E ' ',' 'n ' f',' ll/ c E

.79 n '  ;

'i E g. ym-* my~ ~y ,;; 3 '; s y, . m g p: 5:s Only those explosions of sufficient force to damage permanent structures or equipment required for safe G shutdown within Safe Shutdown Buildings should be considered. De same philosophy is being applied to fires that affect safe shutdown areas. Degraded systems performance or visual observation of damage that could degrade system performance is used as the indicator that the safe shutdown system was actually affected. He inclusion of a " report of visible damage" should NOT be interpreted as mandating a lengthy damage assessment prior to classification. NO attemnt is made in this EAL to nue" the actual mamihvie of damam-m bevond the immediate area. De occurrence of the explosion or fire with reports of evidence of damage (e.g., y deformation, scorching) is sufficient for declaration. m 5 For the purposes of this IC, Safe Shutdown Buildings / areas are considered to be the following locations: y d Control Complex (allelevations) g. 3 Auxiliary Building (all elevations) { o,, Intermediate Building (all elevations) Fuel Handling Building (all elevations) 3 Reactor Building (allelevations) Emergency Service Water Pump House (all elevations) y Electrical Duct Chase Leading to ESW Building g Diesel Generator Building (all areas except the Unit 2 Division 1,2, and 3 DG Rooms) SteamTunnel(allelevations) Diesel Generator Fuel Oil Storage Area Condensate Storage Tank Intake / Discharge Structure Safe Shutdown System / Equipment refers to equipment identified in the Safe Shutdown Capability Report. His is the minimum list of equipment required to achieve and maintain COLD SHUTDOWN (m' cluding all auxiliary

P equipment such as AC/DC power, cooling water and instrumentation). A detailed list is provided in the 1?

" Appendix R Evaluation - Safe Shutdown Capability Report." 6.6.3 FA1 (Cont.) wgz g

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DISCUSSION (Coat.); i ', ?) ; o'^ jl,,,'L'^I5l:'j'JL'qg+i,' 2!f E,l:f'1s <;i'fi ill l

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,c.,. : j' :: m g Safe Shutdown System / Equipment list: (Division 1 and 2 only) G Reactor Protection System Control Rod Drive Hydraulics l Automatic Depressurization System /SRV Reactor Core Isolation Cooling Low Pressure Core Spray m Low Pressure Coolant Injection - A/B/C p-Suppression Pool Cooling m Shutdown Cooling j Safety-Related Instrument Air 9. ca Emergency Service Water g g l Emergency Service Water Screen Wash g. g Emergency Service Water Pump House Ventilation = o E 2 ECCS Pump Room Cooling Systems to w Diesel Generator Building Ventilation = =- R l Stand-by DieselGenerator(DG) = DG FuelOil Storage /fransfer p ElectricalPower Distribution p Emergency Closed Cooling Pump Area Cooling Emergency Closed Cooling 1 Control Complex Chilled Water t MCC, Switchgear and Miscellaneous Electrical Equipment Areas HVAC System Battery Room Exhaust Control Room HVAC and Emergency Recirculation System m in OQ O QO i -w- 6.63 FAI (Cont.) ~ q,sw gg,fffijkiijgh hg/ - a h k$h^ffk:Skfg:m .c. ~ - ~, m[f[%SM [Ihf;; ^], f {'t ' ' ^ k.. ' 7 ^ ~ '~ g g M-ji + wus,>o 9 ~~- m e wr, ^ m~- u., m _..%- ~ s ,s e mm - 4 e 5an

1. NUMARC/NESP-007 (Rev. 2), Unusual Event HA2 G

2. Updated Safety Analysis Report (USAR), Section 9 and Appendix 9A

3. Fire Protection Evaluation Report Sections 4, Rev. 4

4. Appendix R - Evaluation, Safe Shutdown Capability Report Section 4, Rev. 5 m>

5. Plant Administrative Procedure (PAP) 1911, Fue Emergency, Rev. 4 r

m

s m

i a

2.

cn m w A O b M m oo = o U. ,v l [

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00 O Q L i i Sheet 64 of 138 Page 72 EAL Entry Criteria and Bases (Cont 3 i THIS PAGE INTENTIONALLY BLANK 6.7 Category G: Increased Plant Radiation Izvels

.,~~,,' ' -

~, J ~ , ~, /,. ^ ~. - .c e r .,^ ~ ^ ;, } ;- ' .~:. '6.7.1 ' INITIATING CONDITION GU11 '^. i', / ' ',

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- ' 'J": ^ 'f, '.' ?. :: ' i f if ~^ s '~ n' :.:' *' g. ' ' "' : f r * :. ~: ...;, f e : ; ' -f2-i :' a .-J,N ' ' ^ e 3 B 8 Initiating Conditions Entry Criteria G Unit 1 or common area (D21) Health Physics surveys indicate an increase by a fdll radiation momtor reading increases by factor of 1000 times normally expected area a factor of 100 greater than ALERT radiationlevels U alarm setpoint. N U m S >r GUI U m A Unexpected increase in L p plant radiation levels g g E g-g Increase in area in-plant radiation levels CANNOT be attributed to either: V a o o E as u the start-up and operation of plant equipment or systems within design parameters. N g o T the planned movement of radioactive materials. the planned movement of shielding (i.e., plugs, lead shot, etc.) 9 Applicable Modes: 1 2 3 4 5 D ^ gjISCUSsI(Ik! ^ M ~ yg g ~ 2? = The ALERT alarm setpoint on the area radiation monitors (D21) is intended to provide notice to the Control Room ofin-plant radiation levels above normally expected conditions, based on the operation of systems or activities conducted in that given area. 6.7.1 GUI (Cont.) t .~ ';:2 c; +, ;'. / :: ', e v + ::

'c w: a y

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DISCUSSION (Cont.)

~' >, d ': ^'. f ~/ ^ g '~ ~ s_

e: :-

~ a,-

s For event classification purposes, the Unusual Event threshold has been conservatively established at 100 times G

the D21 ALERT setpoint to provide a desirable gradient between the Unusual Event (GUI) and Alert (GA2) classes. For Health Physics survey purposes, NORMAL levels can be considered as the highest reading in the past 24-hour period, excluding the current peak value. rn 'Ihis IC is NOT applicable for alarms resulting from the planned movement of radioactive materials or shielding in the plant or expected increases in radiation levels such as the backwashing of the G36, G41 or N23 filters. gs Unplanned increases in in-plant radiation levels represent a degradation in the control of radioactive material and g represent a potential degradation in the level of safety of the plant. These events tend to have long lead times i d relative to potential for radiological release outside the site boundary thus, the impact to public health and safety B-R is very low. 5 2, te w This EAL escalates to an AI FRT per GA2 if the radiation Icvel increase impairs safe operation of the plant. en M,x,7ac em !?S, ' y .c. ",;s, _ gf:::i %'., +s.R 5 ^0 iU;p% FW ?.>.xo :e - gr ^ f 6, $b '%E!"M / e#c ', :^ s,?^J< -C g~7 ped; Agg g 'y ...,. a igs Nyk Mi$$$ }. ! REFERENCES,^' ^w a gi

• 7.

G-s r-t }'jg ).yf!@mk ij dE 65 j9s *M W_. '-l' ; ',,; I',;^ ;' I j@ " ;, j,j 2 +< s s

1. NUMARC/NESP-007 (Rev. 2), Unusual Event AU2

2. Off-NormalInstruction (ONI) D17, High Radiation Levels Within Plant Unit 1 Rev. 5 i

3. Updated Safety Analysis Report (USAR), Chapter 12.3.4.4 m

m ub 4 . ~. 6.7 Catecorv G: Increased Plant Radiation Levels (Cont 3 ^ 'l h 4.7.2ilNITIATING;;CONDITIONU... f,,.z hsj;;; ^ f' ~ s>, , ', y',:p, 5{C.jj)'..'..-v.a........s... -... : g:$..:. e.v. s- --Li2} ', '~"' 4, j ..."v...,...... s-l q ^ ,e g 3 Cg: /' .. 5:. - ">v s3 /cn g: 'c " ~ ' 4 u .,g....m.s.,... ,, 'l / +-

• - ' 1b, s e "

Yiji$? ' E! " { ' ' WM x - xus s o P. Initiating Conditions Entry Criteria w Uncontrolled decrease in one or mere of the following fuel pools: fdJJ Fuel Storage and Preparation Pool + FuelTransfer Pool U N m Spent Fuel Storage Pool U D Shipping Cask Storage Pool t S F GU2 Upper Spent Fuel Deep Pit U FuelTransfer Canal Uncontrolled fuel pool A p waterlevel decrease L s a with irradiated fuel Irradiated fuelis stored in affected pool. E S outside the RPV E E P., o. remaining covered. V m C E N N k T 9av Applicable Modes: 1 2 3 4 5 D ~

DISCUSSIONj, OR

' a -^ ~ WB ~ F? v v 'r.. O v .s dE These events tend to have long lead times relative to potential for radiological release outside the site boundary. Thus the impact to public health and safety is very low. Classification as an Unusual Event is warranted as a precursor to a more serious event. 6.7.2 GU2 (Cont.) ']

1. [

,'s <^.# 2 ,'s '. ~..,m., . ~: ' :., s',

,N' ' '

k , '. r< '[I s [s .( N' " f- , i' ] [s ['# ' ' ^ q .s s x ' ' '^ ' ' "J +M'r' I~4 ' l P' :,'m f if DISCUSSION (Cont.)' l ' " ,,, "' ', ~.,'Y :O ~ ?, % / ' ',' ., - ~" i,' ; n n.~ v, .~,s_,o o ,s<,'s' s y" ' ~ (A '., 'c 'c. ~ +- s's.' s ' ~- 5 's. 's - (c, % -f es s - B s s;g' g= Oa It is NOT intended that an individual be sent to make a visual observation if it can be verified remotely with G cameras or Health Physics surveys, if performed. Unplasmed increases in in-plant radiation levels represent a degradation in the control of radioactive material and represent a potential degradatice in the level of safety of the plant. This EAL escalates to an ALERT per GA1 and GA2 if the radiation level increase impairs safe operation of the plant. m j%y> bI !hIh diy> > &)g]@y$# @@ fd$,$l6dge.,@di '(/ ~.

,, g,,

~,y -.". ~, : ' 3.._. y gqa. ~.,. (A < yp ~ ;. e y4 g ;iq > < q yy... ,ym ~ ey _.m e s..,., _ x J ' ', '.:... ',, ;!.[ 5 '~ m up w~ ;, 6~~ .; v ..,; u m..- ~ vs 2-

1. NUMARC/NESP-007 (Rev. 2), Unusual Event AU2 g

Q 3. m = oo

2. Updated Safety Analysis Report (USAR), Sections 9.1.2.2 and 9.1.3.3.2

= o 8. w oo rsO r.n o

3:"'

a m 00 O Q Ch 6.7 Cateyorv G: Increas-d Plant Radiation Levels (Cont.) , 'n , ~, - l } - '. ' ;: ; : ': ~ ), d' c :'L ',' ' :-:'c.,', L ' ~ 7 ' y 'e - ' e, ' > y, '; y '1: ' ' ~ 'h:, $ f, N:p'u; 'J': ?~ ~' ~ l m= ' ' 'M 4;' N # r S ', ~ q .,' dE f ' t EU$ INITIATING CON,g DITION GA1 u ,;;y g,

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z S E Initiating Conditions Entry Criteria G Dropping, bumping, or otherwise rough Irradiated fuel bundle suspended from 931 handling of an irradiated fuel bundle. grapple with a decrease in pool inventory. GA1 HIGH alarm on one or more of the following radiation monitors: m Major damage to A Spent Fuel Pool area L m irradiated fuel o E 3 Upper Pool area Fuel Preparation Pool area g y R T pr a FHB ventilation (gaseous) E Containment atmosphere (gaseous) e E P., Applicable Modes: tn w a 1 2 3 4 5 D gn 3 _,. E C D.ISCUSSION?. ^ m , 9 m. ^ Qt, Due to the decreased amount of decay heat present, there is time available to take corrective actions and little potential for substantial fuel damage. In addition, (NUREG/CR-4982)," Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82", July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk ofinjury is low. NRC Information Notice No. 90-08,"Kr-85 Hazards from Decayed Fuel," also presents the following discussion: [ o Q" "In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel onsite, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's protection action guides." 6.7.3 GA1 (Cont.) ..,n ~ ~,, :',q m,~,~,:~,,,',,,

x, 2

,

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y, ,~~ y -.,g'm q:w ':A ~ :c' :;[; ! "':#if 'X > 'M: ~ 3:}::a=:;5's'J'i'l', M5 y b ~ :%;;'::9}y J 3 ~' ' '^%' af E J:.'i ^ s' M: DISCUSSION (Comi.) ' ' q ;b::' '; 3, \\'c  ; J : Q  ;. ys;~ ' ' : ' g. ~ 5o This IC applies to spent fuel requiring water coverage and is concerned with exposures to plant personnel caused G by the rough handling / dropping or uncovery of spent fuel. Permanent area and airborne gas channel radiation monitors in Containment and the Fuel Handling Building (FHB) are utilized as indication for increased radiation levels caused by rough handling or dropping. If the rough handling was done in the Fuel Handling Building, only classify if the FHB alarms are received. [ m i i I k yy Q≫, g gg[y i ES ' ^ ~~ n gygggggggp ggg; ~ 3 ;y y ~ r.a 2.

1. NUMARC/NESP-007 (Rev. 2), Alert AA2.

g w = o

2. Off-Normal Instruction (ONI) D17, High Radiation Levels Within Plant - Unit 1, Rev. 5 m

o 8.

3. NUREG 0818, Entry Criteria for Light Water Reactors

? M lt

4. NUREG/CR-4982, Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82, July 1987 g

l o. I o

5. NRC IE Notice No. 90-08, Kr-85 Hazards from Decayed Fuel C

i i t 5 m 00 5 o 4 L 00 .n s s-, .e-a - ~.. 6.7 Category G: Increased Plant Radiation Levels (Cont ) 's i ' ',;< s,: ' ?- ~ :' ' ... m --...,. :m:::,., a g, No. .O 'ep. t' 0' c - d 'w. '. g> m i ',s y ., -: - -. a. a.;w.s.. y: ~67.45.,....-- y.. ITIATIN.y;C,ON.D. I, TION...L.G. A2. V, <, s' s A,s,, m s 0- . e 'ys =- IN - G

~ >

,e

- ~, s ,g, n p44y s, 9 x 4 x s s on Initiating Conditions Entry Criteria y Control Room area Area radiation levels greater than one or more of the PEI-GM radiation levels of greater N11 Maximum Safe Operating Conditions for Area l GA2 than 15 mR/hr. Radiation. Increases in radiation m levels within Safe B A m Shutdown Buildings j L that impede operation E of systems required to n y R E-g maintain safe n T a. w operations OR to = m O. establish or maintain o 0-COLD SHUTDOWN to t.a M2 F3 Applicable Modes: O Ev 1 2 3 4 5 D w w s.e.n

DINCUSSIDN) #

~ m m s ,^ &4 F# des a ~ s ~ 4 2 'T p3-f e The only area requiring continuous occupancy is the Control Room. The value of 15 mR/hr is derived from the General Design Criteria (GDC) 19 value of 5 rem in 30 days with an expected occupancy time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per ?$ day. ve EXAMPLE: 5000 mR x day x 12 hrs. = 13.9 mR (dose rounded to 15 mR for human factoring 30 days 24 hrs. consideration.) i t 6.7.4 GA2 (Cont.) ,~ .c s s ~. -,s i s c. e,s,4 . - s .w, _, ~

:.s s,

s - s n '/ (;',' e,s s s 4'

. (13,' q ' y ' ;

'*s,. je. ss N ss .s f,,.,, s x c L s.,,.j +; p<> <:.;m ,s DISCUSSION (Cont.)J': F ' ':+' ]l ': ' [' f l's < l:N;'( > f 'i" ':" $~,. # u O ~'W d' l'fi[]~}n?? l [

,n o,jy;,7,a y y ;gs 7 s ;

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, W: - y q^< c y' g 3.:n p, s:g:ps y Section III.D.3 of NUREG-0737," Clarification of TMI Action Plan Requirements," provides that the 15 mR/hr G value can be averaged over the 30 days. However, the value is used here without averaging since a 30-day duration implies an event potentially more significant than an Alert. This IC addresses increased radiation levels that impede necessary access to operating stations or other areas containing equipment that must be operated manually in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial m degradation of the level of safety of the plant. Ihe cause and/or magnitude of the increase in radiation levels is NOT a concern of thisIC. m 3$ l Per the PEI Bases Document, the Maximum Safe Operating Condition values listed under PEI-N11, p( Containment Ixakage Control, provide the highest parameter value at which either: (1) Equipment necessary for i safe shutdown of the plant will fail, or (2) personnel access necessary for the safe operation of the plant will be H-d l precluded. Therefore, the intent of IC GA2 is met by exceeding the Maximum Safe Operating Condition values { SL m y l for Area Radiation. oo N This IC is NOT meant to apply to increases in the Containment radiation monitors as these events area addressed i in the fission product barrier ICs, nor is it intended to apply to anticipated temporary increases due to planned 9 events (e.g., incore detector movement, radwaste container movement, depleted resin transfers, etc.). l

ga s

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~ ? : : + '% : k,:'I'::L'6 'd.s J'%rs ' s.,: sc 't s': i REFERENCES, ' ~ < ~u s, ~, ' s~ s s - uc' c ~. s,' ' ', ' 'g, ~f (,; ,s ,s = s s #, "s - - n e,, - s-- o s s e s ,, <, ',,', - ( 'Q ]'g s ', ,s s s s s

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'7, J,.; y, tq,' j j[s 1 s,

gjj " ~; ^ ,,, q i y s> ,9, x

1. NUMARC/NESP-007 (Rev. 2), Alert AA3.

t

2. NUREG-0737, Cl rification of TMI Action Plan Requirements

[ { a 00 o i i i 1 P 6.7.4 GA2 (Cont.) l-

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~y ; < .:~ o dkog : l';;'z:, e -' - -2, '^ <' ^~~ 1.' '. ' ~ l'i ' ~ ~.' ' ' D ' i ', ' ' ' l. REll'ERENCES (C' nt.)' [, '. . ~, - :'N -y ;- u ' c' -,:: :< ~ ~. ~,~ ',':;- : ' - ' ~: ?' ~ ' :, c -. 9, f

3. Off-Normal Instruction (ONI) D17, High Radiation Izvels Within Plant - Unit 1, Rev. 5 9

4. General Design Criteria 19, Control Room t

i

5. Safety Evaluation 92-161, For Onsite Storage of Low Izvel Waste t

6. Updated Safety Analysis Report (USAR), Chapter 9 and Appendix 9A

7. Plant Emergency Instruction (PEI) N11, Containment Irakage Control, Rev. B

[ i D4 l

8. Plant Emergency Instruction (PEI) Bases Document, Rev. 2/ Update 1 m

9 X a a

s.

v m ta p O b t.o C; oo N= m OD h I T so 00 O 00 [ t Sheet 74 of 138 Page 82 EAL Entry Criteria and Bases (Cont.) THIS PAGE INTENTIONALLY BLANK l 6.8 Category H: Increased Radiation Release to the Environment ,,j, ',f

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- en C~ ,, ',, t, : ~,, ,, =,, ' 'c : ~ ~,g.. ~ ,~, :'::, ;:: a E to Initiating Conditions Entry Criteria Reading greater than TWO times the Routine or as required sample analysis jilli HIGH alarm setpoint on one or more of indicates a release rate greater than two the following plant gaseous effluent times ODCM 3.11.2.1 limits. monitor lasting greater than or equal to 60 minutes: n1> t-Unit 1 Vent ID17-K786 U B1 HU1 OGB Vent Pipe 1D17-K836 The release lasts for equal to or greater N TB/HB Vent ID17-K856 than 60 minutes. U p y Any unplanned release S I 2 of gaseous radioactivity Unit 2 Vent 2D17-K786 U 3-d to the environment that exceeds two times the A [ R ODCM Controllimit Chemistry sample analysis methods L to C; for 60 minutes or CANNOTconfirm within 60 minutes of y greater. receipt of the HIGH alarm that effluent E Z levels are less than two times V 9

s ODCM 3.11.2.1 limits.

E C N T Applicable Modes: 1 2 3 4 5 D E% 6.8.1 HU1 (Cont.) 4 n n., ~ ;- ce e~,-^' f + ~ y,8

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~ ; a . ~;,~ ssa n e,, ~,, + ,,Nr' ,,<i ?<<rs ' l,,' ,\\ ~~ l g,,' '~. l: <~,.,4y', g, * ~5 . - ~ ~ 'i E ^:# ' '9 8' b' 'J3' N '*1M > if /{' - f c 9^ ': !, ' ; ;;,'; Qt c :c : 'i, :-h ' ' N ' O^ G'l' d :: p^3;';"' '; ^ ; t ' c, q'j 8 f[ ' r ^ ! DISCUSSION ' "!'s 3 - '^ ^ ' ' ' ' ' ' y: ' _ ', ' + ' ' ^ %{+ s,' , J '" O'y ;, G:' ' ' - : : $a Releases in excess of'IWO times Off-site Dose Calculation Manual (ODCM) Control 3/4.11.2.1 limit, that G continue for 60 minutes or longer, represent an uncontrolled situation and hence a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is NOT the primary concern here; rather, it is the degradation in plant control implied by the fact that the release was NOT isolated within 60 minutes. It is NOT intended that the release be averaged over 60 minutes. Further, the Emergency Coordinator should m NOT wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release p will exceed 'lWO times the ODCM Control 3.11.2.1 limit for greater than 60 minutes. ma$ g! If an ongoing release is detected and the starting time for Dat release is unknown, the Unusual Event should be g E declared as soon as it has been determined that the release has exceeded TWO times ODCM Control 3.11.2.1 g limit assuming, in the absence of data to the contrary, that the release duration has exceeded 60 minutes. g-g = o 3 m Monitor indications and alarms are based on the ODCM methodology which demonstrates compliance with gg 10CFR20 and 10CFR50 Appendix I requirements. Per CHI-0006, the D17 gaseous effluent (noble gas) HIGH B: = alarm setpoints are 70% of the ODCM Control 3.11.2.1 limits for the gaseous release points. A conservative { value of TWO times the HIGH alarm setpoint (150% limit) was therefore used to provide a quick reference to p Operators for classification purposes. E s ,',, ^

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1. NUMARC/NESP-007(Rev.2),UnusualEvent AU1

2. Off-Normal Instruction (ONI) D17, High Radiation Izvels Within Plant - Unit 1, Rev. 5

p

=

3. Offsite Dose Calculation Manual (ODCM), Section B and Appendix C: Control 3/4.11.2.1, Rev. 5

4. Chemistry Instruction (CHI) 0006, Radiation Monitoring Alarm Setpoint Determination, Rev. 0

6.8 Category H: Increased Radiation Release to the Environment (Cont.) - <, q' M',1, : y ' ' : jv \\, ' ;,' C, e,, ;, S, ~ ; ~ l n ' ' ^ M,'.',O :-, ; e c J,e,, n ~- ^

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s Initiating Conditions Entry Criteria C

Reading greater than 1.2E3 cpm above background Routine or as required sample RU2 for one or more of the following liquid process analysisindicates a release rate monitors lasting at equal to or greater than greater than two times ODCM U HU2 60 minutes: Control 3.11.1.1 limits. N U m ESW Loop A Process ID17-K604 S ESW Loop B Process ID17-K605 The release lasts for equal to or U gs i Any unplanned release ofliquid radioactivity greater than 60 minutes. A to the environment that g j? L exceeds two times the g E., ODCM Controllimit Chemistry sample analysis methods CANNOT E B-d for 60 minutes or confirm within 60 minutes of receipt of the HIGH-V E 2, greater. HIGH alarm that liquid release levels are less than E C; two times the ODCM Control 3.11.1.1 limits. N 5n T 1o O 5 Applicable Modes: s 1 2 3 4 5 D ,y i, ', ' dddl m ~ ed ~ DISCU,SSIDNi ~ N 4 m ;y py m m - = 4 c This IC includes any liquid release for which a radioactive discharge permit was NOT prepared or a release that exceeds the conditions on the applicable permit (e.g., minimum dilution flow, maximum discharge flow, alarm u setpoints, etc.). 4 t l 6.8.2 HU2 (Cont.) j r ,e < ' ~ _ DISCUSSION (Cont.), ',,,, - ^ '8- : '; - < ~ ' , f( l:, 'f ' a,, u, ,' ' l ; < ~ if l ~ ~; ':!', W " c t D,: ' ? J J ' G ^)  :' n ' a. 'f ~ t : s ,e,,, _z, ~.

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n. y, >y n, 3- ,~ g = i Releases in excess of TWO times the ODCM Control 3.11.1.1 limits, that continue for 60 minutes or longer, G j represent an uncontrolled situation and hence a potential degradation in the level of safety. He final integrated dose (which is very low in the Unusual Event emergency class) is NOT the primary concem here; rather, it is the degradation in plant control implied by the fact that the release was NOT isolated within 60 minutes. I It is NOT intended that the release be averaged over 60 minutes. Further, the Emergency Coordinator should 7 NOT wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release m will exceed 'lWO times the ODCM Control 3.11.1.1 limit for greater than 60 minutes.' y m i If an ongoing release is detected and the starting time for the release is unknown, the Unusual Event should be j [ g! j declared as soon as it has been determined that the release has exceeded TWO times ODCM Control 3.11.1.1 g I limit, assuming in the absence of data to the contrary, that the release duration has exceeded 60 minutes. g a. -a = = Monitor indications, derived under FCR 021925 and based on the ODCM methodology, demonstrate compliance { o,, with 10CFR20 requirements. He ESW monitor response is based on an average 1995100% power RCS water g isotopic inventory, decayed to 1.5 days (most conservative mix). g = k Per USAR Chapter 11.5.3, monitoring and sampling are limited to the Emergency Service Water (ESW) and h [ Liquid Radwaste (LRW) liquid effluent pathways. For event classification purposes, concern is limited to the R i ESW Loop A and B process monitors which would provide indication ofleakage from Residual Heat Removal l (RHR) Systems via the non-regenerative heat exchanger. Discharges from the liquid radwaste systems to ESW are considered controlled releases, requiring sampling and evaluation prior to discharging; therefore, releases i l from LRW are NOTconsidered. [ t i L:

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~ ~ /'s ^ l REFERENCES;3^ , y y 5 3 ;; g 'z y y y pp y, li e i m

1. NUMARC/NESP-007 (Rev. 2), Unusual Event AU1

2. Offsite Dose Calculation Manual (ODCM), Section 2 and Appendix C: Control 3/4.11.1.1, Rev. 5 i

t 6.8.2 HU2 (Cont.) ^ e ,;,<, ' + - , ~ r 3 c- ^,,, ;-

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4 1-ss....: ~, - - + ,m x 39 y ' REFERENCES (Co,nt.)/,^ ~ ? ' : a 1,,,~ ,, g=:.:n ;,' ^?Nh,;?*% ' ~ '^l . 3.':n o c f;- > ~, L r, ,c ' ' - s s .4: yy, ,r ~ 1-r ~ -x c' ;m.;; a. c

3. FCR 021925," Effluent LRW Monitor Reading Calculations"

4. Updated Safety Analysis Report (USAR), Chapter 11.53 and Table 11.5-3 c

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a m o m O 3 m Q. -W m oo = o 3r* s T to 00 0 00 M i l 6.8 Catecorv H: Increased Radiation Release to the Environment (Cont.) t I > < ^'ss'svf ( s' s f ' s s, ^ s e~s

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~,, 'p' 'N < ~ '"v ,., 7,: <,, H 6.8.3 ; INITIATING CONDITION HA1, i; :;j"' ;'^ 'u W:;6 6 9; d O.' e ' S','c '.:,' @, p@[cQ'^ '. i 's' " M's ', 7

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n <,.< ~- ,,.t, ' : ;..,, ~ m,. ' '., s . ':, ' : ::c, -, : : a n ~ ',,v '., .,,,, u,s.s :~ :~, v,. > ~ ' 3 y,

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~ ~.. on i Initiating Conditions Entry Criteria G Reading greater than 200 times the Routine or as Portable survey HAl HIGH alarm setpoint OR offscale required sample instruments indicate high on one or more of the following analysis indicates a radiationlevels of equal plant gaseous effluent monitor: release rate greater to or greater than l HA1 than 200 times 10 mR/hr at the Site m Unit 1 Vent ID17-K786 ODCM Control Boundary for greater p OG Vent pipe ID19-K836 3.11.2.1 limits. than 15 minutes. A m Any unplanned release j TB/HB Vent 1D17-K856 L of gaseous radioactivity E Unit 2 Vent 2D17-K786 g-g! to the environment that R R s exceeds 200 times T 3- = = o the ODCM Control Chemistry sample analysis methods The releaselasts for =, p. limit for 15 minutes or CANNOTconfirm within 15 minutes equalto or greater Q g, = greater of receipt of the HIGH alarm that than 15 minutes. m 3 effluent levels are less than 200 times l O ODCM Control 3.11.2.1 limits. o 3 r C I Applicable Modes: 1 2 3 4 5 D c, '; <. ~ ~ ~, 1 s;... .mm, wr-wm m ~ ~ ',. J jig $5mp@y;s+g4s$$,.'?Usitllg m .s ... ;.1..j.sfs....jd(. $!>.y5 s 'L 'J "' ' ^ ' ' ', 's' s > '? ~ ' "f 1DISCUSSIO$ fa.ys' Ei ..a W:.g i e ,s s , s ,, ?. 6.w.f,, dg{$f,ijff": 1 n.* f == r f"s s .m .,, x.. uym :qWq Su e - s- .s '+x s < ~, sE.G, ?s.J '~, ' 'r x mn%, m .v s w B s e v. .<.s e c y y m =e a This event escalates from the Unusual Event by escalating the magnitude of the release by a factor of 100. Prorating 500 mrem /yr (ODCM Control 3.11.2.1 limit) for both time (8766 hr/yr) and the 200 multiplier, the l = associated site boundary rate would be 10 mR/hr. The required release duration was reduced to 15 minutes in i recognition of the increased severity. i 1 6.8.3 HA1 (Cont.) i ~, ,,'*3 ,';,'7;l,; 1 +s ~ ,r < ~

,~,l~~',

q' ~,,, '.~ i"' ' ~ ' ? ' "* c:' ~ c ' '. v ~ v ,;, ~ :: %( ,2-v ~ ,~ ^ 2' ' 5.' if 'c :: c, ', E, ~. ~~ r E! ' ' :,~ ~,, .,' s :'x' ' c: :, z' : ' :;- + - - T'a ' ' [N' ' C' P: c' c ' 5 c s ! DISCUSSION (Cont.). 2 sg

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8 9 am It is NOT intended that the release be averaged over 15 minutes. Further, the Emergency Coordhiator should G NOT wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release [ will exceed 200 times the ODCM Control 3.11.2.1 limit for greater than 15 minutes. f If ongoing release is detected and the starting time for that release is unknown, the Alert should be declared as soon as it has been determined that the release has exceeded 200 times ODCM Control 3.11.2.1 limit, assuming l m [ in the absence of data to the contrary, that the release duration has exceeded 15 minutes. t-i Monitor indications and alarms are based on the ODCM methodology which demonstrates compliance with m 10CFR20. Per CHI-0006, the D17 gaseous effluent (noble gas) HIGH alarm setpoints are 70% of the ODCM j Control 3.11.2.1 limits for the gaseous release points. A conservative value of 200 times the High alarm setpoint g! or offscale high on one or more of the low range vent monitors was therefore used to provided a quicit reference g j to Operators for classification purposes. g. m = o = g$ W w ' M, > ;:;' ': ' J ' \\ ' ?*% % W Ed& YL, E: '< ~ '4' ': ' -l 5 i n' f.. ise A / :' A^~~

gg gp 'fgg isW gg

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- 3 g s '< : % m iREFERENCES, + mgx e ,c op. , Ss g' y s r o 3

1. NUMARC/NESP-007 (Rev.2), Alert AA1 r

s L i

2. Offsite Dose Calculation Manual (ODCM), Section 3 Appendix C: Control 3/4.11.1.1, Rev. 5

3. Off-Normal Instruction (ONI) D17, High Radiation Izvels Within Plant - Unit 1, Rev. 5 l

1

4. Chemistry Instruction (CHI) 0006, Radiation Alarm Setpoint Determination /Rev. Of 1:

= C 6.8 Category H: Increased Radiation Relene to the Environnest (Cont.) a swo y + ~ mm ?js- + wea, 'm:c ~o ..+ v; :: ; c u e, +: m;a vw: ; ~: 6.8.4 " INITIATING CON _DITION HA2M,id; _ i $ d' ~e,- M EM p 'pM e" n : y,.' g y >Q: 9% :. e mx:.,;; _3 ~ [.,: l2 <d '.w f.; 'gj e ,,.,,,...:'q-. :.: ;,,,y,,, ; :.. ;, 4' .. f.;, - _ m _ l';.y:(_ t n p g',;) V., ~: q Me '43 D Gb5 g dy O r_,, m..-.., &, n..;, w,, yg, ;gw..p>, e gpa,m; :ggga:. a w ~gg a. ~ p 3, , m,. y y y g g Initiating Conditions Entry Criteria P-9 Reading greater than 1.2E5 cpm above Routine or as required sample analysis M background for one or more of the indicates a release rate greater than following liquid process monitors lasting 200 times ODCM Control 3.11.1.1 limits. atleast 15 minutes: HA2 m

• ESW Loop A Process ID17-K604 The release lasts for equal to or greater p

Any unplanned release

• ESW Loop B Process ID17-K605 than 15 minutes.

m ofliquid radioactivity to g the envirmnment that A ro L q

r exceeds 200 times the Chemistry sample analysis methods g

E ODCM Controllimit CANNOTconfirm within 15 minutes of g. g R for 15 minutes or receipt of the HIGH-HIGH alarm, that k o, T greater liquid release levels are less than 200 times g ODCM Control 3.11.1.1 limits. R. = r ws ,n O P.' Applicable Modes: 1 2 3 4 5 D $' um.~ USSIONl',:' !'f:', > 1, J> ' sJ > 'N 'N L ' ' ' U ',i'U O+ 5' ' l .' ~,,' ", ,',? ' ' ' ' ' ? c: y.y> -'s ;, L 2, n. m, :: ~: :x 3 ~ ~ ~>

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.s ,/'; x - ^1. ' < ec < m e No ' c d, ' ; arn;s, ; o,., am u aw m 'Ihis IC includes any liquid release for which a radioactive discharge permit was NOT prepared or a release that exceeds the conditions on the applicable permit. (e.g., minimum dilution flow, maximum did-isc flow, alarm 8 setpoints, etc.). 0 l 6.8.4 HA2 (Cont.) 4 r ,( s ^ * ' ' ' ^ ' ' ^ ^ '. ,, ',' '- } l ,, ', y'.[ ] ,^'s (,, s-' y ^ ",,,,p 6.' 'n g s ,s 6'O:j J " "e .-n ? if w ".-. q q 3,;, ;;; ; y ( p 3,. ; 3;;;9 ;;,g?'s'd ('; ' C'5j ij :': - - ,u, n, ~c 1 l DISCUSSION (Cont.j: ' ^ :ce, ^{ ' ,c'

;;, - y

P This event escalates from the Unusual Event by increasing the magnitude of the release by a factor of 100.

G Prorating the 500 mrem /yr (ODCM Control 3.11.1.1 limits) for both time (8766 hr/yr) and the 200 multiplier, the i associated cite boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes ' l in recognition of the increased seventy. l It is NOT intended that the release be averaged over 15 minutes. Further, the Emergency Coordinator should [ NOT wait until 15 minutes has elapsed, but should declare the event as soon as it is determined by that the m l release will exceed 200 times the ODCM Control 3.11.1.1 limit for greater than 15 minutes. p m ( If an ongoing rele.ase is detected and the starting time for that release is unknown, the Alert should be declared as j j soon as it has been determined that the release has exceeded 200 times ODCM Control 3.11.1.1 limit, assuming Q ca =r i in the absence of data to the contrary, that the release duration has exceeded 15 minutes. g g i B-g Monitor indications, derived under FCR 021925 and based on the ODCM, demonstrate compliance with E 2 i 10CFR20, and were adjusted upwards by a factor of 200. 'Ihe ESW monitor response is based on an average g =g, l 1995 100% power RCS water isotopic inventory, decayed to 1.5 days (most conservative mix). g 2 Per USAR Chapter 11.5.3, monitoring and sampling are limited to the Emergency Service Water (ESW) and g Liquid Radwaste(LRW) liquid effluent pathways. For event classification purposes, concem is limited to the g l ESW Loop A and B process monitors which would provide indication of leakage from Residual Heat Removal 1 (RHR) Systems via the non-regenerative heat exchanger. Discharges from the liquid radwaste systems to ESW j are considered controlled releases, requiring sampling and evaluation prior to discharging, therefore, releases from LRW are NOT considered. j t i 2 1 S l i i r 6.8.4 HA2 (Cont.) .~,,, 3',- ,; c ' , _- ' -- ;., z,;, ~ ^ ce ;'. u', ' ' ; c < : ', J,: ;. n -, y REFERENCES ^c' ' y : ; y, :fJ;' - ,,n s ' f ' s . J . ;,, ' ,,, ', y : -+ - 4 s ~ :., ' ^ 'jf g, _':^', ::, l' : '; s d s g i: t c ',' ; t' ' ', ',' + d? :' : ' '< ' ' e f >*'c5 !, :+ > '," t

>' '~

r> '

Wldl?;:;3?

s ss n.- w, s (,, - R

1. NUMARC/NESP-007 (Rev. 2), Alert AA1 ea

2. Off-Site Dose Calculation Manual (ODCM), Section 2 and Appendix C: Control 3.11.1.1, Rev. 5

3. FCR 021925, " Effluent LRW Monitor Reading Calculations"

4. Updated Safety Analysis Report (USAR), Chapter 11.53 and Table 11.5-3 m

m

s w

3" w* b l3. oo n 4 n O 3 Q. wW w oo ? = 1 O Ev T m3 00 0 tO 6.8 Category H: Increased Radiation Release to the Envi1onirent (Cont.) ^ ^n'?l; ::'(' y 6.8.5& INITIATING CONDITION HS1 ' d',1 :^ - ' J'/:: a

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,x '; a :,lo.:,, e ; / * ' ' ', 8 M: +#f'2!:^w u ) n': f P'c ' ?, J 'C J, 'f:' 0 :!@M'Y' M,M's: il C _, 4:.J";5': > :a ;f ? gpe~",,-: e, 'p' ? 3, iu'rc'.' :.s,: W. ' W Ac o Initiating Conditions Entry Criteria G Greater than the listed reading for one or Emergencydose Field survey results HS1 more of the following plant gaseous calculations, using indicate that one or effluent monitors: actualmeteorology more of the indicate that one or following have S

• Unit 1 Vent ID19-N300 3.8E-1 pCi/cc more of the been met at the I

m HS1

• OGB Vent ID19-N400 2.2E0 pCi/cc following are met at Site Boundary:

T p

• TB/HB VentID17-K856 1.6E4 cpm the Site Boundary:

E gi

• Greater than y

Site Boundary dose

• Unit 2 Vent 2D19-N300 6.0E-1 pCi/cc l@

• Greater than 100 mR/hr A

resulting from an actual g orimminent release of 100 mR TEDE Whole Body R g E Greater than E B-a

• Greater than gaseous radioactivity that exceeds 100 mR 500 mR CDE 500 mR CDE A

{ 2, TEDE dose OR 500 mR Child Thyroid Child Byroid m g CDE Child Thyroid E S: dose for the actual or M 3 E h projected duration of the release R P-Emergency dose calculations CANNOT Dose rates are G confirm, within 15 minutes of exceeding expected to E limit, that levels at the Site Boundary are continue for equal N less than 100 mR TEDE and 500 mR CDE to or greater than C Child Byroid dose using actual I hour. Y meteorology. Applicable Modes: E oo* 1 2 3 4 5 D 8 6.8.5 HS1 (Cont.) .~,,v~'+ s w' c ,v, l' ^ :'+ ' b 'O:c a:-: = ': 5 ' ' ; w> v: + - %: U= ,'.,, ?n.ih .b'~:">. ~, ? 'k ' M y 0 4" ' % ;i:l;fJ.'J ' M v :'!:: p. ' W' s L J',;' O ';Al ~: 'n if iDISCUSSION c^ l' ^:, '~';' ' ' " '" ' G ' 'c' '; ? i u y p g. 3 g,;,:-;,3s;:5 7 v't:':;, h 6 ; c. ; ; i ' : c, '(;

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,~ N a Committed Dose Equivalent - child thyroid (CDEct) is conservatively used based on agreement with the State of G Ohio. This usage of a child thyroid dose is consistent with the dose assessment methodology described in Section 7.5.10 of the Emergency Plan. i Effluent monitor readings have been established to quantify the magnitude of the release. These threshold readings are based on 500 mR CDEct as the most limiting dose per EPU/CEI-02 calculations, based on EPI-B7b methodology. In establishing these thresholds, the following inputs were used: (1) one hour release duration, (2) m realistic short term (accident) meteorology per USAR Table 23-24; and (3) Reg. Guide 1.109 Child 'Ihyroid h [ dose factors. g1 Effluent readings shall only be used for the classification of fast breaking events if a dose assessment calculation g CANNOT be completed within 15 minutes but then only until actual dose projections can be made. Effluent i 3, f meter setpoints (for E-Plan classification) are based on "best guess" accident scenarios. The vent monitors are 3-R calibrated to measure Xe 133, plus they provide a rough indication of the actual release. Therefore, dose E o assessment, since it uses current plant values, will be more accurate and should be used. C; w oo is Whole Body dose is considered equivalent to Total Effective Dose Equivalent (TEDE) for emergency dose 1 assessment and event classification purposes. 9 E i s Actual meteorology is specifically identified in the IC since it gives the most accurate dose assessment. [ i l 4 m e ^ i i l 6.8.5 HS1 (Cont.) @ggy$!bbdhkk,g,h@h;;,h net 8,7 m n. ~n a ~ ny n, n -g - m.m,~a mv h i k +, "f l;J,hkhY he#.,) h S$ng is kjk: jjjbgys; iR5FERENCENfhk;d[gJ~dg,

1. D-l 3

i;gggg ggg f7 f0 g. f?F . p' , ' L < Eo

1. NUMARC/NESP-007 (Rev. 2), Site Area Emergency ASI

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2. Emergency Plan Implementing Instructions (EPI) B7b, Manual Offsite Dose Calculations, Attachment 2, Rev. 8

3. Perry Nuclear Power Plant Updated Safety Analysis Report for Unit 1. Table 23-24 and Table 11.5-1 J

4. Emergency Plan for Perry Nuclear Power Plant; Docket Nos. 50-440,50-441, Section 7.5.10, Rev.13 m

5. EPU/CEI-02 Calculations (dated 12/08/95), NUMARC EAL Threshold for Initiating Conditions HS1 and HG1 m

Er

6. 10CFR20, Standards for Protection Against Radiation i

w a-

7. Regulatory Guide 1.109, Calculation of Annual Dose to Man from Routine Releases of Reactor Effluents for the Purpose of g

Evaluating Compliance with 10CFR Part 50, Appendix I g. m = o EE.

8. Guides and Pmtective Actions for Nuclear Incidents (October 1991) m G

m a 3 o

3 i

• C m

e t>t 6.8 Category H: Increased Radiation Release to the Envimninent (Cont.) 6.8.6< INITIATING CONDITION Hb1 S M ddd$d'D'i;h'! d,3' Y; r~ c;;:> c,;, g' : ::MW

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y

s Initiatina Conditions Entry Criteria U

Greater than the reading listed for one or Emergencydose Field survey results Rfd more of the following plant gaseous calculations, using indicate that one or effluent monitors: actual meteorology more of the indicate that one or following have been HG1

• Unit 1 Vent ID19-N300 3.8E0 pCi/cc more of the met at the Site m

• OGB Vent ID19-N400 2.2E1 pCi/cc following are met Boundary:

G j Site Boundary dose . TB/HB Vent 1D17-K8561.6E5 cpm at the Site E g1

• Greater than N

4 resulting from an actual . Unit 2 Vent 2D19-N300 6.0E0 pCi/cc Bourxiary: E orimminent release of 1000 mR/hr E g' d Greater than Whole Body R g gaseous radioactivity that exceeds 1000 mR 1000 mR

• Greater than A

3-E TEDE dose OR TEDE 5000 mR CDE L E SL 1

• Greatre than Chikinyroid to C;

5000 mR CDE Child s Thyroid dose or the 5000 mR CDE E y actual or projected ChikiDyroid M i [ duration of the release E Emergency dose calculations CANNOT Dose rates are R C, confirm, within 15 minutes of exceeding expected to continue G above limit,thatlevels at the Site for equal to or E Boundary areless than 1000 mR TEDE greater than I hour. N and 5000 mR CDE Child Byroid dose C using actualmeteorology. Y Applicable Modes: E 1 2 3 4 5 D i m l 6.8.6 HG1 (Cont.) ' ', ~, ' ' ' th % ' ~, ' ','. ~ ;, ,, w '. ': ~: "'s ^' ? - C'- ~~ N ' ' ', ~;' 5 s <,,,' L : i' ' ' 'l, ; :: ' ' " '. y t - ii ?n$)INCUSSl55Y~, L 3 ~~?*~ t z. n u - ra : [' ;',5 g. .; ' y yg" ' ' g g- ,, u. B E Committed Dose Equivalent - child thyroid (CDEct) is conservatively used based on agreement with the State of G Ohio. This usage of a child thyroid dose is consistent with the dose assessment methodology described in Section 7.5.10 of the Emergency Plan. Effluent monitor readings have been established to quantify the magnitude of the release. These threshold readings are based on 5 R CDEct as the most limiting dose per EPU/CEI-02 calculations, based on EPI-B7b methodology. In establishing these thresholds, the following inputs were used: (1) one hour release duration, (2) en realistic short term (accident) meteorology per US AR Table 23-24; and (3) Reg. Guide 1.109 Child Thyroid g3 dose factors. Effluent readings shall only be used for the classification of fast breaking events if a dose assessment calculation p if E cannot be completed within 15 minutes but then only until actual dose projections can be made. Effluent meter E setpoints (for E-Plan classification) are based on "best guess" accident scenarios. 'Ihe vent monitors are calibrated to measure Xe 133, thus they provide a rough indication of the actual release. 'Iherefore, dose { oi m ti assessment, since it uses current plant values, will be more accurate and should be used. = oo 8 Whole Body dose is considered equivalent to TEDE for emergency dose calculations and event classification Z 9 purposes. ilu Actual meteorology is specifically identified in the IC since it gives the most accurate dose assessment. 2i'% 'fl 6.8.6 HG1 (Cont.) a.. ym-- .~ ~ +,, - ~ ,w og, y, ,,'y 7,'^ ' ^ (' < EU 3 .s ..?.....[ s, s c, ,,,. J ' 4 ;; ;f' REFERENCES " ':s. 7 : ': 4.,

g,

.n;; - 9,. . X ' % ;;m"' >-(3:C q J i ;q,,p J c y g 2 m 'n p a, u., ,;, ; x;; m 3 ;g m 9 E

1. NUMARC/NESP-007(Rev.2),GeneralEmergency AG1 G

2. Emergency Plan Implementing Instructions (EPI) B7b, Manual Offsite Dose Calculations, Attachment 2, Rev. 8 r

3. Perry Nuclear Power Plant Updated Safety Analysis Report for Unit 1. Table 2.3-24 and Table 11.5-1

4. Emergency Plan for Perry Nuclear Power Plant; Docket Nos. 50-440,50-441, Section 7.5.10, Rev.13 m>

5. EPU/CEI-02 Calculations (dated 12/08/95), NUMARC EAL Threshold for Initiating Conditions HS1 and HG1 l

m i

s

6. 10CFR20, Standards for Protection Against Radiation j

r.n 4 2. E

7. Regulatory Guide 1.109, Calculation of Annual Dose to Man from Routine Releases of Reactor Effluents for the Purpose of g

Evaluating Compliance with 10CFR Part 50, Appendix I g-8 E P-o-

8. Guides and Protective Actions for Nuclear Incidents (Octobe.r 1991) to w

llD 00 = o r* L i L it e 00 L 6.9 Category I: Control Room Evacuation y xs

s '

s s - % "c s .:s ~ 4%gW^ s s.W v s w % g p( g:gigg pgjityi:.g!' # N,,' s ' ' ' -s . e ~, ; Ef MC ~. 4 T Rg. M s S:.W.x:v.=, ~ ^ s + v s, 2 (E .x.

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,y 3( v 3, ~yw-3 7 swr gy Bo E Initiating Conditions Entry Criteria o IA_1 IAI A Control Room Entry into ONI-C61. L m evacuation has been E initiated t~ R m j T Applicable Modes: v> b 8 m-1 2 3 4 5 D a a. e n> D3 O 3 m -:...x.: v ~ W...., iD,ISC.CS...S. I..O..N. X. ' s

s....

m u ~> s G 00 v-y a= The Alert condition addresses events which involve a substantial degradation of the level of safety of the plant. 93 Frequently, a distinguishing characteristic of a " substantial degradation" is the need for increased monitoring of, or assistance in monitoring, and direction through the Technical Support Center and/or Operations Support Center is necessary. Therefore, an Alert should be declared when the Control Room must be evacuated. An inability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency per ISI. .ug .w; sq.m C.) y .s.y.s

n. gg u.

n~sv2 - - ~ 3. sw. m nyg ... +. a s-e -tw ww us 7 e

1. NUMARC/NESP-007 (Rev. 2), Alert HA5

2. Off-NormalInstruction (ONI) C61, Evacuation of the Control Room, Rev. 2

6.9 Catecorv I: Control Room Evacuation (Cont.) ~, w, [ u, ,\\, u 'v ,v,e b v O ' 4 4'- ef,i:e, c'1.3'J, ;,r'f / /+p <;J'/6 w v' ( f' / ' ', '[ '(,

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y' e ' 1W'?:%~ '; 0 ~- - O 6.9.2. ; INITI, ATING C,ONDI, TION IS1 s.', :~. f,' s  ;,z,, 'N' :e : ;' J e,,."1 '4;i 5 n ,,s u .u +. ; n ,n ,u ?,', < ^ * > ? '/ : ^ " ~'!,,,,,ys s ' ' 'I .,1 , 1 L',.:p x, ' ); [ Y .,,. [. ' ', 'fg /! ' ' ' ', -,.$ 'z j f',' ,;, b 'i 'f5 'ds'^' y Boo Initiating Conditions Entry Criteria ~ to ISl s I T m E >r m j IS1 Entry into ONI-C61. A R ca 9. Centrol Room E g n evar.uation has been Within 15 minutes of entry into ONI-C61, Operator (s) located at the remote shutdown A g. g initiated, AND plant controls CANNOT maintain RPV water level greater than 0". E 2, control CANNOT be E m w established within M = = 15 minutes. E R o O G a" E N C Y Applicable Modes: 1 2 3 4 5 D m ra CGo w OO 6.9.2 ISI (Cont.) r~- ~ .. n - : -~. hr ~+<_ h ~, a" x> t; *?, m v

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' 'E .':: w ' J ' :^ ' 1;. ~,c, -c':; 4 ^ n w - - o $= This IC and its associated EAL address a condition where evacuation of the Control Room is necessary but G expeditious transfer of safety systems has NOT occurred. Fission product barrier damage may not yet be indicated. The intent ofISI is to ensure that prompt Operator action is taken upon evacuating the Control Room to ensure RPV water level is maintained above TAF, thus preventing possible clad damage. Ioss of RPV inventory below 0"is classified as Site Area Emergency under ASI. A maximum 15 minute time frame for the transfer of control m of" required" systems was established by NUMARC/NESP-007. p m a Escalation of this event, if appropriate, would be by Fission Product Barrier Degradation, Abnormal R adiation 3 levels / Radiological Effluent, or Emergency Coordinator Judgment Initiating Conditions. g ri

s.

e ',w ~,,., ~,, ..,~,, - c n t,a

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w, -l ~ ? REFERENCES, ~ n '? . ~.,,< 1 ' , ', ? '.,, .,, m:: 'r ~ - {'~' n~ - - . P, '^ - y ~_ 8. +++,

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.s ' ~ s s,, pg n ao it

1. NUMARC/NESP-007 (Rev. 2), Site Area Emergency HS2

o O

2. Off-NormalInstruction (ONI) C61, Evacuation of the Control Room, Rev. 2

{

3. Integrated Operating Instruction (IOI) 11, Shutdown from Outside Control Room, Rev. 5

4. Perry IPE MAAP Output Summary Report of Station Blackout with No Injection (MAAP 10_00_70) l t

i as t O.* s -. -....... -. _ -. _ - -. =... 1 1 Sheet 94 of 138 page 102 EAL Entry Criteria and Bases (Cont.) THIS PAGEINTENTIONALLY BLANK 4 6.10 Category J: Loss of Annunciators or Indication s y *, ', g - s 'ys u r ~ q:. s .a rc %';

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~ c': s:'c' p ' ' ' x.;3j s ' y y ('s 6.10.L' INITIATING CONDITION'JUlu ~;~i ~' ':>': ? 6 '; ' '^ J, x* q s',, :'O: # "-,f u p G y s y-q t ^t ~';,' ^;G: y y ' '.,' ' ^ ' ' ~ ', ':,.,',,,

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o ?, Initiating Conditions Entry Criteria o Unplanned loss of most Control Room Unplanned loss of most Control Room ,LQL annunciators for greater than 15 minutes. indication for greater than 15 minutes. U N JU1 U m S y Loss of most U m annunciators or a A 3 indication in the Control Room for g g! L greater than 15 minutes i E In the Shift Supervisor's opinion, increased surveillance is warranted to safely operate E E g V E o the plant. E to u N Sio T Applicable Modes: o o E 1 2 3 s ^ ^ 6Mik _ 4 3"? ,1- ' - o T, " ' U _' s, ^- 4,g ~- s s q-y . DISCUSSION! s ~ ~ '. An' Di P: m#~s ', ~ e - - L' ~ s< ~ This IC and its associated EAL recognize the difficulty associated with monitoring plant conditions without the t use of a major portion of the annunciation orindication equipment. m u li QuantiScation of "most" is left to the Shift Supervisor. It is NOT intended that plant persotael perform a w detailed munt of the instrumentation lost, but rather make ajudgment call with approximately 75% being the threshold. It is estimated that if approximately 75% of the annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. 1 6.10.1 JUI (Cont.) 3 m "qs8 @ 8 N Eh .1 ? R. N ,1 ' l Mi > m Nni; W) i ^1 ', U f ^, ' N

• G ii

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DISCUSSION (Cont;)s m 6a n

N gh M ?' J ,gg ;; g. n y gw 9 7 g, m - 34 + - - B E Control Room panels with annunciators and indications include: G All Unit 1 Control Room Panels in the Operations Areas, = Unit 1 D17, D19, D21 Readout Modules Unit 2 Plant Vent on panel 2H13-P804 (ID17), and 2H13-P884 (2D19) Unit 2 Start-Up Transform r on panel 2H13-P870, and Unit 2 Safety-related batteries on panel 2H13-P877. tt t-- Indications are available at other locations including Control Room back panels; however, using them to safely 21 operate the plant would require increased surveillance, j en 2 II Plant design provides redundant safety system indication powered from separate uninterruptable power supplies. g 2 While failure of a large portion of annunciators is more likely than a failure of a large portion ofindications, the S-E concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of [ 2, specific, or several, safety system indicators should remain a function of that specific system or component m t; operability status. This will be addressed by the specific Technical Specification. 'Ihe initiation of a Technical y Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the i shutdown is NOT in compliance with the Technical Specification action statement, the Unusual Event is based on 9 CUI," Inability to Reach Required Shutdown Within Technical Specification Limits." d The Control Room readouts frotn radiation monitoring systems are included to ensum ^* mtential releases or degraded core conditions can be monitored. Roth the meter and chart recorder (if app. prix ') would be unavailable if the readout modules are out of service such that the process CANNOT be mmiered. Compensatory indications include the Process Computer System and Emergency Response Information System (ERIS). It may include other permanently or temporarily installed monitoring systems if they allow the .e plant Operators to compensate for the failed indications. j E u 6.10.1 JUI (Cont.) w e e, ~ 2 ~ ~ ~p-wm - m ~ mm ~, ; ye. ec DISCUSSIONiC$ntl)li E' ^ Ndy . ;MIfk 3h./ d k 'N g:bcA h N ' ^ (. 3~ B a g ~ oa The D19 Accident Radiation Monitoring System may be started to monitor the 4 plant release points. If G functional, this would qualify as a compensatory indicator for the respective D17 Radiation Monitor (s). Similarly, if no transient is in progress that would auto-initiate the D19 monitor, the D17 system may fulfill the function of a compensatory indicator for the respective D19 monitors. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. m No IC is indicated during COLD SHUTDOWN and REFUELING due to the limited number of safety systems p m required for operation. a$ This event will be escalated to Alert JAl if a transient is in progress during the loss of the annunciation / p indication or the compensatory indications become unav:tilable. g m. e = w (! g> &*i? p O 5 , ji$i??Bic: p Mini.5 in$ B4 a s.

• < x 3

m ..s .....m pp,; .33.,. 4.j;:(, > pgyg, . 44p.> ' ME sp . fa. P _, .s ,s ..c.. m ^" iREFERENCES5 m g. i g w -. ~ g oo en ~, ~ +4 + s .a. s x.# rm

1. NUMARC/NESP-007 (Rev. 2), Unusual Event SU3 h

a r-

2. Off-Normal Instruction (ONI) R61, Imss of Control Room Annunciators, Rev. O i

m CQ O e-o O 1A 6.10 Dtegorv E Loss of Annunciators orIndication (Cont.) s ,3 w w ,'n s l ': s e' s s gja gpsa,. s - w y. -1 's s .,,3 g g~ e i isp ~ [ '^as ' ' ' " ~ ' [; j,sQtc,'J"[,'.'j'il'y$hhgjh.j;;.x:dgy s $103%ilNITIATING CONDITION JAl-,~~, E E . y ':') ? " "s ( 'c : y 1-gg.- h h 9 a s s B 8 Initiating Conditions Entry Criteria G JAL JA1 Unplanned loss of most Control Room Unplannedloss of most Control Room Loss of most annunciators for greater than 15 minutes. indication for greater than 15 minutes. annunciators or m indication in the p Control Room with A m = either: (1)a L 4 e significant transient in E 7 !? progress, OR In the Shift Supervisor's opinion, increased surveillance is warranted to safely operate R g E (2) compensatory the plant. T B-g indicators areN_QI g p. g, available. A significant plant transient is in progress. Compensatory indications i.e., ERIS and y process computer, are NOT available. Applicable Modes: 9 a u 1 2 3 s ~m m: m:- zen - s' s ,,.ixp ' - @SiiI]"$n - O ;.p m'" $ (( '( s 's di:.ldjy!

,

;;g;.

" ejd ] 9' m 2 &;.;x _y -J - ' ' y -.... DIS,CUSSION;S < ^ R::!g:. g'g gg, 3g 39 quy ~ g This IC and its associated EAL recognize the difficulty associated with monitoring plant conditions without the y use of a major portion of the annunciation or indication equipment. It represents an increase in severity above g that described in Unusual Event JUI in that either compensatory indications are NOT available OR a significant transientis in progress. i s 6.10.2 JA1 (Cont.) , e \\w->.,w~.- ~ .m, . ~ ~, '> lt ' \\ >*, v .', ;<>, Q c + 3 ~.

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if DISCUSSION (Cont.)?'J' a n ^

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c'l "l[ i,'a 4 2'c :y', & c -f Tfi: L'.Q 4; ' n ':: o nF&

':;; G ::a F537ri ;y ;'v m 9

B. 5 a Quantification of "most" is left to the Shift Supervisor. It is NOT intended that p'. ant personnel perform a G detailed count of the instrumentation lost, but rather make a judgment call with approximately 75% being the threshold. It is estimated that if approximately 75% of the annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. Control Room panels with annuncir ors and indications include: en All Unit 1 Control Room Panels in the Operations Areas, Unit 1 D17,D19,D21 Readout Modules g1 Unit 2 Plant Vent on panel 2H13-P804 (ID17), and 2H13-P884 (?D19) Unit 2 Start-Up Transformer on panel 2H13-P870, and g y s 2 Unit 2 Safety-related batteries on panel 2H13-P877. 3-2 Indications are available at other locations including Control Room back panels; however, using them to safely [ 2 to C; operate the plant wouki require increased surveillance. R a Plant design provides redundant safety system indication powered from separate uninterruptable power supplies. 1 While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the 93 concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that 5gcific system or component operability status. 'Ihis will be addressed by the specific Technical Specification. 'Ihe initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action statement, the Unusual Event is based on CU1' Inability to Reach Required Shutdown Within Technical Specification Limits." A The Control Room readouts from radiation monitoring systems are included to ensure that potential releases or m degraded core conditions can be monitored. Both the meter and chart recorder (if appropriate) would be {. unavailable if the readout modules are out of service such that the process CANNOT be monitored. g a Compensatory indications include the Process Computer System and Emergency Response Information System (ERIS). It may include other permanently or icmpumily installed monitoring systems if they allow the plant Operators to compensate for the failed indications. 6.10.2 JAl (Cont.) ~ . p u.,:n ~ ~ ~ .: ' _ ; ; g >

' 3, ; -

z ,.; +, .g.. ', - ,.~, ', ^ ,~>,y, a; ~,,., y, g l;'; 4'., };y[ Q '+ ) L';[' Q*~.' f [ ~ 7 l :.';':' ;f ;:;^,q:3,;s;:g 3;3$,l-;' c. ,e: ~ :v - - 2: s , DISCUSSION (Cont.)' . t = ',+ ,,x<- , :m av,;,-: .:,m... 5a The D19 Accident Radiation Monitoring System may be started to monitor the 4 plant release points. If G functional, this would qualify as a compensatory indicator for the respective D17 Radiation Monitor (s). Similarly, if no transient is in progress that would auto-initiate the D19 monitor, the D17 system may fulfill the function of a compensatory indicator for the respective D19 monitors. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. tu A "significant transient" includes response to automatic OR manually initiated functions such as scrams, runbacks p involving greater than 25% thermal power change, ECCS injection, or thermal power oscillations of 10% or m

se greater.

W .m a w No IC is indicated during COLD SHUTDOWN and REFUELING due to the limited number of safety systems S-E required for operation. g. g E o a. ~> a., ',qf,*' ~ , ~. - .,,,'~ ','s~ 5 s i* '6 ',( <,, ' ' ~ ' ' [] J <.w g ss, ^ pih'i.{ ,.,, ~ s ~. ~ - g M

v.,

'm J.:,':f 2' ','a s ' s-a iREFERENCES t:- 4 s' ' :J : ' ' Ja'~ , ~ - ~ ~ ~ a '^ ~ ,q m.s~ ~,, ' L m - ~, + s m~+ %o o

1. NUMARC/NESP-007 (Rev. 2), Alert SA4

{

2. Off-Normal Instruction (ONI) R61, Loss of Control Room Annunciators, Rev. O i

l l in Ooo k a 6.10 Catecory J: Loss of Annunciators orIndication (Cont.) .n , +, -

~

., ~ ., <.-~ ~ + y

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i 510.3$1NITIATING'CONDIT, ION JS1 'T 1 ,,' ::'j', j', '.;# 3 "!>:n 'n, ~, o + g ,g,, gy.. ny;g ;8

y,

~..,.,. ~ on Initiating Conditions Entry Criteria G Loss of most Control Room Loss of most Control Room indication. JS1 annunciators. S m Compensatory indicators,i.e., ERIS and process computer, are NOT available. I y T m JS1 5" E v1 S 5 A g PL Inability to monitor a significant transient in A significant transient is in progress. R 8-5 ~ E E pmgress a. m A e w a E M 9 E E s R G E N Sufficient indication is NOT available to directly monitor plant critical safety parameters C for PEls entered due to the transient. Y Applicable Modes: y 1 2 3 o C 6.10.3 JS1 (Cont.) 9s % Ye y s ^/< ,,4 e=, w,& ,, [ /\\ %, 7, 7, ' ' ,ss s}Qg C q,,,,, g $ v ws N h /* ~ ' + ' >, s ~ ' 'y% V.,, v / ^ '-^ ?~' f ~L \\y,, ~ ' c, * ' ' l,,~ ll, ll ^ t' ' El,: / J 4' Me ?, S'0?., ~ N [, s.. lll

DISCUSSIO,N;

~l'> '^ :- ' ' M i, n <2 4, ' U ' i 'M,,,2-c c,, E C' ^- ' ;t ^ w ,/ ,:,u; , 5 s, vm - ,y n l'> l ~,, ;, e ,, - 2 ,,,~ 2'i..i z ':', ', ', ~ u ::: c,', ; > ..-? '-%% y'[:: '^> ; ' ? y ^ ' ' ' ^; ,1 '. ' ~ - ~ ~,,- 1 o.,- isa. This IC and its associated EAL recognize the inability of the Contml Room staff to monitor plant response to a G transient. A Site Area Emergency is considered to exist if the Control Room staff CANNOT monitor the critical safety functions needed for protection of the public as indicated by an inability to monitor the PEI entry condition indications. This IC and its associated EAL recognize the difficulty associated with monitoring plant conditions without the use of a major portion of the annunciation orindication equipment. m> t~ Quantification of"most"is left to the Shift Supervisor. It is NOT intended that plant personnel perform a m detailed count of the instrumentation lost, but rather make a judgment call with approximately 75% being the j threshold. It is estimated that if approximately 75% of the annunciators or indicators are lost, there is an g y 11 increased risit that a degraded plant condition could go uns.cated. g

2.

g t Control Room panels with annunciators and indications include: ho a.W All Unit 1 Control Room Panels in the Operations Areas, 3 M" Unit 1 D17, D19, D21 Readout Modules 1 l Unit 2 Plant Vent on panel 2H13-P804 (ID17), and 2H13-P884 (2D19) 9 Unit 2 Start-Up Transformer on panel 2H13-P870, and { Unit 2 Safety-related batteries on panel 2H13-P877. Indications are available at other locations including Control Room bacit panels. However, using them to safely operate the plant would require increased surveillance. Critical safety functions are those plant parameters and functions that allow the plant operators to verify they have a coolable core geometry, that core cooling is maintained, and fut Containment is intact. The Perry Plant .e USAR, Chapter 15A.2.2.b states that the safety functions include: [ O o

1. The accommodation of abnormal transients and postulated design basis accidents,

2. The maintenance of Containment integrity,

' 6.103 JS1(Cont.) ~. - .,.;,.y ~ _,:^ m 7 - m: : L : s ~.

T / :' '~, ': ^ ':

> ' ^: >: ~ ' ^ ^ ~ r an..- ~, n ' <

DISCUSSION (Cont.) 1,

'i-)- ',L'O 7^ ' t,,';::,:Jo,. '.'~. s m, 4 ' ^ @' ' '; D j', ^, ~ p "', 5 l w3 J.- 5:

- m ~ :: ~ ~ >.

'..x ; s. w w B -z a E

3. 'Ihe assurance of Emergency Core Cooling, and u

1

4. The continuance of Reactor Coolant Pressure Boundary integrity.

If a significant transient is in progress, entry into one or more PEIs would be required for RPV or Containment control. 'Ihese PEls specify the parameters that must be monitored and controlled. m Compensatory indications include the Process Computer System and Emergency Response Information System (ERIS). It may include other gurgwently or temporarily installed monitoring systems if they allow the plant Operators to compensate for the failed indications. en v w q E The D19 Accident Radiation Monitoring System may be started to monitor the 4 plant release points. If g. functional, this would qualify as a compensatory indicator for the respective D17 radiation monitor (s). Similarly, g. g if no transient is in progress that would auto-initiate the D19 monitor, the D17 system may fulfill the function of g y f c-a compensatory non-alarming indicator for the respective D19 monitors. tz E M 0 Control Rods being fully inserted is compensatory indication for Reactor Power. 8 A "significant transient" includes response to automatic OR manually initiated functions such as scrams, runbacks S., involving greater than 25% thermal power change, ECCS injection, or thermal power oscillations of 10% or l greater. ? 1 Y = I ~ 6.10.3 JS1 (Cont.) q;.:.,, q;: ,7, ~ y .,; y 3 (; y a, -' -,. : :'~ -', ~~ REFERENCES;' ' ' ' d? q ' '; ', : ' :' 8y :p; l ; J'}i; l : ;, [ ^ y ~;, l' ;,'_ [:,, - (;;,-f'j: '4';{:l%',y,';; h N' w w ,,, ~. ~ . ~;;,,>,,,, ; ; ;y, y,~ y.

,~ xy

o r ~ ~,,,, ,, z ' 5 p. ~

1. NUMARC/NESP-007 (Rev. 2), Site Area Emergency SS6

{

2. Updated Safety Analysis Report (USAR), Chapter 15A

3. Off-Normal Instruction (ONI) R61, Loss of Control Room Annunciators, Rev. 0

4. Plant Emergency Instruction (PEI) B13, RPV Contml (ATWS), Rev. A m

5. PlantEmergencyInstruction(PEI)B13,RPV Control (Non-ATWS),Rev. A m

6. Plant Emergency Instruction (PEI) T23, Containment Control, Rev. A

[E en .n W 3. -O = t W I 00 a us i. o3 i u i b Ym OQ O w W N i L 6.11 Category K: less of Communications ,a q,, c.

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^ g s ,'s ,c, ~ > 'a ~ ' ' -: 3 -s s

,.' a ' - -

,,m, - -;, ~,, r, ;' .,s , ~.v ~,, ,,7, s ^ ', 4c.c; ' ' 'J s er ' .)> 6.11.1' INITIATING CONDITION KU1 ~ ' = ^ ~ ,Uc.J'm^ :D a; x 1 J'c:' ~ A::':'? Z:'G:'c : 'n:b:: ~^f M'^~','; tS%': [ s s e s~r v 9,v - ; s;: p 'v,; g, ; y,, >;v ,s ; 'q>, e -. '

- n

y7,: : 3.s 'ga, yg, s n x ',,u:y a,~.:,: ;s s g y-vs

- ~ c y <<y ; s .,s . sx s .sx, s x.,v.s ,.p a Initiating Conditions Entry Criteria Loss of all five Plant Public Address System channels. KU1 U Loss of allof the following Plant Radio System channels: N KU1 U m Channel 1 S h o

• Channel 2 U

m Loss of onsite OR in-o A j plant communications

• Channel 3 capabilities L

g { g ~ s E

v o

e M V E o a. E m N N n T Applicable Modes: 9 sr-J 1 2 3 4 5 D ~ n.~ .y:y * :g.g; 4-e,, dgq (x s 2 s jj;ig v' 4 ^ wgp~i?, g.7., .y

DISCUSSIONi M;. ~.u "i:n i :

m s.b o ~ ' ~ ~.'< m x .m.% .x..., h!'s:ij: ,pg gamg. +. y jyaj;j j,yn:t ~%y:F h;WQy

.y.

s.yf* sm< i:::!g::) < t 5M:W g This IC and its associated EALs recognize a loss of onsite communications capability that defeats the plant staff's ability to perform y routine tasks necessary for plant operations. Use of Plant Radio System channels 4 and 5 is restricted to Site Protection activities. Therefore, credit for Channels 4 and 5 is NOT taken in support of plant operations. 3 6.11.1 KU1 (Cont.) s ;, z < 's REFERENCESS, ^ ^ ~ ^ .J l <,,. y-x - ~,: ~ - ~ s ~ s< ~, - .,x '? ' O'c^''? 0!':1 ( ' ~ - 2 ~ e

~

~ ~ ' ::' ' L ; ^/. : ~ :: 1 ,-.s a m. L+ n ~ c- $D

1. NUMARC/NESP-007 (Rev. 2), Unusual Event SU6 G

2. Plant Administrative Procedure (PAP) 0202, Communications, Rev. 2 m>

C# m n C#2 w 2. 3 Fr ~ l3. b to -W m sn 00 ft us o3v t Tm 90 O w W M 6.11 Category K: Loss of Communications (Cont.) ,.,.~~.,,M'~ w ~ - s< ~ 2:;

y 'n

-:,\\,, <,, .- ~ ~

;3;L. n 3 c ^

t, - ." ^W' < f,s: f m'D cT s"' A+ W ' s e .: :M t- ?' ' : ^ 6.11.2 ~ INITIATING CONDITION KU2 D'. , 4; ',-J ' 2. ' '.'., <: av t-c n <L. ', -..o,. c' p* i. -, a ~ s,, - .v ., a ' <+, >,.7,

' '.s c';^ c' s r,,

, :,s '- ,5 ' +, 9oo Initiating Conditions Entry Criteria G Loss of the State and County Notification Circuit (5-way) reported to the Control ]GJ2 Room. U Loss of affsite long distance calling capability on three or more of the following systems N KU2 circuits for greater than 15 minutes: U m S t-ContralRoom private (259-) lines U g1 Significant degradation Private Branch Exchange, Service Building C'5000") Switch A 3 ofoffsite L p h Private Branch Exchange, Warehouse Building C'6000") Switch communications Company Off-Premise Exchange R' 2 capabilities E B-E; V g j E W G N 5 = T 3 Applicable Modes: 9 Ev 1 2 3 4 5 D ~M- ~ h6eM

M M as M% %

,~ ^ ~ ..-..?.. I OF ^- SR ~ g@~ yd. iD. SCUSSIONj!b~ ' 1 - A*566 36 / ^2 M.s,. 'M ~ m x This IC and its associated EALs recognize a loss of offsite communications capability that significantly degrades m the plant operations staff's ability to communicate with offsite authorities. The loss of offsite communications capability is more comprehensive than that addressed by 10CFR50.72. = u An offsite system circuit refers to one of the four offsite "5-Way" contacts: the State of Ohio, and the counties of Ashtabula, Geauga, and Lake. Testing to determine "5-Way" operability or to initiate circuit restoration actions are governed under PSI-0007. 6.11.2 KU2(Cont.) ~ .a,n, l l' ".::;} 'll::', f ~ u :, l ;' -' j' -:, z,. ; ' iDISCUSSION(Cont.) 'L .~c , ~ - : ' ' ; 't '[ - J a ' a :e w:

r

= ~ n a > '~ ~ . ~ - < 9a 3 ~ All direct (259-) off-site calling capability from the Control Room via private lines refers to: Autodialer at the US console. Private (259-) line on the superphones. Private (259-)!ine at the SAS console. Refer to PSI-0007 for communications failure scenarios and a listing of circuit power supplies. m>r ~- ' 2 2

y;

^ s s s m

g. g>,.,....

3 a s s n ~ Q .4 "p"^"",. ~

/',:; : -

iREFERENCES), m. nn . ~.: ~ a 'j.L: ; :;% ' z, ' c' ' ~ : :: y 2. g-c sm o ~ a a

1. NUMARC/NESP-007 (Rev. 2), Unusual Event SU6 i

g 8 o

2. Preparedness Support Instruction (PSI) 0007, Reporting Emergency Plan-Related Communications

$2 Equipment Problems, Rev. 2 5 M a r.n

3. Plant Administrative Procedure (PAP) 0202, Communications, Rev. 2 3

O

s s

T su 00 O e-o >=* 6.12 Category L: Natural or Destructive Phenomena ;- '5: ' 'n

,, ; 2:;-:L:: ;,; i:'s j', '; ;

,;; t :,,

( ' : - ' x;, , y, ' ::i; ;; ~' y 6.12.1: INITIATI G CONDITION LUl ? 5;;m : 4 ' :; Jc '[' 'i'Q('f/a:' q b^[ ) M ', l'si"'s 'l I'i$ 'Nssi' [ ^ ' ~ . n ':' ;,, ;5:;:a;;;y;., w- ~ ',,, i :, - ' -->;.,~, ; -y., v ' :. u

J:

Boa Initiating Conditions Entrv Criteria G Comrol Room receives Report by High Indications k

Control LEL report from plant plant sustained Room of a Mwn Turbine personnelwho felt an personnel winds greater Trip or failure earthquake.

confirming than 70 mph either of the for equal to m following or greater U within the than N m = LU1 Protected 15 minutes. U N en Area S o g-Natural OR destructive boundary: U $2 phenomena affecting WHITE AMBER Turbine Catastrophic A B-the Protected Area event light (s) on

• tornado casing damage to L

E o boundary indicator Seismic strike penetration. generator $2 P ane or seals. E M light on Monitorin l e a local g Panel train crash V 1 Seismic OH13-E 9 Monitoring P969. N s Panel T OH51-P021. Applicable Modes: 1 2 3 4 5 D m = = w F 6.12.1 LU1 (Cont.) ys ', ;;' 3 L L ; ' ; - 4 '; ' L ;'u; y ::: ; ' 'r# ') R'd%:v '3 "'3 yy:q ~ j'y,';y ~ n;3 2;w;;';; s"ve/Q;?,'Cd'; k

DISCUSSION,

, ' ff ! 1 e'!< 7 %:,: s s > ' 3 -3 ' ?' " 3'2' i fe" Q 1 - f :; h ' j t

'2 - : g'

~ ~ g: % E The method of detection associated with an earthquake of this intensity is based on the condition for a " felt u earthquake" as defined in the EPRI-sponsored " Guidelines for Nuclear Power Plant Response to an Earthquake. [ "Ihese methods include the activation of seismic monitoring instrumentation along with confirmation from plant I personnel who have physically felt the ground motion and recognize the event as an earthquake. ONI-D51 provides indication of a measurable earthquake. i The WHITE event indicator light is received by 0.005 g which is the lowest detectable earthquake for which rn Perry monitors. 'Ihe amber light is received at 2/3 Operational Basis Earthquake (OBE) or 0.05 g and is listed as a backup to the event indicator to ensure declaration. Section 3.7 " Seismic Design" of the USAR describes the t Methodology for measuring the OBE carthquake. 4 i ,::r 2 An earthquake of this magnitude may be sufficient to cause some minor damage to plant structures or equipment g within the Protected Area. Damage is considered to be minor since it does NOT affect physical OR susimal i i integrity. The ewnt is NOT expected to affect the capabilities of plant safety functions. Due to the [ 2, l unpredictable nature of eartha=kes, this may be a precursor to a more serious event and, therefore, represents a to C; potential degradation in the level of safety of the plant. 5 l = A tornado touching down within the Protected Area is an observed event with the potential to cause damage to l structures containing systems or functions necessary for the safe shutdown of the plant. As such, the occurrence C, ( of a tornado strike represents a potential degradation in the level of safety of the plant. If sh im.1 damage is ( confirmed, this event would be escalated to Alert LA1. l If it is determined that the occurrence of the tornado strike has either affected or caused the loss of shutdown cooling functions, then the consequences of the event are assessed under event category B," Loss of Shutdown /Cooldown Functions" or Event Category A," Fission Product Barrier Degradation". The event may then be escalated via these categories if appropriate, y 9 IC LU1 is also intended to address reported crashes such as plane or helicopter crashes OR crashes by trains = which may occur within the Protected Area. As such, the crash represents a potential degradation of the level of safety of the plant. Damage to plant structures and equipment is considered to be minor, with no impact on their physical or stim.imalintegrity. ~._._. 6.12.1 LUl (Cont.) .f af,.,_~ }, '^ ~~, s,. n. ~ g

• /s' : ' ' '

a, is ,~ )' k bu, 'g A s 4,~ ,^ <' DISCUSSIO,N,(Cont.) ~ 't E :f? ' ' ' ' I'OO is 'f ', ' V '. ' e x,:j d(i~/ S 4,,, Q, c b -l if b,\\v,, - g y,~ s< .,..,,i %.g,,, ~- ;, ...,;. ;.r; y, n,'. > ?, nm, e N

s Personal vehicle crashes are NOT istuded since they do not have the potential to impact safe shutdown G

equipment with sufficient force. Two vehicles involved in an accident in the Protected Area does NOT require classification. r If the crash is confirmed to affect a safe shutdown area, the event may be escalated to Alert, LAl. Consideration should also be given to any potential security aspect of the crash under Event Category N," Security Events" for i impact with the security boundary or if an individual was attempting to do damage. m> i r t High sustained winds in excess of 70 mph is a natural and potentially destructive phenomena that may m accompany certain events such as a tornado or hurricane. These sustained high winds may also be produced by j unstable weather conditions. However this event occurs,it may be a precursor to a more serious event and, Q o 2 therefore, represents a potential degradation in the level of safety of the plant. g a' C e m Turbine 'ailure of sufficient magnitude to cause observable damage to the turbine casing or seals increases the g o potential for leakage of combustible fluids and gases (hydrogen cooling) to the Turbine Building. 'Ihe damage Q[ [ must be readily observable and should NOT require equipment disassembly to locate. g M a vs ^ ' < > '^ ~/ I', ' l, ~ ~J ~ ', ' ", ',^ ' O . s. ~,G' 'm, ~ ~ ^'[ ,.,/ ',, > ~,,,, ;,',_ c ~ ; : e ~~ ' ' o j ' ,*^ n .., - :,.,.. '^' ... REFERENCES, 7;^ % :c c,1 - 9 m-u -n . ;~, <,,.'n, ~ v. . m s- . ~ , n, ,,~ +'<,,~2<- n,;7 ?

1. NUMARC/NESP-007 (Rev. 2), Unusual Event HUI i

i

2. Guidelines for Nuclear Power Plant Response to an Earthquake, EPRI l

3. Updated Safety Analysis Report (USAR), Chapters 33.1,33.2 and 3.7

• n i

u i

4. Off-Normal Instruction (ONI) D51, Earthquake, Rev. 4 e

,.a. a... 6.12 Category L: Natural or Destructive Phenomena (Cont.) 1 ~ * ' 'r?,. q ' y M ', '~,?>gt << ?;, ': _},-" [;h;: C N

• N M M #9,f._ f p ;,i.; ;,h,,,e}. Q,,l}' :

, ^ l,'? ,.s .,' ~,, 'L:, ~ '[ ^ ', y l,, , ~., ^ 'S- ,'"S ;, ; <t ;, ^ . ~ ,~ ,,n s. ~.,. v w m r' Ofe ', tS %,"',':'$ d'8'M"? ;p'_;':)';lll@QQy . N'/'. ~,,. L S 'l L. ~j~'yp q_ y;;;;g;-:v -}? lMM)G ~^: 6.12.2' INITIATING CONDITION EAlek' ', G - ' ' ':C'j;{",'j-( ll'-;Q J l g. lR -;, 3 :.;l [ ' z k y i: l 5 1 oa Initiating Conditions Entry Criteria G Control Room receives Report by plant High Greater than Report by report from plant personnel sustained PEI-N11 plant LA1 personnelwho felt an confirming winds with a Maximum personnel earthquake. either of the velocity Safe confirming a following greater than Operating turbine m striking a Safe 90 mph for Value for failure which y LA1 YELLOW REDlight Shutdown equalto or AreaWater resultsin m o seismic on Seismic Buikhng: less than Izvel penetration 3 Natural OR destructive switch Monitorin 15 minutes. (intemal ofthe g g 2 e tornado flooding) turbine A g phenomena affecting indicator g Panel

aircraft, casing.

L B- { Safe Shutdown light 0H13-e Buildings on local P969. barge or E E 2, Seismic train crash R 5 h Monitorng T n= Panel OH51-Missiles 9 i P021. generated E from the turbine failure result in damage to Safe Shutdown Applicable Modes: equipment. m w 00 0 1 2 3 4 5 D -oO f n 6.12.2 LA1 (Cont.) 7' i n:w) ' y >;e a m:,',:n '~$$?'l'S'i'd ~y >^ , ':, :. 9 ,l' -9'l , 2, ,'= :':; = :. Q: ^ ^ :-: ';ilY':NkW:e o , t ',,qt'l. C3*il ' ! " '.:: ~ : ,b .' 'Y '

• IDISCUSSION' ? '

' i = ' I ', l 'l - a u - ~a x. a c + .n. , A' ^~, ~a ,<--6~~',,,T' v, ~.,. ^,~ ~ ~ - ~. s B E Each of these EALs is intended to address events that may have resulted in Safe Shutdown Buildings being G subjected to forces beyond design limits, AND thus damage may be assumed to have occurred to safe shutdown systems. The initial " report" should NOT be interpreted as mandating a lengthy damage assessment pnor to i classification. No attempt is made in these EALs to assess the actual magnitude of the damage. The declaration of an Alert and the activation of the TSC will provide the Emergency Coordinator with the resources needed to perform these damage assessments subsequent to the classification. m> Escalation to a higher emergency class, if appropriate, will be based on the specific system malfunctions, fission m product barrier degradation, abnormal radiological releases, or Emergency Coordinator judgment ICs. [U m =r E f Safe Shutdown Equipment refers to equipment identified in the USAR Appendix 9A. 'Ihis is the minimum list of equipment required to achieve and maintain COLD SHITIDOWN (including all auxiliary equipment such as g. ~ I AC/DC power, cooling water and instrumentation). A detailed list is provided in the " Appendix R Evaluation - g" C l Safe Shutdown Capability Report." to w "N oo o Si 13 6.12.2 LAl (Cont.) .., e,: y. a. >cv. ,,. c,v, - :. 'e ,;, ~,,,.' ' . n+ k ; ' h '!?[^^ ',:b;'!\\'^'ib' '[' >b, '('dYkM> ~ 'b('^'., 9 k iNI.,S.C., USSIO s(Cont.)3 ' >': - - l, ', O ' N[ ' [* y - s -c '~ s s w ss ( D* s o ' $(iih ^ s v ' ,y ss DI s s s ,, [ s, > ', ' ' + 5 'l ;, ' ' s t '; s,( ', ' $')'es s. [s^ 'C '." ' U' ~ l s'[,' ' > ^ s s' aa Safe Shutdown Equipmentlist: (Division 1 and 2 only) G Reactor Protection System Control Rod Drive Hydraulics Automatic Depressurization System /SRV Reactor CoreIsolation Cooling l Low Pressure Core Spray m Low Pressure Coolant Injection - A/B/C Suppression Pool Cooling m Shutdown Cooling j E Safety-Related Instrument Air 9. "2 Emergency ServiceWater g Emergency Service Water Screen Wash g. O Emergency Service Water Pump House Ventilation g o t ECCS Pump Room Cooling System g[ Diesel Generator Building Ventilation g M Stand-by Diesel Generator (DG) = DG FuelOilStorage/fransfer p Electrical Power Distribution 5, Emergency Closed Cooling Pump Area Cooling Emergency Closed Cooling Control Complex Chilled Water MCC, Switchgear and Miscellaneous Electrical Equipment Areas HVAC System Battery Room Exhaust Control Room HVAC and Emergency Recirculation System t T sn 90 O w NN l h i 6.12.2 LA1 (Cont.) n <. > - e >-,. -::., : m: . m ;n v y .w. v' ~ - -:~~ .a ~. m

a,s -

.R

' t' m, +"

il M i" " m:~ b n J ' s ~ '.< ~'

,v-s+.

' ' 'at".n?

DIS,CUSSION (Cont.) 2'

a v L,,: ~ l >

,c, ~ ' ' : .x ~.. ' :, +. ',, D* 1 o m> <>,,n, g

'n % ~, +.

~%,. 'w i j'f: t J,4.,: B i 8 For the purposes of this IC, Safe Shutdown Buildings / Areas are considered to be the following locations: G ControlComplex (allelevations) Auxiliary Building (all elevations) Intermediate Building (all elevations) Fuel Handling Buikhng (all elevations) f ReactorBuilding(allelevations) m Emergency Service Water Pump House (all elevations) Electrical Duct Chase Izading to ESW Building (includes 2 manways per division) m Diesel Generator Building (all areas except the Unit 2 Division 1,2, and 3 DG Rooms) j i SteamTunnel(allelevations) 2.

• 2 l

Diesel Generator Fuel Oil Storage Area g 1 Condensate Stora8e Tank <.n Intake / Discharge Structure g e o. m y ~ An earthquake that exceeds the Operating Basis Earthquake level (0.075 g) is beyond the design basis limits for g M the plant as specified in USAR Section 3.7, Seismic Design. A seismic event of this magnitude can cause damage to safety related systems and functions. ONI-D51 provides indication of a measurable earthquake. p i E i Detection of this event includes activation of seismic monitoring instrumentation along with confirmation from plant personnel who have physically felt the associated ground motion. An evaluation along with a thorough inspection of plant areas and systems will be used to determine the extent of plant damage and will provide the. necessary information to determine if escalation to a higher emergency classification is required. j Maximum Safe Operating Values for " Area Water Izvel," as defined in PEI-N11, are used to quantify the magnitude and significance of plant internal flooding. 'Ihese " area water level" values are all based on equipment j y qualifications, and are identifiable either by installed instrumentation OR water level reference wall-markings in s affected plant areas. O i i 6.12.2 LA1 (Cont.) - m3 g gg . g . !NSlh.;f$gg$2b38$ #g$g! fu. gpg e7; ^ " ' pgp ^ S% Y2 m fdh lM l if ' E2 YN I iREFERENCES gj gsggp g m g y

qq w, w

aw

1. NUMARC/NESP-007 (Rev. 2), Alert HA1 -

G

2. Off-Normal Instruction (ONI) D51, Earthquake, Rev. 4

3. Plant Emergency Instruction (PEI) Nil, Containment Irakage Control, Rev. B

4. Updated Safety Analysis Report (USAR), Chapters 33.1,33.2 and 3.7 m>

t'*

5. Fire Protection Evaluation Report, Section 3 and Table 3-1, Rev. 4 m

s

6. Appendix R - Evaluation, Safe Shutdown Capability Report, Sections 2 and 4, Rev. 5 4

q= w-n s

2.

ma D3 3 O Q. O U 5 = a ws odv T on OQ O w NA. 6.13 Category M: Release of Toxic or Flammable Gasu ",, G ;~ ~j)'L '; h ' J'b's;*!T

';:9 y

;L

' ' Q ' '; ^y 9

,/ n,::

L -

g l, :: C ~ ~,3, ,~ ' t l[w: Y' :l,L ; S- 9 e yO~l4 '. c1, [ J, C< c4. %s, ,^ $13MINITIATING CONDITION MUli,, 0;-: T a wa y,, m .v y,s s 7 v-g q~ y) >"c,- t s - ~, .~ar, s' , v,: f 'r',.

• ~,

~, c'd ' v .A - J..', L ,, y i g '71/v. s s -4 s ,f 4 , v > %- s ^ 5 m ~ Initiating Conditions Entry Criteria w Toxic or explosive gas concentrations Control Roominformed bylocal, ML1 detected within the Protected Area. county, or State officials to evacuate non-essential personnel due to an off-U MU1 site gas release. N U m Release of toxic OR S U m flammable gases a A affecting the Protected 'g en L Area boundary deemed g* g ?. detrimental to the g E B-3 safe operation of the plant Normal operation of the plant is impeded due V { o, to access restrictions. E e -w N N = m Applicable Modes: T 1o O3r 1 2 3 4 5 D s i e {f.h 9 . [,,. I "i ^fj j(. ~ E : < J" ", / ,.g l DISCUSSIONj a ....' 5 '. l,7 g. ~ + ~ .. s) N,l ,^ yg . h. y

i h.Y

'v This IC and its associated EALs are based on gas releases within the Protected Area boundary in concentrations that may affect the health of plant personnel or the safe operation of the plant. This includes gas releases that

p originate both onsite, as well as offsite, and threaten onsite areas.

Ti 9" A toxic gas is considered to be any gas that is dangerous to life or limb by reason ofinhalation or skin contact. A combustible gas, if maintained at a concentration lower than the Lower Explosive Limit (LEL) will not explode due to ignition. 6.13.1 MU1 (Cont.) ,,.,,,,,..,~ ,.c,.,,-c . < @,,,,Q '( -~>,Ns ?^ ' [,, ,,,s, s..,,.,~ f} , a 4 0, . < ; g+, - , ; v;.,'g p .. ~ e j

v,,,, ;' <

.x ,,j c. ." A.:'~: 's 'J ' ' < Ti: ?' i},T': ? ' ':D M-ii^ M' ' 1-;1'~'e r'~ SP;w: if d LDISCUSSION (Cont.) s~, : '

• '- ~"' - +

e en 3 y:: ~. : :e,w:n e: n 3'c:'m a z-n -ae a g,, a~ ,.:,,, ~ n c - '. ;y~ ~ w z

y.,

,e ,, \\ s ' w . ~ ~- 9aD A toxic or flammable gas release is considered to be impeding normal operations due to access restrictions if it is G 6 of sufficient magnitude that access to areas normally accessed to plant operator rounds is restricted. It also includes releases where access to these areas is possible only through the use of protective equipment, such as respirators since this limits the operators visibility and mobility thereby affecting " normal" plant operations. It should NOT be construed to include " confined spaces" that do not require normal access and must be ventilated prior to entry OR situations where the Fire Brigade is using respiratory equipment to protect themselves from the gases released from a fire unless the gases are of such quantity that they also affect personnel not involved in the m fire fighting effort. m i a An offsite event (such as a tanker trucit accident releasing toxic gases) may place the Protected Area within the E us evacuation area. p w a g 2 Site assessment, response, and reportability actions shall be in accuid.nce with PAP-0806. g-C =s o cp < e, s s 3, q g x.,s,' 3, O. 's .. - +, ', q l ~, ~ 7,',0 y. , +~"'y 3,,, y., s, ~s s> s 4 q m ,,~g w ~ ~, < y

• f.; ' p.,, ~ ;'[ ',,.

w gggggggggg - ? % - i;, 9. ~j ^ 3 -. :3 ',9 3' ,s '

, u 'y it

~ ,~ ', 2 2 ' ; s' ';, : ~ ,~,, y ,,~~, +., 1 , ~ + .x ~, s,, c ; > >, nn v, ,o O

1. NUMARC/NESP-007 (Rev. 2), Unusual Event HU3 E

2. Plant Administrative Procedure (PAP) 0806, Oil / Chemical Release Contingency Plan, Rev. 2 i

m . 00 O es N0% i 6.13 Category M: Release oLT_9xic or Flammable Gases (Cont.) m v~~ m p.

u

,,,, e.;,_~: a. n ,,,e. y,,. 6.13I INhTIAbMG CbNDITION mal [ 'h',$ '\\lho ' A ' b$[ : b ' ~ ' h','i 1h d$ w y;;g p'y, ;. (g, ( '::<v.

• +

3: t:; '. yy;;. ; y _ 3 7 ; . 9 ;- ',.; ;. ;!; 7; ;,, n o E Initiating Conditions Entry Criteria w mal MA1 Entry of toxic or flammable gases into Safe Shutdown Buildings or Areas. j i 1 Release of toxic OR flammable gases within Toxic gas in concentrations Flammable gas estimated or Plant personnel NOT tu a Safe Shutdown considered life-threatening determined to be in explosive able to perform actions A Building which concentrations necessary to establish L g1 i and maintain Mode 4 E j jeopardizes operation of systems required to while utilidng R p [ I maintain safe appropriate protective T s 2 operations OR to equipment. 3-l E o establish or maintain a. m cold shutdown en w t g oo O r.n h Applicable Modes: E u 1 2 3 4 5 D 3.s;g. sw - + ~ '.......~ m w'.a e m, ~ ,jj; --, ~- 3 se /- DISCUSSION 0-Wk^* ^ ' 3 e - 4 ~ 4 m r& g g y ,y yyy g ps gg;g. m ;; ~, w y, y This IC and its associated EALs are based on gases that have entered a Safe Shutdown Building and are affecting y safe operation of the plant. The intent of MA1 is NOT to include contiguous buildings or structures (i.e., g warehouse). g u 6.13.2 MA1 (Cont.) sje.f;3 w,, m' e s,

~ '~ s ~,,,'~';,','

<~. ,,, ~ ' ; y, ' ', , ~, o c. . ~, y^' *,, ,r '4,, _h z c ', : l' '

!; W y "'; Q :q yc',n'g glf;'7[J

'g -+8 ~-- <.. DISCUSSI_ON(Cont.), 'c: J;, ' r ?; j':;';; c s.

rg...

.: y m, c -; ', - ;,= p ',.yz y y <,, n.:y; w ,~ $a ~ This IC addresses increased toxic or flammable gas levels that impede necessary access to operating stations or other areas containing equipment that must be operated manually in order to maintain safe operation or perform L a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the Icvel of safety of the plant. The cause of the increase in toxic or flammable gas levels is NOT a concem of this IC. Access to the area must be required, but impeded in order to classify. For example, a toxic or flammable gas m reading in the Intermediate Building IB 599' level pipe chase to radwaste meets the entry condition, but no f i declaration is made since access to this area is NOT needed to safely operate or shutdown the plant. gs N For the purposes of this IC, Safe Shutdown Buildings / Areas are considered to be the following locations: g ~ a Control Complex (allelevations) g-Q Auxiliary Building (all elevations) { o Intermediate Building (all elevations) g Fuel Handling Building (all elevations) m M i Reactor Building (allelevations) ( ^, Emergency Service Water Pump House (all elevations) p Electrical Duct Chase Izading to ESW Building (includes 2 manways per division) El Diesel Generator Building (all areas except the Unit 2 Division 1,2, and 3 DG Rooms) SteamTunnel(allelevations) Diesel Generator Fuel Oil Storage Area l Condensate Storage Tank [ Intake / Discharge Structure t Site assessment, response, and reportability actions shall be in accordance with PAP-0806. y 2 1 53 o .. _ _... ~. 6.13.2 MA1 (Cont.) y n@N am' m anx m~ m ,^ as, - ljde ' ^ % jf g.hh:js$g;t, jjj[* ' 'y e ^"~' 7 s Af84 ^ A 77 3

58) ^ >>

4 %.,:f:? ^ "fs 5 REFERENCES $ ;% ; m;, n,,, y s , ', ; ;, a up y 3 - 7 c s ,, gg m,, 'y Ou .-s"

1. NUMARC/NESP-007 (Rev. 2), Alert HA3

2. Plant Administrative Procedure (PAP) 0806, Oil / Chemical Release Contingency Plan, Rev. 2

3. Appendix R - Evaluation, Safe Shutdown Capability Report, Section 4, Rev. 5

4. Fire Protection Evaluation Report, Section 3 and Table 3-1, Rev. 4 rn>

C# tU

s cn

r O

3* M l3 O" Q. tD "W m m 00 m m o

3Fs m

OQ O W 6 NW Sheet 122 of 138 Page 130 EAL Entry Criteria and Bases (Cont 3 THIS PAGE INTENTIONALLY BLANK ( 6.14 Category N: Security Events t e ' ' ; ' r. : ' 'N' y ',';'x'>'

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~? ': 5: ' ~ ~ ~~ ra'; ' c ' ^ L %%Q M 'c , ', ' s' ' d{#%idi1NITIATING CONDITION NU1:' :: '[,'; $$( '![:', :', '",:' < 4 i ' x; r : - N ~ il q eQ,, ',; e: ~.9 y [ ';u ', ' );,; g ; ; : ' , ; e, (,f, o ' "' ';,.. ; T; ' } l Bo F m Initiating Conditions Entry Criteria G t j NU1 t i U N m NU1 U >r Confirmed security Any security event resulting in the declaration of a SECURITY ALERT in accordance S gs event which indicates to with the PNPP Physical Security Plan. U q A potential degradation in p g a L g thelevel of safety of the B-I3 plant u E g o m a. y cc E B: M a N = T h l Applicable Modes: E u 1 2 3 4 5 D i w

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lg" ~ , gy, ??T W^ 1!% im M s Events which are believed by the Emergency Coordinator to indicate a potential degradation of the level of safety m of the plant should be declared an Unusual Event. Potential degradation of the level of safety of the plant is indicated primarily, but hut exclusively, by exceeding plant Technical Specification LCOs. Precursors of more serious events (e.g., degrading trends) are also included because precursors represent a potential degradation in the level of safety of the plant. 6.14.1 NU1 (Cont.) - s - ' s- , s w m:,. m>: .,.e s- ....s dd.h..,..':y' lj,,,,,. e.,.. ,s s ,,. ',.... s jc s ,4 s' s s : -, c ' l ' g' 'h ' ' Q 'y"' '; : %, ; ',', Q lDISCUSSIOR. (ConQ1'" gm.gh . l'., 5 l < gp;.. ^c.. a a,p~ x g pyyg y s ^ s ~x v v, ,,.: + ' -' - s ',, f, s.. <.;,, ' s m. s c3g OD Security events, which represent a potential degradation of the level of safety of the plant, are addressed by the PNPP Physical Security Plan and would result in the declaration of a Security Alert. 'Iherefore, these items are NOT specifically reiterated here. An increase in the security posture to a Security Emergency will escalate this event to Alert NA1. Security events which do NOT represent at least a potential degradation in the level of safety of the plant are reported under 10CFR73.71 or in some cases 10CFR50.72. m>r s',. ys '4 , ;e m ^ UR;g;g1g - "'7, 7. ; y, W r - ', 1'^ 3 3 3 s' 't' ^ ^ -. c.m...,. iREFERENCESy#.?

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'6.1d.. I.IN. ITIATING, CONDITIO,N NAL,,, ; ~ L' ~ - N ^' l' ' l*: M :c':^24 E ^ , n 'p( :p 3.,, g. y, 9o E Initiating Conditions Entry Criteria es> NAl NA1 A Security event in the Any security event resulting in a declaration of a SECURITY EMERGENCY in L plant Protected A.rea accordance with the PNPP Physical Security Plan. E en R T m Applicable Modes: 4 w 1 2 3 4 5 D g g i 2 ....A.. ' ~ ~' ~ ^ E ~ ~ ^ E o, ~ DISCUSSION,! w m=. g; a g:;,y_ j! - ^ ' , :ws i:!.gic oo o Security events which represent a threat to plant safety are addressed by the PNPP Physical Security Plan. The 1 events that the Security Plan classifies as a Security Emergency are more significant than those classified as a f Security Alert. This increase in the level of concern is analogous to the upgrading from the Unusual Event's C " degradation of the level of safety..." to the Alert's "... substantial degradation of the level of safety...". Intrusion into a Vital Area by a hostile forcc as defined in Site Area Emergency NSI will escalate this event to Site Area Emergency. ~ s ~ iREFERENC.E.. S.).? _......_ ~ en. ^ c u. a

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2. PNPP Physical Security Plan, Rev. 21

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Security event in a plant Intrusion into a plant Vital Area by a hostile force. E 2 g Vital Area A 2-E3 cn mo o CL E e -W M = m m 0 E = R O l O G il E N C Applicable Modes: Y 1 2 3 4 5 D m m OQo W b 6.14.3 NSI (Cont.) m , t. m. m. .r :s, y ^ ~, w we, p~, ' '- DISCUSSIONb M. w m 9 ae %'- ;7 y g. 1, L.m 6

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s, -y 9oa This class of security event represents an escalated threat to plant safety above that contained in Alert NAl in G

that a hostile force has progressed from the Protected Area to the Vital Area. The Vital Area is within the Protected Area and is controlled by key card readers. These areas contain vital equipment which includes any equipment, system, device, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. Equipment or systems which would be required to function to protect health and safety following such, failure, destruction, or release are also considered vital. m l A confirmed explosive device within a vital area is a direct threat to vital equipment designed to protect the y public. If there is conclusive evidence that a vital area has been entered by a hostile force, even though he is no m longer present, the intrusion had been made and a Site Area Emergency is therefore warranted. R Q w a m-l E j For the purposes of this initiating condition, a civil disturbance which penetrates the Protected Area boundary as 3-g well as an individual or group of individuals with known or suspected malicious intent can be considered a hostile g.

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force. However, this hostile force must occupy or gain control of a vital area to meet the criteria for the g j I declaration of a Site Area Emergency. o-to -u m, m v 5. N

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.-gs .,.. g g f 7 ' j f , -y,. . q, 4.);' d + . (, :: 3, , s s , ;s l &l4.4 INIT.IS. TING' CONDITION 'NGC 'Wl:'s:'E -D:N::'i7 O:M-?,:: 981$;4:Q' fNf %g@.; s ' ~W if vv ~ x,~,, s ., ;, y s,.,;;:n r ~,

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n i - : ' x,. ;m ;o +. ;_r;,_ J affq_, __;.;;L :.- > _;;. m Y,~ ;,./2;;9>.. [ N y, <,~, - - u;:.q ::- -^ *" :L, : > Q . ;y';;g',, ;jjg-::7;;3 ;;. lt'f3pg,,. :yz n' l b y s-oo Initiating Conditions Entry Criteria G N.G1 G E N m E t-ngl Loss of physical control of the Control Loss of physical control of the Division 1 R n1 - m Room due to a hostile force or act. and 2 Remote Shutdown Rooms due to a A 3 Security event resulting hostile force or act. L E! o in loss of ability to reach g a I and maintain cold E E3 shutdown M g 7 o. m E tz w R R oo u a G = i E O O N 5 C Y Applicable Modes: t 1 2 3 4 5 D y' ..'* i,,,, ' ~ c ^ ,s { 5 s +'s 5 'y ? ' s s .'>? q f' '"s ^ '. s >*,s,^.- ,[d' s ^ ',. j T . t s< ,s<, ^C c ' :s ,::' d L ':' s ' ~': ' v.

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sgs,, s N ucn ngl encompasses conditions under which a hostile force has taken physical control of areas required to reach and maintain COLD SHIJIDOWN. 6.14.4 ngl (Cont.) n y ,- y, c ,~ <~ , -; n .,,,. ~ <.: ' w : !' li d,' :'M 2.': n J' ~^% ill ~' 5 e, '., '1 J. c;c ;: : {i: 2^'

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~ g B 8 Initiating Conditions Entry Criteria G Dill OU1 Events are in process or have occurred which indicate a potential degradation of the U m N y level of safety of the plant. U m Other conditions j S existing, which in the U judgment of the h A g 51 Emergency L g. G Coordinator, warrant ~ declaration of an g o2 E Unusual Event V $M E 3 N 8 O T g Applicable Modes: 1 2 3 4 5 D !Eif ^ ~ ~._ .M w-N ~ '~ . DISCUSS. IO....N ?.. ~ x g '1:n v s.t.-:: s m Events which are believed by the Emergency Coordinator to indicate a potential degradation of the level of safety y of the plant should be declared an Unusual Event. For those cases where the degradation in the level of safety of g o the plant is tied to equipment or system malfunctions, the decision that the component is degraded should be based upon its functionality and NOT its operability. 6.15.1 OU1 (Cont.) ', : n '~ - w + - v, ' . w :' '

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s ( , AN ,y s v *- " ~ :n w:' D., s 'yw, y f:, s '^, f e, N'^ ' j' / UX 4' Y s .,,%'%,',,p + s s s 1 *-- w A system, subsystem, train, component or device, though degraded in equipment condition or configuration, is G functional ifitis capable of maintaining respective system parameters within acceptable design limits. Releases of radioactive material requiring offsite response or monitoring are NOT expected to occur at the Unusual Event level unless further degradation of safety systems occur. However, if one does occur, it will be classified underIC GUI. m - e ys , e >;, - >,,: x s ~ s s - e ,,- ; s., ~ ,. ~, ,y,, s q;-- -, ' ,t ., ~ ~', ~ s ,y g. e 3, -',1 s 3 ^ m - -J ^ g s^ '.; ^ ~: ' ', ',

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u a-- s s, s. s -.,. s ,3 >f s es- ,s v-en =r O NUMARC/NESP-007 (Rev. 2), Unusual Event HUS g 2 3-w t0 m33 O Q. t'X3 W'"' m un 00 ft un o 3r* a l on 00 O M l o l i t t .~. . ~. - - - - 6.15 Category O: Emergency Coordinator's Judgment (Cont.) J .,[ '.^y^ ,. ~,.., c.,,,v - v, n,,,, >c s, awn : ,ec .,, # f

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. ;,; 8 '.,,, ,,,, '.:, s , ^ s ec! ', d u M '., % '- )( % g ^4':s ; R, . DISCUSSION (Co,ga.4nt.)h, ~;'- ~ w. n.an.w .w+ c ~ c'a',c ' t f ^ ', u~e,; ' ' 2<::r 4 ^4 a ' - c-m- m a :.a (u p ::3 ,,... c 3 w .~. - c. my+ a,,n9 ~~~wa~m, nn 9 O Releases that are expected to be limited to a small fraction of the EPA Protective Action Guideline exposure h levels are addressed underIC HAI. ' g. s.,, 'v, .~' ' : a:-.,.. : : ::. -,,"':, e ~. ,L, > s - .s ,s s s ss s' + s o ?' ,, c a v s,. O2 ' M <

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~ g.; n.: gna Initiating Conditions Entry Criteria G .Q.Sl s I T rn E >r OSI Other conditions exist which indicate an actual or likely major failure of plant functions m j needed for protection of the public. A Other conditions R 2-existing, which in the E 2 g judgment of the A g-G <.n E I 0' Coordinator, warrant E to declaration of a Site M = M Area Emergency E { R Q O G E E N C Applicable Modes: Y 1 2 3 4 5 D m so 00o m.a M 14 t 6.15.3 OSI (Cont.) ,k'- As ~ g.,,, .e -., s,

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up w ' w,.y 's. B aa ~ Radioactive releases may exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Radioactive releases to the general public are addressed underIC HS1 and HG1. g p. ggersyp. wggggp,,g,gfg,. , ~ gs~,r",., pej.m_g,~d. g, , y; ,; e

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~ ggni @ru ~ ~ e' ' ; y , : L', _ en NUMARC/NESP-007 (Rev. 2), General Emergency HG2 f rn 3 P 1 D* OO l2. t m w t M p 3 O i C. W [ w g W M c rsh o3r* s g. I m M i o M Ch ' e md t l i 1 1 PLANT-SPECIFIC EAL GUIDELINES -