ML20065N461
| ML20065N461 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 04/20/1994 |
| From: | UNION ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20065N458 | List: |
| References | |
| NUDOCS 9404270169 | |
| Download: ML20065N461 (87) | |
Text
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O O EMERGENCY ACTION LEVELS INDICATIONS BASES l
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Bases For:
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- 1) Classification of Emergencies EIP-ZZ-00101, Rev. 16 Attachment 1 7Q l
- 2) Radiological Emergency Response Plan Rev. 17 Chapter 4, l Table 4-1 lO i
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Prepared by:
Reviewed by:
Supervisor, Emergency Preparedness J
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1 04/20/94 9404270269 940421 ,
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p EMDIGENCY ACTION LEVEL INDICATIONS DASliS Group 1 Abnormal Radiation Eventa
\ l Initiatina Condition Emeraency Classification A. Any Unplanned Release of Unusual Event l i.' Radioactivity to the l Environment That Exceeds '
2 Times the Radiological Effluent Control Limits in the ODCM (APA-ZZ-01003) for 260 minutes.
MODES: At All Times Indicationa
- 1. all of the following:
- a. A valid alarm and reading on any effluent monitor:
HB-RE-18 (Channel 186)
GT-RE-21B (Channel 213)
GT-RE-10B (Channel 103)
- b. The valid reading is 2 times the Hi HiJ alarm setpoint (trip setpoint) value.
- c. The release cannot be terminated within J 60 minutes of the alarm actuation. ;
- 2. Both of the following:
- a. Confirmed sample analysis indicates that a release exceeding 2 times the applicable 7' values of the ODCM (APA-ZZ-01003), has occurred.
- b. The release cannet be terminated within 60 minutes.
Bases Since Callaway eliminated Effluent Technical Specifications as provided
) in NRC Generic Letter 89-01, we use the Radiological Effluent Control Limits (REC's) in APA-ZZ-01003, our Offsite Dose Calculation Manual (ODCM).
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Any Unplanned Release would be any inadvertent'or accidental release of radioactive material. An Unplanned Release is also a release'via normal o pathways without a release permit or preper authorization, or without proper sampling and analysis, or resulting in significant deviation from
-the requirements of the release permit.
Valid alarms and readings are those~ verified by the operators to be the
!- results of effluent concentrations. Invalid alarms and-readings may be the result of electronic noise, radio frequency interference, electromagnetic frequency interference, or spurious. spikes of unknown b nature. A buildup of radioactivity within the monitor or an-increase in l the ambient background for the monitor would also cause an invalid' alarm.
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l EMERGENCY ACTION LEVEL INDICATIONS BASES I
Group 1 Abnormal Radiation Events The time frame of 60 minutes is used to indicate a definite loss of control. This is also the time used in 10CFR50.72 for a continuing release that would require notification. This loss of control for 260 minutes is of more significance than the level of release in this EAL.
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EMERGENCY ACTION LEVEL INDICATIONS BASES !
' Group 1 Abnormal Radiation Events 1
Initiatino Condition Emeroency ClassificatioD l B. Any Unplanned Release of Alert ..
Radioactivity to the j J' Environment That Exceeds j I
200 Times the Radiological Effluent Control Limits in the ODCM (APA-ZZ-01003) for 215 minutes.
[) MODES: At All Times .l Indications
- 1. All of the following:
- a. A valid alarm and reading on any effluent monitor:
HB-RE-18 (Channel 186) 1 GT-RE-21B (Channel 213)
GT-RE-10B (Channel 103) l
- b. The valid reading is 200 times the Hi Hi alarm
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setpoint (trip setpoint) value,
- c. The release cannot be terminated within 15 minutes of the alarm actuation.
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- 2. Both of the following:
- a. Confirmed sample analysis indicates that a release exceeding 200 times the applicable values of the ODCM (APA-ZZ-01003), has occurred.
- b. The release cannot be terminated within 15 minutes.
Base.g
) Since callaway eliminated Effluent Technical Specifications as provided in NRC Generic Letter 89-01, we use the Radiological Effluent Control Limits (REC's) in APA-ZZ-01003, our Offsite Dose Calculation Manual (ODCM).
This event escalates from the Unusual Event by escalating the magnitude of the release by a factor of 100. The increased level of release is
) the significant factor in this EAL. The duration is reduced to-15 minutes in recognition of the increased level'.
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E}JERQENCY ACTION LEVEL INDJfATIONH3ASFIS Group 1 Abnormci, Radiation Events i
Initiatina Condition Emercency Clancification C. EAB Dose Resulting Site Emergency i From an Actual or Imminent 1 Release of Gaseous Radioactivity Exceeds 100 mrem TEDE or 500 mrem CDE Thyroid for the Actual or Projected Duration of the !
Release.
MODES: At All Times Indicationn Any of the following:
- 1. A valid reading on the Unit Vent monitor GT-RE-21B (Channel 213) indicates >0.1 E+5
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) Ci/aec for 15 minutes.
- 2. A valid dose projection indicaten > 100 mrem TEDE or >500 mrem CDE thyroid dose at the EXCLUSION '
AREA BOU14DARY using inplant rad data or field monitoring. team survey resulto.
- 3. Field survey results at the EAB corresponding to
>100 mrem /hr TEDE for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (or expected to . ,
continue for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) or 3500 mrem /hr CDE thyroid j for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inhalation.
- Declare the event using this indicator only if an 1 actual doao asseaament per Indicator 2 cannot be performed in 15 minuten.
1 Eaneo valid alarma and readinga are those verified by the operatora to be the results of effluent concentrations, Invalid alarms and readinga may be the result of electronic noise, radio frequency-interference, electromagnetic frequency interference, or apurious opikes of unknown nature. A buildup of radioactivity within the monitor or an increase in the ambient background for the monitor would also cause an invalid alarm.
3 The 100 mrem integrated dose in thin initiating condition provides a
> desirable gradient (one order of magnitude) between the Alert, Site Area Emergency, and General Emergency classes. It is deemed that exposures lena than this limit are not consistent with the Site Area Emergency clana description. The 500 mrem integrated thyroid done was established 1 in consideration of the 125 ratio of the EPA Protective Action i Guidelines for whole body and thyroid.
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Actual Meteorology should be used whenever possible aince it given the moat accurate dose annessment.
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[;h4EBQENCY ACTION !.EVEl,INDICATipNS BASES i
! Group 1 Abnormal Radiation Evento Thyroid doses are based upon EPA 400, manual of protective action guides and protective actions for nuclear incidents.
All setpoints were calculated using EIP-ZZ-01211 PC Based Plume Phase l Dose Assessment. This program is based on EPA 400 guidelines. The
) default nuclide mix was used for all setpoint calculations. Annual Average Meteorology was determined from FSAR 2.3-82 for the unit vent monitor and containment High Range Arms, and FSAR 2.3-84 was used for the radwaste vent. In all cases due to the I/NG ratios for the accident types the thyroid was the limiting case.
- 1. The Unit Vent setpoint is calculated using a relative concentration, X/O (sec/m 3) from FSAR Table 2.3-82 " Average Meteorological Relative Concentration Analysis Special Distances, Unit Vent Releases". An average of all directions was taken.
This value, 1.3 E-6 was used with conversion factors for a Main Steam Line Break. These factors are the most limiting for all unit vent accident types. Dased on a 4 hr release
) duration, and using our calculated default nuclide mix, a j'
setpoint of 8.1 E+5 pCi/sec resulted in a 500 mrem CDE thyroid dose at the EAB.
Unit Vent Setpoint = 8.1 E+5 pCi/sec (GT-RE-21B) l l
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O EMERGENCY ACTION LEVEL INDICATIONS BASES Group 1 Abnormal Radiation Events Initiating Condition Emergency Classification D. EAB Dose Resulting General Emergency From an Actual or Imminent Release of Gaseous
$P Radioactivity Exceeds 1000 mrem TEDE or 5000 mrem CDE Thyroid for the Actual or Projected Duration of the Release.
MODES: At All Times C
Indicationn Any of the following:
- 1. A valid reading on the Unit Vent monitor
{} GT-RE-21B (Channel 213) indicates >8.1 E+6 pCi/sec for 15 minutes.
- 2. A valid dose projection indicates >1000 mrem TEDE or >5000 mrem CDE thyroid dose at the EXCLUSION AREA BOUNDARY using inplant rad data or field monitoring team survey results.
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- 3. Field survey results at the EAB corresponding to
>1000 mrem /hr TEDE for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (or expected to continue for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) or >5000 mrem /hr CDE thyroid for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inhalation.
- Declare the event using this indicator only if an g actual dose assessment per Indicator 2 cannot be performed in 15 minutes.
Banen g Valid alarms and readings are those verified by the operators to be the results of effluent concentrations. Invalid alarms and readings may be the result of electronic noise, radio frequency interference, electromagnetic frequency interference, or spurious spikes of unknown nature. A buildup of radioactivity within the monitor or an increase in the anWient background for the monitor would also cause an invalid alarm.
O The setpoints in Indicator 1., are 10 times the values calculated for EAL IC. The 1000 mrem whole body and the 5000 mrem thyroid integrated dose are based on the EPA protective action guidance which indicates that public protective actions are indicated if the dose exceeds 1 rem whole body or 5 rem thyroid. This is consistent with the emergency class description for a General Emergency. This level constitutes the upper level of the desirable gradient for the Site Area Emergency.
Actual Meteorology should be used whenever possible since it gives the most accurate dose assessment.
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EMI))LQENCY ACTION LEVEL INDICATIONS BASES Group 1 Abnormal Radiation Events l
Thyroid doueu are based upon EPA 400, manual of protective action guides and protective actions for nuclear incidents.
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EMERGENCY ACTION LEVEL INDICATIQNS BASES j Group 1 Abnormal Radiation Events b
Initiatinq Condition Emerqency Classification E.* An Unexpected Increase in Unusual Event i Plant Radiation. l l
45 MODES: At All Times Indications Ap_y of the following:
[} 1. Spent Fuel Pool level is decreasing on EC-LI-0039A with Normal makeup being added, and all irradiated fuel assemblies remain covered.
- 2. Refueling Pool level is decreasing on BB-LI-0053A or B with Normal makeup being added, and all irradiated fuel assemblies remain covered.
e 3. Any valid (Confirmed by HP survey) ARM (other than a Group 1,G. Safe Shutdown ARM) >1000 times normal. (Normal levels can be considered as the monitor reading prior to the noticed increase.)
- This Initiating Condition '.s not meant to apply to g, anticipated temporary increases due to planned events (e.g., incore detector movement, radwaste container movement, depleted resin transfers, upper internal movements, etc.)
Banen Valid alarms and readings are those verified by the operators to be the results of effluent concentrations. Invalid alarms and readings may be the result of electronic noise, radio frequency interference, electromagnetic frequency interference, or spurious spikes of unknown nature. A buildup of radioactivity within the monitor or an increase in the ambient background for the monitor would also cause an invalid II alarm.
All of the above events tend to have long lead times relative to potential for radiological release outside the site boundary, thus impact to public health and safety is very low.
Indicator 3 addresses unplanned increases in in-plant radiation levels WB that represent a degradation in the control of radioactive material, and represent a potential degradation in the level of safety of the plant.
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3 EMERGENCY ACTION LEVEL INDICATIO_NS BASES I
. Group 1 Abnormal Radiation Events .
k Initiating Condition Emergency Classification F.* Major Damage to Irradiated Alert
- Fuel or Loss of Water Level
) That Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel.
MODES: At All Times l Indications j-ABY of the following:
- 1. A VALID Hi-Hi Alarm on Fuel Building exhaust monitors GG-RE-27 p_t 28 ' (Channel 273 p_t 283) 1.46 E-3 pCi/cc.
Containment refueling bridge area radiation
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monitor (SD-41) 2150 mr/hr.
- 3. Fuel building area radiation monitor (SD-37 p_r r
- 38) >70 mr/hr.
- 4. Report of visual observation of loss of water 1evel resulting in irradiated fuel being
} uncovered.
- This Initiating condition is not meant to apply to anticipated temport increases due to planned events (e.g., incore detector movement, radwaste container movement, depleted resin transfers, upper internal g movements. ; q 1
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l-Bases This IC applies to spent fuel requiring water coverage.
NUREG-0818, " Emergency Action Levels for Light Water Reactors," forms the basis for these EALs.
For indicator 1, the Hi-Hi' alarm setpoint of'1,46 E-3 is used, which'is the Tech. Spec. required trip setpoint value. This setpoint~is established such that the actual submersion dose rate would not exceed 4 mr/hr in the fuel building. This would be representative of the
)-. conditions required for this EAL.
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__._. m-_____..______m_____.m__._-.. .._.m___.._,
y EMERGENCY ACTIQEMIEdtLD.LCAllONS HASJiS fL Group 1 Abnormal Radiation Events For Indicator 2:
g.ontainment Dose Rate ,
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The Whole Body Conversion Factor for this calculation is an EPA 400 conversion factor from EIP-ZZ-01211, PC Based plume Phase Dose
- l. Assessment. The Tech Spec concentration of 5 E-3 pCi/cc will result in '
the following dose rate to personnel inside containment:
D/R = (CONC) (WBCF) 5 x 10~3 /41 3.04 x 10 -2 mp , g 34 1 cc cc Itr -
/41 j\10-6 m) 3}
,t 152 mR j ._ D/R =
Hr
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This corresponds well to the Tech Spec basis statement'that.the equivalent dose rate is "approximately 150 mR/Iir." Therefore, a. dose rate on SD-41 of 150 mR/Hr would be an indication for declaration of ars Alert (currently set to Alarm at 100 mR/Hr to ' indicate a IIigh Radiation Area).
lD / R = 150 mR / lir ( ARM SD - 41)l
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For Indicator 3:
Fuel Building Done Rates The Whole Body Conversion Factor for this calculation is an EPA 400 conversion factor from EIP-ZZ-01211, PC Based Plume Phase Dose
}. Assessment. A concentration of 1.46 E-3, the Ili-Hi alarm setpoint on GT-RE-27/28 will result in the following dose rate to personnel inside the Fuel Duilding. .,
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1.46 E-3 x 10'2/41 '4.85 x 10-2 mR -
m ,(10-6
'f 1 ccm)
D/R = 3 cc /Ci
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, ,t D/R = l Hr Therefore, a dose rate of 70 mR/Hr on SD-37 or -38 would be an indication for declaration of an Alert (Alarm setpoint is 15 mR/llr per Tech Spec Table 3.3-6, .1 based on criticality monitoring). )
lD / R = 70 mR / Ilr ( ARM SD - 37 or - 3 0 )l U' Indicator 4, eliminates the need.for Spent Fuel Pool & Refueling Pool level indication, as at Callaway indication is not capable of displaying j level as low as the top of a fuel assembly,
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I EMERGENCY ACTION LEVEL INDICATIONS BASES
.O l' Group 1 Abnormal Radiation Events Initiatina condition Emeroency Classification
} G.* -Release of Rad Material, Alert j or an Increase in Rad L Level that Either Impedes
). Safe Operations or the Ability.
to Establish or Maintain Cold shutdown.
MODES: At All Times Indications
) Any of the following.
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- 1. Valid (confirmed by HP) reading on SD-33 (Control Room) >15 mr/hr. f l \
Valid (confirmed by HP) reading on the following j l: 2.
} Safe Shutdown Area ARMS SD-26 PC Changeout Area ')
SD-23 RHR Hx Area Corr. l SD-15 Door to HPl~A Area SD-16 Fire Brigace Locker Area
>1000 times normal (normal levels can be
} considered as the monitor reading prior to the noticed increase).
- This Initiating Condition is not meant to apply to anticipated temporary increases due to planned events (e.g., incore detector movement, radwaste container .{
L movement, depleted resin transfers, upper internal 'l movements, etc.) j Banes valid means that a radiation monitor reading has been confirmed by the operators to be correct.
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 degradation of
)- the level of safety of the' plant. The cause and/or' magnitude of the
. increase in radiation levels is not a concern of this IC. The Emergency coordinator must consider the source or cause of the increased radiation levels and determine if.any other IC may be involved. For example, a dose rate of 15 mR/hr in the control room may be a problem in.itself~.
However, the increase may also be indicative of high dose rates'in the <
containment due to a LOCA. In this latter case, a Site Emergency or General Emergency may be indicated by the fission product barrier matrix ICs.
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b EMERGENCY ACTION LEVEL INDICATIONS BASES j:
Group 1 Abnormal Radiation Events Areas requiring continuous occupancy include the control room. The value of 15 mR/hr is derived from the GDC 19 value of'5 rem in 30 days with adjustment.for expected occupancy times. Although Section III.D.3 of NUREG-0737, " Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is I. used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert.
For Indicator 2, 1000 times normal represents the factor used in the Unusual Event, however these particular monitors are located in areas of required infrequent access to maintain plant safety functions.
This IC is not intended to apply to anticipated temporary increases due
): to planned events (e.g., incore detector movement, radwaste container movement, depleted resin transfers, etc.)
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p j l MERGENCY ACTION LEVEL INDICATIONS BASES i
Group 2 Fission Product Barriers l CONTAINMENT BARRIER EALs l
The Containment Barrier includes the containment building, its connections up to and including the outermost containment isolation p valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and i
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including the outermost secondary side isolation valve.
l Loss Indicators f- 1. Containment Presnure D Rapid unexplained loss of pressure (i.e., not attributab.e to containment spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity.
Containment pressure and sump levels should increase as a result of the mass and e.:ergy release into containment from a LOCA. Thus, sump level or pressure not increasing indicates containment bypass and a loss of containment integrity. ;
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- 2. Containment Isolation Valve Status
'?his EAL is intended to address incomplete containment isolation that allows direct release to the environment. It represents a ;
loss of the containment barrier. It is not intended to address .j y failures during testing.
- 3. SG Release With Primary To Secondary Leakage This EAL addresses SG tube ruptures. Secondary side releases to atmosphere include those from the atmospheric steam dump valves, and main steam safety valves. For larger breaks RCS BARRIER SG
. Tube Rupture " Loss" or " Potential Loss" EALs would result in an 1 Alert. For SG tube ruptures which may involve multiple steam l generators or unisolable secondary line breaks, this EAL would exist in conjunction with RCS BARRIER " Loss" EAL 2 and would result in a Site Area Emergency. Escalation to General Emergency would be based on the addition of a " Loss" or " Potential Lons a of i the FUEL CLAD BARRIER. I J
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k EMERQILN_CX ACTION LEVEL INDICATIONS BAS _QS
.p Group 2 Fission Product Barriers CONTAINMENT BARRIER EALs (cont) :
Potential Loss Indicators
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- 4. Critical Safety Function Status RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment.
Conditions leading to a containment RED path result-from FC.
p barrier and/or Fuel Clad Barrier Loss, Thus,.this EAL is pctmarily a discriminator between Site Emergency and General' Emergency representing a potential loss of the third barrier.
- 5. Containment Pres'.2g The second potential l'.ss EAL represents a potential loas of j containment in that the containment heat removal /depressurization system (e.g., containment sprays, but not including containment venting strategies) are either lost or performing in a degraded manner, as indicated by containment pressure greater than the setpoint at which the equipment tas suppose to have actuated.
- 6. Significant Radioactiv d Dyentory in Ctri The (>l5,000 R/hr) reading is a value which indicates significant fuel damage well in excess of the EALs associated with both loss of Fuel Clad and loss of RCS Barriers. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. Regardless of whether containment is challenged,
)-- this amount of activity in containment, if released, could have such severe consequences.that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. NUREG-1228, " Source Estimations During Incident Response to Severe Nuclear Power Plant _ Accidents,"
indicates that such conditions do not exist when'the amount of '
clad damage is less than 20%. The radiation monitor reading i h corresponding to 20% fuel clad damage was calculated using the '
Westinghouse Owners Group (WOG) " Post Accident Core Damage Assessment Methodology" dated November.1984.. This document was- ,
approved by the NRC for core damage assessment. Based upon a ,
Containment.High Range Area Radiation Monitor (CHARM) ' reading, a .i; percent clad damage (equivalent to percent noble gas release) .can be estimated. Westinghouse makes the assumption thatLany percent. 9
). noble gas release requires an equal percent clad damage, j Conversely, a Radiation Monitor reading can be produced given the ;
percent clad damage. l
-1 An example of the rel*.tionship of the exposure rate of a detector' as a function of time following reactor shutdown is presented in Figure 3-3. The exposure rates are expressed in units of R/hr-MWt.
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EMERGENCY ACTION LEVEL INDICATIONS 13 ASJ5 Group.2 Fission Product Barriers
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CONTAINMENT BARRIER EALs (cont):
Radiation Monitor Reading (R/hr) x CTMT Voluma (ft 3)
R/hr - MWt =
Plant Power (MWt) x 2x206(ft3) where:
) R/hr - MWt = 5.5 from Figure 3-3 for a 20% noble gas release equivalent to 20% clad failure.
CTMT Volume = 2.5x106 ft3 Plant Power = 3565 MWt
)- Solving for Radiation Monitor Reading:
5.5 (3565 MWt) (2x10 6 f3t )
CHARM Reading =
2.5x106 ft3
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= 15686 R/hr
- 7. Core Exit Thermocouples In this EAL, the function restoration' procedures are those
) emergency operating procedures that address the recovery of the core cooling critical safety functions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing.
The conditions in this potential loss EAL represent imminent melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. In conjunction
} with the core exit thermocouple EALs, RCS E'.RRIER indicator 1. and FUEL CLAD BARRIER indicator 1., this EAL would result in the -
declaration of a General Emergency -- loss of two barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no " success" path.
Several accident analyses ( e .. g . , NUREG--1150) have concluded that
}'- function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or'not the procedures will be effective should be apparent within-15 minutes.-The Emergency
) Coordinator should make the declaration as soon as it is i i
determined that the procedures have been, or will be ineffective.
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e EMERGENCY ACTION LEVEL INDICATIONS BASES Group 2 Fission Product Barriers O
The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other GP connections up to and including the primary isolation valves.
Loss Indicators
- 1. RCS Leak Rate The " Loss" EAL addresser conditions where leakage from the RCS is gp greater than available inventory control capacity such that a locs of subcooling has occurred. The loss of subcooling is the fundamental indication that the inventory control systems are inadequate in maintaining RCS pressure and inventory against the mass loss through the leak. Safety injection initiated indicates all available inventory control capacity is in service, gg 2. SG Tube Rupture This EAL is intended to addrecs the full spectrum of Steam Generator (SG) tube rupture eventa in conjunction with Containment Barrier " Loss" EAL 3 and Fuel Clad Barrier EALs. The " Loos" EAL addresses ruptured SG(s) with an unisolable Secondary Lin ? Break corresponding to the loss of 2 of 3 fission product barrie-s (RCS Barrier and Containment Barrier, this EAL will always resul" in
() Containment Barrier " Loss" EAL 3). This allows the direct release of radioactive firsion and activation products to the environment.
Resultant offsite dose rates are a function of many variables.
Examples include: Coolant Activity, Actual Leak Rate, SG Carry over, Iodine Partitioning, and Meteorology. Therefore, dose assessment in accordance with EAL 1B., " Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous
) Radioactivity that Exceeds 1000 mr Whole Body or 5000 mr Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology", is required when there is indication that the fuel matrix / clad is potentially lost.
Indications are consistent with the diagnostic activities of the Emergency Operating Procedures (EOPs). This includes indication of
,b) S/G level increasing uncontrollably, increased secondary radiation levels, and an uncontrolled or conolete depressurization of the ruptured SG. Secondary radiation it. creases are observed via radiation monitoring of Condenser Air Ejector Discharge, SG Blowdown, and SG Sampling System. Determination of the
" uncontrolled" depressurization of the ruptured SG should be based
[, on indication that the pressure decrease in the ruptured steam l J generator is not a function of operator action. This should l prevent declaration based on a depressurization that results from an EOP induced cooldown of the RCS that does not involve the prolonged release of contaminated secondary coolant from the .
affected SG to the environment. This EAL includes unisolable steam l breaks, feed br*.aks, and stuck open safety or relief valves.
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. - - . .. _ . . . - . ~. -- . _ . . . - ~ ~ - _ - . . - . . .
EhiliBOENCY ACTION LEVEL INDICATIONS BASES y Group 2 Fission Product Barriers RCS BARRIER EALs (cont) :
- 3. Containment Radiation Monitorinq
) The (>l E+3 R/hr) reading is a value which indicates the release of reactor coolant to the containment. The reading was calculated assuming the instantaneous release and dispersal of the reactor-coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within T/S) into the containment atmosphere. This reading was calculated using the Westinghouse Owners Group (WOG) " Post Accident Core Damage Assessment
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Methodology" dated November 1984. This document was approved by the NRC for core damago assessment. Based upon a Containment High Range Area Radiation Monitor.(CHARM) reading a percent clad damage (equivalent to percent noble gas release) can be estimated.
Westinghouse makes the assumption that any percent noble gas release requires an equal percent clad damage. Conversely, a Radiation Monitor reading can be produced given the percent clad
), damage.
An example of the relationship of the exposure rate of a detector as a function of time following reactor shutdown is presented in Figure 3-1. The exposure rates are expressed in units of R/hr-MWt.
) Radiation Monitor Reading (R/hr) x CTMT Volume (ft 3)
R/hr - MWt =
Plant Power (MWt) x 2x106(ft3 )
where
). R/hr - MWt = .35 from Figure 3-3 for a 3% noble gas.
release approximately equivalent to our Tech Spec activity limits of 1 Ci/gm DEI-131.
CTMT Volume = 2.5x106 ft3
) Plant Power = 3565 MWt ,
Solving for Radiation Monitor Reading:
.35 (3565 MWt) (2x106 f t3)
CHARM Reading =
} 2.5x106 ft3
= 998 R/hr ;
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- y. This reading will be less than that specified_for Fuel, Clad Barrier EAL #3. Thus, this EAL would be indicative of a RCS leak only.
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9 EMERGENCY ACTION LEVEL INDICATIONS B ASES Group 2 Fission Product Barriers O
Potential Loss Indicators
() 4. Critical Safety Function Status RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings, and these CSFs indicate a potential loss of RCS barrier.
- 5. FCS Leak Rate b The " Potential Loss" EAL is based on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered as any one of three centrifugal charging pumps discharging to the charging header. In conjunction with the SG Tube Rupture " Potential Loss" EAL this assures that any event that results in significant RCS inventory shrinkage or loss (e.g.,
[] events leading to reactor trip and ECCS actuation) will result in no lower than an " Alert" emergency classification. The 50 gpm indicator is based on 1 CCP in service with a 75 gpm letdown orifice in service.
- 6. SG Tube Rupture O
The " Potential Loss" indications are consistent with the diagnostic activities of the Emergency Operating Procedures with indications based on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume control System. This is considered as any on2 of three centrifugal charging pumps discharging to the charging header. In conjunction with the RCS Leak Rate " Potential Loss" EAL this assures that any event that results in significant RCS inventory shrinkage or loss (e.g.,
events leading to reactor trip and ECCS actuation) will result in no lower than an " Alert" emergency classification.
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l b Y EMERGENCY ACTION LEVEL INDICATIONS BASES s( !
Group 2 Fission Product Barriers l
}
FUEL CLAD BARRIER EALs The Fuel Clad Barrier is the zircalloy tubes that contains the fuel pellets.
Loss Indicators
- 1. Critical Safety Function Status RED path indicates an extreme challenge to the safety function.
ORANGE path indicates a severe challenge to the safety function.
Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.
A separate core exit TC value is not used as a loss indicator, as a 1200* TC value is a red path for core cooling' and would be a
) redundant indication.
- 2. Primary Coolant Activity Level Assessment by the NUMARC EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine. spikes and corresponds to about 2% to 5% fuel clad damage. This amount of
} clad damage indicates significant clad heating and thus the Fuel Clad Barrier is considered lost.
- 3. Containment Radiation Monitoring The 73000 R/hr reading is a value which indicates the release of reactor coolant, with elevated activity' indicative of fuel damage, into the containment. The reading was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 .
Ci/gm dose equivalent I-131 into the containment' atmosphere. ,;
Reactor coolant concentrations of'this magnitude are several' times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore '
indicative of fuel damage (approximately 2 -- 5% clad failure
} depending on core inventory and RCS volume). This value was calculated using the Westinghouse Owners Group (WOG) " Post- .
Accident Core Damage Assessment Methodology" dated November 1984. i
'This document was approved by the NRC for core damage assessment. -I Based'upon a Containment High Range Area Radiation Monitor (CHARM) reading a percent clad damage (equivalent to percent noble gas .
release) can be estimated. Westinghouse makes the assumption that
): any percent noble gas release requires an equal percent clad damage. Conversely, a Radiation Monitor reading can be produced 1 given the percent clad damage.
An example of-the relationship of the exposure rate of a detector as a function of time following reactor shutdown is presented in Figure 3-3. The exposure rates are expressed in units of R/hr-MWt.
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EMERGENCY ACTION LEVEL INDICATIONS BASES -'
Group 2 Fission Product Barriers
[ ,
i FUEL CLAD BARRIER EALs (cont):
Radiation Monitor Reading (R/hr) x CTMT Volume (ft )'3 ;
Plant Power (MWt) x 2x106(ft3 )
c where:
R/hr - MWt = 1.1 from Figure 3-3 for a 5% noble gas release equivalent to 5% clad failure. >
CTMT Volume = 2.5x106 ft3 Plant Power = 3565 MWt Solving for Radiation Monitor Reading:
- 1. l (3 56 5 MWt) (2x106 f t3)
CHARM Reading =
2.5x106 ft3
) = 3137 R/hr Conservatively we use 3000 R/hr. '
Eotential Loss Indicators
> RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to the safety function. ,
- 4. Critical Safety Function Status Core Cooling - ORANGE indicates subcooling has been lost and that ,
some clad damage may occur. Heat Sink - RED indicates the ultimate-
) heat sink function is.under extreme challenge and thus-these two items indicate potential loss of the Fuel Clad Barrier. .
A separate core exit TC value is not used as a potential, loss indicator, as a 700" TC value is an orange path for core cooling ;
and would be a redundant indication.
) 5. Core Exit Thermocouples ,
The 700* corresponds to a loss of subcooling that will require at least a Core Cooling " ORANGE path". ,
- 6. Reactor Vessel Water Level I- This level ls approximately at the top of the active fuel and corresponds to the Core Cooling " ORANGE path" values
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l EMERGENCY ACTION LEVELJUDICATIONS BASES l
Group 3 Hazards Affecting Plant Safety Initiatino Condition Emeroency Classification A. Confirmed Security Event Unusual Event Which Indicates'a Potential .
Degradation in the Level of'
]) Safety of the Plant.
MODES: At All Times Indications
,_ Any of the following:
1 1. Bomb device discovered within the plant Protected Area and outside the following Safe Shutdown Areas:
- a. Area 5
- b. Containment
- c. Aux Feed Pump Rooms
] d. Aux Building
- e. Diesel Cenerator Building
- f. UHS Cooling Tower
- g. ESW Pumphouse
- h. Control Building
- i. RWST
- j. Fuel Building
}
- 2. Confirmed report from the Shift Security Supervisor of an attempted entry, sabotage or security threat that cannot be properly compensated for within 10 minutes.
J Basen:
The 10 minute criteria to compensate is derived from 10 CFR 73.71,-
Reporting Of Physical Security Events.
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i PMIRGENCY ACTION LEVEL INDlC_ATIONS BASES Group 3 Hazards Affecting Plant Safety
) !
Initiatina Condition Emeroency Classificatiori B. Security Event in.the Alert Plant Protected Area.
) MODES: At All Times Indicators Confirmed report by the Shift Security supervisor of an intrusion by a hostile force into the plant Protected Area.
l' . j Danest This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event.
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l EMERGENCY ACTION _1 EVEL INDICATIONS DASEE
-i.
Group 3 Hazards Affecting Plant Safety
)'
Initiatino Condition Emercqncy Classification C. Security Event in a Site Emergency Safe Shutdown Area.
} MODES: At All Times Indications Any_of the following
- 1. Bomb device discovered within any of the following areas:
)
- a. Area 5
- b. Containment
- c. Aux Feed Pump Rooms
- d. Aux Building
- e. Diesel Generator Building
- f. UHS Cooling Tower
) g. ESW Pumphouse
- h. Control Building
- i. RWST
- j. Fuel Building
- 2. Confirmed report from the Shift Security Supervisor of an intrusion by a hostile force
) into any of the following areas:
- a. Area 5
- b. Containment
- c. Aux Feed Pump Rooms
- d. Aux Building
- e. Diesel Generator Building
) f. UHS Cooling Tower
- g. ESW Pumphouse
- h. Control Building
- 1. RWST
- j. Fuel Building
)
Bases:
This class of security events represents _an escalated threat to plant safety.above that contained in-the Alert IC in that a hostile force has progressed from the Protected Area to a Safe Shutdown Area. These areas contain' Safe Shutdown Systems as defined per the FSAR Appendix - 5.4 (A) .
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3 EllEXENCY ACTION LEVEL INDICATIONS BASES v
Group 3 Hazardo Affecting Plant Safety Jnitiatina Condition Emeroency Classification D. Security Event Resulting General Emergency ~'
in a Loss of the Ability ,
to Reach and Maintain Cold
]- Shutdown, MODES: At All Times Indications Any of the following:
) 1. Occupation of the Control Room by a hostile force,
- 2. Occupation of the Aux Shutdown Panel by a hostile force.
). 1 Bases:
This IC encompasses conditions under which a hostile force has taken physical control of Safe Shutdown areas required to reach and maintain '
)
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O EMERGENCY ACTION I.EVEL INDICATIONS BASES Group 3 Hazards Affecting Plant Safety Initiating Condition Emergency Classification E. Fire Within Protected Unusual Event Area Boundary Not Extinguished within Gb 15 Minutes of Verification MODES: At All Times Indications
- 1. Fire in or adiacent to any of the following:
D a. Area 5
- b. Containment
- c. Aux Feed Pump Rooms
- d. Aux Building
- e. Diesel Generator Building
- f. UHS Cooling Tower I) 9 ESW Pumphouse
- h. Control Building
- i. RWST
- j. Fuel Building illld
- 2. Not extinguished within 15 minutes of control room
] verification of a fire. I panen:
q The purpose of this IC is to address the magnitude and extent of fires
~'
that may be potentially significant precursors to damage to safety systems. This excludes such times as fires within administration buildings, waste-basketr fires, and other small fires of no safety consequence. This IC applies to buildings and areas adjacent to Safe Shutdown areas or other significant buildings or areas. The intent of this IC is not to include buildings (i.e., warehouses) or areas that are not immediately adjacent to Safe Shutdown areas. These areas contain D Safe Shutdown Systems as defined per the FSAR Appendix 5.4 (A)
Verification of the alarm in this context means those actions taken in the control room to determine that the control room alarm is not sput. sus.
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EMERGrib'CY ACTION LEVEL INDICATIONS BASES V
Group 3 Hazards Affecting Plant Safety
.Initiatino Condition Emergency Classification F. Fire Affecting the Alert Operability of Plant Safety Systems Required
[I to Establish'or Maintain Safe Shutdown.
MODES: At All Times q 1
Jndications
)- 1. Fire in any of the following areas:
- a. Area 5
- b. Containment
- c. Aux Feed Pump Rooms
- d. Aux Building
- e. Diesel Generator Building
- f. UHS Cooling Tower
]) g. ESW Pumphouse
- h. Control Building
- 1. RWST
-J . Fuel Building and
) 2. There is visible damage to permanent structures or equipment, affecting the operability of safety related equipment.
) Bases:
Areas containing functions and systems required for the safe shutdown of the plant are specified per FSAR Appendix 5.4 (A) .
The inclusion of a " report of visible damage" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL 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.
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I EMERGENCY ACTION LEVEL INDICATIONS BASES a
Group 3 Hazards Affecting Plant Safety Initiatino Condition Emergency Jlassification G. Natural and Destructive Unusual Event Phenomena Affecting the
.. Protected Area.
)'."
MODES: At All Times i
Indications .j
&ny of the following ,
l
); 1. a. Response spectrum recorder. operating annunciator 98E alarms in the Control Room and
- b. Verified to be a real event per OTO-SG-00001.
- 2. Report of a turbine rotating component failure resulting in casing penetration or major damage
) to seals causing a rapid loss of lubricating oil or hydrogen.
- 3. Explosion, vehicle crash or tornado in or adiacent to any of the following:
- a. Area 5
- b. Containment
} c. Aux Feed Pump Rooms
- d. Aux Building
- e. Diesel Generator Building
- f. UHS Cooling Tower
- g. ESW Pumphouse
- h. Control Building
- i. RWST
) ~. j. Fuel Building Bases:
The Protected Area Boundary is defined in the site security plan.
-In'dicator 1 was developed on a site-specific basis. Damage may be caused -
to some portions of the site, but should not affect-ability of' safety functions to operate. Method of detection is response validated per OTO-SG-00001. As defined in the EPRI-sponsored " Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a'" felt earthquake" is:
An earthquake of sufficient intensity such that (a) the vibratory ground motion 15 felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the. time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. For most plants with 4
)._ seismic instrumentation,.the seismic switches are set at an !
acceleration of about 0.01g.
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EMERGENCY _6CTION LEVEL INDICATIONS BASES Group 3 Hazards Affecting Plant Safety Indicator 2 is intended to address main turbine rotating component failures of significant magnitude to cause observable damage to the turbir.e casing or to the seals of the turbine generator. Of major concern is the potential for rapid loss of combustible fluids
)) (lubricating. oils) and gases (hydrogen cooling) to the plant environs.
This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk'to non-safety related equipment.
In indicator 3 only those events in or. adjacent to any area containing Safe Shutdown Systems, should be considered. As used here, an explosion
). is a rapid, violent, unconfined' combustion, or a catastrophic failure of-pressurized equipment, that potentially imparts significant energy to near-by structures and materials. No attempt is.made in this EAL to assess the actual magnitude of the damage. The occurrence-of the explosion with reports of evidence of damage (e .g. , deformation, j scorching) is sufficient for declaration.
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IlhlERGENCY ACTION LEVEL INDICATIONS DASES L
Group 3 Hazards Affecting Plant Safety Initiatino Condition Emeroency Classification H. . Natural and Destructive Alert Phenomena Affecting a safe Shutdown Area.
) MODES: At All Times .
Indicationq h of the following:
j' 1. a. Operating basis earthquake annunciator 98D alarms in the Control Room and
- b. Earthquake greater than OBE levels (0.129)in the horizontal and vertical directions as indicated by LIGHT "OSG-AE-1"pr LIGHT "OSG-AE-2"
) 2. a. Report of L tornado, high wind, vehicle crash, explosion, or other natural or destructive phenomena to any of the following Safe Shutdown areas:
- 1. Area 5
- 2. Containment
} 3. Aux Feed Pump Rooms 4 Aux Building .
- 5. Diesel Generator Building
- 6. UHS Cooling Tower
- 7. ESW Pumphouse
- 8. Control Building
- 9. RWST
- 10. Fuel Building and
- b. There is visible damage to permanent structures or equipment, affecting plant operations.
) Bases:
Indicator 1 is based on FSAR design basis. Seismic events of this !
magnitude can cause damage to safety functions.
Indicator 2 specifies areas containing systems and functimos required-for safe shutdown of the plant per FSAR Appendix 5.4 (A) . This indicator
). :is intended to. address such items as plane or helicopter crash.into a- ;
plant vital area.
Each of these EALs is intended to address events that may have resulted 1 in.a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.
The initial " report" should not be interpreted as mandating a lengthy i damage assessment magnitude of the damage. The declaration of an Alert l
). and the activation of the TSC will provide the Emergency Coordinator with the resources needed to perform these damage assessments, l I
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ISTERGENCY ACTION LEVEklNDICATIONS. BASES Group 3 Hazards Affecting Plant Safety -
Initiating Condition Emeroency Classification I. Release of Toxic or Unusual Event Flammable Gases Deemed Detrimental to Safe
)) Operation of the Plant.
MODES: At All Times Indications Any of the following:
) 1. Report or detection of toxic or flammable gases that enter within the Exclusion Area Boundary, that have created a RAZARDOUS ATMOSPHERE per CTP-ZZ-01200.
- 2. Confirmed report by local, County or State Officials of potential evacuation of site
)- personnel as determined from the DOT evacuation tables for selected hazardous materials in the DOT Emergency Response Guide for Hazardous Materials.
Y Banes:
This IC is based on releases in concentrations within the site boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an offsite event (i.e., tanker truck accident releasing toxic gases, y etc.) The evacuation area is as determined from the DOT Evacuation Tables for Selected Hazardous Materials, in the DOT Emergency Response Guide for Hazardous Materials.
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liMERGENCy_ ACTION LEVEL INplCAllONS B ASFJ '
Group 3 Hazards Affecting Plant safety Initiating ConditioD Emergency Clansification J. Release of Toxic or Alert Flammable Gases Within a Facility Structure Which
) Jeopardizes Operation of Systems Required to Establish or Maintain Cold Shutdown.
I MODES: At All Times
- Indications
) Any of the following:
- 1. Report or detection of a toxic or flammable gases, not properly contained, within or adiacent to any of the following Safe Shutdown Areas, that have created a HAZARDOUS ATMOSPHERE per
) CTP-ZZ-01200.
- a. Area 5
- b. Containment
- c. Aux Feed Pump Rooms
- d. Aux Building
- e. Diesel Generator Building
- f. UHS Cooling Tower
): g. ESW Pumphouse
- h. Control Building
- i. RWST
- j. Fuel Building Bases:
This IC is based on gases that have entered a plant structure affecting the safe operation'of the plant. This IC applies to Safe Shutdown Areas.
The intent of this IC is not to include buildings (i.e., warehouses) . or other areas that are not immediately adjacent to Safe Shutdown Areas. It is appropriate that increased monitoring be done to ascertain whether
)- consequential damage has occurred.
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D' D1ERGENCY ACI'lONIJLYELINDICATIONS BASES
.I.
Group 3 Hazards'Affecting Plant Safety Initiatina Condition Emernency Classification K. Control Room Evacuation Alert Has Been Initiated.
) MODES: At All Times Indications Entry into OTO-ZZ-00001, Control Room evacuation is required.
Y Bases:
With the control room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other Emergency Operations Center is necessary.
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EMERGENCY ACTION LEVEL INDICATIONS DASEJ Group 3 Hazards Affecting Plant Safety Ipitiating Condition Emercency Classification L. Control Room Evacuation Site Emergency Has Been Initiated and Plant Control Cannot Be
).- Established.
MODES: At All Times Indications
- 1. Ent'ry into OTO-ZZ-00001, Control Room evacuation
) is required.
and
- 2. Control of the Aux Feed System and a SG PORV for cooldown cannot be established within 15 minutes. '
)
Bades:
Expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated. The time for transfer is based on how quickly control must be reestablished without core uncovering and/or core damage. In cold shutdown and refueling modes,
}- operator concern is directed toward maintaining core cooling such as is discussed in Generic Letter 88-17, " Loss of Decay Heat Removal." In power operation, hot standby, and hot shutdown modes, operator concern is primarily directed toward maintaining critical safety functions and thereby assuring fission product barrier integrity. 7 J ;The 15 minutes is consistent with Westinghouse Response Plan for Immediate Evacuation of the Control Room Time Study. " Plant cooldown established" per OTO-ZZ-00001 would require Aux feed to be initiated and control of SG Power Operated P311ef valves and the Aux feed' pumps to be established from the Aux shutdown panel. .
04/20/94 34
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EMERGENCY ACTION LEVEL INDICATIONS BASES i
Group 4 System Malfunctions Initiatina Condition Emercency Classification A. Unplanned Loss of Most' Unusual Event or-All Alarms (Annunciators) for Greater Than 15 Minutes.
MODES: 1-4 Indications
- 1. Either of the following:
- a. 3 of 4 field power supplies have failed for
)' greater than 15 minutes (loss of all-annunciators) and not a result of planned action, b, All thirteen logic power supplies have failed' for greater than 15 minutes (loss of all annunciators) and not a result of planned action.
) 9I All of the following;
- c. Any combination of power supplies (including Optical Isolat.xs) have failed for greater than 15 minutes,
- d. Any faile? ;ower supply's minimum compensatory aggiors, per OTO-RK-00601, cannot be maintaited.
)E e. The loan does not result from planned action, gases:
This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without-the use of a major portion of the annunciation or indication equipment, since the system is operating at just over 3 amps, and each power supply.
le rated for a maxin.am 3 ' amps, soon after losing the third power supply.
the fourth will fail due to overcurrent.
) Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Due to the limited number of safety systems in operation during cold shutdown, refueling,.and defueled modes, no IC is indicated during these modes of operation.
)' This Unusua'l Event will be escalated to an Alert if.a transient is in progress during the loss of annunciation or indication.
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b EMERGENCY ACTIOS LEVELjNDICATIONS BASES i
Group 4 Dystem Malfunctions Initiatino ConditiQD Emeroency Classification B. Unplanned Loss of All Alert Annunciators With Either a Transient In Progress, or the
)f Plant Computer is Unavailable.
MODES: 1-/
Indications
- 1. Either of the following:
- a. 3 of 4 field power supplies have failed for greater than 15 minutes (loss of all annunciators) and not a result of planned action,
- b. All thirteen logic power supplies have failed for greater than 15 minutes (loss of all annunciators) and not a result of planned
}- action. s All of the following:
- c. Any combination of power supplies (including Optical Isolators) have failed for greater than 15 minutes.
- d. Any failed power supply's minimum compensatory-actions, per OTO-RK-00001, cannot be
} maintained.
- e. The loss does not result from planned action. i and
- 2. Any of the following:
- a. A change in reactor power greater than 10%.
- b. Safety injection initiation.
- c. The plant computer is unavailable.
)
Bases:
This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use'of
) --
- a. major portion of the annunciation or indication equipment during a .
transient. Recognition of the availability of computer based. indication equipment is considered SPDS, plant computer, etc.)
Since the system is operating at just over 3 amps, and each power supply is rated for a maximum 3 amps, soon after losing the third power supply the fourth will fail due to overcurrent.
):
-" Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.
If both a-major portion of the annunciation system and all computer monitoring are unavailable to the extent that additional operating personnel are required to monitor indications, the Alert is required.
04/20/94 36
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EMERGENCY ACTION 1.EVEldNDICATIONS BASES Group 4 system Malfunctions Due to the limited number of safety systems in operation during cold shutdown, refueling and defueled modes. No IC is indicated during these modes of operation.
This Alert will be escalated to a Site Area Emergency if the operating
) Crew cannot monitor the transient in progress.
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r DiERGENCY ACDON LEVEL INDICATIONS B ASE3 Group 4 System Malfunctions Initiatino Condition Emeroency Classification C. Inability to Monitor a Site Emergency Significant Transient in Progress.
)
MODES: 1-4 Indications all of the following:
- 1. a. Either of the following
}. 1) 3 of 4 field power supplies have failed (loss of all annunciators).
- 2) All thirteen logic power supplies have failed (loss of all annunciators) .
EI
- b. Both of the following: 3
- 1) Any combination of power supplies 3 (including Optical Isolators) have failed.
- 2) Any failed power supply's minimum compensatory actions, per OTO-RK-00001, 1
cannot be maintained.
and
) 2. The plant computer is unavailable, and
- 3. .Either of the followings
- a. A. change in reactor power greater than ilo %.
- b. Safety injection initiation.
)
Bases:
This IC and its associated EAL are intended to recognize the inability of the control room staff to monitor the plant response to a transient.
) A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public.
J Since the system is operating at just over 3 amps, and each power supply 1 is rated for a maximum 3 amps, soon after losing the third power supply the fourth will fail due to overcurrent.
)[ . " Planned" actions are included in this EAL since the loss of instrumentation ofLthis magnitude is of such significance during a transient that the cause of the loss is not important.
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04/20/94 38
1 EMERQENCY ACTION LEVEL INDICATIONS BASES Group 4 Byatem Malfunctions Initiating condition Em_ercency classification D. Loss of All Offsite Power Unusual Event to Essential Busses for Greater Than 15 Minutes.
.)
MODES: 1-6 Indications i All of the following:
) 1. Loss of offsite power to NB01 and NB02,
- 2. NB01 and NB02 being supplied by NE01 ap_4 NE02.
- 3. The loss of offsite power has occurred for >15 minutes, p_a s e s :
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 complete Loss of AC Power (Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary
} power losses.
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liMERGENCY ACTION LEVEL INDICAT_[QES BASiiS Group 4 System Malfunctions
).. .
Initiatinq Conditions Emercency Clandification E. Only One AC Source to Alert Essential Busses for
>15 Minutes Such
). .
That Any Additional Single Failure Would Result In Station Blackout.
l MODES: 1-4 Indications
) 1. Losa of gjly 3 of the following power sources;
-q
- a. Offsite power to NB01
- b. Offsite power to NB02
- c. Emergency Diesel NE01
- d. Emergency Diesel NE02 iUld I'.
l' 2. The Losa of all 3 has occurred for >15 minutes.
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The condition indicated by this IC is the degradation of the off-site and on-site power ayatema such that any additional single failure would result
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in a station blackout. The subsequent loss of this single power source would catalate the event to a. Site Area Emergency after an additional 15 minutes.
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E.MERGENCY ACTION LEVELjNDICATIONS BASES Group 4 System Malfunctions Initiatino condition Emeroency Classification F, Loss of All Offsite Site Emergency Power and Loss of All Onsite AC Power to Essential Busses. <
MODES: 1-4 Indications
- 1. Loss of all 4 of the following power sources:
)
- a. Offsite power to NB01
- b. Offsite power to NB02
- c. Emergency Diesel NE01
- d. Emergency Diesel NE02
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- 2. The Loss of all 4 has occurred for >15 minutes. .)
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.J Bases
) Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal'and the Ultimate Heat Sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency.
The 15 minutes ensures the loss is other than a transient or momentary
) power loss.
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O FAfERGENCY ACTION LEVEL INDICAlLONS 11 ASES Group 4 System Malfunctiona O
Iq,1.t i a t i ng condition Emergency Classifi-ation G. Loss of All Vital DC Site Emergency Power
" E8 1-4 O
Indicationn
- 1. Loss of all vital DC power as indicated by less than 106.9 VDC on vital DC busses IIK01, IIK02, IJK03, and NK04.
DJld
- 2. Failure t.o restore power to at least one DC bus within 15 minutes.
O Bases:
Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
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A EMERGENCY AC110N LEVEL INDICATIONS B ASES i Group 4'Bystem Malfunctions 5
0 Initiatina Condition Emeroency Classification H. Prolonged Loss of All General Emergency Offsite Power and Prolonged g Loss of All Onsite AC Power.
4 MODES; 1-4 Indications All of the following:
- 1. Loss of offsite power to NB01 and NB02,
)
- 2. Loss of both Emergency Diesel Generators NE01 and NE02,
- 3. a. Restoration of at least one emergency bus within 4 bours is not likely.
or I) 1 b7 Meet the entry requirements for FRC.1, Red Path for Core Cooling, d
, Dasent lC) Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCs, Containment Heat 2emaval and the !
Ultimate Heat Sink. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore AC power is based on a site blackout coping analysis FSAR Appendix 8.3A " Station Blackout"., perfurred in conformance with 10 CFR 50.63 and Regulatory Guide 1.155, " Station alackcut".
7 This IC is specified co assure that in the unlikely event of a prolonged staticn blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a easonable assessment of the event trajectory.
The likelihood 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 mitigating the event could result in a lons of valuable time in preparing and implementing public protective-actions. In addition, under these conditions, fission product barrier ;
monitoring capability may be degraded. Alchough it may be difficult to i predict when power can be restored, it is necessary to give the ,
Emergency Coordinator a reasonable idea of how quickly (s)he may need to l
- () declare a General Emergency based on two major considerations
- 1. Are there any present indications that core cooling is already' ,
degraded to the point that Loss or 'otential Loss of Fission )
Product Barriers is IMMINENT? j
'l
- 2. If there are no present indications of such core cooling
> (y' degradation, how likely io it that power can be restored in time 4
to assure that a loss of two barriors with a potential loss of the third barrier can be prevented?
04/20/t. 43
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q EMERGENCY ACTION I EVEL INDICATIONS BASILS Group 4 System Malfunctions J
Jnitiatinq Condition Emergency Classification I. Loss of Required DC Power Unusual Event During Cold Shutdown or Rc r ueling Mode for Greater es Than 15 Minutes.
J MODES: 5,6 Indications
]) source as indicated by <106.9 VDC on:
NK01 gr NXO3 (Division 1) 9X NK02 gr NK04 (Division 2).
EEd q
Panes:
The purpose of this IC and its associated EALs is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold shutdown or Refueling operations. This EAL is 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.
Unplanned was not used in this IC am , because as written one Division of Vital DC power can be inc; able for planned maintenance activities. The loss of the remaining operable train would require an Unusual Event. In no instance would maintenance be planned on both divisions.
m' The 106.9 VDC bus voltage in Indi cator 1, is based on the minimum bus voltage necessary for the operation of safety related equipment.
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EMERQENCY ACTION LEVEL INDICATIONS BASES l Group 4 System Malfunctions Initiating Cmdition Emeroency Classification J. Loss of All Offsite Alert Power and Loss of All Onsite AC Power to Essential Busses During
}- Cold Shutdown or Refueling.
-?
MODES: 5,6, and Defueled ,
Indicationg
- 1. Loss of all 4 of the following power sources
)-
- a. Offsite power to NB01
- b. Offsite power to NB02
- c. Emergency Diesel NE01
- d. Emergency Diesel NE02 h
- 2. The loss of all 4 has occurred for >15 minutes.
t Bases,
)T Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and'the Ultimate Heat Sink. When in cold shutdown, refueling, or defueled mode-the event can be classified as an_ Alert, because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for_the Site Area Emergency EAL. Fifteen minutes was selected as a threshold to exclude transient:or momentary
). power lossen.
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04/20/94 45
)-
O EMERGENCY ACTION LEVEL INDICATIONS BASES Group 4 System Malfunctions )
Q Initiating Condition Emergency Classification K.* Inability to Perform a Unusual Event Required Shutdown Within Technical Specification ,
- ) Limits. I MODES
- 1-4 Indications _
- 1. The plant is not brought to a required operating mode within a Technical Specification LCO action ;
-)
statement time.
- It is not intended to declare an Unusual Event due to an unknown condition or failure reculting in exceeding the allowable action statement time. The allowable action statement time is always available
- from the time of discovery.
J Banna-Limiting Conditions of Operation (LCOs) require the plant to be brought
,3 to a required shutdown mode when the Technical Specification required
> configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition.
In any case, the initiation of plant shutdown required by the site Technical Specifications requires a one hour report u:uer 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate Notification of an Unusu '
[] Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications.
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O 04/20/94 46 O
o FMERGENCY ACTION LEVEL INDICATIONS BASES Group 4 System Malfunctions D
Initiating Condition Emerqency Classification L. Inability to Maintain Alert Plant in Cold Shutdown MODES: 5,6
{}
Indications
- 1. Eny of the following:
- a. Complete los:3 of both trains of RHR.
- b. Complete loss of both trains of CCW.
) c. Complete loss of both trains of ESW.
EDd
- 2. Either of the following:
- a. Greater than 200 F on any valid incore
{) thermocouple.*
- b. Uncontrolled temperature rise, with no actions available that will likely pre vent approaching 200 F on any valid incore thermocouple. *
- If a thermocouple is not available, use Wide
]) Range Hot Leg temperature indications:
- BBTI413A - Loop 1
- BBTI423A - Loop 2
- RECORDERS BBTR413 - Loop 1 BBTR423 - Loop 2 g BBTR433 - Loop 3 BBTR443 - Loop 4 Edses:
Indications 1 and 2 indicate a complete loss of Technical Specification required functions to maintain Cold Shutdown.
For PWRs, this IC and its associated EAL are based on concerns raised by G^neric Letter 88-l'l, " Loss of Decay Heat Removal." A number of phenomena such an pressurization, vortexing, steam generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay heat removal system design, and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncovery can occur. NRC analyses show that sequences that can cause core
) uncovery in 15 to 20 minutes and severe core damage within an hour after decay heat removal is lost. Under these conditions, RCS integrity is lost and fuel clad integrity is loct or potentially lost, wnich is consistent with a Site Area Emergency.
" Uncontrolled" meann that system temperature increase is not the result of clanned actions by the plant staff. The intent is to declare the
) ALERT when less than 200 F, only when temperature is increasing and it is known that there is not time to take action to stop the temp from exceeding 200 F.
04/20/94 47
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EMERGE} ICY ACTION LEVEL INDICATIONS BASES ,
Group 4 System Malfunctions D
'Initiatina Condition Emeroency Classification M. Loss of Water Level That Site Emergency Has or Will Uncover Fuel in the Reactor Vessel.
MODES: 5, 6 Indications
- 1. Any of the following:
- a. Complete loss of both trains of RHR.
. b. Complete loss of both trains of CCW.
- c. Complete loss of both trains of ESW.
and
- 2. Either of the following:
D- a Greater than 200 F on any valid incore thermocouple.*
- b. Uncontrolled temperature rise, with no actions available that will likely prevent approaching 200*F on any valid incore thermocouple. *
]) ' and
- 3. a. Water level in the reactor vessel is less than 2.0 inches on BB-LI-0053A or B.
9I
- b. RVLIS (pumps.off) <55%.
- If a thermocouple is not available, use Wide Range
]. Hot Leg temperature indications:
- BBTI413A - Loop 1
- BBTI423A - Loop 2
- RECORDERS BBTR413 - Loop 1 BBTR423 - Loop 2 BBTR433 - Loop 3
} BBTR443 - Loop 4 Bases:
Under the conditions specified by this IC, severe core damage'can occur and reactor coolant system pressure boundary integrity may not be assured.
For indicator 3.a. 2.0 inches is used as the lowest readable level on ,
the instruments within their accuracy. For indicator 3.b. RVLIS (pumps off)'is used If a Reactor Coolant pump is running, void fraction rather than core water level would have to be considered.
This IC covers sequences such as prolonged' boiling following loss of
).
i .
decay heat removal. Thus, declaration of a Site Area Emergency is warranted under the conditions.specified by the IC. Escalation to a general emergency.is via radiological effluence.
04/20/94 48
. . ~ _ - . - . .- -
EMEFGENCY
- ACTION LEVEL INDICATIONS BASES
!, Group 4 System Malfunctions
{
Initiatino Condition Emeroency' Classification N. Complete Loss of Function Site Emergency Needed to Achieve or Maintain Hot Shutdown.
MODES: 1-4 In.dications
- 1. All of the following:
- a. Failure to bring the reactor subcritical with the control rods fully inserted.
- b. Complete loss of all Boron Injection Flowpaths.
2K
- 2. All of the following:
- a. All steam generator levels <10% wide range.
- b. All steam dump valves to condenser (AB-UV-34, 35, and 36) are NOT responding to steam header pressure controller (AB-ZI-34, 35, or 36).
- c. All steam generator steam dump valves to atmosphere are NOT operating properly (AB-PIC-1A, 2A, 3A, and 4A).
)- d. Complete loss of both RHR trains. (A complete loss of ESW or CCW constitutes a complete loss of RHR.)
RI
- 3. all of the following:
) a. The Ultimate Heat Sink (UHS) is inoperable as '
a result of level or temperature.
- b. Complete loss of both UHS Cooling Tower trains.
Bases:
This EAL addresses complete loss of functions, including ultimate heat
)- sink and reactivity control, required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted.
Indication 1. a., control rods, defines the inability to shutdown the reactor normally. s
)
Indication 1.b., defines the inability to add boric acid to the RCS. A complete loss of Boron Injection is definad as a loss of the required Tech. Spec. Boron Injection flowpath(s) .
Indication 2 indicates a complete loss of Heat Sink.
) . Indication 3 indicates a complete loss of the Ultimate Heat Sink.
04/20/94 49 Y :
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b EhLEBGENCY ACTION LEVEL INDICATIONS BASES
'I Group 4 System Malfunctions
)
laitiatino Condition Emergency Classification .
O. Unplanned Loss of All Unusual Event Onsite or Offsite '
Communication Capabilities 'l MODES: 1-6 Indications l
- 1. All of the following:
- a. Complete failure of Plant *;elephone systems '
J b. Complete failure of Paging systems
- c. Complete failure of Plant radios
- d. Complete failure of Plant Emergency Dedicated Phones.
- 2. All of the followings
)
- a. Complete failure of ENS (Red Phone) line ,
- b. Complete failure of Notification and Coordination line (Blue Phone)
- c. Complete failure of Touch-tone telephone sys tem ' (EPABX)
- d. Complete failure of the Sheriff's radio
) system. y Q-gases:
The purpose of this IC and its associated EALs is to recognize a loss of
) communications capability that either defeats the plant operations. staff ability to perform routine tasks necessary for plant operations or the ,
ability to communicate problems with offsite authorities. The-loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72. ,;
Indicator 1, encompasses the total loss of all means of routine
). communications.
Indicator 2, encompasses the loss of all means of communications with offsite authorities.
This EAL is intended to be used only when extraordinary means are being i utilized to make communications possible (relaying of information from 1 jl radio transmissions, individuals being sent to offsite locations, etc.) 1 u
04/20/94- 50
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EMERCgNCY ACTION LEVEL INDICATIONS BASES
'T Group 4 System Malfunctions b
Initiatina Condition Emergency Classification l
P. Fuel Clad Degradation Unusual Event ]
1 MODES: 1-6 1
) Indications
)
- 1. Say of the following:
I
- a. >1.0 pCi/ gram Dose Equivalent I-131 for i greater than a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> continuous period.
) b. Dose Equivalent I-131 activity exceeding the limits of Tech Spec Fig. 3.4-1.
- c. >100/E har pCi/ gram of gross radioactivity.
Bas 111 This IC is included as an Unusual Event because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.
l Indications 1, 2 and'3 are Technical Specification.3.4,8 limits.
) The Alert alarm for the Chemical and Volume Control System Letdown l Monitor (Failed Fuel Monitor) 'SJ-RE-01 was not used as an indicator for high coolant activity. If the monitor alarms, our procedures require ~
sampling to confirm hi activity. Listing it as an' indicator' duplicates the other indicators.
)
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, . - . _ . ~ - - . . - - . - .. - .- .
j JIMEBQENCY ACTION LEVFALNDICATIONS BASES Group 4 System Malfunctions A -,
Initiatino Condition Emergency classification Q. RCS Leakage Unusual Event MODES: 1-4
)- Indication 2
- 1. Any of the followings
- a. Unidentified leakage greater than 10 gpm.
- b. Pressure boundary leakage greater than 10 gpm.
- c. Identified leakage greater than 25 gpm.
}
-i Bases:
This IC is included as an Unusual Event because it may be a precursor of
} more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for "
the unidentified and pressure boundary leakage was selected as it'is observable with normal control room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at-a higher
- value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.
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I EMERGENCY ACTlQN LEVEL INDICATIONS BASES ~ i
! Group 4 System Malfunctions Initiating Condition Emeroency Classification R. Failure of Reactor Alert Protection System Instrumentation to Complete or Initiate an Automatic ;
}-1 Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was Successful.
MODES: 1,2 Indications
- 1. Failure of reactor protection system instrumentation to initiate an automatic trip.
and ,
) 2. Manual reactor trip is successful using either manual trip switch, SB-HS-1 on RLOO3 9L SB-HS-42 on RLOO6.
)
Bases:
This condition indicates failure of the automatic protection system to' trip the reactor. 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 the plant safety has
) been compromised, and design limits of the fuel may have been exceeded.
An Alert is indicated because conditions exist that lead.to potential loss of fuel clad or RCS. A reactor protection system setpoint being exceeded (rather than limiting safety system setpoint being exceeded) is specified here because failure of the automatic protection system is the issue.
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w - . - , , . . - . .- - , . . .- - -
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EMERGENCY ACTION LEVEL INDICATIONS BASEj E NJ Group 4 System Malfunctions Initiatino conditi2n Emergency Classification S. Failure of Reactor Site Emergency Protection System Instrumentation to Complete or Initiate an
Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was NOT Successful.
MODES: 1, 2 Indic_ations
- 1. Failure of reactor protection system instrumentation to initiate an automatic trip.
and
}. .
- 2. Manual reactor trip is NOT successful using manual trip switches SB-US-1 on RLOO3 ,
and SB-HS-42 on RLOO6.
}
Banest Automatic and manual trip are not considered sitccessful if action away from the reactor control console was required to scram the reactor, 1 Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed.
A Site Emergency is indicated because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be viet ad as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure' timely recognition and emergency response.
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04/20/94 54 ,
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!!MERGENQ' ACTION LEVEL INDICATIONS BASES Group 4 System Malfunctions
[)
Emergency Classification I Initiating Condition T. Failure of the Reactor General Emergency Protection System to Complete an Automatic Trip and Manual Trip Was
{} NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core.
MODES: 1, 2 D Indications 611 of the following:
- 1. Failure of reactor protection system instrumentation to initiate an automatic trip.
) 2. Manual reactor trip is NOT successful using manual trip switches SB-HS-1 on RLOO3 iUld SB-HS-42 on RLOOG.
- 3. Meet the entry requirements for FRC.1 or FRH.1, red path summaries for core cooling and heat 3 sink.
Bases:
Automatic and manual trip are not considered successful if action away
[) from the reactor control console is required to scram the reactor.
Under the conditions of this IC and its associated EALs, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed. Although there are capabilities away from the reactor control console, such as emergency boration, the
[) continuing temperature rise indicates that these capabilities are not effective. This situation could be a precursor for a core melt sequence.
The entry requirements for FRC.1 indicate an extreme challenge to the ability to cool the core. The entry requirements for FRH.1 indicate a extreme challenge to the ability to initially remove heat during the
[] early stages of thiu sequence.
In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design (typically 3 to Si power) 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 g fission product barrier matrix declaration to permit maximum off-site intervention time.
04/20/94 55 D
O RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION t- 1. General- No Emergency Coordinator Judgment EAL g The Callaway EAL scheme did not include EALs corresponding to the NUMARC/NESP.007 EALs for classification of events upon the Emergency Coordinator's ,
judgment. The Callaway emergency classification procedure does contain a step which allows the Emergency Coordinator to classify an event based upon his/herjudgment.
Ilowever specific EALs containing guidance as to threshold for the plant conditions and the Potential for radiological releases, upon which this judgment would be based, were not
'O~ specified.
Providejustification for this deviation from the NUMARC/NESP-007 guidance.
Response - Specific guidance as to what thisjudgment is to be based upon has been added
-O to EIP-ZZ-00/01 (enc.5). 7his guidance provides all of the information provided in NUMARC/NESP-007, in thisformat, we won'tprovide EALs with unclear indicators, requiringjudgment, and causing inconsistent classification. We will, however, provide the ,
information necessaryfor the Emergency Coordinator to classify an event based upon his'herjudgment.
O
- 2. EAL 1 A - Any Unplanned Release of Radioactivity to the Environment that Exceeds 2 Times the Radiological Effluent Control Limits in the ODCM (APA-ZZ-01003) for >
60 minutes O The following two NUMARC/NESP-007 EALs were not included in the Callaway EAL scheme.
AUI-3 Valid reading on perimeter radiation monitoring system greater than 0.10 -
mr/hr above normal background for 60 minutes [for sites having telemetered perimeter monitors).
O AUl-4 Valid indication on automatic real-time dose assessment Japability greater than (site-specific value) for 60 minutes or longer [for sites having such capability].
Provide justification for this deviation. If these EALs were not included because these O specific indications are not available at Callaway, you should consider whether other indications are available to determine whether the initiating condition (IC) should be entered.
Response - For EALs I and 2, Callaway does not have a telemeteredperimeter monitor or any type of automatic real-time dose assessmeni capability.
O When any plant emergency is declared, we enter EIP-ZZ-00102, EMERGENCY IMPLEMENTING ACTIONS. Jhisprocedure specifically directs that manualdose assessment be initiated.
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O RESPONSES TO TILE REQUEST FOR ADDITIONAL INFORMATION g 3. EAL 1B - Any Unplanned Itelease ofItadioactivity to the Environment that Exceeds 200 Times the Itadiological Effluent Control Limits in the ODCM (APA-ZZ-01003) for > 15 minutes The following two NUMARC/NESP-007 EALs were not included in the Callaway EAL scheme.
D AAl-3 Valid reading on perimeter radiation monitoring system greater than 10 mr/hr above normal background for 15 minutes [for sites having telemetered perimeter monitors].
AAl-t Valid indication on automatic real-time dose assessment capability greater D than (site-specific value) for 15 minutes or longer [for sites having such capability l, Providejustification for this deviation. If these EALs were not included because these specific indications are not available at Callaway, you should consider whether other indications are available to determine whether the IC should be entered.
D liespanse - For KAls I and 2, Callau av does not hare a telemeteredperimeter monitor or any t)pe of automatic real-time dose assessment capability.
When any plant emergency is declared, u e enter EIP-ZZ-00/02, EA[ERGENCY IAIPI EAIEN11NG A C110NS. 1his procedure .specifically directs that manual dose 3 assessment be imtiated.
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? IIESPONSES TO TIIE ItEQUEST FOlt ADDITIONAL INFOllMATION l
Radioactivity Exceeds 100 mrem Whole llody or 500 mrem Thyroid for the Actual J or Projected Duration of the Release A. It is not clear whether the set points for the emuent radiation monitors were calculated in accordance with the NUM ARC /NESP-007 guidance.
) NUM ARCINESP-007 specifies that the setpoint for the emuent radiation monitors should be calculated using the FS AR source term applicable to each monitored pathway in conjunction with annual average meteorology. The basis for the Callaway EAL IC indicates that a most limiting case regarding the direction of the release was used to calculate the set point. l
)
Provide additional information describing the relationship between the method used to calculate the radiation monitor set point for the Callaway EAL and the NUMARC/NESP-007 method. Providejustification for any deviation between these methods.
)
nespanse - IVe recalculated the X/q values as an average of all directions. 7his resulted in a slighdy less conservative but much more realistic setpc> int. lhis setpoint, along with the above X/q value and selecting a Main Steam Line Break accident (worst casa per our FSAR) results in 500 CDE thyroid at the EAB. Because of the 1/NG ratios the 7hyroid dose is abruys reached before the 100 mrem 1EDE. I have inchuled a copy ofour l
)
Technical Basesfor our PC Based Dose Assessment. 7his document includes a description ofhow this setpoint was calculatedalong with the source terms which are the FSAR source terms corrected to our current core (enc. 4).
y 1here is an issue of the Alert level "200 times the ODCAf limits" exceeding the setpoint at the Site Emergency level. 7he overlap is created because the Site Emergency setpoint is calculated using our K9AR s<mrce terms which have very conservative lodine to Noble Gas ratios.1hese ratiosproduce the 500 CDE thyroid at the EAR when the 1EDE dose is not close to the limit. lhis causes our Site Emergency setpoint to be below the Alert level.
? IVe recogni:e thisproblem, andare revising our source terms at this time. IVe expect to issue these new source terms within 6 months. IVe wil! then utili:e these source termsfor revised serpoint calculations and wiH submit a change request to our EALs.
Due to the NUA1 ARC Alethodah>gy, the Site Emergency setpoint is afactor of 20 higher y
than our current NUREG 0654 Si1E EAfERGENCY EALS, which have the some overhip problem. Until we are able to revise the setpoint our position is improved and wefeel the setpoint is acceptable.
)
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RESPONSES TO THE REQUEST FOR ADDITIONAL INFORMATION B. The Callaway EAL scheme did not include the NUh1 ARC /NESP-007 condition to p perform a dose assessment if the efIluent levels exceed the radiation monitoring set points.
The corresponding NUMARC/NESP-007 provides guidance to initiate a dose assessment using actual source term and meteorological conditions upon exceeding the radiation monitor set point so that the classification will be based upon the best estimate of dose h consequences. An event should be classified based upon the ellluent radiation monitor set points (which were calculated using default . values) only if a dose assessment cannot be performed within 15 minutes.
Providejustification for this deviation from the NUMARC/NESP-007 guidance.
Respouse - A note to indicator I was added to clanfy that an actualdose assessment should be performed. In the event that an actual assessment cannot be made within 15 '
minutes, indicator 1 is used tc declare the event.
C. The following two NESP EALs were not included in the Callaway EAL scheme.
)
ASI-3 A valid reading sustained for 15 minutes orlonger on a perimeter radiation monitoring system greater than 100 mr/hr. [for sites having telemetered perimeter monitors).
ASI-4 Field survey results indicate site boundary dose rates exceeding 100 mr/hr
) field survey samples indicate child thyroid dose commitment of 500 mr for one hour of inhalation.
Providejustification for this deviation. If these EALs were not included because these specific indications are not available at Callaway, you should consider whether other indications are available to determine whether the IC should be entered.
).
Response - For EA L 3 , Callaway does not have a telemeteredperimeter monitor system.
Peryour recommendation EAL 2, was split into separate EALs. 2.for dose assessment and 3.forfieldsurvey residts that indicate 100 mrem /hr 7EDEfor one hour or a CDE thyroid dose of 500 mremfor one hour ofinhahition.
y
- 5. EAL ID - EAU Dose Resulting From an Actual or Imminent Release of' Gaseous Radioactivity Exceeds 1000 mrem Whole Body or 5000 mrem thyroid for the Actual or Projected Duration of the Release
)
A. It is not clear whether the set points for the efiluent radiation monitors were calculated in accordance with the NUMARC/NESP-007 guidance. ,
NUMARC/NESP-007 specifies that the setpoint for the effluent radiation monitors should be calculated using the FSAR source term applicable to each monitored pathway in conjunction
.I with annual average meteorology. The basis for the Callaway EAL 1C indicates that a most-limiting case regarding the direction of the release was used to calculate the 'setpoint.
4 04/20/94
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' RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION '
Provide additional infonnation describing the relationship between the method used to calculate the radiation monitor setpoint for the Callaway EAL and the NUMARC/NESP-007 p ,
method. Providejustification for any deviation between these methods.
Response - We recalculated the Xhj values as an average of alldirections.
this residted in a slightly less conservative but much more realistic setpoint. This setpoint, along with the above X'q value andselecting a Main Steam Line Break accident (worst case ,
per our FSAR) results in 5000 CDE thyroid at the EAB. Because of the UNG ratios the 7hyroiddose is always reached before the 1000 mrem TEDE. Ihave includeda copy ofour Technical Basesfor our PC Based Dose Assessment. 1his document includes a description ofhow this setpoint was calculated along with the source terms which are the FSAR source terms corrected to our current core (enc. 4).
D IL The Callaway EAL scheme did not include the NUMARC/NESP-007 condition to perform a dose assessment if the ellluent levels exceed the radiation monitoring setpoints.
The corresponding NUMARC/NESP-007 provides guidance to initiate a dose assessment J; using actual source term and meteorological conditions upon exceeding the radiation monitor setpoint so that the classification will be based upon the best estimate of dose consequences. An event should be dassified based upon the efiluent radiation monitor se points (which were calculated using default values) onlyif a dose assessment cannot be .
performed within 15 minutes.
) Providejustification for this deviation from the NUMARC/NESP-007 guidance.
Response - A note to indicator 1 was added to clarify that an actual dose assessment should be performed. In the event that an actual assessment cannot be made within 15 minutes, indicator I is used to declare the event.
C. The following two NESP EALs were not included in the Callaway EAl scheme.
AGI-3 A valid reading sustained for 15 minutes orlonger on a perimt ar radiation monitoring system greater than 1000 mr/hr [for sites having telemetered perimeter monitors]. .
AGI-4 Field survey results indicate site boundary dose rates exceeding 1000 mr/hr
- field survey samples indicate child thyroid dose commitment of 5000 mr for one hour ofinhalation.
Providejustification for this deviat_ ion. If these EALs were not included because these 1 specific indications are not available at Callaway, you should consider whether other indications are available to determine whether the IC should be entered.
c Response - For EAL 3, Callaway does not have a telemeteredperimeter monitor system. Peryour recommendation' EAL 2, was split into separate EALs. I 2.for dose
) assessment and 3.forfielJsurvey residts that indicate 1000 mrem /hr IEDEfor one hour .
or a CDE thyroid dose of 5000 mremfor one hour ofinhalation. ,
ll
)
l 5 04/20/94. )
1 1
a b l RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION
) 6. EAL IE - Unexpected Increase in Plant Radiation The Callaway site rpecific EAL lE, " Spent Fuel Pool level is decreasing on EC-LI-0039A with all available installed makeup sources being added, and allirradiated fuel assemblies remain covered" deviates from the corresponding NUMARC/NESP-007 (AU2-1) by including the condition " with all available installed makeup sources being added" in place of the
) NUMARC/NESP-007 EAL condition of" uncontrolled water level decrease."
Provide justification for this deviation.
Elesponse - The indicatorsfor the Spentfuelpool and the Refuelingpool were changed to all
" Normal" as opposed to "all available" makeup sources being addedfor added conservatism.
Rather than use " uncontrolled water level decrease" we p efer to clearly define what that level decrease is to eliminate Operator confusion.
- 7. EAL 2 - Containment llarrier Potential 1,oss Indicators A. The following NUMARC/NESP-007 EAL was not included in the Callaway EAL scheme.
Containment Pressure (site specific) PSIG and increasing.
Providejustification for this deviation.
)
Respanse - Our setpoint would be the same set point that is usedfor FRZ.1 redpath summary, 60 psig, (see ene. 2) which in itself results in a redpath summaryfor containment. 1his set point is displayed in the control room on aplant computer terminalfor the FRZ.1 status tree. A separate indicator wmdd be redundant, andin this case couldcause
) confusion as to why the same set point is looked atfor both indicators. Per ourprocedure Critical Safety Function Status 7 tees (CSF-1), the trees should be continuously scanned, as long as a condition higher than 17%LWE exists. If after any comjdete scan of the trees, no condition higher than YELLME exists, the scanningfrequency nury be reduced with permission of the Shift Supervisor to 10-20 minutes but willcontinue to be monitored. During scanning a 1 review summary attachment is completed indicating the status of each tree and signed by the performer. 1his is then presented to the Shift Supervisor (Emergency Coordinator)for .
sigrutture.
H. Provide additional documentation regarding the derivation of the set point for the
) Containment Darrier Potential loss indicator, "3. Containment Radiation Monitoring:
GT-RE- 59/60.. reading >l 5 E+3 R/hr."
itespanse - Our calculationfor the Containment Barrier Potential Loss indicator was performed using the Westinghouse Owners Group (WOG) " Post Accident Core Damage Assessment Methe>dcdogy" dated November I9M (enc.1).1his document was approved by the NRCfor core c!amage assessment. Based upon a Containment fligh Range Area 6 04/20/94 i
) 1 RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION l l
i
' Response (Cont) - !
) Radiation Monitor (CilARM) reading a percent clad ckimage (equivalent to percent i noble gas release) can be estimated. Westinghouse makes the assumption that any :
percent noble gas release requires an equalpercent clad damage. Conversely, a Radiation Monitor reading can be produced given the percent clad damage.
) An example of the relationship of the exposure rate ofa dete: tor as afunction of time following reactor shutdown is presentedin Figure 3-3 (enc.1). The exposure rates are expressedin units ofJUhr-MWt.
From enc.1:
)
Radiation Alonitor Readmg (Rihr) x CDIT l'olume (p3)
R hr - AIIVt -
6 3 Plant Power ($fIVI) x 2x10 (ft )
) where:
Rihr - AflVt =
5.5from Figure 3-3for a 20% noble gas release equivalent to 20% cladfailure.
CDtT l'olume - 2.5x106p3
) Plant Power -
3565 AflVI i
Solving lbr Radiation Alonitor Reading:
5.5(3565 AfIVt)(2x106p3) ,
) CHARAfReading ~
2.5x100p3
- 156861&hr
)
)
y 7 04/20/94
9 llESPONSES TO Tile ItEQUEST FOlt ADDITIONAL INFOllMATION O C. Provide all of the critical safety function status procedures which are referenced in the fission product barrier EALs.
Response - See enc. 2.
O
- 8. EAL 2 - RCS llarrier Loss Indicators A. The Callaway RCS Barrier loss indicator EAL, "1. RCS Leak Rate: Safety Injection initiated with a loss of subcooling.. " deviates from the corresponding NUMARC/NESP-007 EAL, "RCS Leak Rate: Greater than available makeup capacity as indicated by a
.O loss of RCS subcooling "
The Callaway EAL contains the condition that Safety injection has initiated whereas the NUMARC/NESP-007 EAL does not include this condition.
Provide justification for this deviation.
O iles ponse - Safety injection initiated indicates all available inven:ory control capacity is m service. In all cases if the RCS leak is greater than available makeup capacity S1 will be initiated. By using the SI indicator we eliminate the OperatorJudgment that would be required to determine if the leak is " greater than avaitable makeup capacity" which can be o di[]icult to determine.
II, The Callaway RCS Barrier loss indicator EAL, "2. SG Tube Rupture: a) Any of the following . and b) SG pressure decreasing in an uncontrolled manner" deviates from the corresponding NUMARC/NESP-007 EAL, "(site-specific) indication that a SG is O ruptured and has a non-isolable secondary line break or (site-specific) indication that a SG has a ruptured and a prolonged release of contaminated secondary coolant is occurring from the affected SG to the environment." The Callaway EAL condition that a SG pressure is decreasing in an uncontrolled manner does not appear to have a one to one correspondence to the NUMARC/NESP-007 EAL condition of a non isolab!c O secondary line break or a prolonged release of contaminated coolant.
Provide justification for this deviation.
Response - We have added an indicator so that the manual use of a ruptured SG PORT' for cool down (prolonged release of contaminated secondary coolant) would also meet this O IC. Indicator 2, a) is consistent with our diagnostic activities of the Emergency Operating Proceduresfor a rupturedsteam generator. Indicator 2, b) is again consistent with our EOPsfor afatdted steam generator (unisolable steam line break stuck open safety or PORl' ). 1his is consistent with NESP 007 which states the indicators shouldinclude an uncontrolled or complete deptryssatation of the rupturedSG. It goes on to say \
O.
i 8 04/20/94 4
'O RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION Response (Cont) -
'O the secondary radiation shouldinclude air ejector discharge (our GE-RE-92), SG blowdown (our BM-RE-25), and SG sampling (our SJ-RE-02).
7he sentence that cautions that declaration should not be based on an operator induced cool down, does not apply to Callaway, as we specifically train to recognize operator, vs.
event induced cool down depressuri:ation. Thejinal sentence states that this EAL should
.O encompass steam breaks, feed breaks, and stuck open safety hclef valves. Indicator 2 b) willrepresent all of these.
For consistency, the indicatorfor the CIMT barrier v as also changed to reflect the use of a PORYfor cool down and to reflect afatdted SG. I believe these indicators are now O consistent with the question #3from the Barrier section of the NUMARC Guestions and Answers.
C. Provide additional documentation regarding the derivation of the set point for the RCS Barrier loss indicator, "3. Containment Radiation Monitoring: GT-RE-59/60.. .. reading O M E+3 R/hr."
Response - Our calculationfor the Containment harrier Potential Loss indicator was performed using the Westinghouse Owners Group (WOG) " Post Accident Core Damage Assessment Methodc>h>gv" dated November 1984 (enc.1). This document was approved by the NRCfor core damage assessment. Based upon a Containment High Range Area a Radiation Monitor (CHARM) reading a percent clad damage (equivalent to percent noble gas release) can be estimated Westinghouse makes the assumption that any percent noble gas release requires an equalpercent clad damage. Conversely, a Radiation Monitor reading can be producedgiven the percent clad damage.
An exam [de of the relationship of the exposure rate of a detector as afunction of time following reactor shutdown ispresentedin Figure 3-3 (enc.1). The exposure rates are expressedin units ofRihr-MWt.
From enc.1:
- O Radiation Alanitor Readmg (Ithr) x CDIT Volume (113 )
Ibhr- Atll's ~
6 3 Plant Power (51111) x 2x10 (fl )
where:
lbhr - Alll*t = .35from Figure 3-3for a .3% noble gas release approximately equivalent to our Tech Spec activity hmits of1 pCi/gm del-131..
CDTT Volume - 63 2.5x!O ft Plant Power - 3565 AfII't I
9 04/20/94 p .
O i RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION 1 1
Ihsponse (Cont)-
Solvingfor Radiation Monitor Reading:
.35(3565 MWI)(2x106p3)
CHAlofReading =
6 2.Sx10p3 0
- 9981%r O-
- 9. EAL 2 - Fuel Clad llarrier Loss Indicators A. The following NUMARC/NESP-007 EAL was not included in the Callaway EAL scheme.
Core Exit Thermocouple Readings - Greater than (site-specific) degree F.
g Although the Critical Safety Function Status - Core Cooling Red may use the same indicator as the NUMARC/NESP-007 EAL, e.g. core exit thermocouple > 1200 F, all available indications should be used if available. (Discuss how the critical safety functions are used. Providejustification for this deviation.
O Response - From enc. 2 (Critical Safety Function Status Trees) ifcore exit 7C's are not less than 1200 Deg. F, you go to FR-C.1 redpathfor core cooling. Core exit 7C's are monitored to evaluate this status tree. TVefeel that including a redundant indicator unnecessarily complicates this IC , making it potentially confusingfor the Operators.
O These Critical Safety Function Status 1rees, are displayed on a summary screen on any selectedplant conymter terminal. This includes the control room, the ISC and the EOF.
1he plant computer continuously displays the status of each tree by color on a summary screen, or each individual tree can be displayed. Also a control room crew member continuously scans the trees manually usingprocedure CSF-1.
O B. Provide additional documentation regarding the derivation of the set point for the Fuel Clad 13arrier loss indicator, "3. Containment Radiation Monitoring:
GT-RE-59/60.... reading >3 E+3 R/hr."
O Response - Our calculationfor the Containment Barrier Potential Loss indicator was performed using the Westinghouse Owners Group (WOG) " Post Accident Core Damage Assessment Methodology" dated November 1984 (enc.1). This document was approved by the NRCfor core damage assessment. Based upon a Containment High Range Area Radiation Monitor (CHAi&f) reading a percent clad damage (equivalent to percent O noble gas release) can be estimated. Westinghouse makes the assumption that any percent noble gas release requires an equalpercent clad damage. Conversely, a Radiation Monitor reading can be producedgiven the percent clad damage.
10 04/20/94 O
3-ITESPONSES TO TIIE ItEQUEST FOlt ADDITIONAL INFOllMATION Response (Cont)-
An example of the relationship of the exposure rate of a detector as afunction of time folkming reactor shutd<mn ispresentedin Figure 3-3 (enc.1). The exposure rates are expressedin units ofR/hr-MWt.
) From enc.1:
Radiation Afonitor Readmg (It'hr) x ClAfT l'olume (ft3 )
It'hr - AfIt's -
6 3 Plant Power (AllVI) x 2x10 01 )
D-where:
R:hr - A1IVt -
1.1from Figure 3-3for a 5% noble gas release equh alent to 5% cladfailure.
CIAfT l'olume - 63 2.5x10 ft l,
Plant Power ~ 3565 AfIVt Solv;ngfor Radiation Afannor Readmg:
63 l.l(3363 Afil t)(2xlG )1 )
Clli1RAIReading ~
2.5x106 ;j3
) - 3137R/hr
)
1 1
04/20/94
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RESPONSES TO THE REQUEST FOR ADDITIONAL INFORMATION
- 10. EAL 2 - Fuel Clad Barrier Potential Loss Indicators The following NUMARC/NESP-007 EALs were not included in the Callaway EAL scheme.
Core Exit Thermocoupa Readings - Greater than (site specific) degree F Reactor Vessel Water Level-less than (site-specific) value.
Although the Critical Safety Function Status - Core Cooling Orange may use the same indicators as these NUMARC/NESP-007 EALs, e.g. core exit thermocouple > 700 F and vessel level below..., all available indications should be used. Provide justification for this deviation.
s Response - We willinclude these indicators.
- 11. EAL 3A - Confirmed Security Event Which Indicates a Potential Degradation in the J Level of Safety of the Plant The following NUMARC/NESP-007 EAL was not included in the Callaway EAL scheme.
11U4-2 Other security events as determined from (site - specific) Safeguards Contingency Plan" Provide justification for this deviation.
Response - 1here were no additional events identified to be include 1 with these security ems. We will however include in our description ofEC typejudgment ems that other security events not covered here could warrant an emergency declaration.
i O
l 12 04/20/94 l 9
. , ,' ' ,0')
b.
RESPONSES TO THE REQUEST FOR ADDITIONAL INFORMATION' t
y 12. EAL 3B - Security Event in the Plant Protected Avea
[ The following EAL was not included in the Callaway EAL scheme.
IIA 4-2 Other security events as determined from (site - specific) safeguards Contingency Plan"
} Provide justification for this deviation.
Response - 1here u ere no additional events identified to be included with these security EALs. We will however inchide in our description ofEC typejudgment EALs that other security events not covered here could warrant an emergency
} declaration.
- 13. EAL 3C - Security Event in a Safe Shutdown Area The following EAL was not included in the Callaway EAL scheme.
[ IISI-2 Other security events as determined from (site - specific) safeguards Contingency Plan" ;
Providejustification for this deviation.
L Response - There were no additional events identified to lie included with these :
. security EALc We will however include in our description ofEC typejudgment l EALs that other security events not covered here could warrant an emergency declaration.
f 14.' EAL 3F- Fire AITecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown The Callaway EAL for this IC, " Fire in any of the following areas: . and there is visible damage to permanent structures or equipment, afTecting the operability of redundant trains of .
Safety related equipment" deviates from the corresponding NUMARC/NESP-007 EAL, " Fire -
)
or explosion in any of the following areas . . and a rfected system parameter indications show degraded performance or plant personnel report visible damage to permanent structures or .;
equipment within the specified area." in that the Callaway scheme includes the condition that i
redundant trains are affected.
y- A fire does not need to affect redundant trains of safety related equipment to meet the r threshold specified in the NUMARC/NESP-007 guidance for this EAL.
Provide justification for this deviation.
y Response - We have revisedindicator 2, to eliminate our requirement ofaffecting redundant trains.
i !
l;
! 04/20/94 13
~
v
RESPONSES'TO THE REQUEST FOR ADDITIONAL INFORMATION A I j 15. EAL 3G - Natural and Destructive Phenomena Affecting the Protected Area -
The following 'NUMARC/NESP-007 EAL was not included in the Callaway EAL scheme.
1I01-4 Yehicle crash into 1,lant structures or systems within the protected area boundary. ;
D. The licensee states that although NUMARC/NESP-007 specifies that a vehicle crash into -
l_ a plant safety system is an unusual event, since at Callaway safety systems are located in l
vital areas, this EAL is classified at the Alert level.
The basis for the NUMARC/NESP-007 EAL (HU4-1) is: "EAL 4 is intended to address such item:; as plane or helicopter crash, or on some sites, train crash, or barge crash that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant." Furthermore, the NUMARC/NESP-007 basis specifies thati"if the crash is confirmed to affect a plant vital area, the event may be escalated to Alert." In l j- consideration of the basis for the 'NUMARC/NESP-007 EAL it is appropriate for the q l Callaway EAL scheme to include an unusual event EAL for a vehicle crash. ,
):. Add a vehicle crash EAL or provide additionaljustification for this deviation.
L l
Res90use - We have revised indicator 3,for this IDIL to include a vehicle crash.
)
i 16. EAL 311 - Natural and Destructive Phenomena Affecting a Safe Shutdown Area
IIA 1-6 Turbine Failure generated missiles result in any visible structural damage to or penetration of any of the following plant areas..."
)
L The Lices. ees basis for not including a site-specific EAL which corresponds to this v NUMARC/NESP-007 EAL is that turbine generated missiles cannot affect safety systems due to the configuration of the main turbine. Provide additional information to support ,
this supposition.
D Response - Per our FSAR SIIE ADDENDUM 3.5 (enc. 6), the reactor building,
( auxiliary building andfuel buildings are all outside of the trajectory of a turbine l
\ generated missile. "The annualprobability ofa turbine missile damaging a critical
!I component at the Callaway Plant isfound to be 1.98 x 10-8. This value is sufficiently low that no specificprotective measures are requiredfor turbine missiles. " We consider "no l specificprotective measures" to include no specific EAL required. 1 1
i 14 04/20/94 l
- i
- RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION
- 17. EAL 3I- Release of Toxic or Flammable Gases Deemed Detrimer.tal to Safe 3 Operation of the Plant Provide additional information regarding the " limits" specified in this EAL, (i.e.
" amounts in excess oflimits for atmospheric contaminants per CTP-ZZ-01200").
Response - This indicator was changedfrom " limits of atmospheric contaminants" O to creating a ha:ardous atmosphere per C1P-ZZ-01200. A ha:ardous atmosphere is definedin CIP-ZZ-01200 as .
An atmosphere that may expose employees to the risk ofdeath, incapacitation, impairment of ability to self-rescue, injury or acute illnessfrom one or more of thefollowing causes:
(1) Flammable gas, vapor, or mist in excess of 10% ofits lower explosive limit (LEL).
(2) Airborne combustible dust at its LEL.
(3) Atmospheric oxygen concentration below 19.5% or above 23.5%.
(4) Atmospheric concentration of any substancefor which a dose orpermissible O exposure limit (PEL) has been listed in subpart G, Occupational Heahh and Environmental Control, or in Subpart Z, Toxic and Ha:ardous Substances of 29 CPR 1910 and which couhlresult in employee exposure in excen of its dose or PEL.
O 18. EAL 3J - Release of Toxic or Flammable Gases Within a Facility Structure Jeopardizes Operation...
Provide additional information regarding the " limits" specified in this EAL, (i.e.
"IDLH concentration per CTP-ZZ-01200 and LEL per CTP-ZZ-01200").
O Response - This indicator was changed to " creating a HAZARDOUS A1AIOSPHERE"per CIP-ZZ-01200. A ha:ardous atmosphere is definedin C1P-ZZ-01200 as :
An atmosphere that may expose employees to the risk ofdeath, incapacitation, O impairment of abihty to self-rescue, injmy or acute illnessfrom one or more of thefollowing causes:
(1) Flammable gas, vapor, or mist in excess of10% ofits lower explosive limit (LEL).
(2) Airborne combatible dust at its LEL.
(3) Atmospheric o , gen concentration below 19.5% or above 23.5%.
D (4) Atmospheric concentration of any substancefor which a dose orpermissible exposure limit (PEL) has been listed in subpart G, Occupational Heahh and Environmental Control, or in Subpart Z, Toxic and Ha:ardous Substances of 29 CFR 1910 and which could result in employee exposure in excess ofits dose or PEL.
O 15 04/20/94
k
- l. :
1 RESPONSES TO THE REQUEST FOR ADDITIONAL INFORMATION-l
- 1 3 19. EAL 3L - Control Room Evacuation flas Been Initiated and Plant Control '.
Cannot Be Established The Callaway EAL scheme for Group 3 L, "1. Entry into OTO-ZZ-00001 Control -
room evacuation is required and 2. The Aux Shutdown Panelis manned within 15 - !
minutes," deviates from the corresponding NUMARC/NESP-007 EAL, "a. Control 1
room evacuation has been initiated and Control of the Plant cannot be established .
! per (site-specific) procedure" in that the Callaway scheme does not specify that control has been established. ,
i The condition that the ASP is manned within 15 minutes does not correlate to the j condition that plant control has been established.
Providejustification for this deviation.
Response - We changed this indicator to include " Control of the Aux FeedSystem ,
and a SG PORYfor cooldown cannot be established within 15 minutes". The 15
)
l minutes is consistent with the Westinghouse Response Planfor Immediate
- Evacuation of the Control Room time study.
- ~ ,
i b
! .J I
I i
)
r l
l< l i
j 16 04/20/94
RE'SPONSES TO THE REQUEST FOR ADDITIONAL INFORMATION .
} 20. EAL 4A - Unplanned Loss of Most or All Alarms (Annunciators) for Greater than 15 minutes The following NUMARC/NESP-007 EALs were not included in the Callaway EAL scheme under this IC.
SU3-lb Compensatory non-alarming indications are available, and 7-Providejustification for this deviation.
Response - Per the NESP-007 document, compensatory non-alarming indications include computer basedinformation such as SPDS At Callaway we no longer have
~) separate computer systems. Our plant computer includes SPDS, IUUS and the NSSS! BOP systems. A_]_I of our computer syst:ms are part of the Plant Computer System.
In the Alert or Site Emergency level, we have a " plant computer is unavailable" indicator which equates to " compensatory non-alarming indications unavailable". A
" plant computer is available" indicator is not needed at the Umtsual Event level if a
). " plant computer is unavailable"is included at the Alert level. 7he plant computer has to be available to be at the UnusualEvent level.
SU3-Ic In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased sun >cillance to safely operate the unit (s).
) Providejustification for this deviation.
Provide information regarding the conditions specified in the Callaway EAL scheme to indicate that the loss of most or all annunciatorsc In particular, describe how the Callaway EAL conditions relate to the corresponding NUMARC/NESP-007 EAL.
Response - Per the NESP EAL the emergency is only declaredifin the opinion of the Shift Supervisor, the loss requires increased surveillance. By removing this indicator, .
we assume that ay loss requires increasedsurveillance. This is consistent with our Off _
normalprocedure OTO-RK-00001 (enc. 7). Which lists required compensatory actions y for as loss ofan annunciator.
In regards to our discussion about whether there are otherpossible events that could ,
disable our annunciators such as system grounds. Our system includes ground detection on power supplies, inputs and outputs. Since our system is highly sectionalized, it would be extremely difficult topostulate a situation where a cardfailure or ground could
} . disable most annunciators. On either of these events an alarm is received and we would enter 070-RK-00001 andincrease surveillance ofaffected. systems. In the unlikely h event that afailure did occur that is not covered here, the EC could declare an emergency on ECjudgment per EIP-ZZ-00101. I willinclude this example in the body '
of the procedure EIP-ZZ-00101 EMERGENCYIMPLEMEN11NG AC170NS for EC y Judgment EALs.
17 04/20/94
RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION 1;
2L EAL 4B - Unplanned Loss of Most or All Alarms (Annunciators) With Either a Transient In Progress...
).
The following NUMARC/NESP-007 EALs were not included in the Callaway EAL scheme under this ICL SA4-lb Compensatory non-alarming indications are tLnavailable, and
) Provide justification for this deviation.
Response - This indicator should be unavailable vs. the available listedinyour requestfor additional information.
Per the NESP-007 document, compensatory non-alarming indications include
> computer based information such as SPDS. At Callaway we no longer have separate computer systems. Ourplant computer inchides SPDS, RRIS and the NSSSIBOP systems. ALI of our computer systems are part of the Plant Computer <
System. In the Alert or Site Emergency level, we have a " plant computer is unavailable" indicator which equates to " compensatory non-alarming
) indications unavailable". A " plant computer is available" indicator is not needed at the Unusual Event levelif a " plant computer is unavailable"is included at the Alert level. The plant computer has to be available to be at the Unusual Eventlevel.
) SA4-1.d.2 In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased surveillance to safely operate the unit (s).
Provide justification for this deviation.
) Response - Per the NESP EAL the emergency is only declared ifin the opinion of the Shift Supervisor, the loss requires increasedsurveillance. By removing this indicator, we assume that ay loss requires increased surveillance. This is consistent with our Offnormalprocedure OTO-RK-00001, which lists required compensatory actionsfor ag loss of an annunciator.
y _
In regards to our discussion about wheiher there are otherpossible events that '
could disable our annunciators such as system grounds. Our system includes ground detection onpower supplies, inputs and outputs.' Since our system is highly sectionalized, it would be extremely difficult to postulate a situation where
. a cardfaihtre or ground could disable most annunciators. On either of these
)'
events an alarm is received and we would enter 010-RK-00001 and increase surveillance ofaffectedsystems. In the unlikely event that afailure did occur that is not covered here, the EC could declare an emergency on ECjudgment per EIP-ZZ-00101. I willinclude this example in the body of the procedure EIP-ZZ-00101 EMERGENCYIMPLEMENTING ACTIONS for EC.hidgment y
EALs.
18 04/20/94'
).
I RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION-1
- 22. EAL 4C - Inability to Monitor a Significant Transient in Progress M The following NUMARCINESP-007 EALs were not included in the Callaway EAL scheme under this IC.
SS6-lb Compensatory non-alarming indications are tLn.available, and Provide justification for this deviation.
)E Res90nse - 1his indicator should be unavailable vs. the available listedin your requestfor additionalinformation.
Per the NESP-007 document, compensatory non-alarming indications include computer based information such as SPDS. At Callaway we no longer have
)~ separate computer systems, Ourplant computer includes SPDS, RRIS and the NSSS/ BOP systems. All ofour computer systems arepart of the Plant Computer System. In the Alert or Site Emergency level, we have a " plant computer is unavailable" indicator which equates to " compensatory non-alarming indications unavailable" A " plant computer is available" indicator is not D. needed at the Urusual Event levelifa ' plant computer is unavailable"is included -
at the Alert leve The plant computer has to be available to be at the Umisual Event level.
SS6-1.c In the opinion of the Shift Supervisor, the loss of the
) annunciators or indicators requires increased surveillance to safely operate the unit (s).
Provide justification for this deviation.
Response - Per the NESP EAL the emergency is only declaredifin the opinion 5 of the Shift Supervisor, the loss requires increasedsurveillance. By removing this indicator, we assume that as loss requires increasedsurveillance. This is consistent with our Offnormalprocedure OTO-RK-00001. Which lists required ,
compensatory actionsfor a_rjy loss of an annunciator, y in regards to our discussion about whether there are otherpossible events that could disable our anmmciators such as system grounds. Our system includes ground detection on power supplies, inputs and outputs. Since our system is highly sectionali:ed, it would be extremely difficult to postulate a situation where a cardfailure or groundcould disable most annunciators. On either of these y events an alarm is received and we would enter OTO-RK-00001 andincrease surveillance ofaffectedsystems. In the unlikely event that afailure didoccur that is not covered here, the EC coulddeclare an emergency on ECjudgment per EIP-ZZ-00101. I willinclude this example in the body of the procedure EIP-ZZ-00101 EMERGENCYIMPLEMENTING ACTIONS for EC Judgment -
j EALs.
19 04/20/94. . .,
)
8 RESPONSES TO TIIE REQUEST FOR ADDITIONAL INFORMATION g 23. EAL 4P - Fuel Clad Degradation The following NUMARC/NESP-007 EAL was not included in the Callaway EAL scheme under this IC.
SU4-1 (Site Specific) radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable g limits.
The licensee states that the failed fuel monitor was not used in its EAL because it would duplicate the chemistry sample indication. However, all available indicators should be used to determine whether an IC is met. In addition, the failed fuel 9 monitor indication may be available before the chemical sample indication.
Modify this IC to incorporate the NUMARC/NESP-007 EAL or provide additional justification for this deviation.
nesponse - Upon receiving an alarm on SJ-RE-01 (failedfuel monitor), we are O required byprocedure to have chemistry confirm the alarm by sample. At Callaway, our monitor isolates on a Containment Isolation signalfrom a safety injection. Since there are situations where it is not on-line and since we ahsays sample to validate an alarm, we would wait to declare the emergency untilthe sample results are obtained. Including this as a separate indicator is not D necessary, as an alarm willrequire the sample, which will require the declaration, if the Tech. Spec. limits are exceeded. Infact it would confuse the Operators, suggesting that we declare an emergencyfrom apossible spurious alarm. Again once we validate the alarm we will know if we are exceed 7'echnical Specification limits.
D D
i O
i 20 04/20/94 0
1 CIIANGES TO CALLAWAY'S EAL SUBMITTAL 3 As a result of training all of the Operators, a few changes were identified that we would like to make to our original submittal.
1: Since there are no Radwaste accident events that could reach the required levels to declare the Site or General Emergencies, we would like to delete the Radwaste Vent O Monitor Gil-RE-10. Attached for your review is enc. 3, which shows the calculated EAB values for Radwaste accidents.
- 2. We would like to add the asterisk used in IC IE and IG to IF as well. There will be planned events such as lifling the Upper Internals for refueling where the setpoints O could be exceeded.
- 3. We would like to label the Area Radiation Monitors in IC IG.
- 4. Per the NESP question and answer for IIazards, question #9, there was some O confusion as to what a bomb in a Vital (Safe Shutdown Area) would be classified.
The answer was to include it in IC 3C, which we would like to do. I missed this point in our original submittal.
- 5. In IC L and M in Group 4, there are times during refueling, that core exit TC's are O not available. We would like to add additionalindicators for this situation.
w 3
e 1 04/20/94 J
Enclosure 1 .
f) '*. Page 1 of 6 3.3 CONTAINMENT RADIATION MONITORS AND CORE DAMAGE -
86 .
Post accident radiation monitors in nuclear plants can be used to estimate the
~
xenon and krypton concentrations in the containment. .
C) An analysis has been made to correlate these monitor readings in R/hr to estimate gaseous radioactivity concentrations. For this analysis the following assumptions were cade:
- 1. Radiogases released from the fuel are all released to containment.
() .
- 2. Accidents were considered in which 100% of the noble gases, 52% of noble-gases, and 0.3% of the noble gases were released to the containment.
'O
- 3. Halogens and other fission products are considered not to be significant contributors to the containment monitor readings.
A relation can be developed which describe's'the gamma ray exposure rate of a O detector with time, based on the amount of noble gases released. .The exposure rate reading of a detector is dependent on plant specific parameters: the operating power of the core, the ef ficiency of the monitor, and the volume .
seen by the monitor. The plant specific response of the detector as a _
C) function of time following the accident can be calculated from the instantaneous gamma ray squrce strengths due to noble gas release. Table 3-2, and the plant characteristics of the detector. The gaena ray source strengths presented in Table 3-2 are based on 100 percent release of the noble gases.
To determine the exposure rate of the detector based on 52 percent and 0.3
()
percent noble gas release, 52 percent and 0.3 percent, respectively, of the .
gamma ray source strength are used.
Alternately, the energy rates in Kev / watt-sec given in Table 3-2 can be c) expressed in terms of an instaneous flux by assuming the energy is absorbed in 3
a em of air. These energy rate values, in Nev/ watt-sec-cm , when divided by discrete values of Mev/ photon and the gamma absorption coef ficient for air,
-5 y,' considered as a constant (3.5 x 10 g,-1), provide values of the photon flux, photons / watt-cm -sec, as shown in Table 3-2A. The discrete O' 2 values of Mev/ photon were cotained by using the average values of the energy groups, Mev/garra, from Table 3-2. *
=
56
- Rur. f
y ,. . , Enclosure 1 Page 2 of 6 TABLE 3-2 e
- l' 4 INSTANTANEOUS GAMMA RAY SOURCE STRENGTHS DUE TO A 100 PERCENT l
RELEASE OF NOBLE GASES AT VARIOUS TINES FOLLOWING AN ACCIDENT
[nerav Group Source Strenath at Time Af ter Release (Nev/ watt-sec)
O. 1 Hour 2 Hours 8 Hours Hev/ gamma 0 Hours 0.5 Hourji, 8 0 8 0.20 - 0.40 1.2 x 10 9
3.0 x 10 8 2.6 x 10 2.4 x 10 2.0 x 10 9 0 0 0 7 0.40 - 0.90 1.5 x 10 3.4 x 10 2.6 x 10 1.9 x 10 5 9 x 10 6
0.90 - 1.35 1.3 x 10 0
9.4 x 10 7
6.7 x 10 4.7 x 10 7 9.8 x 10 8 0 7 1.35 - 1.80 1.8 x 10 0
3.4 x 10 2.1 x 10 1.4 x 10 2.9 x 10 0 0 0 7 1.80 - 2.20 1.4 x 10 0
5.4 x 10 3.6 x 10 2.4 x 10 5.2 x 10 0 0 0 8 2.20 - 2.60 1.3 x 10 0
8.5 x 10 7.1 x 10 5.3 x 10 1.1 x 10 5.0 x 10 5 2.60 - 3.00 4.0 x 10 0
6.6 x 10 6
5.1 x 10 6 3.5 x 10 6 6 6 4 3.00 - 4.00 3.5 x 10 8
6.3 x 10 5 4.5 x 10 2.6 x 10 9.7 x 10 4.00 - 5.00 3.1 x 10I 4.4 x 10 4 3.6 x 10 2 0 0 0 0 5.00 - 6.00 0 0 0 Q
Mev/carmia 1 Day 1 Week 1 Month 6 Months 1 Year 0 6 O 0.20 - 0.40 1.3 x 10 3.0 x 10 1.5 x 10 0 0 4 4 4 4 O.40 - 0.90 1.1 x 10 7
1.5 x 10 1.5 x 10 1.5 x 10 1.4 x 10 0
0.90 - 1.35 1.8 x 10 5 0 0 0 5 0 0 1.35 - 1.80 5.5 x 10 0 0 5 0 0 Y 1.80 - 2.20 9.9 x 10 0 0 6 0 0 0 2.20 - 2.60 2.0 x 10 0 3 0 0 2.60 - 3.00 8.5 x 10 0 0 0 0 3.00 - 4.00 0 0 0 0 0 0 b 4.00 - 5.00 0 0 0 0 0 5.00 - 6.00 0 0 l
\
! 57 l
l Enclosure 1 0 '
Fage 3 of 6 TABLE 3-2A .
e
~
INSTANTANEOUS GAMMA RAY FLUXES DUE TO 100% RELEASE OF NOBLE GASES AT VARIOUS TIMES FOLLOWING AN ACCIDENT Enerav Group Photon Flux at Time After Relea<e (photons /cm -watt-sec)
O Mev/camma 0 Hours 0.5 Hours 1 Hour 2 Hours 8 Hours I3 13 1.1 x 10l '
13 13 0.3 2.7 x 10 2.4 x 10 2.2 x 10 1.8 x 10 I# I3 13 I3 3.9 x 10 12 0 0.65 1.0 x 10 2.3 x 10 1.7 x 10 1.3 x 10 I I 1.2 x 10 12 2 Il 1.13 3.3 x 10 2.4 x 10 1.7 x 10 2.5 x 10 I 12 3.8 x 10 12 2.5 x 10 II 5.3 x 10 Il 1.58 3.3 x 10 6.2 x 10 I 12 12 12 II 2.0 2.0 x 10 7.7 x 10 5.1 x 10 3.4 x 10 7.4 x TO I 12 12 12 O 2.4 1.5 x 10 1.0 x 10 8.4 x 10 6.3 x 10 1.3 x 10 0 0 10 9 2.8 4.1 x 10 6.7 x 10 5.2 x 10 3.6 x 10 5.1 x 10 2 9 10 10 8 3.5 2.9 x 10 5.3 x 10 3.8 x 10 2.2 x 10 8.1 x'10 0 6 4.5 1.9 x 10" 2.8 x 10 2.3 x 10 0 0 Mev/camma 1 Dav 1 Week 1 Month 6 Months 1 Year I3 12 1.4 x 10" 0 0.3 1.2 x 10 2.7 x 10 0 9 ll 1.0 x 10' 1.0 x 10' 0.65 7.3 x 10 1.0 x 10' 1.0 x 10' 1.13 4.5 x 10 0 0 0 0 10 0 0 1.58 1.0 x 10 0 0 10 0 2.0 1.4 x 10 0 0 0 9 10 0 2.4 2.4 x 10 0 0 0 7 0 2.8 8.7 x 10 0 0 0 3.5 0 0 0 0 0 4.5 0 0 0 0 0 9
- 58
m Enclosure 1 i
)[. .Page 4 of 6 .l In general, values below 0.3% releases are indicative of clad failures, values l c between 0.3% and 52% release are in the fuel pellet overtemperature regions,
[~ while values between 52% release and 100% release are in the core mel.t
)~ regime. To represent the release of the normal operating noble gas activity ,
in the primary coolant as obtained from ANS 18.I I I, 1.0 x 10~3 % of the
~
gamma ray source strength is used. In actual practice it must be recognized f that there is overlap between the regimes because of the nature in which core
) heating occurs. The hottest portion of the core is in the center due to flux distribution and hence greater fission product inventory. Additionally heat transfer is greater at the core per'phery due to proximity of pressure vessel walls. Thus conditions could exist where there is some molten fuel in the
) center of the core and overtemperature conditions elsewhere. Similar conditions can occur which lead to overtemperature in the central portions of the core, and clad damage elsewhere. Thus, estimation of extent of core damage with containment radiation readings must be used in a confirmatory
) sense - as backup to other measurements of fission product release and other indicators such as pressure vessel water levels and core exit thermocouples.
An example of the relationship of the exposure rate of a detector as a ,
function of time following reactor shutdown is presented in Figure 3-3. The
)
exposure rates, which are expressed in units of R/hr-MWt, are representative of a point located 57.5 feet below the apex of the containment dome of a containment volume of 2 x 10 6 ft.3 No objects or components shield the detector from the noble gas sources yhich are assumed to be uniformly ,
) distributed throughout the containment building.
l The methodology of using the relationship of containment radiogas monitors readings shown in Figure 3-3 is:
{
- 1. Determine time lapse between core shutdown and radiation reading. j f
l'
- 2. Record containment monitor reading in R/hr at this time.
l
- 3. Correct the monitor reading for specific plant power via the relationship:
t
+
, Radiation Monitor Readinq Plant Power (MWt) i e
- _. i 59 f
l
i Enclosure l' '
- Page 5 of 6 ,
c
)~
) ,
?
~
1000.05
- 7 3.- -
t I 100% Noble Gas Release {
100.0= :
y i I 52% Noble 10.0 l Gas Release ,
s n 5 Do.g i
$1L 3 g!
5 sa , <<g& 5-
- w
.5 5 -
I
%e %
T ,
E= I ;
- . p 1.0-1 ;l 0.3% Noble Ges (
~
Rel ease 5 j .
g .
i
) r 1.0-2:3 t
I. .
ANS 18.1 Nomal Operating s
- 1.0-34 3 -
! Noble Gas Release y
l 1.0-4_ _-
?'
. 4.
I
) 1.0 10.0 100.0 1000.0 i^
TIME AFTER ACCIDENT (HOURS)
FIGURE 3-3 PERCENT NOBLE GASES IN CONTAINMENT FOR
, 6 3 CONTAINMENT VOLUME OF 2 x 10 FT i 3
. 1 60
. Enclosure 1
~
Page 6 of 6 .
4 4. Determine core damage regime f rom Figure 3-3 at the time interval j ascertained in step 1. .
For plants which have the same monitor characteristics as the monitor described above, except for the cnntainment volume which differs from 2 x 106 ft , Figure 3-3 can be used provided a correction is made to the exposure r' ate (R/hr) as follows.
3 Radiation Monitor Readine (R/hr x Containment vol. (ft 1 R/hr-MWt Plant Power (MWt) x 2 x 10 ft s ,
O
.) e a
e O
J O
O I
6T O
f r' Enclosure-2.
Page 1 of 3~ g g -
O ,Proced. No. CRITICAL SAFETY FUNCTION Attachment Rev.
- CSF-1 STATUS TREES- 2 '1BO
,r . CORE COOLING
.o CORE COOLING .
40 GO TO FR-C.1 GO TO FR-C.1 4
+i CORE EXIT TC'sE' ----- ~ - ' ~ ~-'--'~'l N O~l ,
- O b- !!.
Y E Sj RVLIS (PUMPS OFF) NO GREATER THAN 48%
YES
= 0 E M. 2
~
E IT TC*5 LE'S'S O O
THAN 198 DEG. F.
YES l AT LEAST ONE tNOl pyt s (pungs opp) g=QG0TOFR-C.2 0-
~
O RCP RUNNING YES GREATER THAN 48 % YES
! . e
- e.- L-G0 TO FR-C.3 1 i i u m 4 30 TO FR-C.2 O g 17 1
"nurs'TJ.J"' l! NO f w . . on, nzuznun, ;
$ si YES i !!
O _ o RCS SUBC00 LING - NORE SUBC00 LED r.1 .
i
~ 6, GO TO FR-C.3 THAN INSTRUMENT ERROR. USE PAGE
,2 of 4. PAGES 3 of 4 and 4 ef 4
- \0' l
- . ___j V MY BE USED IF SUBC00 LING NETERS jv- i NOT AVAILABLE. (USE APPLICABLE I _ j CURVE FOR CTNT CONDITIONS). , ,_j C9F S A ~~
i G,
O
+
Page _.L of 4 O
i _ Enclosure 2 Page 2-of'3 Proced. No. Crit! cal SAFETY ruwcticN STATUS TSEE5 Attachment Rev.
CSF-1 5 180 i CONTAINMENT
).
CONTAINMENT D'
GOTOFR-Z.1 I
) NO CONTAIMENT PRESSURE LESS THAN bl PSIG. YES y = = = = = = = 4 GD TO FR-2.1 1 i E u E CONTAIRENTPRESSURE NO LESS THAN 27 PSIG. YES l
sp = == 4 GO TO FR-2. 2 m a 1 I u l
) CONTAINMENT RECIRC. NO 1 SUMP LEVEL LESS THAN 138 INCHES YES l
....3GOTOFR-2.3 1
CONTAlmENT RADIATION NO LESS THAN 3 R/HR. YES ggg g47 3 I
)--
4 Page 1 of 1
Enclosure 2 j.. -Page'3 of 3-Proced.-No. CRITICAL SAFETV FUNCTION STATUS TSEES Attachment Rev.
OSF-1 3 1BO HEAT SINK y
HEAT SINK
)
CDTOFR-H.1 TOTAL FEED FLOW TO NO SG's CREATER THAN 388.888 LBM/HR. YES
)
tARROW RANGE LEVEL IN AT LEAST DNE SG NO c.,.eI.00TOFR-H.2 Y GREATER THAN 4% 1 f 5-
[35% FOR ADVERSE CTMT) YES 3RESSURE IN ALL NO
) "
l y
.o.iGOTOFR-H.3 s
~
tARROW RANGE LEVEL NO
) IN ALL SG's LESS THAN 78% YES _
o3: 00TOFR-H.4 PRESSURE IN ALL NO
) SG's LESS THAN.
1180 PSIG YES e e 4};00 TO FR-H.5
) tARROW RANGE LEVEL IN ALL NO SG's GREATER THAN 4 %
(35% FOR ADVERSE CTNT) YES c g;- 347 ll C's U
)
i Page _L. of 1
~
l
""" f g
CALLAWAY - SP TABLE 15.7-4 RADIOLOGICAL CONSEQUENCES OF A WASTE GAS DECAY TANK RUPTURE
,s a)
Doses (reml Exclusion Area Boundary (0-2 hr)
D Thyroid 8.85E-2 Whole body 3.29E-2 Low Population Zone Outer Boundary (duration)
D Thyroid 1.16E-2 Whole body 4.28E-3 g
3 9
O D-
,e Rev. OL-2 6/88 e
, _ _ . . _ . - . . - _ _ _ . . m . . _ _ . _. _ ., _ _ _
' Enclosure 3
,. Page: 2 of. 2 - ,
CALLAWAY - SP
-TABLE 15.7-6 RADIOLOGICAL CONSEQUENCES OF A I) LIQUID RADWASTE TANK FAILURE Doses frem)
Boron Recycle Tank i O Exclusion Area Boundary (0-2 hr)
Thyroid 4.25E-2 Whole-body 5.10E-3 >
Y) Low Population Zone Outer Boundary (duration)
Thyroid 5.56E-3 Whole-body 6.65E-4 O
Primary Evaporator Bottoms Tank Exclusion Area Boundary (0-2 hr)
Thyroid 2.63E-1 Whole-body 6.11E-5 Low Population Zone ;
Outer Boundary (duration) ,
Thyroid 3.47E-2 Whole-body 8.09E-6 0
O
- O . .
1 Rev. OL-2 6/88 -
O i
.. - .