ML022910276
| ML022910276 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 09/30/2002 |
| From: | Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML022910276 (43) | |
Text
Committed to Nuclear Excellence DAEC EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMO (TAM-5Q To:
NRC-NRR Document Control Desk US NRC Washington DC 20555 Re:
Entire EAL Basis Document (Table of Contents Rev) (Copy 91)
PSM
Title:
n/a Distribution Date:
Effective Date of Change:
Return by:
09/30/2002 10/07 /2002 10 /21 /2002 Please perform the following to your assigned manual. If you have any questions regarding this TAM please contact Don A. Johnson at 319-851-7872.
REMOVE Rev. 9 EAL Table of Contents Revision EAL EBD-S (PWR: 19080)
Rev. 3 INSERT Rev. 10 Rev. 4 PERFORMED BY:
Print Name Please return to:
Sign Name K. Dunlap PSC/Emergency Planning 3313 DAEC Rd.
Palo, IA 52324 Date To be completed by DAEC EP personnel only:
Date TAM returned:
EPTools updated:
0
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EAL BASES DOCUMENTRe.1 INDEX Page 1 of I PROCEDURE TITLE REV#
REV. DATE Introduction 1
2/01/2000 Definitions 1
2/01/2000 Organization of Basis Information 2
8/812002 EBD-A Abnormal Rad Levels/Radiological Effluent 3
8/8/2002 Category EBD-F Fission Product Barrier Degradation 3
11/20/2000 Category EBD-H Hazards and Other Conditions Affecting 3
8/8/2002 Plant Safety Category EBD-S System Malfunction Category 4
9/30/2002
7 Effective Date:
/
I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County EmergencyManagement officials prior to implementation.
Documentation"of State and County Emergency Management approval is via NEP-A"-o t Approved by:.,,
Manager, Emergency Planning-Date:
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Table of Contents, SUI - Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes..... S-4 SUJ2 - Inability to Reach Required Shutdown Within Technical Specification Limits:... S-6 SU3 - Unplanned Loss of All Safety System Annunciation or indication in the Control Room for Greater Than 15 Minutei Z ZI SU4 - Fuel Clad Degradantion........
S-9 SUS - RCS Leakage...
5....
-13 SU6 Unplann d............................-"
SU6 - Unplanned Loss of All Onsite4 of OTslte dommnuniations'Capabilities............ S-15 SU7 - Unplanned Loss'of Required DC Power During Cold Shutdown or Refuel Mode For Greater Than 15 M inutes.........................................................................
S-17 SA.I - Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Conditions....................................
S-19 SA2 - Failureof Reactor Protection System lnstrijmentation to Complete or Initiate an
- Automatic Reactor ScrariOhce a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful
-21 SA3 - Inability to Maintain Plant in Cold Shutdown.................................................. S-23 SA4 - Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progressý,or (2)
Compensatory Non-Alarming Indicators are Unavailable...............................
S-25 SA. - AC Power Capability'to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout................................................
.............................S-27 SS1 - Loss of All Offsite Power and Loss of All Onsite AC Power to Essential BussesS-28 SS2 - Failure of Reactor Protection System Instrumentation to Complete or Initiate on Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful.............................
S-29
,1[
7 S~EBD-S
- -l.......
i-SYSTEM MALFUNCTION: cATEPORY,-;,,q Rev.
4' a g e 3 SS3 -Loss of All Vital DC Power..........
S..
-31, SS4 - Complete Loss, of Function Needed. to Achieve or Maintain.Hot Shutdown... S-32 SS5 - Loss of Water Level in the Reactor Vessel That.HIas or Will Un-cover Fuel in the Reactor Vessel.................
................................... S-34 S_6 - Inability to Monitor a Sign..ificnt. Transient'"
o"......
36 Piogress SGI - Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power......................................
I
.. S-38ý1 SG2 - Failure of the Reactor Prote6tihn'S"yrtii'to' ompletean Automatic Scram and "Manual Scram was NOT Successful anridhere, is, Indication of an Extreme Challenge to the Ability to Cool'the' Core...........................
........................ S-40 I n-.
44 T**
- 44.
3 jf
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g of 41
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SUI Loss of All Offsite Power to Essential BusseS for GreaternThan 15 Minutes EVENT TYPE: Loss of Power,
OPERATING MODE APPLICABILITY: All EAL THRESHOLD VA4LUE:
The following conditions exist:.
- 1. Unplanned loss of power to both Startup (1X3) and Standby (1X4) trahsformers is expected to last for greater than'15 minutes.
AND
- 2. Emergency Busses 1A3 and 1A4 are powered by their respective Standby Diesel Generators.
UNPLANNED - The loss of power is not the result of a planned evolution.
This event is a precursor of a more serious Station Blackout condition and is thus considered as a potential degradation of the level of safety of the plant. It is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration of an Unusual Event in accordance with this EAL.
The intent of this EAL is to declare an UNUSUAL EVENT when offsite power has been lost and at least one of the emergency diesel generators has successfully started and energized at least one ECCS bus.
Sul
~FERENCES:,
~
Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power UFSAR Section 8.2, Offsite Power System NEI Methodology for Development of Emergency Action
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SU2 Inability to Reach Required Shutdown Within Technical Specification Limits:
EVENT TYPE: Tech. Spec. LCO Action Statement Time Limits Expired OPERATING MODE APPLICABILITY: Run, Startup, H6t Shutdown EAL THRESHOLD VALUE:
The following conditions exist
- 1. Plant is NOT brought to, required operating. mode within the Technical Specifications LCO Action Statement.Tirme.
Limiting Conditions for Operations (LCO) require the plant to be brought to a specific condition when an LCO has been entered. Depending on the circumstances this may or may not be an emergency or a precursor to-a more 'serious event. In any case when a plant initiates a shutdown due to. having entered. an LCO action statement a one, hour report must be made under 10CFR50.72(b) non-emergency events. The plant is within its safety envelope when being shutdown within the allowable action staterient time of a Technical Specification. An immediate classification of UNUSUAL EVENT should be made when the plant is NOT brought;to-,the requir, cL mode, within the' allowable action, statement time. of any Technical Specificatiorn LCO. -Declaration is bas'ed on the time at which the LCO Action Statement specified time, period elapses and is NOT related to' how long a condition may have existed.-
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REFERENCES:
- 1. DAEC Technical Specifications
- 2. NEI Methodology for Development of Emergency Action Levels NUMARCAVESP-007 Revision 4, May 1999 SU2
EBD-S
- 'SYSTEM MALFUNCTION.C.ATEGORY'.."..
- ;
.'.,... Rev. 4 Page
'of 41 SU3 Unplanned Loss of AIhSafety-System-Annunciation in'the Control Room for Greater Than 15 Minutes',
EVENT TYPE:' lnstrurmentation/Commu'nication :
or Indication OPERATING MODE APPLICABILIUTY:YRun;,Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist:.
I'
- 1. Unplanned loss of most' or "al Cn CO5-Annunciatorsa or-lindicators associated with Critical Safety Functions for greater than 15 miinutes.
AND
- 2. Compensatory non-alarming indications are'available.
AND
- 3. In thee.opinion of the 2Operations ]Shift,Mana~er,,,the--Iossof rannunoiators or indicators
-requires increased surveillanceto safely operate the'uhit:
DAEC EAIL INFORMATION:
.,.-,;,lvrv,....,,,:.
Control room panels I C03,A 1C04, and 1C05 ýcbntaih the ann~inciators associated with safety systems at. DAEC.-Therefbre,'the.DAEC:_ EAL;addresses'unplanned loss of most annunciators on these;panels.,.Comnpensatory nbnWalarnmig indications includes the plant process-computer, SPDS, -plant, recorders, -or plant-insirument. 'displayg:in,the control room. Unplanned loss of annunciators or indicators excludes scheduled maintenance and testing activities.
Under the conditions of concern, entry into AOP 302.2, Loss, of Alarm Panel Power, would be made. 'The procedure -requires alerting operators on shift to the nature of the lost annunciation.
It further requires that operators be attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation. Therefore, the generic criterion related to specific opinion of the Operations Shift Manager that additional operating personnel will be required to safely operate the unit is not included in the DAEC EAL because the concern is addressed by the AOP.
SU3
MOST - 75% of safety system annunciators or indicatorsý are lost OR a significant risk that a degraded plant condition could go undetected exists. The use and definition of MOST is not intended to require a detailed count of lost annunciators or indicators but should be used as a guide to assess the ability to monitor the'6perationof the plant.
Unplanned loss of critical safety function indicators (i.e., EOPIEAlC'0irameters) for greater than 15 minutes may preclude operators from taking actions to mitigate a transient Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker. 1D13.-'
.- r Indications of loss of annunciators associated with safety systems include:
125 VDC charger, battery, or system annunciators on'control room panel 1C08
- Loss of "sealed in" annunciators at affected panels Failure of affected annunciator panels shiftily testing by plant operators
- Expected alarms are not received
- Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and 1C05) --
REFERENCES:
- 1. Operating Instruction (01) No. 31 [.2unciato r Sys t-
- 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
- 3. Abnormal Operating Procedure (AOP) 302.2; Loss of Alarm Panel Power
- 3. NE! Methodology for Development of Emergency Action Leveli NUMARC/NESP-O07 Revision 4, May1999 SU3
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'D O C M EBT-S SYSTEM MALFUNCTION CAT7EGORY,;
Rev 4=
Page 9 of 41 SU4 Fuel Clad Degradation EVENT TYPE: Coolant Activity -.
OPERATING MODE APPLICABILITY: All-,
,,c; EAL THRESHOLD VALUE:, :,
I One of the following:
- 1. Valid pretreat radiation monitor (RM-4104, readinggreater, than 4E+3 mR/hrn
-OR:.-
.pc:5x:
- 2. Reactor Coolant sample activity value, ndicating greater than 1.2-jiCi/mIl -dose
-equivalent 1-131 1..,
There are no significant deviations fIomr'the gene 6 EALs.': These EALs-are precursors of more serious fuel clad, degradation and are thus considerid as indicating -a potential degradation of the level of safety of the plant.,...Thus, it-is possible to be operating within Technical Specification LCO Action Statement time,limits for iodine spikes.and make a declaration of an Unusual Event. ;DAEC. mode,app!icability.for these EALs are consistent with the Tech Specs,
-i j
EALt 1 addresses valid.pretreat,rad moift6.rexceeding (RM-4104) 5bdVe,4-E+3 mR/hr.
The calculation supporting this valie is des-ribed below. ;Valid means -that the pretreat rad monitor reading is determined to be 'operable in accordance With the Technical Specifications or-has-been verified by other independent methods such'as indications-,'.
displayed on the control panels, reports from plant personnel, or coolant-sinpling results.:-.:
This reading would be displayed on Control Room panelsAlC-02 and 1C-10 on recorder" RR-4104.'
As specified in the generic methodology, DAEC-EAL-2 addresses --coolant -samples.',
exceeding technical. specification 3.4.6, coolant activity less than or equal-:to 1.2 giCi/ml* '
dose equivalent 1-131.
SU4
Radiological Engineering CalcuLlatio'n' 94-014A and UFSAR Table'15.4-1 were reviewed to" determine a suitable EAL threshold for the pretreat rad monitor reading corresponding to the Tech Spec 3.4.6 coolant activity limit of 1.2 piCVml of dose equivalent 1-131. Using the condenser noble gas source term for the-control rod. drop accident of 2.38 E +06 Curies shown on UFSAR Table,15.4-1 'anidthe oGnhd6seir ftee volumre of 55,000 cubic feet, an initial noble gas concentration in the condenser offgas line is determined. Because the offgas flow rate is veiy-small (about 50 st~ndlkrdicubid feet per minute) compared to the total condenser free volume, 'dilution of'"e noble'gas' concentration due to.
offgas flow is not considered in the calculati6ffthown~bel6w&'
Decrease in the noble' gas source term due to decay of short-lived noble gas radioisotopes and offgas flow dilution effects are addressed by rounding~down thevalue'1alculated as sh'own below.
Calculation 94-014A. used, an;-exposurePrate.arethod-based on' using a source term consisting of a defined,'mixture. of' noblt,& ases, and 'iodine from the control rod drop accident as described in the DAEC UFSAR, Section 15.4. -The'calculation assumed that the activity is released._instantly andimmediýtely. reached in equilibrium with the reactor coolant inventory. Using this calculation, usingdose correction factors (DCFs) for child
(
thyroid dose. from F*g-(Guide 1.109, and-adjusting for the specific gravity (0.736) of saturated water at 1050;,ps~ia..(fluid.conditiong assumed in the calculation) to adjust for standard conditions, the, Ir-13,1 dose equivalent (inunits ofjiCi/ml assuming 1 c6 equals I ml) is determined& for' this! event...'*.This ;result is; then linearly' scaled for-rad monitor' readings corresponding to' the-Tech. Spec,3.4A6& allowable primar*,6.olant; activity of 1.2 1iCVml 1-131 dose equivalent,-4.e.,,,the, relative: mixtur6of,'noble` gasbs and iodine is-assumed to remain constant. 1-129 is ignored because it has no effect on the calculation result..
Isotope DCF,'-,
Concentration'-
-v Correction Factor',
-131 DEQ (1Cicc)
(mrem/pci)-y (jC,/cc)-
[DCFToPE'i DCF&I' 31] /
v'~~~
0.736-1-131 4.39 E-03'
'2 1.6 E+01 "1.4E+00*
, 2.2 E+01 1-132 5.23 E-05.
2.2 E+01 '
1.6 E-02 -i 3.6 E-01 1-133 1.04 E-03' 3.1 E+01 3.2 E-01 1.0 E+01 1-134 1.37 E-05 3.4 E+01 4.2 E-03 1.4 E-01 1-135 2.14 E-04 2.9 E+01 6.6 E-02 1.9 E+00 TOTAL
,3.4 E+01 Therefore, for this event, a coolant activity of 34 pCilcc 1-131 dose equivalent is calculated. Scaling the results for 1.2 liCi/cc 1-131 dose equivalent, a suitable condenser
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I'
'I,
EBD-S
-SYSTEM MALFUNCTION -CATEGORY.-
Rev. 4 Page 11 of 41 source.term,and corresponding initial concentration.in the offgas flow is then determined.
This is then-converted to a pretreat rad monitor reading by use of the monitor efficiency factor:.
f.. 4,,...
Pretreat Rad Monitor fRM-4104) Reading NG concentrationcd damne"-N,G con cr.7trationRoDLRO X [1.2.Ci/cc /34 jiCi/cc]
= [2.38 E +6 Ci.x I,E+6pCiI/CI/,[5,5.E+4,ft3 2.83 E+4 cc/fte] X [1.2 pCi/cc/34 pCi/cc]
= 1529 ICi,x ý0.03530:= 54.0 tCi/cc, Pretreat rad monitor reading = NGx*oncenlration X Rad monito. efficiency Rad monitor efficiency= B9.2 mRfh*,,,-pCiicc, ther4efore:
Pretreat rad monitor reading--=,: 89.2X :54.0,= 4800 mR/hr To account for isotopic decay and dilution ffccts of offgas flowv,:rcund down to 4E+03
, m R/hr-.,.
The calculation results were also reviewed to:determ'ne.f siutabl.values for the main,
steam line (MSL) radiation monitors could be developed. - As showin' above,, the rod drop accident corresponds to coolant activity -of 34,p.lCi/cc :1-131, dose equivalent.
As determined, by.the 'reference calculation,;thila "corresponds' to IaWMSL" radiation monitor reading of about 5.7-R/hr. Scaling the rsulis fof".1.2 pCirmil-131' dose equivalent:
MSL Reading Corresponding to 1.2 pCi/ml 1-131 dose equivalent
"-" [12 pCi/cc]/[34 pClcc))I X 5.7Rhr'l:O.2-R/hi*= 200rn*R/hr 200 mR/hr is at the lower end of the'nornmal M.$L.mpnitor readings-during full power.
Because this value is not distinguishable, and hydrogenj water chemistry system malfunctions that result in increased production of N-16 can also result in increased main steam line radiation le'els-,
i.t s not appropriate, at DAEC to use the main steam line monitor readings.
SU4
'I-
EBD-S:- -
A I. I SYSTEM MALFUNCi'iON CATEGORY Rev. 4 Page 12of41 I
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 672.2, Offgas Radiation/Reactor Coolant High Activity
- 2. Technical Specification 3.4.6, Coolant Chbrnistry
- 3. Radiological Engineering Calculation No. 94-014A, Main Steam Line Radiation Monitor Setpoint Calculation, August 29,1994
- 4. Surveillance Test Procedure (STP) No. 3.4.6-01, Reactor Coolant Gamma and Iodine Activity
- 5. Annunciator Response Procedure (ARP) IC03A, Reactor and Containment Cooling and Isolation
- 6. Annunciator Response Procedure (ARP)"iC05B, Reactor Control
- 7. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 3
- 1.**
- 1 SU4 pr"t I
-i SU5 RCS Leakage EVENT TYPE: Coo ant Leak OPERATING MODE APPLICABILITY: Run, Stariup, Hot Shutdown, Cold Shutdown EAL THRESHOLD VALUE: --
One of the following:
- 1. -Unidentified or pressure boundary leakage greater than 10 gpm.
OR~
- 2. Identified leakage greater than 25 gpm.
OR 2
- 3. -Valid indication of Main Steamline Break.
EAL Threshold Values J and 2 are precursors of more serious RCS -barier challenges.
and are thus considered as.a potential degradation of the level of safety of the plant.
Thus, it is possible to be operating within Techn,-al Specification LCO Action Statement time limits and make a declaration of an Unusunl Event in accordance with these EALs.
Credit for the action statement time limit should only be given when leakage exceeds technical specification limits but has not yet exceeded the Unusual Event EAL thresholds described above. In addition, indication of main steam line break has been added here as discussed in NUMARC Methodology for Development of Emergency Action Levels NUMARCIVESP-007 Revision 2 Questions and Answers, June 1993, Fission Product Barrier-BWR section. This was in response to question 4 which states that the main steam line break with isolation can be classified under System Malfunctions.
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
)
~SU5 ITELBAE'O UM EBD-S SYSTEM MALFUNCTION CATECORY.
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"Page 13of41
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ýNTI EBD-S SYSTEM MALFUNCTION CATEGORY Rev. 4 Page 14of41 The DAEC Tech Spec Section 3.4.4 coolant systemleakageLCO limits are: (1): 5 gpm.
unidentified leakage, (2) < 25 gpm total leakage averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and (3) < 2 gpm increase in unidentified leakage within 'the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in Mode 1. Total leakage is defined, as the sum 'of-identified and unidentified leakage.
DAEC EAL Threshold Value 1 uses the generic value of 10 GPM for unidentified leakage or pressure boundary leakage.
The 10 gpm value for the -unidentified or pressure boundary leakage was selected as it is observable with normal control room indications.
DAEC EALThreshold Value 2 uses identified ý!eakage set' at-a higher value due to the lesser significance of identified leakage in comparison, to unidentified or pressure boundary leakage.
REFERENCES:
- 1. Technical Specification 3.4.4, Coolant Leakage
(
- 2. Surveillance Test Procedure No. (STP) 3.0.0.0-01, Reactor Coolant System Leak Rate Calculation
- 3. Operating Instruction No. (01) 920, Drywell Sump System
- 4. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Recirculation
- 5. Alarm Response Procedure'(ARP) 1C04C, Reactor Water Cleanup and Recirculation
- 6. UFSAR Section 5.2.5; Deteciion:*of,. Leakage' through' Reactor Coolantr'Presure Boundary
- 7. UFSAR Section 15.6.6, Loss-of-Coolant-Accident.
- 8. NEI Methodology for Developmdnt of Emergency Action 'Levels'NUMARC/NESP-O07 Revision 4, May 1999 SU5
EBD-S
-,SYSTEM MALFUNCTION,AT EGORY,,, ;-, V "
Rev. 4 Page 15of41l SU6 Unplanned Loss of.All,Onsiteor;Offsite Communications
,oCap i
EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:,:-,.
a' One of the following groups of.communication losses:,
1 4 4
44
- 4,jý'-
J,
- 1. Loss of ALL of the following onsite communication capabilities affecting the ability to
,.perform routine operation:
Plant Operations Radio System
- Plant Paging System In-plantTelephones,.,
)-g;.*,'-f Sound Power Telephones O R.
4.,
47,.'.,
.,,4-:
- ,,r,.*.:*-..
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- 2. Loss of ALLoffie following offsitecommunicationscapability:
"* All telephone lines (commercial) n..'-.
" Microwave PhoneSystem,,
"* FTS-2000 phone system (ENS & HPN)
" Cellular Phones" DAEC EAL INFORMATION:
There is no significant deviation from the generic EAL. The communications methods used at DAEC are described in the Emergency Plan.
In-plant and external agency telephone communication methods include PABX lines, direct-ring lines, and NRC telephones which are extensions for the Emergency Notification System. There is also a microwave system to provide backup emergency telephone communications.
The availability of one method of,ordinary offsite communication is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when SU6
extraordinary means (relaying of infonnatioh' from' r1/2dio-tirnsmissionis, individuals' being'4 sent to offsite locations, etc.) are being utilized to make communications possible.
The DAEC plant operations radio system is a UHF system with consoles located in the" Control Room, Technical Support Center, Operational Support Center, and the Central Alarm Station.
Hand-held transceivers are. used hn,this, system to provide simplex, communications within the plant and onsite.
The DAEC Radiological Survey Radio System is an 800 MHz trunked/conventional repeater system' that, provides base-to portable communications throughout the DAEC EPZ. A. secondary high-band system provides back-up capability for the 800 MI-lz radio.. Consoles are located in the Technical Support Center and the Emergency Operatiors Facility at the IES Tower. The DAEC Security (backup radiological: survey)- R3fi,. O;3stern provides base-to-portable security communication within the plant and with the Linn County Sheriffs Office using a mobile relay (repeater) type base station and two VHF frequencies. Control consoles are located in the Secondary Alarm Station, Central Alarm Station, Security Control Point, Technical Support Center, and Emergency Operations Facility. The DAECalso has a base station licensed for operation in,-the Police Racd QS3rvice, on the law enforcement state-wide, point-to-point VHF frequency. The transmitter and one control console are located at the Secondary Alarm Station and in the Central Alarm Station.
This station is for communications with Iowa Department of Public Safety radio station, Linn County, Sheriffs office, and the Benton County Sheriffs office. This point-to-point channel is also used by the Linn County Emergency Management and other. public-safety organizationrs throughout the state of Iowa.
REFERENCES:
- 1. Emergency Plan, Section F, Emergency Communications
- 2. NE! Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 K t SU6
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AAA A A, EBD-S SYSTEM MALFUNCTION -CATEGORY.^) "-,.
Rev. 4 Page 17 of 41 SUT7 Unplanned, Loss-ofRequired DO PowerDuring Cold or RefuelModeFor Greater Than.15 Minutes EVENT TYPE: Loss of Power.,.
OPERATING MODE APPLICABILT: 'Cold Shutdown, Refdel *-
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EAL THRESHOLD.VALUE:.;.,A*
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2 The following conditions exist:
A,..
AY
,f Shutdown
.1
- 1. Unplanned loss of Division I (1D1) and DRisid&I'A(1D2)`125 VDC busses based on bus voltage indications. * -
- A-,
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- AA A,
AND 6
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- 2. Failure to restore power to at-leastone reqirrdýirDC. bus,'Wtlin' 1,5 mijinutes from'time of loss.
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DAECEAL INFORMATION:.'
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There is no significant deviationifrom the' genednoEAL.XJUfi[lannedJoss'of Div. I a nd Div.
I1. 125 VDC busses excludes scheduled maintenance and testing adtivities. Under the conditions of concern, AOP 302.1, Loss of 125 VDC Power, would be entered. The DAEC EAL's address the loss of both divisions of the 125 VDC systems consistent with AOP 302.1.
The 125 VDC systnm'.is divided, into two'indei p dn tediAion*
Division' I(1D1)Aand Division II (1 D2) - each with separate AC and DC (battery) power supplies. Loss of both 125 VDC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. These EAL's are intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per SA3 "Reactor Coolant temperature to exceed Technical Specification limit of 212 F or UNCONTROLLED temperature rise approaching the Technical Specification limit of 212".
SU7 I
Bus voltage is based on the lminimum bus i
oitaged ecey sry for theoperation 'of safety 1
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
- 2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power ',*
- 3. Technical Specification 3.8, Electric Power Systems
- 4. UFSAR Section 8.3; Onsite Powee Systeln,-:"
- 5. UFSAR Table 8.3-6, Plant Battery System - DC Power, Instrumentation, and Control, Principle DC Loads (125V)
- 6. ARP IC08A A-9
- 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999.
SU7
SYSTEM MALFUNCTION CATEGORY F
F SA1 Loss of All Offsite Power-and Loss of.All Onsite AC P(
- -,*Essential Busses During Cold Shutdown Conditions EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel, Defueled EAL THRESHOLD VALUE:
The following conditions exist:.
"::J -
- 1. Loss of power to both Startup (1X3) and St4ricdby.lX4) transformers.
AND
- Ir;'v, v
- 2. Failure of "A" Emergency Diesel Generator IG-31 and
£"Ij Emei Generator to supply power to emergency busses 1A3 and 1A4.
AND.
- 3. Failure to restore power to at least oneemeehncy blu, 1A3 6r. A4j wit from the time of loss of both offsite and onsite AC power.
Under the conditions of concern, entry into AOP 301.1, Station Blackout, w under Tab 1. Indications/alarms related to station blackout are displayed or panel 1 C08 and are listed in the procedure under "Probable Indications."
The loss of both offsite and onsite AC power to the emergency buses whei Shutdown, Refuel or Defueled modes, compromises safety systems require heat removal and is a substantial degradation of the level of safety of the p ALERT is declared in Cold Shutdown and Refueling modes due to the less to the protection of the health and safety of the public because of the mucl available to restore power and decay heat removal systems.
15 minutes was selected to exclude transient or momentary power losses.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout
- 2. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical F
-BD-S
- ev. 4
'age 19 of 41
)wer to rgency Diesel hin 15 minutes fould be made n control room n in Cold d for decay ilant. An severe threat "longer time
ýower SA1 l
SYSTEM MALFUNCTION-CATEGORY, EBD-S
(
Rev. 4 Page 20 of 41 I
- 3. Technical Specifications Section 3.8, Electrical Power Systemns-
- 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 SAI
.~~SYSTEM MALFUNCTION,3ATREGORY,,1:
- i '\\v. ',
Rev. 4 S.....
- ; 'Page 21 of 41 SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Rehctor Scram Once a Reactor Protection System Setpoint'Has Been Exceeded andr Manual Scram Was Successful EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY:, Power Operation, Startup EAL THRESHOLD VALUE:
The following conditions must exist to declare this EAL:
- 1. Auto Scram Failure AND
- 2. Operator actions to reduce power are SUCCESSFUL as indicated by either:
- a. ALL Rods Full-in, OR
The condition of concern is failure of the Reactor Protection System (RPS),to.scram -the_--*
reactor when a valid scram signal is present. This condition 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 and thus plant safety has been compromised and design limits of the fuel may have been exceeded.
The EAL evaluation should occur after.'operators have taken actions from the'main control room to insert a manual scram and reduce reactor power. Permissible actions include all actions that can be performed quickly from the main control room by-on-s-shift operators.(e.g., use of the Manual Scram pushbuttons, ARI, placing the Mode Switch in
)
Shutdown, individual scram test switches, etc.). It is not appropriate to delay the EAL SA2
SYSTEM MALFUNCTibKNCAtEG0RY, Rev. 4 Page 22 of 41 evaluation until other time, consuming.actiohs ar&e corýpleted such as manual rod.
insertion or completion of in-plant EOP Support Procedures for rod insertion (e.g.,
venting the over-piston areas of individual CRDs).
Operator actions are considered successful if any of the following results are achieved:
- All control rods inserted to at least position 02 - this is defined in' EOPs as the Maximum Subcritical Banked Withdrawal Position and is'the lowest control rod position to which all control rods may be withdrawn in a bank and the reactor will none the less remain shutdown under all conditions, irrespective of reactor coolant temperature and any boron which may have been injected into the RPV.
- Determination that the Reactor is uShutdown under ALL c6nditions without boron" this can be determined by relying on the Technical Specification demonstration of adequate shutdow'n"mrgi.."..
- One control rod is out beyond position 00 AND
- All other control rods are at position 00 For other combinations of rod pattemrs, and boron -cofcentrtation, reactor engineering will need to perform a shutdown margin calculation.
Reactor power is below the APRM Downscale Alarm Setpoint on-ALL valid APRM instruments.
Note - If the mode switch is in Siartup and the rods are. flly inserted (i.e., the reactor is shutdown) prior to the automatic signal, faiiure,_ then declaration of ai Alert would niot be required. In this case, the event would be reported under '10 CFR 50.72 (b) (2) (1) as a four hour report.
REFERENCES:
- 1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram
- 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999
(
SA2
-SYSTEM MALFUNCTIONGAT-EQORY. 1 -'r SA3 Inability to MaintainPlant in Cold ;Shutdown EVENT TYPE: Inalityto Maintain Shutdown Conditions "OPERATING MODE APPLICABILITY: Cold ShOtdown:'Refuel
.1-w EAL THRESHOLD VALUE:,'"
One of the following:
- 1. Loss of Decay Heat Removal s tem &required to maintain Cold Shutdown.
- AND,
~
~
~
~
- 2. With CONTAINMENT CLOSURE not"n'"
d,!t cstabon b-,,d no sahhd temperature conditions ~exist that either:
- a.
Cause reactor coolant temperature tb e6ceed the-Tebhriical Specification limit of 212 'F.
OR t-c.
.2
.w
- b.
Result inan UNCONTROLLED Iemperature rise bpproaching the Technical Specification limit of,212oF.;:
- i.
- DAEC EALJINFORMA4TION: -' '
Under the conditions, of concem.for EAL Threshold Value 1, AOP 149, Loss of Decay Heat Removal,- woud6lee enteored under" a,6 t;'--Loss 'of "Shutdown Cooling:
Indications/alarms related fo 'lo6
'bf shutdown cooling "are -dsp ayed on control room panels 1 C03 and 1 CO5 and are listed ir'the iOroclu6' id "lProbable Indications." The procedure requires that shutdown cooling be re-established.
The procedure provides curves of maximum water heat up rates which provide an upper bound of the heatup until -an estimated,,time toboil calculation can be completed by Engineering.
The DAEC EAL is written to imply an RCS temperature rise above 212 OF that is not allowed by plant procedures.
This 'corresponds to the inability to maintain required temperature conditions for Cold.Shutdown.
"Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff. The wording SA3 I
i, iz
is also intended to eliminate 'miri6r'cooling inltenuptibios -ccuriing at the transition between Hot Shutdown and Cold Shutdown or temperature changes that are permitted to, occur during establishment of altemate. cors cooling so that an ufnecessary declaration of an Alert does not occur. The uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for event starting from temperatures m.uch lower than the cold shutdown temperature limit.,
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal
- 2. DAEC Technical Specifications
- 3. Surveillance Test Procedure (STP) 3.4.9-01, Heatup and Cooldown Rate Log
- 4. NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, Sepf er
'ibr i99'
- 1. NEI Methodology for Development of Emegency Action Levels NUMARC/NESP-007 Revision 4, May 1999 SA3
EBD-S SYSTEM MALFUNCTION CATECORY t',
' Rev. 4 Page 25 of 41 SA4, Unplanned Loss of Most orAilSafety System Annunciation or z Indication in Control RoomWith Either (1) a significant
,Transient in Progress, or (2) Compensatory Non-Alarmin g Indicators-are Unavailable EVENT TYPE: instrumentation/Communicatio6W OPERATING'MODE APPLICABILITY:. Run, Startup, Hot Shutdown EAL THRESFHOLD VALUE:
Q,.
The folowing coditions exit
- 1. Unplanned loss of most or-all 1C03, 4C04-tradA C05 Annunciators or indicators associated with Critical Safety Functions for-grppter,%apn-15,minutes..
AND V","
1.. In the opinion of the Operations Shift Manager, the loss of all annunciators or indicators
,requires increased surveillance to safely operate the unit.
AND
- 2. Eith'er of the following conditions exist:
- a. A significant plant transient in progress.
- b. Loss of all indication needed to monitor criticality, core heat removal, OR Fission Product Barrier status.
-r Control room panels 1C03, 1C04, and IC05 contain the annunciators associated 'with safety systems at DAEC. Therefore, the DAEC EAL addresses unplanned loss of annunciators on these panels. Compensatory non-alarming indications includes the'plant process computer, SPDS, plant recorders,- or, plant' instrument displays in the control room. Unplanned loss of annunciators.or indicators excludes scheduled maintenance and testing -activities. "Significant transient includes response to automatic or manuially initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater.
SA4
SYSTEM MALFUNCTIONT-CA'EGORY I
I EBD-S Rev. 4 Page 26 of 41 L
-.l.---I Under the conditions of concernm, entry-into AOP 302.2,* Lossof Alarm Panel Power, would be made. The procedure-requires alerting operators on shift to the nature of the lost annunciation.
It further requires that operators be attendant and responsive to abnormal indications that relate to those systems and comlnonents that have lost annunciation. Therefore, the generic criterion related to specific opinion of the Operations, Shift Manager that additional operating perscnnel will be required to,safely operate the unit is not included in the DAEC EAL because the"concem'is addressed by the AOP.
MOST - 75% of safety system annunciators or indicators are lost OR a significant risk that a degraded plant condition could go undetected exists. The use 'nd definition of MOST is not intended to require a detai;ed, ccunt cf lost annunciators or indicators but.
should be used as a guide to assess the ability*Lo monitor the operation of the plant.
Unplanned loss of critical safety function indicators (i.e., EOP/EAL parameters) for greater than 15 minutes may preclude opegratois irofn-taking actions to mitigate a transient.
Annunciators on 1C03, 1C04, and ICO5 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D13. Therefore, DAEC'does not specify a loss of "most" annunciators as specified in the generic methodology.
Indications of loss of annuhciators, associated with safety isystbmrs include:
S125 VOC charger, battery,jor system annunciators on control r6oo rpanel IC08 Loss of "sealed in" annunciators-at affected panels Failure of affected annunciator, panels shiftily testin g'by plant operators Expected alarms are not received'
- Computer point ID B350 indicates UNSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and 1C05)
REFERENCES:
- 1. Operating Instruction (01) No. 317.2 Annunciator System
- 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
- 3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power
- 4. NEI Methodology for Development of Emergency Action Levels NUMARC1VESP-007 Revision 4, May 1999 SA4
EBD-S SYSTEM MALFUNCTION CATEGORY,,
'Rev.
4
[ *Page 27 of 41 SA5 AC.Power Capability to Essential Busses Reduced to a Single Po0We'r-Surce for Greater Than 15 Minutes Such That Any Additional Single'Failure Would-Result in Station Blackout EVENT TYPE" Loss of Power-,
OPERATING MODE APPLICABILII"T:R-titt8ifHdpotShutdowh',
EAL THRESHOLD VALUE:
The following conditi6ns exist:
- 1. Unplanned loss of. power to,. both StartXJp.*Xýy)--ald'--Standby (1X4) transformers is expected to last for greater than 15 mirfgt.
- 2. Onsite power capability, has been degaded;opnetan.of ernrncybusses powered from either A Diesel Generator (1G-31) or B Diesel Generator (GG-21),
and any additional single failure will result in a Station, Blackout.
DAECEALINFORMATION:
The DAEC EAL is written to addressthe-underling.concem,,i.e.,. only one AC power source remains and if'it -1s lost,`
Station E"lacýouf wil1 occuir: Udnler the conditions of -*
concern, entry into AOP 301, Loss of Esseltial.,.ectfical,Power, would b6 made under Tab 1, Lose :of 'One,Essenhtial `4i*6)*;Bus, 1and r under Ta",L6's of Offsite Power.
Indications/alarms related to degradedTAC po a.e'displayed
- 1on, control room panel lC08 andar listedinAO'301 ire aP l lndi.ti's.".'
At DAEC, the Essential Buses of-concern a~e4160VBus~s :1A3 aid1l4. Each of these4 buses feed their associated 480V and 120V AC busses through step dowvn'transformers.
Onsite'power sources at DAEC include theA and B Diesel GeneratorsIlG-31 and IG, 21, respectively.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 301, L'oss of Essential Ele6trical Power
- 2. UFSAR Chapter 8 Electrical Power
- 3. Technical Specifications Section 3.8. Electrical Power Systems.
NEI Methodology for Developmenit of EALs NUMARC/NESP-O07 Revision4, May 1999-"
SA5
,*.*-,*'E A** EB E,
'Q4*
EBD-S, SYSTEM MALFUNCTICN CATEGORY' Rev. 4 Page 28 of 41 SSI Loss of All Offsite Power and Loss of All Onisite AC Power to' Essential Busses:
EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist:
- 1. Loss of power to both Startup (iX3) ar~d S16i6idby (1X4) transformers.
AND
- 2. Failure of both A Diesel Generator (IG-31) AND B Diesel Generator (1G-21) to supply power to emergency busses.
I,,,
AND
- 3. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.
There is no significant deviation' from the generic EAL. In accordance with the generic guidance, DAEC is using a threshold cfý,i15.mlnutes for Station Blackout to exclude transient or momentary power losses.
Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made, under Tab 1. Indications/alarms related to station blackout are 'displayed an control room panel 1 C08 and are listed in the procedure under "Probable lndications."
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout
- 2. Technical Specifications Sectio n 3.8, Electrical Power Systems
- 3. UFSAR Chapter 8, Electric Power
- 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 SS1
EBD-S
,-SYSTEM MALFUNCTION.CATEGORY,'
Rev. 4 Page 29 of 41 SS2-Failure of Reactor Protection System.Instrumentation to -_ 2" Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and "Manual Scram Was NOT Successful' EVENT TYPE: RPSFailure
'T OPERATING MODE APPLICABILITY: Power Operation, Startup' EAL THRESHOLD VALUE:
Failure of automatic scram and actions taEkkqrby4qgerptqrs.in the.Control Room to shut down the reactor OR reduce reactor power below the APRM downscales have been INEFFECTIVE.
The following conditions must exist to declare this-EAL:-
I.
Ill J--I *VV' L.--.J AND_-
i-
- C' v",
- 2. Operator~actions to reduce power are UNSUCCESSFULas indicated by either:,,
-OR
~
~
m"~-,~
This EAL addresses conditions where failure ofan. automaticscramn'has occurred and, manual actions performed in the, Control "Room,,to reduce r6actor power, have been unsuccessful.
Under the conditions of concern for this EAL, the reactor may be pro'ducin-g moreheati than the maximum decay heat load for which safety systems are designed.:A' Site Area Emergency is warranted because conditions exist that may. lead to the potential loss of the fuel cladding or primary containment. Although this EAL ýmay be viewed as redundant' to the Fission Barrier Table, its inclusion is necessary to better assure timely.recognition and emergency response.
SS2 0
F
B'
~EBD-S SYSTEM MALFUNCTION:bATEGORY-Rev. 4 Page 30 of 41 The EAL evaluation should occur after operators havei taken& actions from the main control room to insert a manual scram'and reduce reactor power.- Permissible actions include all actions that can be performed, quickly from the main control room by on-shift operators (e.g., use of the Manual Scram pushbuttons, ARI, placing the Mode Switch in Shutdown, individual scram test switches' etc.).' It is'not 'apropriate to 'delay the' EAL evaluation until other time consuming actions are completed such as manual rod insertion or completion of in-plant EOP Support Procedures for rod insertion' (e.g.,
venting the over-piston areas of individual CRDs).
The purpose of the ATWS EOP, is to, maintain adequate core cooling, shutdown the reactor and cooldown 'the RPV to cold' shutdown '66nditiohs."
The ATWS-EOP is implemented when it cannot be determined that control rod insertion alone will assure that the reactor will remain shutdown under all conditions.
Reactor power above the APRM dovnscale setpoint is indicative of power generation above the decay heat leVels which primary c'ntaiment is designed to suppress.
Furthermore, if reactor power is above the APRM downscale setpoint, it is likely that the core bulk boiling boundary would be above that which provides suitable stability margin for operation lat'higlh *o*ers P
f?
n lkw tows.
Exceeding the Boron Injection Initiation Temp6ratur; -(BIIT) limit (EOP Graph 6) is an indirect indication' that the reactd*fis',t tpovwer and that excessive decay heat is beinrg added to the suppression pool.
The higher the reactor power level is; the idore heat energy will be rejected to the torus thus requiring a lower toribý ternpeiature for initiati6n'of boron injection if the Heat Capacity Limit is not to be exceeded before reactor shutdown is achieved.
As long as the core remains submerged (the preferred method of core cooling), fuel integrity and RPV' integrity are 'not directly. challenged even' under failure-to-scram conditions. However, a scram failure coupled with an' MSIV isolation results in rapid heatup of the torus due to the steam discharged from the RPV via SRVs.
The challenge to the primary containment will thus become a limiting factor.
REFERENCES:
- 1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram
- 3. NEI Methodology for Development of EALs NUMARC/VESP-007 Rev 4, May1999 j
SS2
z A-i a-SS3.Loss of All Vital DC Power.
EVENT E: Loss of Power OPERATING MODE APPLICABILITY:,Ru-g;Startup, Hot Shutdown EALTHRESHOLD VALUE:"-'
The following condition exists:
]L
- 1. Loss of both'divisions of the Vital 2501125V'IDC~systemrbadpq on APP,302.1 and AOP" 388 for greater than 15 mineit' "o"
Under the conditions of concern, AOP' 302.',
Losso1 VDC Power, would be entered under Tab 3, Complete Loss of 125 VDC." C6nhieqently, the DAEC EAL addresses loss of both divisions of the 125 DC.system consistent, P,,0with O,
At DAEC, the 250V/125V DC Systems ensurepower is avyailabte;for the reactorto be shutdown safely and maintained in a safe condition. The 125V ystem is divided into two independent divisions - Divjsiopd fl,and,.Divisiori, Tn,with y
separate" DC: power supplies.
These powier, supplies consist.ob serving systems such as RCIC, RHR, EDGs, and HPCI. -
Complete loss of both 125V DC-Divisions' -could compromise the abilityto monitor and control the removal-of decay heat during cold shutdown or refueling operations...
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 302.1; Loss of 125 VDC Power
- 2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power
- 3. -Technical Specification 3.8, Electrical Power Systems
- 4. UFSAR Section 8.3, Onsite Power Systems,
- 5. UFSAR Table 8.3-6, Plant Battery System - DC Power, Instrumentation, and Control, Principle DC Loads (125V)
- 6. NEI Methodology for Development of Emergency Action Levels NUMARC1VESP-007 Revision 4, May 1999 SS3
SYSTEM MALFUN'CTION'CATEGORY, :'
A Rev. 4 Page 32 of 41 SS4 Complete Loss of Function Needed to Achieve'or Maintain Hot Shutdown EVENT TYPE: Inability to Maintain Shutdown Conditions' OPERATING MODE APPLICABILITY: Ruh, Startup,
.i-Ht,Shutdown.
EAL THRESHOLD VALUE:
- 1. EOP Graph 4 Heat Capacity Limit is exceeded.
- 2. Reactor CANNOT be brought subcriti*a-l.
This EAL addresses complete loss of functions,- including ultimate heat sink and reactivity control, required for hot shutdown6with thý..eactor at-pressure.and temperature. Under these conditions, there is ani actual mrajofailJre of a systemI intended for protection of the public.
The reactivity condition' criteria is addressed by 'maintenance of required shutdown margin.
If inadvertent criticality could not be eliminated by performing the actions of AOP 255.1, AOP 255.2, or the AWVS EOP, it corresponds to a failure of a system intended for the protection of the public and thus classification as a Site Area Emergency is warranted.
This EAL represents an escalation from the conditions of concern in SA3, Inability to Maintain Cold Shutdown, because the reactor is at operating pressure and temperature and decay heat levels are higher.
Per DAEC Technical Specifications, the following systems are necessary to achieve or maintain Hot Shutdown conditions:
"* Reactor Protection System Instrumentation
"* Core and Containment Cooling Systems Instrumentation
"* Reactivity Control
"* Standby Liquid Control System
"* Core and Containment Cooling Systems SS4 Ii
EBD-S SYSTEM MALFUNCTION-,CATEGORY,.
Rev. 4 Page 33 of 41
"* Primary System Boundary
"* Auxiliary Electrical Systems Loss of instrumentation is addressed by SS6,.lnability to Monitor a Significant Transient in Progress: The Auxiliary Electrical Systermis aiddressed by SSI,' St'ation'Blackout, and SS3, Loss of all Vital DC Power anltheref.ore they are,not coyer edhere. Failure of the primary system boundary is' cov'red 'byti'iio"Fssion 'BaMi6'r Table" and SU5 RCS Leakage.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Reiyioval
- 2. Abnormal Operating Procedure (AOP) 255.1, Control Rod -Movement/Indication
'Abnormal
- 3. Abnormal Operating Procedure (AOP) 255.2, Power/Reactivity Abn6rmral Change
- 6. Emergency Operating Procedure ALC -Alternate Level Control
- 8. NEi Methodology for Developmentbof En*efeit6y A'-tion L. evlsNUIMARC/NESP-O07 Revision 4, May 1999' At-i S
jI*
4 I,...-.t
-'4-
- 4
1=
,lf SS4
SS5 Loss of Water Level in the Reactor VesseiThat Has or Will Uncover Fuel in the Reactor Vessel EVENT TYPE: Inability to, MaintainShutdown Conditions, OPERATING MODE APPLICABILITY:.Cold Sohu~tdown, Refuel EAL THRESHOLD VALUE:
The following conditions exist:
- 1. Loss of Reactor.Vessel Water Level'=s iridi.ated by:
- a. Loss of all decay-heat rmoval, c.;ing as determined-by AOP 149 under Tab 1, Loss of Shutdown Cooling.
AND
- b. RPV level below 15 inches indicating that the core is or will be uncovered.
The DAEC EAL is writtdn-in t6rms'of the'gen6ral cfncerh that no cooling water source is lined up or available for injection into theýRV knd water level is decreasing below the top of the active fuel (TAF).! Under the'cohditionsof ccrner for, EAL Threshold Value 1, AOP 149, Loss of Decay Heat Removal, would be entered under Tab 1, Loss of Shutdown Cooling. Indications/alarms related to loss of shutdown cooling are displayed on control room panels 1C03 and 1C05 and are listed in the procedure. Consistent with the value used in the EOPs, the EAL uses an indicated RPV level of 15 inches for the water level corresponding to TAF.
The conditions address concerns raised by the NRC AEOD Report AEOD/EG09, "BWR Operating Experience Involving Inadvertent Draining of the Reactor Vessel", dated August 8,1986. This report states:
In broadest terms, the dominant cause of inadvertent reactor vessel draining are related to the operational and design problems associated with the residual heat removal system when it is entering into or exiting from the shutdown cooling mode. During this transitional period water is SS5
b-EBD SYSTEM MLUCINCATEGORY,'*;-.-/*-
Rev. 4 Page 35 of 41
.drawn fromrthe reactor vessel, cooled by RHR heat exchangers (from the "cooming provided by the servic"IWater system), ai'd Jreturned to the reactor vessel. First there are piping and valves in the residual heat removal system which are common to both the shutdown cooling mode and other modes of operation such as1low pressure c6olarit injection and, suppression pool coolirig. These valves, when improperly positioned provide a drain path fdr'th' ie'ret o66la*t to flow` frbrh the reactor vessel to the suppression pool or the radwaste system. Second, establishing or exiting the shutdown cooling mode of operation is entirily manual making such evolutions vulnerable to personnel and procedural errors. Third; -"
there is no comprehensive valve interlock arrangement for all the residual heat removal system valves that.could be.'activated,during shutdown cooling. Collectively, these factors have contributed to the repetitive occurrences of the operational evints-irvolving the inadvertent draining of the reactor vessel.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal
- 4. NRC AEOD Report AEOD/EG09"l, VWR;Op~erting Experience Involving Inadvertent Draining of the Reactor Vessel,,August 8,,,IQ83,W, 5., &NE!
Methodology for-Developmentof Emergenr, y Action Levels NUMARCG/ESP-007 Revision 4,,May 1999 SS5
EBD-S SYSTEM MALFUN-CTION-CATEGORY ',:[
- Rev. 4 Page 36 of 41 I
SS6 Inability to Monitor a Significant Transient in' Progress EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist:
- 1. Significant transient in progress and ALL of the following:`
- a. Loss of annunciators on P anels IC03,,lC04 and 1C05 AND
- b. Compensatory non-alarming indications are unavailable.
AND
- c. Indications needed to monitor' criticality, OR core heat removal, OR Fission Product Barrier status.are unavailable..
The DAEC EAL is written in terms of a significant transient in progress with loss of both safety system annunciators and loss of compensatory non-alarming instrumentation. The DAEC EAL structure, which addresses all the key points in the generic EAL, better assures that the condition of concern for this EAL will be readily recognized.
Significant transient includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or undamped thermal power oscillations greater than 10%.
Compensatory non-alarming indications include the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. These indications are needed to monitor (site-specific) safety functions that are of concern in the generic EAL.
SS6
"r
'tAL E.
ED-S
-, SYSTEM MALFUNCTIOI CATE'=GORY-,:,,-" -< :o*'-
Rev. 4 Page 37 of 41 Control room panels 1C03, 1C04, and - C05, contain the annunciators associated with safety systems at DAEC. 'Annunciators on r1C03, %C604 and 1005 share.a common power supply from 125 VDC Division I that isfed through circuit breaker' 1D133.
Therefore, DAEC does not specify a loss of "most" annunciators as specified in,the.:
generic methodology.
Indications of loss of annunciators associated with safety systems include:
- 125 VDC charger, battery, or system annunciators on control room panel 1C08
- Loss of "sealed in" annunciators at -affected,panels,:,-,..
- Failure of affected annunciator panels shiftily testing by plant operators Expected alarms are not received
- Computer point ID B350 indicates "NSS*AINN"C LOSS TRBL" (Loss of DC power to panels 1C03, 1C04, and 1C05)
REFERENCES:
- 1. Operating Instruction (01) No. 317.2, Annunciator System
- 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
- 3. Abnormal Operating Procedure (AOP) 302.2, Loss ofAlarm Panel Power
- 4. NEI Methodology for Development of Emergency Action Levels NUMARCINESP-007 Revision 4, May 1999
,C
.. I SS6
MEN
-EBD-S.
SYSTEM MALFUN-FbTN'6NICEGORY Rev. 4 Page 38 of 41 SG1 Prolonged Loss of All Offsite Power and, Prolonged Loss of All Onsite AC Poweir" EVENT TYPE: Loss of Poweir OPERATING MODE APPLICABILIl': Run; Stirtuip, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist:
- 1. Loss of voltage on buses 1A3 and1A4..) -
AND ANY ONE OFTHE, FOLLW0 '..
- a. Restoration of pov~
to'. efther Bus 1A3 or 1A4 is not likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- b. RPV level is indetemi nate'..
.P..W. -
- c. RPV level is below +15 inches.
There is no signficant deviation frothe generic,EAL. Under prolonged Station Blackout (SBO) conditions, fission product býaiiier" monitoring capability may be degraded.
Although it may be difficult to predict when power can be restored, it is necessary to give the EC/OSM a reasonable idea of how quickly a General Emergency should be declared based on the following considerations:
Are there any present indications that core cooling is already degraded to the point where a General Emergency is IMMINENT (i.e., loss of two barriers and a potential loss of the third barrier)?
If there are presently no indications of degraded core cooling, how likely is it that power can be restored prior to occurrence of a General Emergency?
'ELME-DcAN EBD-S SYSTEM MALFUNCTION CATEGORY Rev. 4 Page 39 of 41 The first part of this EAL c'orresponds.to tre'shold conditioi-ns for Initiating Condition SS1, Statio'n Blackout - namely, entry into AOP 301.1, Station Blackout. The second part of the EAL addresses the conditions-that Will escalate the SBO to, General Emergency.
Occurrence of any of the following is sufficient for escalation: (1) SBO coping capability exceeded, or (2) loss of drywell cooling,that continues to make RPV water level measurements unreliable, or (3) indicationh gof inadequate core "cooling*: Each of the'se conditions is discussed below:
- 1.
SBO Copinq Capabilitv Exceeded DAEC has a SBO coping duration of four hours...
eThmlikelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of niitigatig the eveht could result in a loss' of valuable time in preparing and implementing public protective actions.
- 2.
RPV Water Level Measurements Remaining Unreliable Flashing of the reference leg water will result in erroneously.high RPV water 'level readings giving a false indication of actual water inventory and 'potentially indicating adequate core co01ing when it may not exist. EOP Graph 1, RPV Saturation Temperature, defines the conditions under which RPV level instrument leg boiling may occur.,
- 3.
Indications of Inadequate Core Coolinq DAEC uses the RPV level that is used for the Fuel Clad "potential loss" condition -in the Fission Product Barrier Matrix. This is RPV lyvel below' +15 inches.'-
REFERENCES:
- 1. "Abnormal Op erating Procedure (AOP) 301i.1, Statio(n Blackout'
- 2. 'Letter NG-92-0283, John F. Franz,'Jr. to Dr. Thomas E. Murley,-Response'.to Safety
-Evaluation by NRC-NRR"Station Blackout Evaluation Iowa ElectricLightfarid Power
,.Company Duane Ambid Energy Center," February "10, 1992
- 4. Emergency Operating Procedure (EOP)-ALC - Alternate Level Control
- 5. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007,.
Revision 4, May,1999 SGI
A4 EBD-S SYSTEM MALFUNC' iC-(AtEGORY Rev. 4 Page 40 of 41.
SG2 Failure of the Reactor Protecti6n Systeimtto *Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Power Operation, Startup EAL THRESHOLD VALUE:
Failure of automatic and 'manual scrarmnAND 'conditions exist that no longer assure adequate core cooling or adequate decay heat removal.
The following conditions must exist to dlecIar this EAL:
- 1. In AFWS EOP AND
- 2. Loss of adequate core cooling cr decaybieat removal capability as indicated by either:
- a. RPV level cannot be restored and maintained above the Minimum Steam Cooling RPV Water Level (i.e., SAG Entry Required),
OR 7";
- b. HCL Curve (EOP Graph,4) exceeded.
This EAL addresses conditions where failure of an automatic scram has occurred and manual actions performed in the Control Room to reduce reactor power have -been unsuccessful AND a subsequent loss of adequate core cooling or decay heat removal capability occurs.
If either of these challenges exist during an ATWS, a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time.
The purpose of the ATWS EOP is to maintain adequate core cooling, shutdown the reactor and cooldown the RPV to cold shutdown conditions.
SG2 t-
ý B
0 EBD-S SYSTEM MALFUNCTION CATEGORI Y'-"!.};
Rev. 4 Page 41 of 41 implemented when it cannot be determined that control rod insertion alone will assure that d a l c diti6ns.
the reactor will remain shutdown u er all,66n
.s If injection with all available -Preferred andAltemate AT'WS'Injection -Systems fails to provide sufficient injection to restore and maintain level above -25 inches (Minimum Steam Cooling RPV Water Level), adequate core cooling is threatened and submergence of the core is attempted by flooding the primary containment. This is accomplished by transfer to and implementation of the DAEC Severe Accident Guidelines (SAGs).
The Heat Capacity Limit (EOP Graph 4) is d"6efie:i to be'the highest"torus temperature at which initiation of RPV depressurization will not result in exceeding the Primary Containment Pressure Limit (the PCPL is 53 psig at the DAEC) before the rate of energy transfer from the RPV to the primarypco,ntainrnent is within the capacity of the containment vent.
Control of torus temperature relative to.-the Hest,Capacity Limit is.directed in the Primary Containment Control Guideline, EOP 2. If the abtidris being tak6ri'in EOP 2 to preserve torus heat capacity are inadequate or not effective, RPV pressure must be reduced in order to remain below the Heat Capacity Limit. Therefore, actions in the RPV pressure control section of the ATWS EOP must accommodate these requirements. Failure to do so may lead to failure.of the containment:or.,dss, cofequiprnent necessary for the safe shutdown of the plant.
REFERENCES:
- 2. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May1999 b