ML20083K661
| ML20083K661 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom, Limerick  |
| Issue date: | 04/26/1995 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20083K655 | List: |
| References | |
| NUDOCS 9505150054 | |
| Download: ML20083K661 (321) | |
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ENERGY i PEACH BOTTOM ATOMIC POWER STATION Emergency Preparedness 1
-r UPGRADED EMERGENCY ACTION LEVELS ,
NRC Submittal Copy I Revision a i
April 26,1995 l
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ENERGY i -
PEACH BOTTOM ATOMIC POWER STATION Emergency Preparedness I
i UPGRADED EMERGENCY ACTION LEVELS l NRC Submittal Copy I Revision a J
April 26,1995
l PBAPS EAL TechnctJ Bisis Manual REv a. Apnl 26.1995 Page 1 of 130 PBAPS EAL Technical Basis Manual Table of Contents Section t - Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Section ll - Acronym s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Section lli - EAL Technical 3 asis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.0 Reactor Fuel
. 1.1 Co olant Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1.2 Containment High Radiation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 1.3 Irradiated Fuel or New Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 2.2 Re acto r P owe r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.0 Primary Containment 3.1 Primary Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 4.0 Secondary Containment 4.1 Secondary Containment Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 4.2 Main Ste am Lin e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 .
5.0 Radioactivity Release 5.1 Effluent Release and Dose .................................. 67 5.2 I n - Pla nt R adi atio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . 77 6.0 Loss of Power 6.1 Loss of AC or DC Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 ,
7.0 Intemal Events 7.1 - Technical Specifications & Control Room Evacuation . . . . . . . . . . . . . . . . 91 ,
7.2 Loss of Decay Heat Removal Capabililty . . . . . . . . . . . . . . . . . . . . . . . . . 94 -
7.3 Loss of Assessment / Communications Capabililty . . . . . . . . . . . . . . . . . . . 97 8.0 Extemal Events 8.1 S ecu rity Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 105 8.2 Fire / Explosion and Toxic / Flammable Gases . . . . . . . . . . . . . . . . . . . . . . 110 8.3 Ma n - Made Eve nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117 ,
8.4 N atu ral E ve n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 120 9.0 Other 9.1 General . . . ............................................. 127
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PBAPS EAL Technical Basis Manud REV a. Apnl 26,1995 Page 2 of 130 Section 1 - Introduction 1
This manual contains the technical basis for the Emergency Action Levels as utilized in ERP-101, Classification of Emergencies. The format and use of this manualis as follows. 1 1
- 1. Heading and Sub-Heading l There are nine major headings each containing one or more sub-headings. These are as J
follows: )
1.0 Reactor Fuel 1.1 Coolant Activity 1.2 Containment High Radiation 1.3 Irradiated Fuel or New Fuel 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level 2.2 Reactor Power 3.0 Primary Containment 3.1 Primary Containment 4.0 Secondary Containment 4.1 Secondary Containment Temperature 4.2 Main Steam Line 5.0 Radioactivity Release 5.1 Effluent Relese and Dose 5.2 In-Plant Radiation 6.0 Loss of Power 6.1 Loss of AC or DC Power 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation 7.2 Loss of Decay Heat Removal Capabililty 7.3 Loss of Assessment / Communications Capabililty 8.0 External Events 8.1 Security Events 8.2 Fire / Explosion and Toxic / Flammable Gases 8.3 Man-Made Events 8.4 Natural Events 9.0 Other 9.1 General
l PBAPS EAL Technical Bosio Manual REV a, April 26.1995 Page 3 of 130
- 2. Emergency Classification Level and Number identification The classifications range from Unusual Event through Alert, Site Area Emergency to General Emergency. For each sub heading, there may not be an EAL in every classification level. Each EAL is individually and uniquely numbered. No two numbers are the same.
- 3. INITIATING CONDITION The Initiating Condition or IC (as described in NUMARC NESP-007) is contained in this section. ICs are a predetermined subset of condititions where either the potential exists for a radiological emerency or such an emergency has occurred. Additionally, ICs are the means by which EALs for different nuclear power plants are standardized.
- 4. EAL Each Emergency Action Level exactly as it is contained in ERP-101.
- 5. OPCON The operational condition (OPCON) that the EAL is applicable in is contained here. There are six OPCONs (1,2,3,4 and 5 and defueled) that are used. PBAPS also uses mode switch position. These positions are stated below and are Run, Startup, Shutdown and Refueling. It should be noted that these OPCONs are entry level conditions. The EAL is applicable if the plant was in the OPCON at the start of the event. Subsequent positons of the mode selector switch should be ignored for purposes of classification.
OPCON (MODE) EODE SWITCH POSITION Ii12141416101 , cJn m ais141slol Startup hitisl41siol Shutdown (hot) 111:1s14Islol Shutdown (cold) 1 0 :1:141s1o1 Refueling I1]tl4]Al5]DI N/A (defueled)
- 6. BASIS The technical basis of each EAL is contained in this section. This includes any necessary calculations and also includes escalation references.
- 7. DEVIATION Any deviations from the NUMARC NESP-007 methodology are contained in this section.
If there are no deviations, NONE is used.
- 8. REFERENCES All applicable references used in developing the technical basis for each EAL are contained in this section.
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I PBAPs EAL Technied Basis Manual i REV a April 26,1995 l Page 4 of 130 :
Section 11- Acronyms AC - Attemating Current ADS - Automatic Depressurization System APRM - Average Power Range Monitor ARI -
Alternate Rod insertion ARM -
Area Radiation Monitor ATWS - Anticipated Transient Without Scram BRP -
Bureau of Radiation Protection CDE - Committed Dose Equivalent CFM -
Cubic Feet Per Minute CFR - Code of Federal Regulations CRD -
Control Rod Drive CS - Core Spray DBA -
Design Basis Accident DC -
Direct Current DEI -
Dose Equivalent lodine .
EAL - Emergency Action Level ECCS -
Emergency Core Cooling Systems EDG -
Emergency Diesel Generator EPA - Environmental Protection Agency ERP-C - Emergency Response Procedure - Common FC -
Fuel Clad (Barrier)
FTS -
Federal Telephone System GPM -
Gallons Per Minute HCTL -
Heat Capacity Temperature Limit HPCI -
High Pressure Coolant injection IC -
Initiating Condition IRM -
Intermediate Range Monitor KV -
Kilovolt LCO -
Limiting Condition for Operation LOCA -
Loss of Coolant Accident LPCI - Low Pressure Coolant injection MPH -
Miles Per Hour mR/hr -
Milli Roentgen Per Hour MSIV -
Main Steam isolation Valve NFPB -
Normal Full Power Background NPSH -
Net Positive Suction Head NRC -
Nuclear Regulatory Commission NUMARC -
Nuclear Management and Resources Council ODCM -
Offsite Dose Calculation Manual OPCON -
Operating Condition PBAPS -
Peach Bottom Atomic Power Station PEMA -
Pennsylvania Emergency Management Agency PC - Primary Containment (Barrier)
PSIG - Pounds Square Inch Gauge RC -
Reactor Coolant (Barrier)
RCIC - Reactor Core isolation Cooling RCS - Reactor Coolant System
f PBAPS EAL Technicci Basis Manu"J REv a, Apnl 26,1993 Page 5 of 130 RHR -
Reactor Protection System RPV -
Station Blackout SJAE -
Source Range Monitor SRV -
Top of Active Fuel TPARD -
Total Protective Action Recommendation Dose TRIPS -
Transient Response implementation Plan Procedures pCi/cc -
Micro Curie Per Cubic Centimeter Cl/gm -
Micro Curie Per Gram UFSAR- Updated Final Safety Analysis Report VDC -
Volts Direct Current
l PBAPS EAL Technical Bass Manual REV c. Apnl 26,1995 Page 6 of 130 i
f Section lli - EAL Technical Basis We
PBAPs EAL Technical Basis Manual REv c. April 26,1995 PIga 7 of 130 1.0 Reactor Fuel 1.1 Coolant Activity UNUSUAL EVENT - 1.1.1.a IC Fuel Clad Degradation EAL Reactor Coolant activity > 4 pC//gm Dose Equivalent lodine 131 OPCON til21sidislol BASIS Coolant activity in excess of Technical Specifications (> 4 Cl/gm) is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition. This level is chosen to be above any possible short duration spikes under normal conditions. An Unusual Event is only warranted when actual fuel clad damage is the cause of the elevated coolant sarnple (as determined by laboratory confirmation). However, fuel clad damage should be assumed to be the cause of elevated Reactor Coolant activity unless another cause is known, e.g., Reactor Coolant System chemical decontamination evolution (during shutdown) is ongoing with resulting high activity levels.
This event will be escalated to an Alert when Reactor Coolant activity exceeds 300 pCl/gm Dose Equivalent lodine 131 per EAL Section 1.1.2.
DEVIATION The OPCON applicability [1,2,3,4) is a deviation from NUMARC [all) in that the Technical Specifications only require only 1,2,3,4.
REFERENCES Technical Specification Section 3.6.B NUMARC NESP 007, SU4.2
PBAPS EAL Technical Bads Manual REv O, Apnl 26,1995 Page 8 of 130 1.0 Reactor Fuel 1.1 Coolant Activity UNUSUAL EVENT - 1.1.1.b IC Fuel Clad Degradation EAL SJAE Radiation (Offgas Monitor) > 2.5x10' mR/hr OPCON fil 218 l*15 Iol BASIS The steam Jet air ejector (Offgas) radiation monitor RR-2(3)-17-152 in the Control Room would be one of the first Indicators of a degrading core. The high-high alarm is set at the Technical Specification limit of 2.5x10' mR/hr. This instrument takes a sample before the recombiner. This indicator of elevated activity is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition.
This event will be escalated to an Alert when RCS activity exceeds 300 pCl/gm Dose Equivalent lodine 131 per EAL Section 1.1.2. or when Containment radiation is measured in 1.2.2.
DEVIATION None REFERENCES Technical Specifications Section 3.8.C.7.a NUMARC NESP-007, SU4.1
PBAPs EAL Technied Basis ManucJ REV c. April 26,1995 Page 9 of 130 1.0 Reactor Fuel 1.1 Coolant Activity ALERT - 1.1.2 IC Loss of Fuel Clad EAL Reactor Coolant activity > 300 pC//gm Dose Equivalent lodine 131 OPCON 18I2131d15101 BASIS A reactor coolant sample activity of greater than > 300 pCi/gm was determined to indicate significant clad heating and is indicative of the loss of the fuel clad barrier. This concentration is well above that expected for lodine spikes and corresponds to 2.6% clad damage. 2.6% fuel clad damage is based upon NUREG-1228 core damage analysis.
Calculation of 300 pCi/cc equivalance to percent fuel clad damage is as follows (for purposes of this calculation, cc and gm are considered equivalent):
lodine isotope Dose Factors Cl/MWe Values (Time After Shutdown = 0)
(Rea Guide 1.109) (NUREG-1228) 1-131 4.39E-3 85000 1-132 5.23E-5 120000 1-133 1.04E-3 170000 1-134 1.37E-5 190000 1-135 2.14E-4 150000 Time After Shutdown (T = 0) Ratios R,3, = 120000/85000(I-131) = 1.41(I-131)
Ri33 = 170000/85000(1-131) = 2.00(I-131)
R,u = 190000/85000(I-131) = 2.24(I-131)
R,33 - 150000/85000(I-131) = 1.76(1131)
Equation for Dose Equivalent lodine (del,3,)
isi ist i tai i2 i isi 3 m si m is 131 as DE/i3, =
DFi3, Solve for A,3, assuming del,3, = 300 pCi/cc
PBAPS EAL Techrned Basis Manud REV a. April 26,1995 Page 10 of 130 l
, A 33,4.39E-3+1.41 Ai33 5.23E-5+2.00Ai3,1.04E-3+2.24A i3,1.37E-5+1.76A,332.14E-4 )
4.39E-3
,6.95E-3A33, 4.39E-3 Therefore: A,3, = 189 pCl/cc l-131 l Clad damage fraction (NUREG-1228, Table 4.1) = .02 Full Power = 1150 MWe Clad Activity I 131 = (Ci/MWe) (MWe) (Clad Damage Fraction)
= (85000Ci/MWe) (1150MWe) (.02)
= 1.96E6 Cl Reactor Water Volume = 2.67E8 cc (ERP-C-1410)
Total Coolant Activity 1131 = (A,3,) (Rx Water Volume) (Ci/ Cl)
= (189 pCl/cc) (2.67E8cc) (1.0E-6Cl/ pCl)
= 5.05E4Cl Percent Clad Damage = Total Coolant Activity / Clad Activity 1131
= (5.05E4) / (1.96E6)
= 2.6%
This event will be escalated to an Site Area Emergency when additional fission product barriers are lost per EAL Section 1.1.3.
DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #1 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents, Table 2.2 Reg. Guide 1.109, Table E-9 ERP-C-1410, Page 15
PBAPs EAL Technied B: sis Manual REV a. April 26,1995 Prgi 11 of 130 1.0 Reactor Fuel 1.1 Coolant Activity SITE AREA EMERGENCY - 1.1.3 IC Loss of Fuel Clad and (Loss of Reactor Coolant System or Containment)
EAL Reactor Coolant activity > 300 pCl/gm Dose Equivalent lodine 131 AND Identified breach of Primary Containment (Tech Specs Section 3.7) OR Drywell Pressure
> 9 psig OPCON h l 21 a ldis101 BASIS This EAL Indicates a loss of the fuel clad barrier and the loss of primary containment or loss of the reactor coolant barrier. A reactor coolant sample activity of > 300 pCi/gm Dose Equivalent lodine 131 was determined to indicate significant clad heating and be indicative of the loss of the fuel clad barrier. Technical Specifications 3.7 define conditions which must be met to consider the primary containment barrier intact. Intentional venting of the containment is also included.
These conditions include:
- drywell and pressure suppression chamber are intact and;
- all primary containment penetrations required to be closed during accident conditions are either capable of being closed by an operable containment automatic isolation valve system or closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position;
- at least one door in each airlock is closed and sealed; and,
- all blind flanges and manways are closed.
The continual increase in TORUS pressure, or the sudden increase in pressure, to 9 psig indicates the presence of a large breach in the reactor coolant pressure boundary and subsequent release of high energy reactor coolant into the containment and thus a loss of the Reactor Coolant barrier.
This event will be escalated to a General Emergency through the loss or potentialloss of the third fission product barrier per EAL Sections 4.1.4, 4.2.4 and 1.2.4.
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PBAPs EAL Technical Basis Manud REV e. April 26,1995 Page 12 of 130 DEVIATION An exception to the NUMARC Methodology was taken in that NUMARC states that the site-specific drywell pressure should be based on the drywell high pressure alarm setpoint which indicates a LOCA (50gpm leak). The high drywell pressure alarm is 2 psig and can be reached by a small primary system leak and/or loss or drywell cooling, which are addressed in EALs 3.1.1.a and 3.1.1.b, respectively. The value of 9 psig was selected in that it is larger than experience shows of blown packing and recirc seal leaks. The value of 9 psig is more representative of a LOCA condition and this torus pressure is in the TRIPS for actions to protect '
the containment.
REFERENCES NUMARC NESP-007, FC EAL #1, PC EAL #5 and RC EAL #2 Technical Specifications Section 3.7 T-102, Primary Containment Control, PC/P-6 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents f
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PBAPs EAL Technicd Basis Manud REV a, April 26,1995 P4g313 of 130 1.0 Reactor Fuel 1.2 Containment High Radiation ALERT - 1.2.2 IC Loss of Fuel Clad EAL Containment high range radiation > 3x10' R/hr OPCON litrialdislol
+
BASIS The reading of 3x10' R/hr (from shine) on a containment radiation high range monitor RI-8(9)103,A,B,C,D has been calculated to correspond to 300 pCl/gm Doce Equivalent lodine 131 within the RCS.This is based on a direct ratio of 9000 R/hr shine with 100% clad failure to 2.6%
clad failure. The calculation is as follows:
Since: 300 Cilgm Dose Equivalent lodine 131 = 2.6% Clad Damage Drywell Rad of 9000 R/hr shlne = 100% Clad Damage (2.6%)(9000 R/hr)/(100%) = approximately 250 R/hr Containment Dose Rate rounded to 300 R/hr for human factors for 2.6% Clad Damage OR 300 Ci/gm Dose Equivalent lodine 131 The value of 300 R/hr is used for human f actors of the logarithmic meter and strip paper recorder.
This reading is significantly higher than readings expected during normal operations and also significantly lower than the readings expected with this coolant concentration released into containment. This concentration indicates significant clad heating and is indicative of the loss of the fuel clad barrier. This corresponds to approximately 2.6% clad damage based upon ,
NUREG-1228 core damage analysis (detailed calculations are contained in the Basis for EAL 1.1.2).
This event will be escalated to a Site Area Emergency with the loss of an additional fission product barrier per EAL Section 1.2.3.
DEVIATION None
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_i- PBAPS EAL Technical Basis Manual it - RE/ 0, April 26,1996 Page 14 of 130
- g. REFERENCES NUMARC NESP 007, FC EAL #4 NUREG 1228 - Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents GE Assumptions - Mitigation of Core Damage I
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PBAPs EAL Techrmd Basis Manu*J REv a, April 26,1995 Page 15 of 130 1.0 Reactor Fuel 1.2 Containment High Radiation SITE AREA EMERGENCY - 1.2.3.a IC Loss of Fuel Clad and Containment EAL Containment high range radiation > 3xff R/hr AND Identified breach of Primary Containment [ Tech Specs Section 3.7]
OPCON lil2181dislol BASIS This EAL Indicates the loss of the fuel clad barrier and the loss of the containment barrier. The reading of 3x10' R/hr (from shine) on a containment high range radiation monitor (RI-8(9)103,A,B,C,D) has been calculated in EAL 1.2.2 to correspond to 300 pCl/gm Dose Equivalent lodine 131 within the Reactor Coolant. This is based on a direct ratio of 9000 R/hr shine with 100% clad failure to 2.6% clad failure. This reading is significantly higher than readings expected during normal operations and also significantly lower than the readings expected with this coolant concentration released into containment. This concentration indicates significant clad heating and is indicative of the loss of the fuel clad barrier. This corresponds to approximately 2.6% clad damage based upon NUREG-1228 core damage analysis (detailed calculations are contained in the Basis for EAL 1.1.2).
Technical Specifications 3.7 define the conditions which must be met to consider the primary containment barrier intact. Intentional venting of the containment is also included. These conditions include:
+ drywell and pressure suppression chamber are intact and;
- all primary containment penetrations required to be closed during accident conditions are either capable of being closed by an operable containment automatic isolation valve system or closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position;
- at least one door in each airlock is closed and sealed; and,
- all blind flanges and manways are closed.
This event will be escalated to a General Emergency when additional barriers are lost. ,
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P9APS EAL Technical Basis Manual REV a. Apnl 26,1995
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DEVIATION None REFERENCES NUMARC NESP 007, FC EAL #4 and PC EAL #5.
Technical Specifications Section 3.7 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents GE Assumptions - Mitigation of Core Damage i
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PBAPS EAL Technicd Basis Manud REv a. Apnl 26,1995 Ptgs 17 of 130 1.0 Reactor Fuel 1.2 Coctainment High Radiation SITE AREA EMERGENCY - 1.2.3.b IC Loss of Fuel Clad and Reactor Coolant System EAL Containment high range radiation > 8x1# B/hr OPCON 1812131418E BASIS The reading of 8x10' R/hr on a containment high range radiation monitor RI-8(9)103A,B,C,D indicates the loss of both the fuel and Reactor Coolant System fission product barriers. The reading was calculated assuming an instantaneous release of the Reactor Coolant volume into the Primary Containment (direct reading not shine) at a coolant concentration of 300 Cl/gm Dose Equivalent lodine 131. This calculation is as follows:
Using attachment 5(A) figure 1, Curve 3 (1%) of ERP-C-1410 Time after Shutdown = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 30,000 R/hr Extrapolating to 2.6%
(30,000 R/hr/1%)(2.6) = 78,000 R/hr This is rounded conservatively to 80,000 R/hr for human factors considerations 2.6% clad damage is based upon NUREG-1228 core damage analysis, and by virtue of its release into containment, the loss of the Reactor Coolant barrier (detaileo calculations are contained in the Basis for EAL 1.1.2).
This event will be escalated to a General Emergency when the monitor reading exceeds a value 5
corresponding to 20% fuel clad damage (6x10 R/hr) or the containment barrier is lost per EAL section 1.2.4.
DEVIATION None l
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E PBAPS EAL Technical' Basis Manud REV 0, April 26,1996
-. Page 18 of 130 ,
REFERENCES
, NUMARC NESP 007, FC EAL #3 and RC EAL #3 NUREG 1228 - Source Term Estimation During Incident Response to Nuclear Power Plant Accidents ERP C-1410, pg 10, Attachment 5(A) t t
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PBAPs EAL Technictl Basis Manual REV a. Apnl 2G,1905 Ptg319 of 130 1.0 Reactor Fuel 1.2 Containment High Radiation GENERAL EMERGENCY - 1.2.4 IC Loss of Fuel Clad and Reactor Coolant System and Potential Loss of Containment EAL Containment high range radiation > 6x10' R/hr OPCON lil21*Idl5fol BASIS 5
A containment high range radiation monitor 9RI-8(9)103A,B,C,D reading 6x10 R/hr indicates significant fuel damage, well in excess of that required for the loss of the Reactor Coolant System and Fuel Clad barriers. The reading was calculated assuming an instantaneous release of the Reactor Coolant volume into the Primary Containment (direct reading not shine) where the value corresponds to a release of approximately 20% of the gap region. This calculation is as follows:
Using attachment 5(A) figure 1, Curve 3 (1%) of ERP-C-1410 Time after Shutdown = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 30,000 R/hr Extrapolating to 20%
(30,000 R/hr/1%)(20) = 600,000 R/hr Regardless of whether containment is challenged, this amount of activity in containment, if released, could have severe consequences and it is prudent to treat this as a potential loss of containment and declare a General Emerge ;cy. NUREG-1228, " Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents" states that releases of this magnitude are not possible if plant systems function as designed and any accident with a release of 20% or greater of the gap region must be considered severe.
t DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4, RC EAL #3 and PC EAL #3 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents ERP-C 1410, pg 10, Attachment 5(A) l
PBAPs EAL Technied Basis Manual REv a. April 26.1795 Page 20 of 130 1.0 Reactor Fuel 1.3 Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.3.1.a IC Potential increase in Refuel Floor Radiation due to Lowering Water Level Covering Fuel EAL Unexpected RPV level decrease to e 100"when reactor cavity is flooded up and fuel pool ,
gates are in place OPCON hl21*l4 5lol l
BASIS During refueling operations, RPV level indication is read on Panel 005 (Instrument LT-70). This instrument has a range of 0 - 500" with normal level at 479".
An unexpected level decrease below 100", which is approximately 22 feet above the fuel indicates the cavity is being drained. This is validated with visual observation of a decreasing reactor cavity level if the RPV head is removed.
The value of 100" is calculated from determining 22' above the fuel in the RPV. Technical Specifications requires 22' above the fuel, which equals 264" by (22')(12"/1') = 264". Top of Active Fuel = -172", so 264"- 172" = 92" or conservatively 100" This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to public health and safety is very low. Classification as an Unusual Event is warranted as a precursor to a more serious event.
This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor per EAL section 1.3.2.a.
DEVIATION l
None !
REFERENCES NUMARC NESP-007, AU2.1 Technical Specifications 1
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PBAPS EAL Technied Basis Manud REV c, April 26,1995 Psgs 21 of 130 1.0. Reactor Fuel 1.3 Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.3.1.b IC Potential increase in Refuel Floor Radiation due to Lowering Water Level Covering Fuel EAL Skimmer Surge Tank low level alarm AND Visual observation of a water level decrease below the fuel pool skimmer surge tank inlet OPCON 1812181415101 BAS!S A drop in the Spent Fuel Pool level or the RPV [when in refueling and flooded up with the gates removed] will result in a control room annunciator Fuel Pool Cooling and Cleanup System Troubla Alarm. This Control Room alarm directs an operator to be dispatched to a local alarm panel which will identify the Skimmer Surge Tank low level alarm. This alarm is validated with visual observation of a decreasing Spent Fuel Poollevel. If the spent fuel poollevel decreases below the inlet to the skimmer surge tank, without a planned event such as removing a large piece of equipment, there must be a leak in the spent fuel pool or the RPV. This event has a long lead time relative to potential for radiological release outside the site boundary, thus the irspact to public health and safety is very low. Classification as an Unusual Event is warranted as a precursor to a more serious event.
This everit will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor per EAL section 1.3.2.
i DEVIATION None REFERENCES NUMARC NESP-007, AU2.1, AU2.2 l
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PBAPS EAL Technicd Basis Manust REv a. April 26,1993 Page 22 of 130 1.0 Reactor Fuel 1.3 Irradiated Fuel or New Fuel ALERT - 1.3.2.a IC Major Damage to irradiated Fuel or Loss of Water Level that Resulted or Will Result in the Uncovering of irradiated Fuel Outside the Reactor Vessel EAL Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1)
L_
OPCON 1212181415101 BASIS Unexpected radiation levels which are at least 100 times higher than the normal background will generally indicate a fuel handling accident or loss of water covering the irradiated fuel. Readings may be from refuel floor Area Radiation Monitors or taken during a qualified radiological survey.
Table 1-1 monitors are as follows:
Table 1-1 Refuel Floor ARMS 3-7 (7-9) Steam Separator Pool 3-8 (7-10) Refuel Slot 3-9 (7-11) Fuel Pool 3-10 (7-12) Refueling Bridge Offsite doses during these accidents would be well below the EPA Protective Action Guidelines and the classification as an Alert is therefore aporopriate. This radiation level could also be caused by an inadvertent criticality and is included even though the probability of this event occurring is low. Radiation increases above 500 mR/hr which were expected should not cause an Alert to be declared during a planned evolution. Additionally, surveys which identify " hot spots" greater than 500 mR/hr should not cause an Alert to be declared.
This event will be escalated to a Site Area Emergency by way of offsite doses per EAL Sections 5.1.3. I DEVIATION None I
_ . _ - - - - . -- - - _ . 1
k I
' PBAPS EAL Technical Basis Manusi REV a, April 26,1995 ,
Ptgs 23 of 130 l REFERENCES NUMARC NESP 007, AA2.1, AA2.2, AA2.3, and AA2.4 NUREG 1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant :
- Accidents c
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PBAPS EAL Technical Bais Manual REv ?, Apnl 26,1995 Page 24 of 130 1.0 Reactor Fuel 1.3 Irradiated Fuel or New Fuel ALERT - 1.3.2.b IC Unexpected increase in Plant Radiation or Airborne Concentration EAL A fuel handling accident causing a HiHi alarm from the Refueling Floor Vent Exhaust Monitor OPCON D.12131416101 .
BASIS The Refueling Floor Vent monitor (RR 2(3)-17-456A,B,C,D) alarm is indicative of a release rate which may exceed Technical Specifications from the refuel floor. Although the release rate does not in itself warrant the declaration of an Alert, this alarm could indicate that a fuel handling '
accident has taken place causing major damage to irradiated fuel. The event should be validated by a report from the scene of a fuel handling accident or a radiological survey. It is not intended to indicate planned or expected alarms.
This event will be escalated to a Site Area Emergency due to offsite doses per EAL Section 5.1.3.
DEVIATION None REFERENCES NUMARC NESP-007, AA2.1 EPA-26-24-2
l PBAPs EAL Technical Basis Manud REv c. Apnl 26,1995 Pige 25 of 130 2.0 Reactor Pressure Vessel l
2.1 Reactor Water Level l
ALERT - 2.1.2 \
IC Potential Loss of Reactor Coolant System EAL l RPV level < -160 "
l OPCON lil21*ldislol BASIS As reactor water level decreases from it's normal band, a multitude of automatic actions should i
have occurred. These automatic actions include a reactor scram, trip of the recirculation pumps, isolation of selected systems, and automatic start of both HPCI and RCIC. Some of these l activities are designed to limit inventory loss from the RPV and others are designed to provide inventory makeup to the RPV. Continued decrease in RPV level to -160" indicates that either these actions were unsuccessful in limiting inventory loss from the RPV or that there is a breech in the primary systems which has caused the RPV to depressurize or is at a rate which can not l be overcome by the aforementioned actuations. This RPV level (-160") value is also used for various other system isolations and actuations including: initiation set point for the low pressure Emergency Core Cooling Systems; start signal for the Emergency Diesel Generators; containment I isolation signal; and, as a permissive to the Automatic Depressurization system.
Reactor water level decreasing to the low, low, low reactor water level (-160") set point on level indicator LI-2(3)-02-3-091 or LI-2(3)-02-3113 is indicative of a major plant transient. Although actual core uncovery does not begin until RPV level has decreased to -172" (Top of Active Fuel),
this value has been selected to characterize a loss of coolant event. When -160" is reached an automatic isolation of the MSIVs will occur. Decay heat generated from the fuel after this isolation occurs will need to be removed and/or transferred to the containment. Core Submergence is the preferred method of heat removal from the nuclear fuel. Maintaining RPV level above 2/3 core height ensures that the Reactor Fuel Cladding integrity will remain intact. RPV level decreasing to -160" is an abnormal event and signifies the possibility that level may continue to degrade to the point where submergence no longer occurs, thus this event signifies a potentialloss of reactor coolant barrier. If a LOCA has occurred, this event should be declared even if reflood to > -160" is successful.
Under an Anticipated Transient without Scram (ATWS) scenario, it is possible that actions will be taken to intentionally lower RPV level to below -160". Additionally, there are provisions under an ATWS scenario to permit the MSIVs to remain open even though RPV level has been reduced T nelow -160". These events will be classified under EAL Section 2.2, Reactor Power and should not be classified under this EAL.
PBAPS EAL Technical Bais Manual REV c. April 26,1995 Page 26 of 130 l
This event will be escalated to a Site Area Emergency based upon a loss or potentialloss of the Fuel Clad or Containment barriers per EAL Sections 1.1,1.2 and 2.1.
DEVIATION None -
REFERENCES l
. NUMARC NESP-007, RC EAL #5 T-101, RPV Control t
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PBAPS EAL Technical Basia Manual REV e, Aprd 26,1995 I I
Page 27 of 130 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level SITE AREA EMERGENCY - 2.1.3.a IC Potential Loss of Fuel Clad and Loss of Reactor Coolant System EAL
~
RPV level cannot be restored above -772 "
OPCON h i r l a l d i s lol
~
BASIS Reactor Pressure Vessel Water level cannot be restored above the Top of Active Fuel (TAF) s - 1 172 inches as indicated on RPV Fuel Zone Level Instruments Ll-2(3)-02-3-091 or LI-2(3)-02 113. Core submergence ensures adequate core cooling. This EAL identifies "cannot be recovered" to ensure that emergency blowdown can be utilized with low pressure ECCS to recover above. When RPV level decreases below the top of active fuel the ability to remove the decay heat generated from the nuclear fuel becomes suspect and the Fuel Clad Fission Product barrier can no longer be considered intact. Sustained partial or total core uncovery can result in the release of a significant amount of fission products to the reactor coolant. Sustained core uncovery indicates potentialloss of fuel clad and loss of reactor coolant system.
This event signifies a potential loss of fuel cladding integrity as the ability to effectively remove heat from the nuclear fuel cannot be assured. The assumption can also be made that for level to decrease to below the top of active fuel, there is an inability to restore and maintain core submergence and a breach in Reactor Coolant system integrity. The combination of Emergency Core Cooling Systems (ECCS) and normal water sources are sufficient to restore level when the reactor coolant system is intact, thus there is a loss of Reactor Coolant system integrity when RPV level falls below and cannot be restored to above the top of active fuel.
Core uncovery is expected to occur under large break LOCA scenarios. This inventory loss and subsequent RPV pressure reduction provides a low pressure ECCS initiation signal and permits the injection of these low pressure water soumes. When deciding on the ability to restore RPV level consideration should be given to the availability of low pressure sources (including non-ECCS sources), the injection status of these systems (IA awaiting start permissives, injecting as full flow, etc.), and the trend of indicated RPV level.
Prior to concluding that RPV level cannot be restored, consideration must be given to injection :
system availability and status, reactor pressure and rate of depressurization, and trend of the rate at which RPV level is decreasing. Ample time should be allotted to analyze the ability of injection sources to restore levelin an expeditious manner (after completion of the blowdown stage of the
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PBAPs EAL Techrwcal Basis Manud REV c. Apnl 26,1995 Page 28 of 130 postulated LOCA). Loss of core submergence by itself should not be utilized to classify this event if there is indication that injection systems will be successful in recovering RPV water level.
Under an Anticipated Transient without SCRAM scenario is possible that actions will be taken to intentionally lower RPV level at or below - 172 inches. These events will be classified under EAL Section 2.2, Reactor Power and should not be classified under this EAL.
Escalation to a General Emergency would be based on the inability to restore RPV water level for a sustained period of time per EAL Section,1.2.4 or 2.1.4.
DEVIATION r
None REFERENCES NUMARC NESP-007, FC EAL #2 and RC EAL #4 T-101, RPV Control, RC/L-7 T-111, Level Restoration / Steam Cooling T-117, Level / Power Control T-116, RPV Flooding i
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l PBAPs FAL Techrucal Basis Manual REV c. April 26,1995 Page 29 of 130 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level SITE AREA EMERGENCY - 2.1.3.b IC Potential Loss of Fuel Clad and Loss of Reactor Coolant System EAL RPV level cannot be determined .
OPCON . I 1218ldislol BASIS Inability to determine Reactor Pressure Vessel level may be due to reference leg boil-off, instrument power failure, or conflicting information on uncontrolled parameter oscillations. TRIP procedure guidance will require the flooding of the Reactor Pressure Vessel, thus ensuring core submergence. Based on differences in calibration and design,all ranges of level instruments may not indicate exactly the same; this operational difference is expected and is not to be used for deciding that conflicting RPV level indication exists. Multiple indications of level Instruments pegged high is indication that the level is above the range and that it is known, also visual observation during refueling is indication of RPV water level.
If Indeterminate Reactor Pressure Vessellevelis due to one of the reasons mentioned above, adequate core cooling would be rapidly assured using the guidance provided in the TRIP Procedures; however, if water level cannot be determined, it is conservative to assume that water level is actually below the top of active fuel and that both the Reactor Coolant System and Fuel Clad Fission Product Barriers are potentially lost.
Operator attention should be given to the possibility that under depressurized conditions, there is the possibility that gases may come out of solution and result in distorted RPV levelindications.
Operators should be attentive to observe multiple level indications (particularly those which utilize separate reference legs) to ensure that actual RPV level is known and displayed. Unexplained and/or sudden changes in specific level indications may be a result of degassification of the coolant contained in the level instrumentation.
Escalation to a General Emergency would occur if minimum RPV flooding pressure cannot be established per EAL Section 2.1.4.b.
DEVIATION None
- I
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PBAPS EAL Technical Basis Manual REV o, April 26,1996 Page 30 of 130 i-REFERENCES t
NUMARC NESP-007, FC EAL #4 and RC EAL #5 T-101, RPV Control, RC/L-1 ,
T-112, Rapid Depressurization i T-117, Level / Power Control T-116, RPV Flooding l
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PBAPs EAL Technied Basis Manu11 REv a. Apnl 26,1995 Page 31 of 130 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level GENERAL EMERGENCY - 2.1.4.a ,
IC Loss of Fuel Clad, Loss of Reactor Coolant System and potential Loss of Containment EAL RPV level cannot be restored above -226 "
OPCON 15121*ldl5Iol BASIS Core submergence is the preferred method of core cooling and as such, the failure to reestablish RPV water level above the top of active fuel for an extended period of time could lead to significant fuel damage. This condition, -226 " as read on instruments LI-2(3)-02-3-091 or Ll2(3)-
02-3113, could be indicative of a large break Loss Of Coolant Accident (LOCA) (where ECCS Systems are designed to maintain level at 2/3 core height) or a small LOCA with the inability of emergency core cooling systems to reflood the RPV. The value of -226" was chosen as it represents 2/3 core height.
The time basis for deciding whether or not vessel flooding can be accomplished is dependent on the rate of reactor depressurization, the availability of low pressure ECCS systems, and the rate of RPV inventory loss. Indications such as RPV level trend, ECCS injection flow rates, containment parameter trends, and low pressure ECCS system operability should be considered in rendering the decision as to the ability to reflood the RPV.
Calculations in the UFSAR indicate that reactor fuel should be recovered minutes after a design basis LOCA. The inability to reflood the reactor following a LOCA may indicate severe ECCS degradation and/or multiple failures, including the possibility that jet pumps have failed.
Ample time must be allotted for ECCS systems to reflood the RPV. For events starting from power operation, the failure to rapidly reflood could result in some core melting. Even under these conditions vessel failure and containment failure with resp! tant release to the public would not be expected for some time. Sustained operation with water level below the top of active fuel represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions. Partial core submergence may provide adequate cooling although it cannot be assumed.
The inability to restore level may indicate the presence of an break in one of the jet pumps. This failure would prevent core reflooding with even all injection systems intact when combined with a large loss of coolant accident.
- ._. - - -_ ___~ . - . __ _. .__ , . -..
PBAPs EAL Technical Basis Manual REV a. Apnl 26,1995 Page 32 of 130 Prior to concluding that RPV level cannot be restored consideration must be given to injection ;
system availability and status, reactor pressure and rate of depressurization, and trend of the rate '
at which RPV levelis decreasing. Ample time should be allotted to analyze the ability of injection sources to restore levelin an expeditious manner (after completion of the blowdown stage of the postulated LOCA). Loss of core submergence by itself should not be utilized to classify this event if there is indication that injection systems will be successful in recovering RPV water level.
The failure of the fuel cladding (due to overheating) together with the loss of reactor pressure !
vessel integrity (which would occur during a LOCA) and the introduction of a large amount of ;
energy into the primary containment indicates that all three fissiori product barriers are in jeopardy ;
or lost, i DEVIATION None ,
REFERENCES NUMARC NESP-007, FC EAL #2, RC EAL #4 and PC EAL #4 T-101, RPV Control T-111, Level Restoration / Steam Cooling, LR-11 T-112, Rapid Depressurization T-117, Level / Power Control T-116, RPV Flooding ,
i l
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l PBAPS EAL Technied Basis Manud REV a. Apnl 26,1995 Page 33 of 130 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level GENERAL EMERGENCY - 2.1.4.b IC Potential Loss of Fuel Clad and Loss of Reactor Coolant System and Containment EAL RPV level cannot be determined AND RPV Flooding cannot be established per T-116 OPCON I11213I415101 BASIS The decision to enter RPV Flooding is made when RPV water level cannot be determined. This judgement consists of evaluating all plant indications which can influence the ability to maintain adequate core ecoling. Entry to RPV flooding requires rapid RPV depressurization. The minimum RPV Flooding Pressure is defined as the lowest differential pressure between the RPV and the Torus at which steam flow through the SRVs will be sufficient to remove all of the generated decay heat. Operation at the minimum reactor flooding pressure requires that a sufficient amount of water reach the core to carry away decay heat by boiling, which in tum requires that RPV water level increase. So RPV flooding not established requires containment flooding.
Inability to determine Reactor Pressure Vessel level may be due to reference leg boil-off, instrument power failure, or conflicting information on uncontrolled indication oscillations. TRIP procedure guidance will require the flooding of the Reactor Pressure Vessel, thus ensuring core submergence. Based on differences in calibration and design, all ranges of level instruments may not indicate exactly the same; this operational difference is expected and is not to be used for deciding that conflicting RPV levelindication exists. Levelindication pegged high is indication that the level is above the range and that it is known, also visual observation during refueling is indication of RPV water level.
If indeterminate Reactor Pressure Vessel level is due to one of the reasons mentioned above, adequate core cooling would be rapidly assured using the guidance provided in the Emergency Operating Procedures; however, if water level cannot be determined, it is conservative to assume that water levelis actually below the top of active fuel and that both the Reactor Coolant System and Fuel Clad Fission Product Barriers are potentially lost.
The minimum RPV flooding pressure will ensure that adequate core cooling exists independent of RPV level indication. Failure to establish the differential pressure between the RPV and the Torus in a timely manor can jeopardize the ability of the reactor coolant system to dissipate the decay heat generated.
i PBAPs EAL Techrscal Bads Manual REv a April 26,1995 Page 34 of 130 Eventual clad failure may occur due to overheating of the nuclear fuel if RPV flooding pressure cannot be established in a timely maner. The heat produced from the fuel can cause addit'onal core damage if the cause of the RPV level problem was caused by a LOCA, then both the Clad and the Reactor Coolant have been lost. This will occur with heat being added to the containment. Thus there is a loss of the Fuel Clad and Reactor Coolant barriers with a potential loss of the Containment barrier.
Ample time must be allotted for determining the f ailure of ECCS systems to pressurize the RPV.
Control room indications such as RPV level (used for trending), RPV Pressure, ECCS injection flow rates, Containment parameters, and injection system operability should all be used to gauge the effectiveness of the RPV Flood, if the loss of level indication was caused by reference leg flashing, then level indicators can still be utilized to monitor the trend in RPV level. Actual RPV level will never be higher than indicated
- level, in the event that the loss of level indication is only a result of degassification of the coolant contained in the level instrumentation piping, then it is anticipated that flooding pressure can be obtained.
RPV water level below the top of active fuel for a sustained period of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions. For events starting from power operation, some core melting can be expected.
Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time.
DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4, RC EAL #5 and PC EAL #5 T-101, RPV Control T-111, Level Restoration / Steam Cooling, LR-11 T-112, Rapid Depressurization T-117, Level / Power Control T-116, RPV Flooding
l PBAPS EAL Technicil Basis Manuti REv a Apnl 26,1995 Page 35 of 130 2.0 Reactor Pressure Vessel 2.2 Reactor Power ALERT - 2.2.2 IC Failure of Reactor Protection System instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful EAL Failure of Automatic RPS SCRAM to reduce reactor power < 3%
OPCON lil21*ldl8101 BASIS Entry into this EAL is based on a reactor parameter actually exceeding a RPS setpoint and the reactor is not brought to < 3% power conditions and maintained at that state with automatic RPS functions. The parameter must exceed the RPS setpoint by a significant margin eliminating minor setpoint drifts which are accounted for in the Technical Specification Margin of Safety.
Subsequent manual scram actions were successful in bringing the reactor to < 3% conditions.
Confirmation indications include control room annunciators, APRM power, !RM/SRM power level, SRM period, and Control rod position indication.
A failure nf the Reactor Protection System (RPS) to initiate and complete a reactor scram may indicate that the design limits of the nuclear fuel has been compromised. RPS is designed to automatically detect and generate a reactor scram signal when a limiting safety system setpoint is reached or exceeded. Control rod insertion following a scram signal is designed to be passive (i.e. system deenergizes, control rod motive energy source is previously charged).
Assuming that shutdown (< 3% power) conditions cannot be established / maintained, an automatic scram signal failure followed by a successful manual scram would still constitute a scram failure and should be classified under this event.
Although the reactor may be brought initially < 3% power based on partial control rod insertion, there is a possibility that positive reactivity may be introduced by a number of factors. Xenon decay and factors associated with cooldown, lower fuel temperature (doppler), lower moderator temperature, and a lower presence of steam bubbles (volds) may all contribute to cause a power increase.
Suberitical conditions can be assured even with the most reactive control rod fully withdrawn from )
the core if the remaining 184 control rods fully insert. Any other control rod pattern resulting from partial control rod insertion must be carefully analyzed and/or monitored to detect the possibility l of recriticality or local criticality. !
l 1
I PBAPS EAL Technect! Basis Manual REV a. April 26.1995 Page 36 of 130 Due to the buildup of Xenon in areas of the core that have previously been operating at high power levels, attention should be applied to the possibility that control rods which previously had low worth (e.g. peripheral control rods) may now have significant control rod worth.
When the reactor is not shutdown as identified in the TRIPS, then entry into this EAL is warranted.
When partial control rod insertion occurs following a scram signal (either manual or automatic) judgement should be applied as to whether classification should occur. Multiple control rods failing to insert beyond notch position 02 may require actions to fully insert the control rods.
However, the reactor has been made suberitical, and for all intent the reactor will remain subcritical. TRIP guidance will govern the insertion of these control rods.
This EAL would be escalated with a failure of both manual and automatic scram signals with APRM power remaining above 3% per EAL Section 2.2.3.
DEVIATION None REFERENCES NUMARC NESP-007, SA2 T-100, Scram T-101, RPV Control, RC 1
PBAPS EAL Technical Basis Manual REV a. Apnl 26.1995 Page 37 of 130 2.0 Reactor Pressure Vessel 2.2 Reactor Power SITE AREA EMERGENCY - 2.2.3 IC Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful EAL Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 3%
OPCON 11121 141 1o1 BASIS A valid automatic and/or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication is greater than 3%. The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automaticly). The system is
" fail safe", that is it deenergizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydraulic failure).
A failure of the Reactor Protection System to shut down the reactor (as indicated by reactor power remaining above 3%) is a degraded plant condition that together with suppression pool temperature approaching 110*F requires the injection of boron poison to shut down the reactor.
The RPV Control Trip Procedure establishes 3% power coincident with loss of scram capability as the initiating condition for various plant responses to ATWS events. With Reactor Power less than 3% the heat being generated in the core can be removed from the RPV and containment while actions are taken to bring the ri actor suberitical.
A manual scram is defined as any set of actions by the reactor operator (s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor suberitical (i.e. mode switch to shutdown, manual scram push buttons, or manual ARI initiation).
Taking the mode switch to shutdown as part of the actions required by trip procedure is considered a manual scram action although the mode switch in shutdown will generate a scram signal.
While the plant is being shutdown, significant heat is being generated in the core and the heat up rate of the Torus (due to heat rejection through SRVs) can increase which could approach the Torus temperature limit prior to shutting down. As the Torus heat increases towards the limiting temperature, the probability of causing a major over-pressure event increases substantially.
' PBAPS EAL Technical Basis Manual REv m, Apnl 26.1995 Page 38 of 130 After an ATWS event, there is a potential that the Main Steam Isolation Valves will remain open.
There is additional guidance in the Trip procedures to ensure that the MSIVs remain open even if RPV level is intentionally lowered to below the normal MSIV isolation level. This situation would allow the plant to remove heat and provide makeup through the normal steam / feed cycle. If this path is not available, or becomes unavailable during the transient, heat rejection will be to the Torus.
With Standby Liquid Control initiated and with partial or no control rods insertion there is a possibility that the neutron flux profile in the reactor core may become uneven or distorted.
Localized clad damage is possible if localized power levels increase significantly.
With reactor power remaining above 3% containment Integrity is threatened as the ability of systems to remove all of the heat transferred to the containment may be exceeded. As the energy contained in the containment increases there may be a degradation in the ability to remove heat generated by the "at power" reactor core. There is therefore a potential loss of the containment or the fuel cladding (caused by overheating).
This event will be escalated based on Torus Temperature exceeding 180 F per EAL Section 2.2.4.
DEVIATION None REFERENCES NUMARC NESP-007, SS2 T-100, Scram T-101, RPV Control, RC/L-2 T-117, Level / Power Control l
I PBAPS EAL Technical Basis Manual REV a. Apnl 26,1995 Page 39 of 130 2.0 Reactor Pressure Vessel 2.2 Reactor Power GENERAL EMERGENCY - 2.2.4 1C' Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core EAL Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 3%
AND Torus Temperature is > 180'F OPCON I11218I815101 BASIS A valid automatic or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication is greater than 3%. In addition control room instrumentation indicates that Torus temperature is > 180 F.
Failure of all automatic and manual trip functions coincident with a high Torus temperature will place the plant in a condition where reactivity control capability is Jeopardized and heat removal capability is severely limited.
ECCS systems which may be used to cool the core, transfer heat from the reactor, and/or supply cooling water to the reactor all take a suction of the Torus. Operation with sustained high Torus temperatures may render these systems inoperable due to NPSH considerations.
The RPV Control Trip Procedure establishes 3% power coincident with loss of scram capability as the initiating condition for various plant responses to ATWS events. The timely initiation of Standby Liquid Control (prior to Torus temperature reaching 110 F would bring the reactor to below 3% power before Torus temperature approaches the heat capacity temperature limit curve limitations.
Under ATWS conditions, it is important to assure continuous, stable steam condensation capability. An elevated Torus temperature of 180#F would result in unstable steam condensation should rapid reactor depressurization occur (ADS activation).180 F is the TORUS heat capacity temperature limit (HCTL). Maintaining the ability to condense steam will preclude the pressurization of the containment and prevent possible containment failure.
PBAPS EAL Techrwcal Basis Manual REVQ Apnl 26,1995 Page 40 of 130
, Containment over-pressurization, which would be an eventual result of sustained operation with heat being added to the containment and Torus temperature above 180 F would result in the loss of containment integrity and the inability to remove the heat generated from the fuel. Fuel clad failure would result from the overheating of the fuel.
DEVIATION None REFERENCES NUMARC NESP-007, SG2.1, SG2.2 T-100, Scram T-101, RPV Control T 117, Level / Power Control, RC/L-2
l PBAPs EAL Technical Basis Manud REV c, Apnl 26,1995 Page 41 of 130 3.0 Primary Containment 3.1 Primary Containment UNUSUAL EVENT - 3.1.1.a IC Reactor Coolant System Leakage EAL Drywell Pressure > 2 ps/g AND Indication of a Leak into Containment OPCON h l 213 l A j $lD]
BASIS 1 Reaching 2 psig in the Drywell is likely indication of a primary system leak. Upon receipt of the 2 psig drywell pressure signal on Pressure Recorder PR-2(3)-508, system response includes a reactor scram, ECCS initiation (including HPCI), tripping of the drywell cooling fans and isolation of the cooling water to the drywell. These actuations may mask the trend in drywell pressure.
For example, the scram will result in less heat being added to the containment and the cooling water isolation will result in no heat being removed. Trip procedure guidance will include establishment of cooling for the containment.
There will be control room annunciation indicating the presence of a high drywell pressure condition (.75 psig) prior to the threshold value of 2 psig being reached. Actions initiated as a result of the indication of elevated drywell pressure conditions include stopping energy addition to the containment (e.g. Nitrogen make up). maximizing drywell cooling, and possibly lowering drywell pressure by venting (if permitted by procedures and plant conditions). Continued increase in drywell pressure or a sudden increase in drywell pressure may indicate a leak of high energy reactor coolant into the containment.
Indication of a Leak into Containment was added to qualify the pressure indication to avoid declaring an emergency for situations where the pressure increase is clearly not due to a primary j system leak. For example, an emergency declaration is not appropriate if the high drywell pressure is a result of a loss of Drywell Cooling. Indication of a leak should be determined by observing other containment indications such as sump level, ambient radiation, ambient temperature, and status of cooling systems.
This event will escalate to an Alert based on containment pressure reaching 9 psig per EAL Section 3.1.2
PSAPS EAL Technical Bats Manu2 REV e, April 26,1995 Fage 42 of 130 DEVIATION .
None REFERENCES NUMARC NESP-007, SU5 T-100. Scram T-101, RPV Control T-102, Primary Containment Control P
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l PBAPs EAL Technied Basis Manuj REV a. Apnl 26,1995 Page 43 of 130 3.0 Primary Containment 3.1 Primary Containment UNUSUAL EVENT - 3.1.1.b IC Reactor Coolant System Leakage EAL Unidentified Primary System Leakage > 10 ppm into the Drywell O R_
Identified Primary System Leakage > 25 ppm into the Drywell
=
OPCON h l 21 s idislol BASIS Utilizing the leak before break methodology, it is anticipated that there will be indication (s) of minor reactor coolant system boundary integrity loss prior to this fault escalating to a major leak or rupture. Detection of low levels of leakage while pressurized is utilized to monitor for the potential of catastrophic failures.
This EAL is included as an Unusual Event because it may be a precursor of more serious conditions and, as a result, it is considered to be a potential degradation of the level of safety of the plant. The value of 10 gpm unidentified leakage is significantly higher than the expected pressurized leak rate from the reactor coolant system. The 10 gpm value for the unidentified pressure boundary leakage was selected as it is twice the Technical Specification value, indicating an increase beyond that assumed in Safety Analysis, it also is observable with normal control room indications. The EAL for identified leakage is set at a higher value (25 gpm) due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.
Technical Specification LCO required actions would necessitate a plant shutdown and subsequent depressurization, unless the source of the leak can be isolated, identified, and/or stopped.
Actions initiated by plant staff would include close monitoring of the calculated break size such that any sudden or gradualincrease in leak rate would be identified. A slow power reduction and gradual depressurization would be necessitated due to the possibility that a sudden power and/or pressure surge could potentially worsen the break or cause a catastrophic failure.
The leak rate of 10 gpm may cause a high drywell pressure indication Other indications of a leak of this magnitude would include an increase in drywell temperature or radiation.
This event will escalate to an Alert based upon high Drywell pressure per EAL Section 3.1.2.
- PSAPS EAL Technical Bais knud REV a, April 26.1995 Fage 44 of 130 DEVIATION None REFERENCES NUMARC NESP-007, SU5 Technical Specifications 3.6.C.1 T-101, RPV Control T-102, Primary Containment Control i
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PBAPs EAL Technicd Basis Mand REV a. April 26,1995 Page 45 of 130 3.0 Primary Containment 3.1 Primary Containment ALERT - 3.1.2 IC Loss of Reactor Coolant System EAL Drywell Pressure > 9 psig OPCON lil21*I*181ol BASIS If drywell pressure exceeds 9 psig, there is a clear indication that a leak of sufficient magnitude exists that prevents drywell pressure stabilization.
Providing that RHR was not required to maintain reactor water level, these systems would normally be placed into the Torus cooling / spray mode following the receipt of the high pressure ,
annunciators. This additional cooling, together with the reduction of heat being added to the containment (due to the reactor scram), should result in a reduction in primary containment pressure. The value of 9 psig is obtained from T-102, in that it is the pressure of the Torus where Torus sprays are initiated. For this EAL it will be used as the drywell pressure indicative of a loss of reactor coolant system. This value being used for the drywell is conservative in that it will be reached before the Torus. The continual increase in Torus pressure, or the sudden increase in pressure to 9 psig, indicates the presence of a large breach in the reactor coolant pressure boundary and subsequent release of high energy reactor coolant into the containment and thus a loss of the Reactor Coolant barrier.
This event will escalate to a Site Area Emergency based on a loss or potential loss of other barriers per EAL Sections 1.0,2.0 and 3.0.
DEVIATION An exception to the NUMARC Methodology was taken in that NUMARC states that the site-specific drywell pressure should be based on the drywell high pressure alarm setpoint which indicates a LOCA (50gpm leak). The high drywell pressure alarm is 2 psig and can be reached by a small primary system leak and/or loss of drywell cooling, which are addressed in EALs 3.1.1.a and 3.1.1.b, resp 6ctively. The value of 9 psig was selected in that it is larger than experience shows of blown packing and recirc seal leaks. The value of 9 psig is more representative of a LOCA condition and this torus pressure is in the TRIPS for actions to protect the containment.
Ih PSAPS EAL Techrucal Basis Manual REVQ AprH 26,1996 Page 46 of 130 REFERENCES NUMARC NESP 007,RC EAL #2 T-101, RPV Control T-102, Primary Containment Control, PC/P-6 i
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PBAPs EAL Technical Bais Manud REV e, Apnl 26,1995 Page 47 of 130 3.0 Primary Containment 3.1 Primary Containment SITE AREA EMERGENCY - 3.1.3 IC Loss of Reactor Coolant System and Containment EAL Containment Failure indicated by a rapid, unexplained drop in Containment Pressure following initial pressure rise above 9 psig OPCON 1'1218141s101 BASIS This EAL represents a condition where both the Reactor Coolant System and Containment Fission Product Barriers cannot be considered intact. Rapid, unexplained loss of pressure (i.e.,
not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity. Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, drywell pressure increasing to > 9 psig then dropping (without reason) indicates a loss of containment integrity.
Other indications such as Reactor Enclosure radiation levels, reactor enclosure area ,
temperatures, stack radiation levels, and containment isolation status should be used to confirm the loss of containment integrity, if possible. l This Site Area Emergency declaration is based on the loss of Reactor Coolant System boundary as indicated by higher than expected containment pressurization followed by the loss of primary containment integrity as indicated by the rapid unexplained decrease in containment pressure.
This event would escalate to a General Emergency based on loss of the fuel clad per EAL Sections 1.0,2.0 and 3.0.
DEVIATION None REFERENCES NUMARC NESP-007, RC EAL #2 and PC EAL #1 ON-110, Loss of Primary Containment T-101, RPV Control T-102, Primary Containment Control T-103, Secondary Containment Control
I PBAPs EAL Techrdcal Bats Manual REV c. April 26,1995 Page 48 of 130 3.0 Primary Containment 3.1 Primary Containment GENERAL EMERGENCY - 3.1.4.a IC Loss of Reactor Coolant System and Containment wth potential loss of Fuel Clad EAL Containment Pressure > 60 psig OR Containment Venting via T-200 is required OPCON 11121:141s101 BASIS Should Primary Containment Pressure exceed 60 psig, the Containment Vent Pressure has been exceeded. This condition is reflective of failure of all Containment Pressure Control Systems and a loss of the pressure suppression mode. At this pressure the potential exists for uncontrolled and unpredictable breach of primary containment integrity and release of radioactivity to the environment.
This condition is also indicative of loss of the Reactor Coolant System fission product boundary since the only mechanism to pressurize containment to this pressure would be through the loss of the Reactor Coolant System.
This pressure is well above the maximum pressure expected to be present in primary containment l during a design basis Loss of Coolant Accident. Before reaching this pressure procedural ,
i guidance is provided to vent the primary containment regardless of offsite dose consequences.
A controllea, monitored radiological release is preferred to an unisolable, uncontrolled, and potentially unmonitored release which could result from a containment failure. Entry into this EAL is also warranted when the primary containment is vented at pressures lower than 60 psig.
Containment failure could also result in subsequent loss of adequate core cooling due to loss of source water for ECCS, loss of ECCS pumps or piping integrity, or loss of Reactor Pressure Vessel support and integrity.
Thus this event reflects the loss of the Reactor Coolant and Containment Barriers and the potential loss of the Fuel Clad Barrier.
DEVIATION None
.. - .-. - . . . - .-. . . - . . ~ . . . = ,
PBAPS EAL Technical Bats Manual REV c, Apnl 26,1995 -
Page 49 of 130
' REFERENCES NUMARC NESP-007, FC EAL #4, RC EAL # 5, PC EAL #2 T-104, Radioactwity Release Control
.T-112, Emergency Blowdown T-102, Primary Containment Control, PC/P-18 i
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l PBAPS EAL Tochtscal Basis Manual REVQ April 26,1995 Page 50 of 130 3.0 Primary Containment 1
3.1 Primary Containment !
GENERAL EMERGENCY - 3.1.4.b I IC Loss of Fuel Clad and Reactor Coolant System and a potentialloss of Containment EAL Drywell Hydrogen 2 6 %
AND Drywell Oxygen :t 5 %
OPCON 11121 1d18101 BASIS This EAL is indicative of a potential loss of the Primary Containment fission product barrier, as well as the loss of the fuel clad and the RCS barriers.
Hydrogen gas concentrations exceeding 6% in the containment is representative of degraded fuel clad [approximately 15%], reactor coolant system leak and a challenge condition to containment.
This level of hydrogen is not expected to indicate an uncoolable core condition. If oxygen concentration exceeds 5%, an explosive mixture would exist and provide the potential for detonation. Hydrogen gas concentrations of this level are as a result of extensive zirc-water reaction, indicating loss of the fuel clad and loss of the reactor coolant system. It is unlikely that containment oxygen concentrations would exist due to the containment being inerted with nitrogen at power operations. The significant mitigative action is in the TRIPS is to require requiro Emergency Blowdown on the following conditions H 2 2 6% and O,2 5%.
A General Emergency is warranted because a LOCA and fuel clad damage have occurred with a potential for loss of containment due to an explosive mixture. This accident has produced significant levels of hydrogen, combined with high levels of oxygen would potentially produce an exalosive mixture. With levels of hydrogen and oxygen this extreme there is a precursor to pc ' ntial loss of containment and therefore warrants a General Emergency.
DEVIATION None
PBAPS EAL Tochincal Basis Manud REV c. April 26,1995 Page $1 of 130 REFERENCES NUMARC NESP 007, FC EAL #4, RC EAL #5, PC EAL #1 >
T 102, Primary Containment Control NUREG/BR 0150 Vol 1, Rev 3, Response Technical Manual (RTM-93) l I
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f PBAPS EAL Techrucal Bads Manual REV c, April 26,1995 Page 52 of 130 This page intentionally left blank
PBAPs EAL TechnH:11 Basis Manual REV a, Apnl 26,1995 i Prge 53 of 130 4.0 Secondary Containment 4.1 Secondary Containment Temperature ALERT - 4.1.2 IC Potentialloss of Reactor Coolant System and Potentialloss of Containment EAL !
An Unisolable Primary System Leak is discharging into Secondary Containment AND A T-103 Temperature Action Levelis exceeded in ONE area requiring a SCRAM OPCON lilalaisisiol BASIS This EAL represents a challenge to both the Reactor Coolant and Containment barriers. The case of single area exceeding their Temperature Action Levels indicates that there is a potential bypass of primary containment, as well as the potential loss of the reactor coolant pressure boundary by either a breech in high energy piping without isolation or interfacing systems LOCA.
Increase in temperature in only one area indicates that the size of the leak is small enough to not cause a direct flow path to the environment. Temperature Action Levels limits are located in T-103.
TRIP guidance stipulates that when the Temperature Action Levels limit has been exceeded for ONE areas that the reactor be manually SCRAMMED.
There are two ways that the temperatures in the Se' adary Containment can reach these levels; l
l 1.e. primary leak into secondary and a fire within th condary containment. As the temperatures i rise above normal conditions, the plant staff will isolate the containment and all systems, except i those required for shutdown and cooling, at the Temperature Action Levels isolation levels. If the temperatures continue to rise to the Temperature Action Levels it is indicative that an unisolable leak has occurred.
If an instrument line ruptures and cannot be isolated, these temperatures could reach the action levels and an ALERT is warrented.
This event will be escalated to a Site Area Emergency based upon verification of more than one area exceeding Temperature Action Levels per EAL Section 4.1.3.
DEVIATION None
l PBAPS EAL Technical BaJs Mann:1 REV e, April 26,1996 Page 64 of 130' !
REFERENCES NUMARC NESP-007, RC EAL #5, PC EAL #5 i T-103, Secondary Containment Control, SSC/T-1 ;
T-101, RPV Control T-112, Emergency Blowdown I
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i PBAPs EAL Technmal Basis Manud REV a, Apnl 26,1995 Pag) 55 ct 130 4.0 Secondary Containment 4.1 Secondary Containment Temperature SITE AREA EMERGENCY - 4.1.3 IC Loss of Reactor Coolant System and Containment EAL An Unisolable Primary System Leak is discharging into Secondary Containment AND T-103 Temperature Action Levels are exceeded in TWO OR MORE areas requiring an Emergency Blowdown per T-112 OPCON 11121:141s1o1 BASIS This EAL represents a challenge to both the Reactor Coolant and Containment barriers. The case of multiple areas exceeding their Temperature Action Levels indicates that there is a bypass of primary containment, as well as the loss of the reactor coolant pressure boundary by either a breech in high energy piping without isolation or interfacing systems LOCA. Temperature Action Levels limits are located in T-103.
TRIP guidance stipulates that when the Temperature Action Levels limit has been exceeded for TWO OR MORE areas that the reactor be manually SCRAMMED and that an emergency blowdown be performed. The increase in Reactor Enclosure temperature in multiple areas may be an indication of a wide spread problem which may pose an immediate and direct threat to the Reactor Enclosure integrity, equipment and continued safe operation of the plant.
There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e. primary leakinto secondary and a fire within the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except ,
those required for shutdown and cooling, at the Temperature Action Levels isolation levels. If the temperatures continue to rise to the Temperature Action Levels it is indicative that an unisolable leak has occurred. i This event will be escalated to a General Emergency based upon verification of Fuel Clad degradatiuon per EAL Section 4.1.4.
DEVIATION None
. PSAPS EAL Technical Basis Manual l REV a April 26,1995 l i
Page 56 of 130 REFERENCES ,
NUMARC NESP-007, RC EAL #5 and PC EAL #2 - l T-103, Secondary Containment Control, SSC/T-1 !
T-101, RPV Control T-112, Emergency Blowdown :
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Page 57 of 130 4.0 Secondary Containment 4.1 Secondary Containment Temperature GENERAL EMERGENCY - 4.1.4 IC Loss of Fuel Clad, Reactor Coolant System and Containment EAL I
An Uninotable Primary System Leak is discharging into Secondary Containment AND T-103 Temperature Action Levels are exceeded in TWO OR MORE areas requiring an Emergency Blowdown per T-112 AND T-103 Radiation Action Levels are exceeded in the same TWO OR MORE areas OPCON 1i1218I416101 BASIS This EAL represeres a challenge to the Reactor Coolant, Containment and Fuel Clad barriers. The case of multiple areas exceeding their Temperature Action Levels indicates that there is a bypass of primary containment, as well as the loss of the reactor coolant pressure boundary by either a breech in high energy piping or an interfacing systems LOCA. T-103 Identifies secondary containment areas with their Temperature Action Levels and their Radiation Action Levels. For Radiation levels would not increase above these action levels unless fuel clad damage had occurred.
TRIP guidance stipulates that when the Temperature or Radiation Action Levels have been exceeded for two or more areas that the reactor be manually scrammed and that emergency blowdown be performed. The increase in reactor enclosure temperature or radiation in multiple areas may be an indication of a wide spread problem which may pose an immediate and direct threat to Reactor Building integrity, equipment and continued safe operation of the plant.
There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e. primary leak into secondary and a fire within the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except those required for shutdown and cooling, at the T-103 Temperature Action Levels. If the temperatures continue to rise to the Temperature Action Levels, it is indicative that an unisolable leak has occurred, and if the radiation levels rise above the Radiation Action Levels it is an indication that fuel damage has occurred.
DEVIATION None
PBAPS EAL Technical Bais Manual REV c, April 26,1995 Page 58 of 130 REFERENCES NUMARC NESP-007, FC EAL #4, RC EAL #5 and PC EAL #2 T-103, Secondary Containment Control T-101, RPV Control T-112, Emergency Blowdown
l PBAPs EAL Technicd Basis Manual REV a. April 26,1995 Page 59 of 130 4.0 Secondary Containment 4.2 Main Steam Line UNUSUAL EVENT - 4.2.1 IC Fuel Clad Degradation EAL Main Steam Line HiHi Radiation (10xNFPB)
OPCON lil218Idislol BASIS Main Steam Line High-High Radiation alarm (2(3)-252,A,B,C,D and 2(3)-251,A,B,C,D) > 10 times normal full power background maybe indicative of minor fuel cladding degradation and warrants the declaration of an Unusual Event. This levelis set to preclude most spurious events including resin intrusion.
The main steam line high-high radiation condition requires a manual Main Steam isolation Valve closure and a reactor scram. This transient may result in the introduction of fission product gases (previously contained in the gap area) to be suddenly released into the coolant due to the rapid down power transient and subsequent collapse of voids in the coolant.
This level of steam line activity is indicative of the release of gap activity to the coolant however, this level is not Indication of a major failure of the fuel clad. The mechanics that caused main steam line radiation to increase to this level indicate there is a degradation of fuel clad integrity.
This event will escalate to an Alert based on the breach in the main steam line together with a failure of the MSIVs to isolate the main steam lines per EAL Section 4.2.2.
DEVIATION None REFERENCES NUMARC NESP-007, SU4.1 T-099, Post Scram Recovery T-100, Scram
I PBAPs EAL Technied cuis Manud REV a. Apnl 26,1995 Page 60 of 130 4.0 Secondary Containment 4.2 Main Steam Line !
1 ALERT - 4.2.2 BC Potential Loss of Reactor Coolant System and Containment EAL Failure of one or more Main Steam Lines to isolate on any MSIV Closure Signal l
OPCON 181218I815101 BASIS The MSIV Closure Signal statement identifies that the MSIVs should be closed. The reason for the isolation signalis not relative to the basis of this EAL. The fact that the MSIV's failed to isolate the main steam line indicates the potential loss of the reactor coolant pressure boundary as well as a bypass potential of the primary containment.
Control room personnel should be attentive to indications as to if and where the steam is flowing.
Depending on the nature of the isolation signal. Radiation levels in the turbine enclosure and elsewhere may stay elevated. Actions should be taken to isolate the MSIVs as soon as possible.
10 CFR 100 release calculations assume the operability of the MSIVs and as such a steam line break may result in the release of radioactive materialin excess of the limits specified in 10 CFR 100.
This EAL is set below the threshold for determining that the RCS and Containment Fission Product Barriers can no longer be considered intact; however, it represents a serious condition in that two of the three Fission Product Barriers are affected. A direct pathway from the reactor to the environment would result if there is a breach in the steam system.
Escalation of this event would be through identification of a significant leak downstream of the MSIVs indicative of actual or potential loss of both RCS and Containment Fission Product Barriers per EAL Section 4.2.3.
DEVIATION None
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I
' PBAPS EAL TechnerJ B: sis Manual l REV e, April 26,1995 {
Page 61 of 130 l REFERENCES !
t NUMARC NESP-007, RC EAL #5, PC EAL #2 !
10 CFR 100 T-102, Primary Containment Control - l T-103, Secondary Containment Control l T-104, Radioactivity Release Control I
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PBAPs EAL Techrucal Bais Manud REv c. April 26,1995 Page 62 of 130 4.0 Secondary Containment 4.2 Main Steam Line SITE AREA EMERGENCY - 4.2.3 IC Loss of Reactor Coolant System and Containment EAL Main Steam Line Break discharging into the Turbine Building AND Vent Stack > 1x10' pCI/cc
~
OPCON lil218ldisjoi BASIS This event signifies that there is a direct path established for the transfer of main steam to inside the Turbine Building. Assumptions made in dose calculations regarding radioactive material transport (i.e. hold up, plate out, scrubbing, and retention) may be invalid. Additionally the transport time associated with a radiological release may be significantly shortened and there may be a higher percentage of short lived radioisotopes in any release. As both the reactor coolant pressure boundary and the primary containment are degraded; the extent of radioactive release is dependent on fuel clad integrity. Should a rapid reactor depressurization occur as a result of this event then there is a potential that a large amount of radioiodine may be released.
The Steam Line Break discharging into the Turbine Building is indicative of lost Reactor Coolant System and lost Containment Fission Product Barriers. The term " discharging" indicates that the break cannot be isolated. The Vent Stack > 1x10'* Ci/cc provides confirmation of the Main Steam Line Break and inFcation that the break is large enough to be significant. This value is based on a dose assessment calculation using average annual meteorological conditions in order ,
to obtain a maximum projected offsite TPARD Dose of 100 mrem.
l The following are the inputs and assumptions of the calculation-Software: MESOREM, Jr ver 8.0 Stability Class: E Vent Stack Flow: 225,000 cfm Wind Direction: 22 degrees
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l Release Duration: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Wind Speed: 6.30 mph I Accident Type: DBA LOCA Delta Temperature: 1.50 'F i l
This event will be escalated to a General Emergency based on the indication that the fuel clad integrity has also been threatened or lost per EAL Section 4.2.4.
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'l PBAPS EAL Technicrtl Basis Manud ,
REV t. April 26,1995 (
Page 63 of 130
- DEVIATION None REFERENCES NUMARC NESP-007, PC EAL #2, RC EAL #5 T-103 Secondary Containment Control
' T-104, Radioactivity Release Control t
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l PBAPS EAL Technical Bads Manual REV o. Apnl 26,1995 Page 64 of 130 4.0 Secondary Containment 4.2 Main Steam Line GENERAL EMERGENCY - 4.2.4 IC Loss of Fuel Clad, Reactor Coolant System and Containment EAL Main Steam Line Break discharging into the Turbine Building AND Vent Stack > 1x10 pCl/cc OPCON lil213Idl81N BASIS This event signifies that there is a direct path established for the transfer of main steam to inside the Turbine Building. Assumptions made in dose calculations regarding radioactive material transport (i.e. hold up, plate out, scrubbing, and retention) may be invalid. Additionally the transport time associated with a radiological release may be significantly shortened and there may be a higher percentage of short lived radioisotopes in any release. As both the reactor coolant pressure boundary and the primary containment are degraded; the extent of radioactive release is dependent on fuel clad integrity. As there is indication of fuel clad integrity loss there is established a path for fission products to escape directly to the environs (bypassing primary and .
secondary containment). Thus all three fission product retention barriers have been compromised. l The Steam Line Break discharging into the Turbine Building is indicative of lost Reactor Coolant l System and lost Containment Fission Product Barriers. The term " discharging" indicates that the i break cannot be isolated. The Vent Stack > 1x10' pCi/cc provides confirmation of the Main Steam Line Break and indication that there is loss of the Fuel Clad Barrier. This value is based on a dose assessment calculation using average annual meteorological conditions in order to obtain a maximum projected offsite TPARD Dose of 1 rem.
The following are the inputs and assumptions of the calculation:
Software: MESOREM, Jr ver 8.0 Stability Class: E Vent Stack Flow: 225,000 cfm Wind Direction: 22 degrees Release Duration: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Wind Speed: 6.30 mph Accident Type: DBA LOCA Delta Temperature: 1.50 F DEVIATION None
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PBAPS EAL Technical Bads Manual REV a. April 26,1995
' Page 65 of 130 REFERENCES NUMARC NESP 007, FC EAL #4, RC EAL #5 and PC EAL #2 Technical Specification T 103, Secondary Containment Control T 104, Radioactivity Release Control '
T-112, Emergency Blowdown l
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PBAPS EAL Technical Bais Manual REV c, April 26,1995 Page 66 of 130 This page intentionally left blank
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PBAPs EAL TechncJ Basis Manud REv a. April 26,1995 Page 67 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose UNUSUAL EVENT - 5.1.1.a IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer EAL Main Stack, Vent Stack, or Torus Hardened Vent Rad monitor continuously in HiHi Alarm OR known Unmonitored Release continuously in progress for > 60 minutes AND Calculated maximum offsite dose rate exceeds 0.114 mrem /hr TPARD OR 0.342 mrem /hr child thyroid CDE based on a 60 minute average i
OPCON l i l 21
- l d l 5 bl BASIS Releases in excess of 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE that continue for > 60 l minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The final integrated dose is very low and is not the primary concern. Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.
It is not intended that the release be averaged over 60 minutes, but exceed 0.114 mrem /hr i TPARD or 0.342 mrem /hr CDE limits for 60 minutes. For a monitored release, this EAL includes l a 60 minute average for the dose projection with the release point radiation monitor above the i HiHi alarm set point for the entire 60 minutes. For an unmonitored release [ie. blowout panel, inoperable radiation monitors), this EAL entry requires a known continuous unrnonitored release lasting for greater than 60 minutes. Also, it is intended that the event be declared as soon as it is determined that the release will exceed 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE for greater than 60 minutes.
An indication or report is considered to be valid when it is verified by:
- 1. An instrument channel check
- 2. Indications on related or redundant instruments
- 3. By direct observation by plant personnel The response to the radiological monitor HiHi alarm condition in the control room is dose assessment entering their procedures to determine if dose projections are necessary (NOTE: the HiHi alarm is conservatively set to be significantly below an Unusual Event value]. If determined, dose projections utilizing the monitoring readings and actual meteorology will be performed. This projection will determine which classification, if any, is warranted.
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PBAPS EAL Technical 3 asia Manual REV e. April 26,1995 Page 68 of 130 The Total Protective Action Recommendation Dose (TPARD) is calculated by dividing the yearly ;
allowable Technical Specification limit (500 mrem /yr) by the number of hours per year (8760 ,
hr/yr), and then multiplying by a factor of 2 times Technical Specifications (ODCM). ,
TPARD = 2x(Tech Spec Limit)/(hours per year)
= 2(500 mrem /yr)/(8760 hr/yr)
= 0.114 mrem /hr :
The Committed Dose Equivalent (CDE) in calculated by dividing the yearly allowable Technical Specification limit (1500 mrem /yr) by the number of hours per year (8760 hr/yr), and then -
multiplying by a factor of 2 times Technical Specifications (ODCM). .
CDE = 2x(Tech Spec Limit)/(hours per year)
= 2(1500 mrem /yr)/(8760 hr/yr)
= 0.342 mrem /hr ,
This event will be escalated to an Alert when effluents increase per EAL Sections 5.1.2.a.
DEVIATION None REFERENCES NUMARC NESP-007, AU1.1, AU1.2 Offsite Dose Calculation Manual 4
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5.0 Radioactivity Release 5.1 Effluent Release and Dose l UNUSUAL EVENT - 5.1.1.b IC Any Unplanned Release of Liquid Radioactivity to the Environment that Exceeds Two l Times Radiological Technical Specifications for 60 Minutes or Longer EAL Report indicates Liquid Release exceeds TWO TIMES Tech Specs (T.S. 3.8.8.1) for > 60 minutes OPCON lil21slaisiol BASIS Releases in excess of two times technical specifications that continue for > 60 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The final integrated dose is very low and is not the primary concem. Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.
It is not intended that the release be averaged over 60 minutes, but exceed two times technical specifications limits for 60 minutes. Further, it is intended that the event be declared as soon as it is determined that the release will exceed two times technical specifications for greater than 60 minutes.
An indication or report is considered to be valid when it is verified by:
- 1. An instrument channel check
- 2. Indications on related or redundant instruments
- 3. By direct observation by plant personnel The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm setpoints, etc).
This event will be escalated to an Alert when effluents increase per EAL Sections 5.1.2.b.
DEVIATION None
- i. .
PSAPS EAL Tochtml Beds Manual i
REV e, April 26,1995 Page 70 of 130 REFERENCES , ,.
i NUMARC NESP-007 AU1.2 Offsite Dose Calculation Manual - -!
T-104, Radioactivity Release Control _.
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I PBAPS EAL Technied Bisis Manuil
'- REv t Apnl 26,1995 Page 71 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose ALERT - 5.1.2.a IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EAL Main Stack, Vent Stack, or Torus Hardened Vent Rad monitor continuously in HiHi Alarm OR known Unmonitored Release continuously in progress for > 15 minutes AND Calculated maximum offsite dose rate exceeds 11.4 mrem /hr TPARD OR 34.2 mrem /hr child thyrold CDE based on a 15 minute average OPCON 1112181415101 BASIS Releases in excess of 11.4 mrem /hr TPARD or 34.2 mrem /hr CDE that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The primary concern is the final integrated dose [100 times greater than the Unusual Event) and the degradation in plant control implied by the fact that the release was not isolated within 15 minutes.
This EAL includes a 15 minute average for the dose projection. For a monitored release, this EAL includes a 15 minute average for the dose projection with the release point radiation monitor above the HiHi alarm set point for the entire 15 minutes. For an unmonitored release [ie blowout panel, inoperable radiation monitors), this EAL entry requires a known continuous unmonitored release lasting for greater than 15 minutes. Also, it is intended that the event be declared as soon as it is determined that the release will exceed 11.4 mrem /hr TPARD or 34.2 mrem /hr CDE for greater than 15 minutes.
An indication or report is considered to be valid when it is verified by:
- 1. An instrument channel check
- 2. Indications on related or redundant instruments
- 3. By direct observation by plant personnel The response to the radiological monitor HiHi alarm condition in the control room is dose assessment entering their procedures to determine if dose projections are necessary (NOTE: the HiHi alarm is conservatively set to be significantly below an Unusual Event value). If determined, dose projections utilizing the monitoring readings and actual meteorology will be performed. This ,
projection will determine which classification, if any, is warranted. j i
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i PBAPS EAL Technical Basis Manual REV c. April 26,1995 Page 72 of 130
. The Total Protective Action Recommendation Dose (TPARD) is calculated by dividing the yearly allowable Technical Specification limit (500 mrem /yr) by the number of hours per year (8760 hr/yr), and then multiplying by a factor of 200 times Technical Specifications [ODCM]. ,
TPARD = 200x(Tech Spec Limit)/(hours per year)
= 200(500 mrem /yr)/(8760 hr/yr)
= 11.4 mrem /hr The Committed Dose Equivalent (CDE) is calculated by dividing the yearly allowable Technical Specification limit (1500 mrem /yr) by the number of hours per year (8760 hr/yr), and then multiplying by a factor of 200 times Technical Specifications [ODCM).
CDE = 200x(Tech Spec Limit)/(hours per year)
= 200(1500 mrem /yr)/(8760 hr/yr)
= 34.2 mrem /hr This event will be escalated to a Site Area Emergency when actual or projected doses are determined to exceed 10CFR20 limits per EAL Sections 5.1.3.
DEVIATION None REFERENCES NUMARC NESP 007 AA1.1, AA1.2 Offsite Dose Calculation Manual I
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i PBAPS EAL Technicci Basis Manut!
REV a. Apnl 26,1995 Page 73 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose ALERT - 5.1.2.b IC Any Unplanned Release of Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EAL Report indicates Liquid Release exceeds TWO HUNDRED TIMES Tech Specs (T.S.
3.8.B.1) for > 15 minutes OPCON l'I218ldislol t
BASIS Releases in excess of two hundred times technical specifications that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The primary concern is the final integrated dose [100 times greater than the Unusual Event) and the degradation in plant control implied by the fact that the release was not isolated within 15
- minutes, it is not intended that the release be averaged over 15 minutes, but exceed two hundred times technical specifications limits for 15 minutes. Further, it is intended that the event be declared as soon as it is determined that the release will exceed two hundred times technical specifications for greater than 15 minutes.
An indication or report is considered to be valid when it is verified by:
- 1. An instrument channel check
- 2. Indications on related or redundant instruments
- 3. By direct observation by plant personnel The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm setpoints, etc).
This event will be escalated to higher classifications based on plant conditions.
DEVIATION None
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l PBAPS EAL Technical Batis Manuel REV a, April 26,1995 j Page 74 of 130 REFERENCES ;
NUMARC NESP-007 AA1.2 Offsite Dose Calculation Manual T-104, Radioactivity Release Control !
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PBAPs E AL Technics.1 Basis Manu11 REV a. Apnl 26,1995 !
Pags 75 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose SITE AREA EMERGENCY - 5.1.3 Boundary Dose Resulting from an Actual or Imminent Release of Gaseous IC Radioactivity Exceeding 100 mR Whole Body or 500 mR Child Thyroid for the Actual )
or Projected Duration of the Release using actual meteorology EAL Projected offsite dose exceeds 100 mrem TFARD, OR, Projected offsite dose exceeds 500 mrem child thyroid CDE, OR, Actual offsite whole body dose rate exceeds 100 mrem /hr OPCON lil2151dl5lol BASIS An actual or projected dose of 100 mrem Total Protective Action Recommendation Dose (TPARD) is based on the 10 CFR 20 annual average population exposure limit. TPARD is the sum of Extemal Dose Equivalent + Committed Effective Dose Equivalent + 4-day Deposition Exposure.
This value also provides a desirable gradient (one order of magnitude) betwecn the Site Area Emergency and General Emergency classifications. The 500 mrem integrated child thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TPARD and Child Thyroid Committed Dose Equivalent (CDE). Actual meteorology is used since it gives the most accurate dose projection.
Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with Pennsylvania Emergency Management Agency (PEMA) / Bureau of Radiation Protection (BRP).
This event will be escalated to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines per EAL Section 5.1.4.
DEVIATION None REFERENCES NUMARC NESP-007, AS1.1, AS1.3 and AS1.4 EPA 400
l PBAPS EAL Technical Ba 's Manual REV c. April 26,1995 Page 76 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose GENERAL EMERGENCY - 5.1.4 IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mR Whole Body or 5000 mR Child Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology ;
EAL Projected offsite dose exceeds 1000 mrem TPARD, OR Projected offsite dose exceeds 5000 mrem child thyroid CDE, OR Actual offsite whole body dose rate exceeds 1000 mrem /hr OPCON lil2181dl51ol BASIS Actual or projected dose exceeding 1000 mrem Total Protective Action Recommendation Dose (TPARD) or 5000 mrem child thyroid CDE is greater than the EPA Protective Action Guidance.
TPARD is the sum of External Dose Equivalent + Committed Effective Dose Equivalent + 4-day Deposition Exposure. At these doses, public protective actions are required, consistent with the emergency class description of a General Emergency.
Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with Pennsylvania Emergency Management Agency (PEMA) / Bureau of Radiation Protection (BRP).
Actual meteorology is used since it gives the most accurate dose projection.
DEVIATION ,
None REFERENCES NUMARC NESP-007, AG1.1, AG1.3 and AG1.4 EPA-400
PBAPs EAL Technical Ba:!s Manual REV a. April 26,1995 Page 77 of 130 5.0 Radioactivity Release 5.2 in-Plant Radiation l i
UNUSUAL EVENT - 5.2.1 IC Unexpected increase in Plant Radiation or Airbome Concentration l
EAL Inplant radiation level > 1x10' mR/hr requiring T-103 entry i
OPCON 1512181dislol BASIS The value of 1x10' mR/hr indicates a radiation level of approximately 1000 times normal.
Unplanned increases in in-plant radiation levels represent a degradation in the control of radioactive material and represent a degradation in the level of safety of the plant. Planned evolutions which cause elevated radiation levels are not covered by this EAL. T-103 identifies the systems that interface with the reactor coolant system.
An area monitor reading is considered to be valid when it is verified by:
- 1. an instrument channel check indicating the monitor has not failed;
- 2. Indications on related or redundant instrumentation; or
- 3. direct observation by plant personnel This event will be escalated to an Alert when radiation levels increase in areas required for the safe shutdown of the plant resulting in impeded access per EAL Section 5.2.2.a.
DEVIATION None REFERENCES NUMARC NESP-007, AU2.4 T-103, Secondary Containment Control
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PBAPS EAL Te chncal Basis Manual REv a. Apnl 26,1995 Page 78 of 130 5.0 Radioactivity Release 5.2 In-Plant Radiation ALERT - 5.2.2.a IC Release of Radioactive Material or increases in Radiation Levels Within the Facility That Impeder Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL ,
inplant radiation level > 9x10' mR/hr for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requiring T-103 entry OPCON lil2laidislol BASIS The EAL addresess radiation levels which would impede operation of systems required to maintain safe operations or to establish or maintain cold shutdown. Radiation levels could be indicated by ARM or radiological survey. T-103 lists the areas in the plant that locate systems interfacing with the reactor coolant system.
An area monitor reading is considered to be valid when it is verified by:
- 1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or,
- 3. Direct observation by plant personnel.
Unplanned increases inplant radiation levels represent a degradation in the control of radioactive materials and represent a degradation in the level of safety of the plant. Planned evolutions which cause elevated radiation levels are not covered by this EAL.
Access to the areas listed in T-103 may be necessary to perform manual actions to achieve or malatain cold shutdown.
This event will be escalated to a Site Area Emergency when loss of control of radioactive materials cause significant offsite doses per EAL Section 5.1.3.
DEVIATION None
PBAPS EAL Technicil Ba-Js Manual REV e, April 26,1995 Page 79 of 130 REFERENCES NUMARC NESP-007, AA3.2 T-103, Secondary Containment Control
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PBAPS EAL Technkd Basis Manual REv O. Apnl 26,1995 Page 80 of 130 5.0 Radioactivity Release 5.2 In-Plant Radiation ALERT - 5.2.2.b IC Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations EAL Control Room area radiation level > 15 mR/hr for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OPCON 111 18I415101 BASIS The EAL address radiation letels which would impede operation of systems required to maintain safe operations or to establish or maintain cold shutdown. Radiation levels could be indicated by ARM or radiological survey.
The Control Room general area level is set at 15 mR/hr and was chosen because continuous occupancy is required. This is consistent with General Design Criteria 19, which addresses continuous occupancy of the Control Room for 30 days after a design accident. Additionally, since the Control Room is shielded this radiation level represents a serious loss of control of radioactive material.
This event will te escalated to a Site Area Emergency when loss of control of radioactive materials cause significant offsite doses per EAL Section 5.1.3.
DEVIATION )
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REFERENCES I
NUMARC NESP-007 AA3.1 l
'1 PBAPs EAL Technicd Basis Manud REV a, Apnl 26,1995 I 1
Page 81 of 130 6.0 Loss of Power !
UNUSUAL EVENT - 6.1.1.a l lC Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EAL Loss of ALL Offsite Power for >f5 minutes OPCON 1112181415101 BASIS This EAL addresses the loss of off-site AC power supplying the station. Off-site power is fed through transformers 2 startup and 3 startup to busses 2 emergency aux and 3 emergency aux.
Loss of off site power will cause a reactor scram and a containment isolation. All four (4) emergency Diesel Generators will be available to carry the essential loads for each unit (the four Diesel Generators are shared between each unit). Balance of Plant systems that would assist in plant operations (i.e. condensate pumps, etc.) may be unavailable due the loss of power.
Fifteen (15) minutes has been selected to exclude transient or momentary power losses.
However, an Unusual Event should be declared in less than 15 minutes if it can be determined <
in less than 15 minutes that the power loss is not transient or momerttary.
Although no fission product release baniers are directly affected by the loss of offsite power, the plant is more vulnerable to a complete loss of AC Power with the reliance on the emergency Diesel Generators to power emergency systems to remove heat and maintain the reactor core submerged and cooled.
Escalation of this event to an Alert would be based on having a loss of all offsite AC power coincident with onsite AC power being reduced to a single power source in Modes 1,2, and 3 or l having a loss of all offsite and onsite AC power in Modes 4 or 5 per EAL Section 6.1.2.
DEVIATION None REFERENCES NUMARC NESP-007, SU1 SE-11, Station Blackout I
PBAPS EAL Technscal Basis Manual REv c. Apnl 26.1995 Page 82 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power UNUSUAL EVENT - 6.1.1.b IC Unplanned Loss of Required DC Power During Cold Shutdodwn or Refueling Mode for Greater than 15 Minutes EAL Loss of ALL safety related DC Power indicated by < 107.5 VDC for > 15 minutes OPCON H18181415101 BASIS The purpose of this EAL 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. The safety related 125 volt DC Distribution Panels are as follows:
Unit 2 Unit 3 20D21 30D21 20D22 30D22 20D23 30D23 ;
30D24 :
20D24 107.45 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. The value of 107.5 VDC will be used for human factors concerns. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is near the minimum voltage selected when battery sizing is performed.
This event will escalate to an Alert if the loss results in the inability to maintain cold shutdown, per EAL Section 7.2.2.
DEVIATION ,
None REFERENCES NUMARC NESP 007, SU7 SE-13, Loss of a 125/250 VDC Safety Related Bus
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PGAPs EAL Techncol Basis Manual REV a. Apnl 26.1995 Page 83 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power ALERT - 6.1.2.a IC AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout EAL Loss of ALL Offsite Power for > 15 mlnufes AND Only ONE 4 KV Emergency Bus is available OPCON fil21al818101 BASIS This EAL is intended to provide an escalation from " Loss of offsite Power for greater than 15 minutes " This condition is a degradation of the offsits and onsite power systems such that any additional failure would result in a station blackout.
Depending on the 4 KV AC bus that remains energized there is a disparity in the systems that may be available. The ability to remove heat from the containment via Torus cooling may be lost due to the need to operate the 1 available RHR pump in other than Torus cooling (i.e. LPCI). As such there is a decrease in the systems available to remove heat transferred to the containment and there is an ongoing release of energy from the reactor to the containment (via SRVs, HPCI and/or RCIC operation).
Fifteen (15) minutes has been selected to exclude transient or momenbry power losses.
However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary.
The ability to cool the nuclear fuel, remove decay heat, and control containment parameters is severely limited. Should equipment be unavailable prior to the loss of power, functions necessary to maintain the plant in a cold shutdown condition may be threatened.
Escalation of this event would be based on the loss of the remaining Emergency Diesel Generator per EAL Section 6.1.3.
DEVIATION None
PBAPS EAL Technical Basia Manual REV c, April 26,1995 Page 84 of 130 HEFERENCES NUMARC NESP-007, SAS SE-11, Station Blackout
PBAPs EAL Technicil Basis Manual REV a. April 26,1995 Page 85 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power ALERT - 6.1.2.b IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode EAL Loss of ALL Offsite Power AND ALL 4 KV Emergency Busses are unavailable for > 15 minutes OPCON l81818141 0 01 BASIS Control room annunciators would indicate that all offsite and onsite AC power feeds have been lost. Loss of all AC power compromises all plant safety systems requiring electric powerincluding RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal, High Pressure Service Water, and Emergency Service Water. Although Instrumentation (supplied through instrument invertors) and DC power loads would be available, their operability would be limited to the amount of stored energy contained in their respective batteries, instrumentation, communication equipment, and in-plant lighting and ventilation will be significantly hampered by the loss of all AC power.
Fifteen (15) minutes has been selected to exclude transient or momentary power losses.
However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary.
When in cold shutdown, refueling, or defueled mode this event can be classified as an Alert, because of the significantly reduced decay heat, lower temperature and pressure. It is assumed that the plant will be maintained in a cold shutdown condition and that it the plant is not able to be maintained in this mode then escalation to Site Area Emergency would be appropriate and be based on EAL Sections 2.1 or 5.1.
DEVIATION None j REFERENCES NUMARC NESP-007, SA1 SE-11, Station Blackout 1
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PBAPS EAL Technical Ba2s Manual REV c. Apol 26.1995 Page 86 of 130 6.0 Loss of Power l
I SITE AREA EMERGENCY - 6.1.3.a IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EAL Loss of ALL Offsite Power AND ALL 4 KV Emergency Busses are unavailable for > 15 minutes OPCON lil218tdl8101 BASIS Control room annunciators would indicate that all offsite and onsite AC power feeds have been lost. Loss of all AC power compromises all plant safety systems requiring electric powerincluding RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal, High Pressure Service Water, and Emergency Service Water. Although instrumentation (supplied through instrument invertors) and DC power loads would be available, their operability would be limited to the amount of stored energy contained in their respective batteries. Instrumentation, communication equipment, and in-plant lighting and ventilation will be significantly hampered by the loss of all AC power.
Fifteen (15) minutes has been selected to exclude transient or momentary power losses.
However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary.
Escalation of this event would be based on the time that the Emergency Diesel Generator are unavailable per EAL Section 6.1.4.
DEVIATION None REFERENCES NUMARC NESP-007, SS1 SE-11, Station Blackout
l PBAPs EAL Technical Basis Manutt REv a, April 26.1995 P gi 87 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power SITE AREA EMERGENCY - 6.1.3.b IC Loss of All Vital DC Power EAL Loss of ALL safety related DC Power indicated by < 107.5 VDC for > 15 minutes OPCON 1112181815101 BASIS:
A loss of all DC power compromises the ability to monitor and control plant functions.125 volt DC system provides control power to engineered safety features valve actuation, diesel generator auxiliaries, plant alarm and indication circuits as well as the control power for the associated load group. If 125 volt DC power is lost for an extended period of time (greater than 15 minutes) critical plant functions such as RPS Logic, Alternate Rod Insertion, Emergency Service Water Indication,4KV Breaker Controls, HPCI, RCIC and RHR pump controls required to maintain safe plant conditions may not operate and core uncovery with subsequent reactor coolant system and primary containment failure might occur. The 125 volt DC Main Distribution Panel Busses are as follows:
Unit 2 Unit 3 20D21 30D21 20D22 30D22 20D23 30D23 20D24 30D24 Loss of all DC Power causes the loss of the following equipment:
. Altemate Rod Insertion . ADS HPCI - RCIC
- Normal EDG Control - Normal Recirculation Pump Trip
- Contairement Instrument Gas Compressors
. Other 4KV Circuit Breakers (RHR, CS, CRD)
Loss of ADS creates a loss of low pressure ECCS due to the inability to depressurize the reactor.
In addition, loss of these buses will eventually lead to MSIV closure and reactor trip due to the loss of the Containment Instrument Gas Compressor as a result of suction valve closure.
Subsequent to MSIV closure, much of the equipment noted above will be required for plant stabilization and shutdown.
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PBAPS EAL Techrncal Basis Manuil REV a April 26,1995 Page se of 130 A sustained loss of DC power will threaten the ability to remove heat from the reactor core, ,
resulting in eventual fuel clad damage. The loss of DC power will also result in the loss of the ability to remove heat from the containment. SRVs will remain operable in the relief mode and the heat addition to the containment could result in a loss of the primary containment as a fission product release barrier.
107.45 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This EAL uses 107.5 VDC for human factors concerns. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is near the minimum voltage selected when battery sizing is performed.
DEVIATION None REFERENCES NUMARC NESP-007, SS3 T-101, RPV Control T 102, Primary Containment Control SE-11, Station Blackout 4
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PBAPs EAL Technical Basis Manud REV a. April 26,1995 Dage 89 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power GENERAL EMERGENCY - 6.1.4 IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power EAL Loss of ALL Offsite Power AND ALL 4 KV Emergency Busses are unavailable for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> OPCON lil:Iaidisio]
BASIS 10 CFR 50.2 defines station blackout (SBO) as complete loss of AC power to essential and non-essential buses. SBO does not include loss of AC Power to busses fed by station batteries through inverters, nor does it assume a concurrent single failure or design basis accident.
Successful SBO coping maintains the following key parameters within given acceptable limits:
- 1. Reactor water level > -172" (TAF)
- 4. Containment pressure < 60 psig, design limit
- 5. Torus temperature < 200 F, HPCl/RCIC lube oil temperature concem when suction aligned to Torus
- 6. Drywell temperature
<200'F indefinitely
<250 F 99 days
<320*F 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />
<340 F 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Successful extended SBO coping depends on ability to keep HPCl/RCIC available for injection, and ability to maintain RPV depressurized for low pressure injection should HPCI and RCIC become unavailable. Control power for HPCI, RCIC and SRVs is provided by 125V DC. The parameters listed above can be maintained as long as the batteries are intact. Two hours is the earliest the batteries would fail, and thus is the basis for the time limit in this EAL.
The significance of a station blackout relative to the loss of fission product release barriers is that all three barriers will eventually be lost due to the inability to remove heat from the fuel and the containment. Although the RCS will be intact the longest, eventually SRVs will operate in the relief mode due to RPV overpressurization and if the containment has already failed then there is a l direct bypass of the RCS boundary. <
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' DEVIATION None
. REFERENCES '
NUMARC NESP-007, SG1 SE-11, Station Blackout T-101, RPV Control T-102, Primary Containment Control .
T-104, Radioactivity Release Control i
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REv a. Apnl 26,1995
< Page 91 of 130 7.0. Internal Events 7.1 Technical Specification & Control Room Evacuation UNUSUAL EVENT - 7.1.1 IC Inability to Reach Required Shutdown Within Technical Specification Limits EAL Unable to bring the Plant to the required Mode within Tech Spec LCO action times OPCON h l 2181 *l 5161
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BASIS Limiting Conditions of Operation (LCOs) require the plant to be brought 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 under 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 Unusual Event is required when it is determined that the plant cannot be brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO-specified action stateme'nt time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, SU2 Technical Specifications
PBAPS EAL Techrucal Basis Manu:2 REV c. Apnl 26,1995 Page 92 of 130 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation ALERT - 7.1.2 IC Control Room Evacuation Has Been Initiated EAL Control Room evacuation procedures have been initiated OPCON lil21sidislol BASIS Control Room evacuation requires establishment of plant control from outside the control room (local control and remote shutdown panel) and support from the Technical Support Center and/or other emergency facilities as necessary. Control Room evacuation represents a serious plant situation since the level of control is not as complete as it would be without evacuation. The establishment of system control outside of the Control Room will bypass many protective trips and interlocks. In addition, much of the instrumentation and assessment tools available in the Control Room will not be available.
This event will be escalated to an Alert if control cannot be established within fifteen minutes per EAL Section 7.1.3.
DEVIATION None REFERENCES NUMARC NESP-007, HAS SE-10, Alternate Shutdown SE-1 Plant Shutdown from the Remote Shutdown Panel
PBAPS EAL Technical Basis Manu1J REV a. April 26,1993 Ptga 93 of 130 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation SITE AREA EMERGENCY - 7.1.3 IC Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EAL Failure to establish Altemate/ Emergency Control of the Plant within 15 minutes after evacuation of the Control Room OPCON !i12121415101 BASIS Transfer of safety system control has not been performed in an expeditious manner but it is unknown if any damage has occurred to the fission product barriers. The 15 minute time limit for transfer of control is based on a reasonable time period for personnel to leave the control room, arrive at the remote shutdown area, and reestablish plant control to preclude core uncovery and/or core damage. During this transitional period the function of monitoring and/or controlling parameters necessary for plant safety may not be occurring and as a result there may be a threat to plant safety.
This event will be escalated based upon system malfunctions or damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, HS2 SE-10, Alternate Shutdown SE-1, Plant Shutdown from the Remote Shutdown Panel l
PBAPs EAL Technical Basis Manu:.1 REv c. April 26.1995 Page 94 of 130 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability UNUSUAL EVENT - 7.2.1 IC Inability to Maintain Plant in Cold Shutdown EAL Uncontrolled Reactor Coolant temperature increase to > 212 *F OPCON IIIsl*141slol BASIS: .
This EAL addresses complete loss of normal functions required for core cooling during refueling and cold shutdown modes. Uncontrolled means that system temperature increase is not the result of planned actions by the plant staff.
This EAL is concerned with the ability to keep the reactor core cooled less than 212 'F. The criteria of uncontrolled Reactor Coolant temperature increase > 212 F is met as soon as it becomes known that sufficient cooling cannot be restored in time to maintain the temperature <
212 *F regardless of the current temperature.
This event will be escalated to an Alert if attemate decay heat removal capability cannot be established per EAL section 7.2.2.
DEVIATION This EAL has been created to be a precursor to NUMARC NESP-007, SA3 which calls for an Alert to be declared for a loss of shutdown cooling function and reactor coolant temperature rises above 200 'F. The reactor core can continue to be properly cooled as long as Torus cooling is available. While this event clearly warrants the declaration of an Unusual Event, PBAPS does not feelit meets the threshold for an Alert.
REFERENCES NUMARC NESP-007, SA3 GP 12, Core Cooling Procedure Technical Specifications
PBAPS EAL Technic;l Basis Manual REV c. April 26,1995 P:g3 95 of 130 7.0 Internal Events i 7.2 Loss of Decay Heat Removal Capability ALERT - 7.2.2 IC Inability to Maintain Plant in Cold Shutdown EAL
~ Uncontrolled Reactor Coolant temperature increase to > 212 *F AND inability to establish alternate decay heat removal capability OPCON hlal*1dl8101 BASIS This EAL addresses complete loss of normal functions required for core cooling during refueling and cold shutdown rnodes. Uncontrolled rneans that system temperature increase is not the result of planned actions by the plant staff.
This EAL is concerned with the ability to keep the reactor core temperature less than 212 F. The criteria cf uncontrolled Reactor Coolant temperature increase > 212 F is met as soon as it becomes known that sufficient cooling cannot be restored in time to maintain the temparature <
212 F, regardless of the current temperature. The inability to establish Alternate methods of decay heat removal indicates that either alternate methods are unavailable to cool the core in the RPV or when the steam is transferred to the Torus, Torus cooling is unavailable. Loss of Torus cooling will result in a continuing, uncontrolled increase in reactor coolant temperature.
This event will be escalated to a Site Area Emergency if boiling of the reactor coolant reduces level and produces high radiation and/or fuel damage as identified in EAL sections 1.2,1.3,2.1, and 5.1.
DEVIATION None REFERENCES NUMARC NESP-007, SA3 GP-12, Core Cooling Procedure Technical Specifications
PBAPS EAL Technical Basis Manual REv O, Apnl 26,1995 Page 96 of 130 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability SITE AREA EMERGENCY - 7.2.3 IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EAL Loss of Main Condenser as a heat sink AND Loss of TORUS Heat Sink capabilities as evidenced by T-102 legs [T/T, T/L, PC/P, or DW/T] requiring an Emergency Blowdown OPCON l11218I41s101 BASIS:
This EAL addresses complete loss of functions required to reach cold shutdown from MODE 1, 2 or 3.
The normal method for rejecting heat during operation is via the Main Condenser. If the Main Condenser is not available, heat may be rejected directly to the TORUS utilizing SRVs. The number of SRVs required to reduce pressure will be dependent upon reactor pressure and power.
A low TORUS level would result in Heat Capacity Temperature Limit (HCTL) being exceeded if a full power blowdown occurred at water levelin the TORUS. A high TORUS temperature would result in the TORUS being at the HCTL whereby it can no longer function as a heat sink. If the TORUS Level is at a high level the TORUS cannot handle a full power blowdown. T-102 requires an Emergency Blowdown before these TORUS conditions are reached to ensure the transfer of the energy to the TORUS. Without an Emergency Blowdown, reactor pressure cannot be reduced to the shutdown cooling pressure interlock of 75 psig and shutdown cooling cannot be established. Once the interlock is cleared, shutdown cooling can be utilized to reduce reactor coolant temperature to below 212 F.
DEVIATION None REFERENCES NUMARC NESP-007, SS4 T-102, Primary Containment Control, SP/L-8
l PBAPs EAL Technical Basis Manual REV c. April 26,1995 Pcgo 97 of 130 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability l UNUSUAL EVENT - 7.3.1.a f
IC Unplanned Loss of Most or All Safety System Annunciation or Indication in The Control Room for Greater Than 15 Minutes EAL Loss of All Annunciators in the Control Room for > 15 minutes I
OPCON l'l 181dlelol BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. in the opinion of the Shift Manager this loss of annunciators requires increased surveillance to safely operate the plant. Although loss of ALL annunciators is specified, if a large portion of annunciators or significant annunciators, as determined by the shift manager, are lost this EAL would then be appropriately entored. It is not intended that a detailed count of j instrumentation be performed, but that only a rough approximation be used to determine the severity of the loss. The Plant Monitoring System is available to provide compensatory indication.
Fifteen minutes is used as a threshold to exclude transient or momentary power losses.
Unplanned loss of annunciators excludes scheduled maintenance and testing activities. Control Room panels with annunciators and direction for response are included in ON 123, Loss of Control Room Annunciators.
Reportability of Technical Specification imposed shutdowns, or the inability to comply with Technical Specification action statements is covered in EAL section 7.1, Technical Specifications.
This EAL is not applicable in cold shutdown or refueling modes due to the limited number of ,
safety systems required for operation. '
This event will be escalated to an Alert if a transient is in progress or if compensatory indications become unavailable per EAL Section 7.3.2.
DEVIATION None
~
PSAPS EAL Tochtwcal Ba:Is Manud REV a, April 26,1995 Page 98 of 130 REFERENCES NUMARC NESP-007, SU3 ON-123, Loss of Control Room Annunciators AIT A0004447, EP Self Assessment on Salem Loss of Annunciators 1
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PBAPs EAL Technical Basis Manual REV a. Apnl 26,1995 Pig) 99 of 130 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability UNUSUAL EVENT - 7.3.1.b IC Unplanned Loss of All Onsite or Offsite Communications Capabilities EAL Loss of ALL Onsite communications (Table 7-1)
OR Loss of ALL Offsite communications (Table 7-1)
OPCON lil218141slol BASIS This EAL recognizes a loss of communication ability that significantly degrades the plant operations staff's ability to perform tasks necessary for plant operations or the ability to communicate with offsite authorities. This EAL is separated into two groups of communications, Onsite and Offsite. A complete loss of either group is so severe, that the Unusual Event declaration is warranted. Table 7-1 is identified as follows:
Table 7-1 Communications Onsite Offsite Site Phones (GTE System) X X OMNI System X X Plant Public Address X Station Radio X NRC (FTS-2000) X PA State Police Radio X Load Dispatcher Radio X PECO Dial Network X There is no escalation to an Alert for loss of communications, although there is escalation to higher classifications if other communications for plant assessment is lost.
DEVIATION None REFERENCES NUMARC NESP-007, SU6 Nuclear Emergency Plan
PBAPs EAL Technical Basis Manu:J REv a. April 26,1995 Page 100 of 130 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability ALERT - 7.3.2 IC Unplanned Loss of Most or All Safety System Annunciation or Indication In Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable EAL Loss of All Annunciators in the Control Room for > 15 minutes AND Significant Plant Transient (Table 7-2) is in progress OR Plant Monitoring System is unavailable -
OPCON 1512181416101 BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Manager this loss of annunciators requires increased surveillance to safely operate the plant. Although loss of ALL annunciators is specified,if alarge portion of annunciators or significant annunciators, as determined by the shift manager, are lost this EAL would then be appropriately entered. This EAL represents an increase in severity above 7.3.1.a in that the Plant Monitoring System can not provide compensatory indication, or that a significant transient is in progress. Table 7-2 significant plant transients include response to automatic or manually initiated actions including:
Table 7-2 Plant Transients SCRAM l Recirc runbacks > 25% thermal power thermal power oscillations of 10% or greater stuck open relief valves ECCS injection Fifteen minutes is used as a threshold to exclude transient or momentary power loses. Control Room panels with annunciators and direction for restoration is included in ON 123, Loss of Control Room Annunciators.
Reportability of Technical Specification imposed shutdowns, or the inability to comply with Technical Specification action statements is covered in section 7.1, Technical Specifications. ,
1 l
This EAL is not applicable in cold shutdown or refueling modes due to the limited number of safety systems required for operation.
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PBAPs EAL Technicil Basis Manual REV a, April 26,1995 Page 101 d 130 This event will be escalated to a Site Area Emergency if a transient is in progress, the Plant Monitoring System is unavailable and a loss of annunciators occurs per EAL Section 7.3.3.
DEVIATION None REFERENCES NUMARC NESP-007, SA4 [
ON 123, Loss of Control Room Annunciators l
C
l PBAPs EAL Technical Bads Manual REV c, April 26.1995 l Page 102 of 130 7.0 Internal Events l 7.3 Loss of Assessment / Communication Capability i SITE AREA EMERGENCY - 7.3.3 ,
IC Inability to Monitor a Significant Transient in Progress EAL Loss of All Annunciators in the Control Room for > 15 minutes AND Significant Plant Transient (Table 7-2) is in progress AND Plant Monitoring System is unavailable OPCON 111:181*15101 BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal <
annunciators, in the opinion of the Shift Manager this loss of annunciators requires increased surveillance to safely operate the plant. Although loss of ALL annunciators is specified, if a large portion of annunciators or significant annunciators, as determined by the shift manager, are lost this EAL would then be appropriate!y entered. This EAL represents an increase in severity above 7.3.2 in that the Plant Monitoring System can not provide compensatory indication, and that a significant transient is in progress. Table 7-2 significant plant transients include response to automatic or manually initiated actions including:
Table 7-2 plant Transients SCRAM Recire runbacks >25% thermal power change thermal power oscillations of 10% or greater stuck open relief valves ECCS injection Planned maintenance or testing activities are included in this EAL due to the significance of this event. Control Room panels with annunciators and the restoration is included in ON 123, Loss of Control Room Annunciators.
DEVIATION None
t PBAPS EAL Technical Basis Manud l REV t, Apnl 26,1995 Page 103 of 130 ,
l REFERENCES 1
.NUMARC NESP-007, SS6
- ON-123, Loss of Control Room Annunciators i
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PBAPS EAL Technical Bads Manual REV c. Apnl 26,1995 Page 104 of 130 This page intentionally left blank
i PBAPs EAL Technmal B: sis Manual REv a, Apnl 26,1995 Page 105 of 130 8.0 External Events 8.1 Security Events UNUSUAL EVENT - 8.1.1 IC Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EAL Confirmed security threat directed towards the station as evidenced by :
- Credible sabotage or bomb threat within the Protected Area, QR
- Credible intrusion and attack threat to the Protected Area, OR
- Attempted intrusion and attack to the Protected Area, OR
- Attempted sabotage discovered within the Protected / Vital Area, QR.
- Hostage / Extortion situation that threatens normal plant operations OPCON 11121.51416101 BASIS A security threat that is identified as being directed towards the station and represents a potential degradation in the level of safety of the plant. A security threat is satisfied if physical evidence supporting the threat exists, if information independent from the actual threat exists, or if a specific group claims responsibility for the threat. The Shift Manager will declare an Unusual Event subsequent to consulting with the Manager, Nuclear Security to determine the credibility of the security event.
Security threats which meet the threshold for declaration of an Unusual Event are:
- 1. Credible sabotage or bomb threat within the Protected Area
- 2. Credible intrusion and attack threat to the Protected Area
- 3. Attempted intrusion and attack to the Protected Area
- 4. Attempted sabotage discovered within the Protected / Vital Area
- 5. Hostage / Extortion situation that threatens normal plant operations Secu,ity events which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or 10 CFR 50.72 and will not cause an Unusual Event to be declared.
This event will be escalated to an Alert based upon a hostile intrusion or act within the Protected Area per EAL Section 8.1.2.
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l PBAPS EAL Technical Basis Manual ;
REV c. April 26,1995 i Page 106 of 130 j DEVIATION None REFERENCES
' NUMARC NESP-007, HU4.1 and HU4.2 Safeguards Contingency Plan Physical Security Plan 1
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l PBAPS EAL Technical Basis Manual REV c. Apnl 26.1995 Pige 107 of 130 8.0 External Events 8.1 Security Events ALERT - 8.1.2 IC Security Event in a Plant Protected Area EAL Confirmed hostile intrusion or act within the Protected Area as evidenced by
- Actual attack and intrusion into the Protected Area, OR
- Suspected bomb, sabotage or sabotage device discovered in the Protected / Vital Area l
(
OPCON I1121 1415l01 BASIS This class of security event represents an escalated threat to the level of safety of the plant. This event is satisfied if physical evidence supporting the hostile intrusion or attack exists. The Shift Manager will declare an Alert subsequent to consulting with the Manager Nuclear Security to determine the validity of the entry conditions.
Security threats which meet the threshold for declaration of an Alert are: l l
- 1. Actual attack and intrusion into the Protected Area j
- 2. Suspected bomb, sabotage or sabotage device discovered within the Protected Area This event will be escalated to a Site Area Emergency based upon a hostile intrusion or act in plant Vital Areas per EAL Section 8.1.3.
DEVIATION None REFERENCES NUMARC NESP-007, HA4.1 and HA4.2 Safeguards Contingency Plan Physical Security Plan
r PBAPs EAL Techrucal Bats Manual REv a. Aprv 26,1995 Page 108 of 130 8.0 External Events 8.1 Security Events SITE AREA EMERGENCY - 8.1.3 IC Security Event in a Plant Vital Area EAL Confirmed hostile intrusion or act in plant Vital Areas as evidenced by :
. Actual attack and intrusion into a Vital Area, OR
- Confirmed bomb, sabotage or sabotage device discover 6d in a Protected / Vital Area e
OPCON filrl314lslol BASIS This class of security event repres6nts an escalated threat to plant safety above that contained in an Alert in that a hostile intrusion or attack has progressed from the Protected Area to a Vital Area. The Vital Areas are within the Protected Area and are generally controlled by key card readers. These areas contain vital equipment which includes any equipment, system, device or material, the failure, destruction or release of could directly or indirectly endanger the public health and safety by exposure to radiation. Equipment or systems which would be required to function to protect health and safety following such failure, destruction or release are also considered vital.
Security threats which meet the threshold for declaration of a Site Area Emergency are:
- 1. Actual attack and intrusion into a Vital Area
- 2. Confirmed bomb, sabotage or sabotage device discovered in a Vital Area l This event will be escalated to a General Emergency based upon the loss of physical control of I the Control Room or Remote Shutdown Capability DEVIATION None REFERENCES NUMARC NESP-007, HS1.1 and HS1.2 Safeguards Contingency Plan Physical Security Plan l
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PBAPS EAL Technied Basis Manu:1 REV a. Apnl 26,1995 Page 109 of 130 8.0 External Events 8.1 Security Events i GENERAL EMERGENCY - 8.1.4 IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold Shutdown EAL Security event resulting in the actual loss of physical control of the:
Control Room ,OR Remote Shutdown Capability OPCON 111218141510I BASIS This class of security event represents conditions under which a hostile force has taken physical control of areas required to reach and maintain cold shutdown. Loss of Remote Shutdown Capability would occur if the control function of the Remote Shutdown Panels was lost.
Security events which meet the threshold for declaration of a General Emergency are physical loss of the Control Room or the Remote and Alternate Shutdown Panels. ,
This situation leaves the plant in a very unstable condition with a high potential of multiple barrier f failures.
DEVIATION None
- REFERENCES NUMARC NESP-007, HG1.1 and HG1.2 Safeguards Contingency Plan Physical Security Plan e
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PBAPs EAL Techncil Bats Manual REV c. Apnl 26,1995 Page 110 of 130 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.a IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection EAL Fire within ON-114 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of verification of alarms OPCON 1812131415101 BASIS This EAL addresses verified fires in Plant Vital Structures that house safety systems. These fires may be precursors to damage to safety systems contained in these structures. Verification that a fire exists is by operator actions to confirm that fire alarms received in the Control Room are not spurious or by any verbal notification by plant personnel. Fifteen minutes has been established to allow plant staff to respond and control small fires or to verify that no fire exists.
Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure inner Screen Structure Emergency Cooling Tower This event will be escalated to an Alert if the fire damages redundant trains of plant safety systems required for the current operating condition per EAL Section 8.2.2.a.
DEVIATION None REFERENCES NUMARC NESP-007, HU2 l
1
l PBAPS EAL Technied Basis Manud REV a. Apnl 26,1995 Page 111 of 130 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.b IC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the Plant EAL Normal plant operation is impeded by:
Toxic or flammable gas concentrations confirmed within the Protected Area OR Control Room informed by Local, County or State Officials to evacuate personnel within the protected area due to an offsite gas release ,
OPCON litzlaldl$bl BASIS This EAL addresses toxic / flammable gas releases within the Protected Area in concentrations high enough to affect health of plant personnel or the safe operation of the plant. This includes releases that originate both on-site and off-site. A toxic / flammable gas is considered to be any substance that is dangerous to life or limb by reason of inhalation or skin contact. A gas release is considered to be impeding normal plant operations if concentrations are high enough to restrict normal operator movements. it also includes areas where access is only possible with respiratory equipment, as this equipment restricts normal visibility and mobility. It should not be construed to include confined spaces that must be ventilated prior to entry or situation involving the Fire Brigade who are using respiratory equipment during the performance of their duties unless it also affects personnel not involved with the Fire Brigade.
A combustible gas maintained at a concentration lower than the Lower Explosive Limit will not explode due to ignition.
An off-site event (such as a tanker truck accident or train derailment releasing toxic gases) may place the Protected Area within the evacuation area. This evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials.
DEVIATION None i
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I PBAPS EAL Technical Bads Manu2 REV c. April 26,1995 Page 112 of 130 REFERENCES NUMARC NESP-007, HU3.1 and HU3.2
l PBAPs EAL Technied Basis Manud REv a, April 26,1995 Pig 113 of 130 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report by plant personnel confirming the occurrence of an explosion in a Plant Vital Structure (Table 8-1)
OPCON lil218ldl51ol BASIS This EAL addresses verified explosion in Plant Vital Structures that house safety systems and exhibit sufficient force to cause damage to Plant Vital Structures or safety systems. As used here, Explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment (Process storage tanks, gas cylinders, heat exchangers, etc.) that potentially imparts sufficient energy to damage nearby structures or materials. No attempt is made to assess the magnitude of the damage. The occurrence of the explosion with a report of damage (deformation / scorching) is sufficient for declaration. Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure '
Emergency Cooling Tower Any security aspects of this event should be considered under EAL Section 8.1, Security Events.
This event will be escalated to an Alert if the explosion damages one or more redundant trains of plant safety systems required for the current operating condition per EAL Section 8.2.2.
DEVIATION None
+ REFERENCES NUMARC NESP-007, HU1.5
l PBAPs EAL Techmcal Bais Manud REV e, Apnl 26,1995 Page 114 of 130 l l
8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases ;
ALERT - 8.2.2.a IC Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EAL Fire or explosion within Plant Vital Structures (Table 8-1) damages safe shutdown systems causing or jeopardizing plant shutdown OPCON 1112181416101 BASIS This EAL recognizes that damage has occurred to safe shutdown systems required to achieve or maintain cold shutdown in that equipment required for a particular safety function is either inoperable or incapable of performing its function. This EAL is entered if the fire causes an automatic shutdown of the plant or the fire damages safe shoutdown equipment jeopardizing a plant shutdown. The safe shutdown systems are housed in the Plant Vital Structures.
Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower Degraded system performance or observation of damage that could degrade system performance is used as the indicator that the safe shutdown system was actually affected. A report of 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 damage. The occurrence of the fire or explosion with reports of damage (e.g., deformation, scorching) is sufficient for declaration.
This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.
DEVIATION None i
t PBAPS EAL Technical Basis Manual REV a, April 26,1995 Page 115 of 130 REFERENCES NUMARC NESP-007, HA2 SE-10
l PBAPs EAL Technicd bus Manual REV c. Apnl 26.1995 Page 116 of 130 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases ALERT - 8.2.2.b IC Release of Toxic or Flammabte Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Plant personnel are unable to perform actions necessary for safe shutdown without appropriate personal protection equipment due to toxic or flammable gas concentrations confirmed within Plant Vital Structures (Table 8-1)
OPCON 1112131416101 BASIS This EAL recognizes that toxic / flammable gases have entered Plant Vital Structures and are affecting safe operation of the plant by impeding operator access to the safety systems that must be operated manually in these structures. The cause and/or magnitude of the gas concentrations is not a concern, but rather that access is required to an area and is impeded. Plant Vital Structures that must be accessed are as fc!!aws:
Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure inner Screen Structure Emergency Cooling Tower i
This event will be escalated to higher classifications based upon dan 4 ow ve ences covered ;
under other various EAL Sections. l l
DEVIATION 1 None REFERENCES NUMARC NESP-007, HAS.1 and HA3.2
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l PBAPs EAL Technied Basis Manu"J REv a, April 26,1995 Page 117 of 130 8.0 External Events 8.3 Man-Made Events UNUSUAL EVENT - 8.3.1.a IC Destructive Phenomena Affecting the Protected Area EAL Confirmed report of any vehicle crash affecting Plant Vital Structures (Table 81) '
OPCON lil213ldl51ol BASIS This EAL address crashes of vehicles affecting Plant Vital Structures with no damage occurring to safe shutdown systems. Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure inner Screen Structure Emergency Cooling Tower if there is evidence of damage by a vehicle crash including aircraft and large motor vehicles, such as a crane or missile impacts including flying objects from offsite, onsite rotating equipment or turbine failure causing casing penetration, an Alert classification should be made under EAL 8.3.2.
DEVIATION None REFERENCES NUMARC NESP-007, HU1.4
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PBAPs EAL Technical Bais Manual REV c, April 26.1995 Page 118 of 130 8.0 External Events 8.3 Man-Made Events UNUSUAL EVENT - 8.3.1.b IC Destructive Phenomena Affecting the Protected Area EAL Turbine failure resulting in casing penetration OR Damage to Generator releasing hydrogen requiring immediate plant shutdown OPCON lil2ialdislol BASIS Turbine failure of sufficient magnitude to cause observable damage to the turbine casing or seals of the turbine generator increases the potential for leakage of combustible fluids and gases (Hydrogen cooling) to the Turbine Enclosure. The damage should be readily observable and should not require equipment disassembly to locate.
This event will be escalated to an Alert based upon damage done by missiles generated by the failure or by radiological releases per EAL Sections 8.3.2 or 5.0.
DEVIATION An exception to the NUMARC Methodology was taken in that NUMARC states all operating mode applicability for this EAL. This EAL addresses main turbine rotating component failures, which do not occur during shutdown conditions. Therefore, only MODE 1,2 and 3 will be used for this EAL.
REFERENCES NUMARC NESP-007, HU1.6
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l PBAPS EAL Techned B, sis Manual REV a. Apnl 26,1995 Page 119 of 130 .
8.0 External Events i 8.3 Man-Made Events ALERT - 8.3.2 IC Destructive Phenomena Affecting the Plant Vital Area EAL ,
Confirmed report of damage to Plant Vital Structures (Table 8-1) from either a vehicle crash or missile impact ,
OPCON lilrialaisloi BASIS This EAL address crashes of vehicles or missile impacts that have caused damage to Plant Vital Structures, and thus damage may be assumed to have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification. The evidence of damage is sufficient for declaration. A vehicle crash includes aircraft and large motor vehicles, such as a crane. Missile impacts including flying objects from off-site, on-site rotating equipment or turbine failure causing casing penetration. Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Power Block .
Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower This event will be escalated to higher classifications based upon damage consequences covered !
under other various EAL Sectionc. i DEVIATION None REFERENCES NUMARC NESP-007, HA1.5 and HA1.6
. - _ _ . _ 4_ .,--
PBAPs EAL Technical Basis Manud REV c. Apnl 26.1995 Page 120 of 130 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.a IC Natural and Destructive Phenomena Affecting the Protected Area EAL Earthquake >.01 g OPCON lilalaldtslol BASIS This EAL addresses a sensed earthquake. The magnitude of .01g is the lowest detectable earthquake measured on PBAPS seismic instrumentation per SO 67.7.A. An earthquake of this magnitude may be sufficient to cause minor damage to plant structures or equipment within the Protected Area. Damage is considered to be minor as it would not affect physical or structural integrity. This event is not expected to affect the capabilities of plant safety functions.
This event will be escalated to an Alert if the earthquake reaches an Operating Basis Earthquake per EAL Section 8.4.2.
DEVIATION None REFERENCES NUMARC NESP-007, HU1.1 SE-5, Earthquake and Bases UFSAR, section 1.6 1
PBAPS EAL Technic 11 Basis Manud REv ct, Apnl 26,1993 Page 121 of 130 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report of a Tornado within the Site Boundary OR Wind speeds > 75 mph as indicated on site Metrological data for > 15 minutes OPCON 11121:1415101 BASIS 1 A tornado touching down within the Protected Area or wind speeds > 75 mph within the owner controlled Area are of sufficient velocity to have the potential to cause damage to Plant Vital Structures. The value of 75 was selected to maintain consistency with plant value and to coincide with the Beaufort Scale for Hurricane wind speed winds of 73-136 mph. These conditions are indicative of unstable weather conditions and represent a potential degradation in the level of safety of the plant. Verification of a tornado will be by direct observation and reporting by station personnel. Verification of wind speeds > 75 mph will be via meteorological data in the control room. For purposes of this EAL, sustained is > 15 minutes.
i This event will be escalated to an Alert if the tornado or high wind speeds cause damage to Plant .
Vital Structures per EAL Section 8.4.2. If it is determined that the tornado or high wind speeds '
have caused a loss of shutdown cooling, then escalation will be by EAL Section 7.3, Loss of Decay Heat Removal Capability.
DEVIATION None REFERENCES l NUMARC NESP-007, HU1.2 and HU1.7 i
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l PeAPS EAL Techrd".al Basis Manud REv a. April 26.1995 Page 122 of 130 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EAL High River level > 112' OR Low River level < 98.5' -
OPCON lil2181415lol BASIS High River level of greater than 112 feet on instrument LI-2(3)278A,B,C or Li 2(3)278A,B,C is indication of the river being in flood. By procedure the units will SCRAM and be brought to cold shutdown.
Low River level of less than 98.5 feet is indication of loss of Conowingo Pond and loss of circulation water pumps. Procedures require the unit to be SCRAMMED and brought to cold shutdown.
This event will be escalated to an Alert classification based continuation of the river situation as included in EAL 8.4.2.c.
DEVIATION None REFERENCES NUMARC NESP-007, HU1.7 SE-4, Flood SE-3, Loss of Conowingo Pond
l PBAPS EAL Technictl Basis Manu"J REv a, April 26,1995 '
Page 123 of 130 8.0 External Events 8.4 Natural Events ALERT - 8.4.2.a IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL EarthquaK9 >,05 g OPCON 1112181415101 BASIS This EAL addresses an earthquake that exceeds the Operating Basis Earthquake level of .05g and is beyond design basis limits. An earthquake of this magnitude may be sufficient to cause damage to safety related systems and functions.
The Max Crediable Earthquake for PBAPS is 0.12g per UFSAR section 1.6, therefore this EAL is conservative and warrents an Alert classification.
This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, HA1.1 SE-5, Earthquake and Bases UFSAR section 1.6 1
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l 4 PBAPS EAL Technical Basis Manud REV Q April 26,1995 Page 124 of 130 8.0 External Events 8.4 Natural Events ALERT - 8.4.2.b IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Tornado or wind speeds > 75 mph causing damage to Plant Vital Structures (Table 81)
OPCON 111218Id15101 BASIS This EAL addresses events that have caused damage to Plant Vital Structures, and thus damage may be assumed to have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification The evidence of damage is sufficient for declaration. Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, HA1.2 and HA1.3 i
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l PBAPs EAL Technied Basis Manu J j REV a. Apnl 26,1995 Pags 125 of 130 l
l 8.0 External Events
! 8.4 Natural Events l
ALERT - 8.4.2.c IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL l
High River level > 176' l OR Low River Level < 92.5' i 1
l i
I OPCON 1i121 1415101 j BASIS High River level > 116 feet is indication of the river being in flood. This level is capable of causing flooding that can affect Plant Vital Structures. No attempt should be made to determine the magnitude of flooding. This is a long lead time event but this levelis ground elevation of the reactor building and intake pump structure so classification as an Alert Event is appropriate. The evidence of flooding is sufficient for declaration.
Low River level < 92.5 feet is indication of loss of Conowingo Pond and loss of circulation water I pumps. Procedures require the unit to be SCRAMMED and brought to cold shutdown and utilization of the ECW pump and Emergency Cooling Tower.
This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.
DEVIATION l
None 1
REFERENCES l
NUMARC NESP-007, HA1.7 SE-4, Flood
( SE-3, Loss of Conowingo Pond
f PBAPS EAL Technical Basis Manual REV c April 26,1995 Page 126 of 130 This page intentionally left blank i
PBAPs EAL Technicci Basis Manu:.1 REV e. Apnl 26,1993 Page 127 of 130 9.0 Other 9.1 General UNUSUAL EVENT - 9.1.1 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of an Unusual Event EAL Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant OPCON 1112181415101 BASIS This EAL allows the Shift Manager to declare an Unusual Event upon the determination that the level of safety of the plant has degraded. Where the degradation is associated with equipment or system malfunctions, the decision that it is degraded should be made upon functionality, not operability. A system, subsystem, train, component or device, though degraded in equipment condition or configuration, should be considered functional if it is capable of maintaining respective system parameters within acceptable design limits. Examples include:
Internal flooding in excess of sump handling capability affecting safety related areas of the plant.
Releases of radioactive materials requiring off-site response or monitoring are not expected to occur at this level unless further degradation of safety systems occurs. However, if one does occur, it will be classified under section 5.0, Radioactivity Releases.
DEVIATION None REFERENCES NUMARC NESP-007, HUS and HU1.3
=_ __
PBAPS EAL Technical Basis Manual REV a. April 26,1995 Page 128 of 130 9.0 Other 9.1 General ALERT - 9.1.2 lC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of an Alert EAL Events are in progress or have occurred which indicate an actual or potential substantial degradation of the level of safety of the plant OPCON lil21*laisiol BASIS This EAL allows the Shift Manager to declare an Alert upon the determination that the level of safety of the plant has substantially degraded but is not explicitly addressed by other EALs.
This includes a determination by the Shift Manager that the TSC and OSC should be activated and command and control functions should be transferred for the event to be effectively mitigated.
Transfer of command and control functions is used as an initiator since an event significant to warrant transfer is a substantial reduction in the level of safety of the plant. Other examples are:
Internal flooding affects the operability of plant safety systems required to establish or maintain cold shutdown.
Releases that are expected will be limited to a small fraction of the EPA Protective Action Guidelines and will be classified under section 5.0, Radioactivity Releases.
DEVIATION None REFERENCES NUMARC NESP-007, HA6, FC EAL #5, RC EAL #6, PC EAL #6 i
)
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PBAPs EAL Technical Basis Manu:j REV a, Apnl 26,1995 Page 129 of 130 9.0 Other 9.1 General SITE AREA EMERGENCY - 9.1.3 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of Site Area Emergency EAL Events are in progress or have occurred which indicate an actual or likely major failure of plant functions needed for protection of the public OPCON lil2131dislol BASIS This EAL allows the Shift Manager to declare a Site Area Emergency upon the determination of an actual or likely major failure of plant functions needed for protection of the public, but is not explicitly addressed by other EALs, Releases are not expected to result in exposure levels which exceed the EPA Protective Action Guidelines except within the site boundary and will be classified under section 5.0, Radioactivity Releases.
DEVIATION .
None REFERENCES NUMARC NESP-007, HS3, FC EAL #5, RC EAL #6, PC EAL #6
PBAPs EAL Techrucal Basis Marmal REV e. Apnl 26,1995 Page 130 cf 130 9.0 Other i
9.1 General GENERAL EMERGENCY - 9.1.4 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of General Emergency EAL ,
Events are in progress or have occurred which indicate an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity OPCON hl218141slol '
BASIS This EAL allows the Shift Manager to declare a General Emergency upon the determination of an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity, but is not explicitly addressed by other EALs.
Releases may exceed the EPA Protective Action Guidelines for more than the immediate site area and will be classified under section 5.0, Radioactivity Releases.
DEVIATION None REFERENCES NUMARC NESP-007, HG2, FC EAL #5, RC EAL #6, PC EAL #6 m #e.w = - w w w . ----. w- m--m e m--. -- - >--e-- -.
, PBAPs EAL Ttbk '
REV a. April 26.1995 !
Page 1 of 20 PBAPS EAL Tab!e Table of Contents '
1.0 Reactor Fue! !
1.1 Coolant Activity . . . . . . . . . . . . . . . ............................ 2 1.2 Containment High Radiation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.3 Irradiated Fuel or New Fue! . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2 R e acto r P ow e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.0 Primary Containment 3.1 Primary Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.0 Secondary Containment 4.1 Secondary Containment Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.2 Main Steam Lin e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 5.0 Radioactivity Release 5.1 Effluent Release and Dose ................................... 10 5.2 in- Plant R adiatio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 6.0 Loss of Power 6.1 Loss of AC or DC Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 7.0 Internal Events 7.1 Technical Specifica3ons & Control Room Evacuation . . . . . . . . . . . . . . . . . 13 7.2 Loss of Decay Heat Removal Capabililty . . . . . . . . . . . . . . . . . . . . . . . . . . 14 ,
7.3 Loss of Assessment / Communications Capabililty . . . . . . . . . . . . . . . . . . . . 15 ,
8.0 Extemal Events ,
8.1 S ecu rity Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 '
8.2 Fire / Explosion and Toxic / Flammable Gases . . . . . . . . . . . . . . . . . . . . . . . 17
? 8.3 Man-Made Events . . . . . . . . . . . . . . . .......................... 18 8.4 N atu ral E ve n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 9.0 Other 9.1 G e n e ral . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 MODE MODE SWITCH POSITION tilais141slol Run m alelais101 Startup I11:1alaistol Shutdown (Hot) m ais141 sial Shutdown (Cold)
Illslalal a101 Refueling m alsla m ol N/A (Defueled) ;
I
l PBAPS EAL Table REV a. Apnl 26,1995 Page 2 of 20 1.0 Reactor Fuel 1.1 Coolant Activity CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 1.1.1.a lil Islatelol Reactor Coolant activity > 4 C//gm Dose Equivalent lodine 131 1.1.1.b til2181dl*lpi SJAE Radiation (Offgas Monitor) > 2.5x10' mR/hr ALERT 1.1.2 lil21*I4151ol Reactor Coolant activity > 300 pC//gm Dose Equivalent lodine 131 SITE AREA 1.1.3 til2181*Istol EMERGENCY Reactor Coolant activity > 300 pCI/gm Dose Equivalent lodine 131 AND Identified breach of Primary Containment (Tech Specs Section 3.7) OR Drywell Pressure > 9 psig GENERAL None EMERGENCY l
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PBAPS EAL Table REV a, Apnl 23,1995 Page 3 of 20 l
1.0 Reactor Fuel i l
1.2 Containment High Radiation CLASSIFICATION EMERGENCY ACTION LEVEL l UNUSUAL EVENT None ALERT 1.2.2 l il 21:I d i s lo]
Containment high range radiation > 3x1f R/hr SITE AREA 1.2.3.a lil21:141elol EMERGENCY Containment high range radiation > 3x1# R/hr AND Identified breach of Primary Containment [ Tech Specs Sect!on 3.7) ,
1.2.3.b lil21sidisfoi Containment high range radiation > 8xff R/hr GENERAL 1.2.4 l il 21s l 41 sI61 EMERGENCY Containment high range radiation > 6xff R/hr
. - , . . - - . -,-2. -_, .
I PBAPs EAL Table REV a. Apnl 26,1995 Page 4 of 20 1.0 Reactor Fuel 1.3 Irradiated Fuel or New Fuel CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 1.3.1.a IYitl*1dl 5 Irl Unexpected RPV level decrease to < 100"when reactor cavity is flooded up and fuel pool gates are in place 1.3.1.b l il r l 314 l sLJ Skimmer Surge Tank low level alarm AND Visual observation of a water level decrease below the fuel pool skimmer surge tank inlet ALERT 1.3.2.a fil21sl415lol Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1) 1.3.2.b 11121:1415101 A fuel handling accident causing a HiHi alarm from the Refueling Floor Vent Exhaust Monitor SITE AREA None EMERGENCY GENERAL None EMERGENCY Trbl41-1 Refuel Floor ARMS 3-7 (7-9) Steam Separator Pool 3-8 (7-10) Refuel Slot 3-9 (7-11) Fuel Pool 3-10 (7-12) Refueling Bridge
)
i PBAPS EAL Ttble REv a. April 26,1995 Page 5 of 20 2.0 Reacter Pressure Vessel 2.1 Reactor Water Level l
CLASSIFICATION EMERGENCY ACTION LEVEL !
l UNUSUAL EVENT None l
ALERT 2.1.2 111 131415I01 RPV level < -160 "
SITE AREA 2.1.3.a 1:12151415101 EMERGENCY RPV level cannot be restored above -172 "
2.1.3.b 1812131415101 RPV level cannot be determined i GENERAL 2.1.4.a 1:12131415101 EMERGENCY RPV level cannot be restored above -226 "
2.1.4.b 18l21314151#1 RPV level cannot be determined AND RPV Flooding cannot be established per T-116 m - - -
l PBAPS EAL Table REV a. April 26,1995 Page 6 of 20 2.0 Reactor Pressure Vessel 2.2 Reactor Power CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None ALERT 2.2.2 l il:1.1141slol Failure of Automatic RPS SCRAM to reduce reactor power < 3%
SITE AREA 2.2.3 lil2181dislol EMERGENCY Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 3%
GENERAL 2.2.4 li hlal41*lol EMERGENCY Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 3%
AND Torus Temperature is > 180'F 6
e
l PBAPs EAL Ttble REV a, Apnl 26,1995 Page 7 of 20 3.0 Primary Containment 3.1 Primary Containment CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 3.1.1.a lil21sisi+1ol Drywell Pressure > 2 ps/g AND Indication of a Leak into Containment 3.1.1.b 111213I418101 Unidentified Primary System Leakage > 10 ppm into the Drywell O_E Identified Primary System Leakage > 25 ppm into the Drywell ALERT 3.1.2 lil2131*Islol Drywell Pressure > 9 psig SITE AREA 3.1.3 lil21sistslol EMERGENCY Containment Failure indicated by a rapid, unexplained drop in Containment Pressure following initial pressure rise above 9 psig GENERAL 3.1.4.a til21sI'Islol EMERGENCY Containment Pressure > 60 ps/g 96 Containment Venting via T-200 is required 3.1.4.b lil21sI*Islol Drywell Hydrogen 2 6 %
AND Drywell Oxygen 2 5 %
e=
. 1 PBAPs EAL Ttble REV a. April 26,1995 Page 8 of 20 4.0 Secondary Containment 4.1 Secondary Containment Temperature CLASSlFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None ,
ALERT 4.1.2 [D21814181ol .
An Unisolable Primary System Leak is discharging into Secondary Containment AND A T-103 Temperature Action Levelis exceeded in ONE area '
requiring a SCRAM 3 SITE AREA 4.1.3 lil2181dl*161 EMERGENCY An Unisolable Primary System Leak is discharging into Secondary Containment AND T-103 Temperature Action Levels are exceeded in 1WO OR MORE arcas ,
requiring an Emergency Blowdown per T-112 GENERAL 4.1.4 l il 218141slol EMERGENCY An Unisolable Primary System Leak is discharging into Secondary Containment AND T-103 Temperature Action Levels are exceeded in TWO OR MORE areas requiring an Emergency Blowdown per T-112 AND T-103 Radiation A%co Levels are exceeded in the same TWO OR MORE areas I
l PBAPS EAL Ttbb REV a. Apnl 26,1995 Page 9 of 20 4.0 Secondary Containment 4.2 Main Steam Line CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 4.2.1 lil:IsI41*lol Main Steam Line HlHi Radiation (10xNFPB)
ALERT 4.2.2 1i121 1415101 Failure of one or more Main Steam Lines to isolate on any MSIV Closure Signal SITE AREA 4.2.3 lil21sid181ol EMERGENCY Main Steam Line Break discharging into the Turbine Building AND Vent Stack > 1x10 pCl/cc GENERAL 4.2.4 1112131415101 EMERGENCY Main Steam Line Break discharging into the Turbine Building AND Vent Stack > 1x10 pCl/cc l
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PBAPS EAL Ttble REV a. Apnl 26.1995 Page 10 of 20 5.0 Radioactivity Release 5.1 Effluent Release and Dose CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 5.1.1.a fil2lal4151ol Main Stack, Vent Stack, or Torus Hardened Vent Rad monitor continuously in HiHi Alarm M known Unmonitored Release continuously in progress for > 60 minutes AND Calculated maximum offsite dose rate exceeds 0.114 mrem /hr TPARD OR 0.342 mrem /hr child thyrold CDE based on a 60 minute average 5.1.1.b 1112181415101 Report indicates Liquid Release exceeds TWO TIMES Tech Specs (T.S.
3.8.B.1) for > 60 minutes ALERT 5.1.2.a lil2181415lol Main Stack, Vent Stack, or Torus Hardened Vent Rad monitor continuously in HiHi Alarm E known Unmonitored Release continuously in progress for > 15 minutes AND Calculated maximum offsite dose rate exceeds 11.4 mrem /hr TPARD OR 34.2 mrem /hr child thyroid CDE based on a 15 minute average 5.1.2.b I112131415101 Report indicates Liquid Release exceeds TWO HUNDRED TIMES Tech Specs (T.S. 3.8.B.1) for > 15 minutes SITE AREA 5.1.3 iil21sI41slol EMERGENCY Projected offsite dose exceeds 100 mrem TPARD, OR Projected offsite dose exceeds 500 mrem child thyrold CDE, OR Actual offsite whole body dose rate exceeds 100 mrem /hr GENERAL 5.1.4 lil21314161ol EMERGENCY Projected offsite dose exceeds 1000 mrem TPARD, OR Projected offsite dose exceeds 5000 mrem child thyroid CDE, OR Actual offsite whole body dose rate exceeds 1000 mrem /hr NOTE: CDE - Committed Dose Equivalent TPARD= Total Protective Action Recommendation Dose
I PBAPS EAL Table REV a. Apnl 26,1995 Page 11 of 20 5.0 Radioactivity Release 5.2 In-Plant Radiation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 5.2.1 1112181415101 Inplant radiation level > 1x1f mR/hr requiring T-103 entry ALERT 5.2.2.a lil218I*Islol inplant radiation level > 9x1f mR/hr for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requiring T-103 entry 5.2.2.b lil2181dlslol Control Room area radiation level > 15 mR/hr for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SITE AREA None EMERGENCY GENERAL None EMERGENCY
l PBAPS EAL Tcble REV a. Apnl 26,1995 Page 12 of 20 6.0 Loss of Power 6.1 Loss of AC or DC Power CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 6.1.1.a lil2ial41*lol Loss of ALL Offsite Power for >f5 minutes 6.1.1.b hitislaislol Loss of ALL safety related DC Power indicated by < 107.5 VDC for > 15 minutes ALERT 6.1.2.a lil21sl4161pl Loss of ALL Offsite Power for > 15 minutes AND Only ONE 4 KV Emergency Bus is available 6.1.2.b 14!18131415101 Loss of ALL Offsite Power AND ALL 4 KV Emergency Busses are unavailable for > 15 minutes SITE AREA 6.1.3.a lil21sld181ol EMERGENCY Loss of ALL Offsite Power AND ALL 4 KV Emergency Busses are unavailable for > 15 minutes 6.1.3.b 11121 141$101 Loss of ALL safety related DC Power indicated by < 107.5 VDC for > 15 mlnufes GENERAL 6.1.4 1i121 14151o1 EMERGENCY loss of ALL Offsite Power AND ALL 4 KV Emergency Busses are unavailable for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> m
d PBAPS EAL Ttble REV a. Apnl 26,1995 Page 13 of 20 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 7.1.1 lil:Isl41.iol Unable to bring the Plant to the required Mode within Tech Spec LCO action ,
times
)
4 ALERT 7.1.2 lil21slaisiol Control Room evacuation procedures have been initiated SITE AREA 7.1.3 lil21sI41slol EMERGENCY Failure to establish Alternate / Emergency Control of the Plant within 15 minutes after evacuation of the Control Room GENERAL None EMERGENCY l
l t
I PBAPS EAL Table REV a, Apnl 26,1995 P igs 14 of 20 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability CLASSlFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 7.2.1 tiltioidisiti Uncontrolled Reactor Coolant temperature increase to > 212 *F ALERT 7.2.2 liist 141slol Uncontrolled Reactor Coolant temperature increase to > 212'F AND Inability to establish alternate decay heat removal capability l
SITE AREA 7.2.3 1 121814161p1 EMERGENCY Loss of Main Condenser as a heat sink AND Loss of TORUS Heat Sink capabilities as evidenced by T-102 legs [T/T, T/L, PC/P, or DWR] requiring an Emergency Blowdown GENERAL None EMERGENCY l
m_ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ - _ - _ _
l PBAPS EAL Tabla REV a Apnl 26,1995 Page 15 of 20 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 7.3.1.a lil21*I41*lol Loss of All Annunciators in the Control Room for > 15 minutes 7.3.1.b Lil 21
- l d i s lol Loss of ALL Onsite communications (Table 7-1)
OR Loss of ALL Offsite communications (Table 7-1)
ALERT 7.3.2 lil213lAl4]D]
Loss of All Annunciators in the Control Room for > 15 minutes AND Significant Plant Transient (Table 7-2) is in progress OR_ Plant Monitoring System is unavailable SITE AREA 7.3.3 lil21:141siol EMERGENCY Loss of All Annunciators in the Control Room for > 15 minutes AND Significant Plant Transient (Table 7-2) is in progress AND_
Plant Monitoring System is unavailable GENERAL None EMERGENCY ,
Tcble 7-1 Communications Table 7-2 Plant Transients Onsite Offsite Site Phones (GTE System) X X SCRAM OMN! System X X Recire Runbacks > 25% thermal power Plant Public Address X Thermal power oscillations > 10%
Station Radio X Stuck open relief valve (s)
NRC (FTS-2000) X ECCS injection PA State Police Radio X Load Dispatcher Radio X PECO Dial Network X e
l PBAPs EAL Table REv a. April 26,1995 Page 16 of 20 8.0 External Events 8.1 Security Threats CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 8.1.1 lil2iaI415lal Confirmed security threat directed towards the station as evidenced by :
- Credible sabotage or bomb threat within the Protected Area, OR
+ Credible intrusion and attack threat to the Protectad Area, OR
+ Attempted intrusion and attack to the Protected Area, OR
+ Attempted sabotage discovered within the ProtectedNital Area, OR
- Hostage / Extortion situation that threatens normal plant operations ALERT 8.1.2 1i121 1415101 Confirmed hostile intrusion or act within the Protected Area as evidenced by
- Actual attack and intrusion into the Protected Area, OR
- Suspected bomb, sabotage or sabotage device discovered in the ProtectedNital Area SITE AREA 8.1.3 lilelalelslol EMERGENCY Confirmed hostile intrusion or act in plant Vital Areas as evidenced by :
- Actual attack and intrusion into a Vital Area, OR
- Confirmed bomb, sabotage or sabotage device discovered in a ProtectedNital Area GENERAL 8.1.4 l'Irlalaisiol EMERGENCY Security event resulting in the actualloss of physical control of the:
Control Room OR_ Remote Shutdown Capability l
l 4
l PBAPs EAL Table REV a. April 26,1995 Page 17 of 20 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 8.2.1.a lil21sidislol Fire within ON-114 Plant Vital Structures (Table 81) which is not extinguished within 15 minutes of verification of alarms 8.2.1.b l'l21*ldl6lol Normal plant operation is impeded by:
Toxic or flammable gas concentrations confirmed within the Protected Area OR Control Roem informed by Local, County or State Officials to evacuate personnel within the protected area due to an offsite gas release 8.2.1.c lil2181dl5lol Report by plant personnel confirming the occurrence of an explosion in a Plant Vital Structure (Table 8-1)
ALERT 8.2.2.a lil2181dl6lol Fire or explosion within Plant Vital Structures (Table 8-1) damages safe shutdown systems causing or jeopardizing plant shutdown 8.2.2.b 1112181415101 Plant personnel are unable to perform actions necessary for safe shutdown without appropriate personal protection equipment due to toxic or flammable gas concentrations confirmed within Plant Vital Structures (Table 8-1)
SITE AREA None EMERGENCY GENERAL None EMERGENCY Tabli 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower
l PBAPs EAL Ttble REV a. April 26,1995 Page 18 of 20 8.0 External Events ,
8.3 Man-Made Events CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 8.3.1.a t il 2l a l 41 s bl Confirmed report of any vehicle crash affecting Plant Vital Structures (Table 81) 8.3.1.b lil21*ldl*IRI -
Turbine failure resulting in casing penetration OR Damage to Generator releasing hydrogen requiring immediate plant shutdown ALERT 8.3.2 miririfil51 Confirmed report of damage to Plant Vital Structures (Table 8-1) from either a vehicle crash or missile impact SITE AREA None EMERGENCY GENERAL None EMERGENCY Trbh 8-1 Plant Vital Structures Power Block -
Diesel Generator Building -
Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower
- i f1 PBAPs EAL Table REV o, April 26,1995 !
Page 19 of 20 8.0 External Events 8.4 Natural Events CLASS!FICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 8.4.1.a 1i12131415101 Earthquake >.07 g 8.4.1.b lil2Isl4Islol Report of a Tornado within the Site Boundary OR Wind speeds > 75 mph as indicated on site Metrological data for > 15 minutes 8.4.1.c 1112181alslol High River level > 112' OR Low River level < 98.5' ALERT 8.4.2.a lil2Isl41slol Earthquake >.05 g 8.4.2.b lil213ldislol Tornado or wind speeds > 75 mph causing damage to Plant Vital Structures (Table 81) 8.4.2.c 1812131415101 High River level > 1f 6' OR Low River Level < 92.5' SITE AREA None EMERGENCY GENERAL None EMERGENCY Tabb 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure inner Screen Structure Emergency Cooling Tower
PBAPS EAL Ttble REV a. Apnl 26,1995 Page 20 of 20 9.0 Other 9.1 General CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 9.1.1 lil:Islaiolol Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant ALERT 9.1.2 1812131416101 Events are in progress or have occurred which indicate an actual or potential substantial degradation of the level of safety of the plant SITE AREA 9.1.3 lil213ldislol EMERGENCY Events are in progress or have occurred which indicate an actual or likely major failure of plant functions needed for protection of the public _
GENERAL 9.1.4 lil:Isidislol EMERGENCY Events are in progress or have occurred which indicate an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity
_-m- _ _ - _ _ . _ _ _ _ _ - _ _ _ _ _ _ _ _
i PBAPS EAL NUMARC Companson REV a. Apnl 26,1995 ,
Page 1 of 8 l PBAPS EAL NUMARC Comparison Table of Contents Section 1 - NUMARC EAL Versus PBAPS EAL Comparison Matrix . . . . . . . . . . . . 3 Section 11 - Summary of Fission Product Barriers . . . . . . . . . . . . . . . . . . . . . . . . . 8 r
x 1
.. = -.- . .
PBAPs EAL NuMARc Companson REv a. Apnl 26,1995 Page 2 of 8 Section 1 - NUMARC EAL Versus PBAPS EAL Comparison Matrix NUMARC PBAPS COMMENTS / EXCEPTIONS AU1.1 5.1.1.a AU1.2 5.1.1.a 5.1.1.b AU1.3 N/A PBAPS does not have telemetered perimeter monitors.
AU1.4 N/A PBAPS does not use automatic initiation of real time dose assessment.
AU2.1 1.3.1.a 1.3.1.b AU2.2 1.3.1.b AU2.3 N/A PBAPS does not have dry fuel storage capabilities.
AU2.4 5.2.1 AA1.1 5.1.2.a AA1.2 5.1.2.a 5.1.2.b AA1.3 N/A PBAPS does not have telemetered perimeter monitors.
AA1.4 N/A PBAPS does not use automatic initiation of real time dose assessment AA2.1 1.3.2.a 1.3.2.b AA2.2 1.3.2.a AA2.3 1.3.2.a This EAL is addressed by utilizing radiation levels which could be caused by uncovering the fuel.
AA2.4 1.3.2.a PBAPS does not have level indication on the Spent Fuel Pool. This EAL is addressed by utilizing radiation levels which could be caused by uncovering the fuel.
AA3.1 5.2.2.b AA3.2 5.2.2.a AS1.1 5.1.3 PBAPS contains this within AS1.3 & 1.4, but does not specifically address.
,- - _ . - . . - . _ - . - . . - .. . .. . ~ . . _ . _ - . . _ - - . . - . . - . . . _ . ,
PBAPs EAL NUMARC companson REV a. Apnl 26.1995 Page 3 of 8 NUMARC PBAPS COMMENTS / EXCEPDONS AS1.2 N/A PBAPS does not have telemetered perimeter monitors.
AS1.3 5.1.3 AS1.4 5.1.3 AG1.1 5.1.4 PBAPS contains this within AS1.3 & 1.4, but does not specifically address.
AG1.2 N/A PBAPS does not have telemetered perimeter monitors.
AG1.3 5.1.4 AG1.4 5.1.4 HU1.1 8.4.1.a HU1.2 8.4.1.b HU1.3 9.1.1 HU1.4 8.3.1.a HU1.5 8.2.1.c HU1.6 8.3.1.b This EAL addresses main turbine rotating component failures, which do not occur during shutdown conditions. Therefore, only MODE 1,2 and 3 will be used for this EAL.
HU1.7 8.4.1.b PBAPS does not have any EAL addressing hurricanes or 8.4.1.c seiches which are not applicable. High winds and floods are addressed.
HU2 8.2.1.a ,
HU3.1 8.2.1.b HU3.2 8.2.1.b HU4.1 8.1.1 !
HU4.2 8.1.1 HUS 9.1.1 H A1.1 8.4.2.a HA1.2 8.4.2.b l
HA1.3 8.4.2.b i
_ _ _ _ _ _ _ _ _m . _ __ _ _ ____.___. _ _ _ _ _ _ . . . , . _ _ , _ _ _ . . _ _ . . __ _ . . . _ . . . _ _ _ _ __ _ ___ _ _ _ _ _ . , _ __
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l PBAPs EAL NUMARC Companson REV a. Apnl 26.1995 Page 4 of 8
<- NUMARC PBAPS COMMENTS / EXCEPTIONS HA1.4 N/A PBAPS has no instrumentation other than meteorological which can be used to measure the impact of natural or destructive phenomena.
HA1.5 8.3.2 HA1.6 8.3.2 HA1.7 8.4.2.c HA2 8.2.2.a HA3.1 8.2.2.b HA3.2 8.2.2.b HA4.1 8.1.2 HA4.2 8.1.2 HA5 7.1.2 HA6 9.1.2 HS1.1 8.1.3 HS1.2 8.1.3 HS2 7.1.3 HS3 9.1.3 HG1.1 8.1.4 HG1.2 8.1.4 HG2 9.1.4 SU1 6.1.1.a SU2 7.1.1 SU3 7.3.1.a SU4.1 1.1.1.b 4.2.1 SU4.2 1.1.1.a PBAPS Technical Specifications only require only MODES 1,2,3,4.
SU5 3.1.1.a 3.1.1.b
PBAPs EAL NuMARc Companson REv a. Apnl 26.1995 Page 5 of 8 NUMARC PBAPS COMMENTS / EXCEPTIONS SU6 7.3.1.b SU7 6.1.1.b SA1 6.1.2.b SA2 2.2.2 SA3 7.2.1 EAL 7.2.1 was created to be a precursor to SA3 which calls 7.2.2 for an Alert to be declared for a loss of shutdown cooling function and reactor coolant temperature rises above 200 *F.
The reactor core can continue to be properly cooled as long as Torus cooling is available. While this event clearly warrants the declaration of an Unusual Event, PBAPS does not believe it meets the threshold for an Alert.
SA4 7.3.2 SAS 6.1.2.a SS1 6.1.3.a SS2 2.2.3 SS3 6.1.3.b SS4 7.2.3 SS5 N/A PBAPS has not identified a specific EAL for this condition.
Both conditions stated in this reference are addressed i appropriately in EAL sections 7.2.2 and 2.1.2,2.1.3 and 2.1.4.
SS6 7.3.3 1
SG1 6 1.4 l SG2.1 2.2.4 SG2.2 2.2.4 FC.1 1.1.2 1.1.3 FC.2 2.1.3.a 2.1.4.a FC.3 1.2.3.b l
PBAPs EAL NUMARc Companson REV a, Aprd 26,1995 Page 6 of 8 NUMARC PBAPS COMMENTS / EXCEPDONS FC.4 1.2.2 1.2.3.a 1.2.4 2.1.3.b 2.1.4.b 3.1.4.a 3.1.4.b 4.1.4 4.2.4 FC.5 9.1.2 9.1.3 .
9.1.4 RC.1 N/A This barrier failure is not explicity addressed at PBAPS.
There are several EALs, including Reactor Pressure Vessel' level and Primary Containment pressure which adequately address this situation.
RC.2 1.1.3 There is a deviation in that a high drywell pressure 9 psig 3.1.2 was selected duen to the fact that the alarm value of 2 psig 3.1.3 can be reached by a small primary system leak and 9 psig is larger than experience shows of blown packing and recirc seal leaks. It is more representative of a LOCA condition and this torus pressure is in the TRIPS for actions to protect the containment.
RC.3 1.2.3.b 1.2.4 RC.4 2.1.3.a 2.1.4.a RC.5 2.1.2 2.1.3.b 2.1.4.b 3.1.4.a 3.1.4.b 4.1.2 4.1.3 4.1.4 4.2.3 4.2.4
PBAPS EAL NUMARC Companson REV a Apnl 26,1995 Page 7 of 8 NUMARC PBAPS COMMENTS / EXCEPTIONS RC.6 9.1.2 9.1.3 9.1.4 PC.1. 3.1.3 3.1.4.b PC.2 3.1.4.a 4.1.3 4.1.4 4.2.2 4.2.3 4.2.4 t PC.3 1.2.4 i
PC.4 2.1.4.a PC.5 1.1.3 ;
i 1.2.3.a I 2.1.4.b 4.1.2 PC.6 9.1.2 9.1.3 9.1.4 1
1 l l i l i 1
PBAPS EAL NUMARC Comparison REV a. Aprd 26,1995 Page 8 of 8 Section II- Summary of Fission Product Barriers The following summarizes the EALs whiCh resulted from the analysis performed of the fissbn product barrier methodology of NUMARC NESP-007 for Peach Botom Atomic Power Station UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 9.1.1 Emergency Director 1.1.2 RPV Coolant Acnvity > 300 pCi'gm DEI 1.1.3 RPV Coolant Achvity > 300 pCi'gm DEI 1.2.4 Cont Rad > 6x10' Rhr (FC4 bss and Judgement (FC 1 loss) (FC-1 bss) and cont integrity bss (PC- RC-3 loss and PC-3 potentialloss) 5 loss) or Drywell > 9 psig (RC-2 loss) 1.2.2 Cont Rad > 3x10' Rhr (FC-4 bss) 2.1.4 a RPV <-226* (FC-2 loss and RC-4 bss 1.2.3.a Cont Rad > 3x10' Rhr (FC4 bss) and and PC 4 potendalloss) 2.1.2 RPV <-160r* (RC-5 potential loss) bss containment integrity (PC-5 loss) 2.1.4 b RPV level unknown and RPV fboding 31.2 Drywell > 9 psig (RC-2 bss) 1.2.3.b Cont Rad > 8x10' Rhr (FC-3 bss and not establiished (FC4 potenbal loss.
RC-3 loss) RC-5 loss and PC-5 loss) 4.1.2 RCS discharging into secondary cont (RC-5 potential bss and PC-5 potenbal 2.1.3 a RPV <-172* (FC-2 potentialloss and 3.1.4 a Cont Press > 60 psig or venting loss) RC-4 loss) required (FC-4 potentialloss and RC-5 loss and PC-2 loss) 4.2.2 Failure of MSIVs (RC 5 potentialloss 2.1.3.b RPV level undetermined (FC4 and PC-2 potentalloss) potential bss and RC-5 loss) 3.1.4 b Drywell H2 > 6% and O2 > 5% (FC-4 loss and RC-S loss and PC 1 potennal 9 1.2 Emergency Director Judgement 3.1.3 Cont Press >9 psig with rapid drop (RC- bss) 2 loss and PC-1 bss) 4.1.4 RCS discharging into secondary cont 4.13 RCS discharging into secondary cont and indication of fuel damage (FC4 (RC-5 bss and PC-2 loss) loss and RC-5 loss and PC-2 loss) l 42.3 MS break into Turbine Building (RC-5 4 2.4 MS break into Turbine Builddng with bss and PC-2 bss) indication of fuel damage (FC-4 loss and RC-5 loss and PC-2 bss)
, 9.1.2 Emergency Director Judgement 9.1.4 Emergoney Director Judgement l
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h ENCLOSURE 2 LIMERICK GENERATING STATION UNITS 1 AND 2 EMERGENCY ACTION LEVEL SUPPORTING DOCUMENTATION t
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A M
N PECO ENERGY LIMERICK GENERATING STATION i-Emergency Preparedness
-r UPGRADED EMERGENCY ACTION LEVELS NRC Submittal Copy l Revision a i.
April 26,1995 4
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PECO ENERGY LIMERICK GENERATING STATION Emergency Preparedness UPGRADED EMERGENCY ACTION LEVELS NRC Submittal Copy Revision a l
April 26,1995
LGS EAL Technical Basis Manual REv a. Apnl 26,1995 Page 1 of 130 LGS EAL Technical Basis Manual Table of Contents Section l - introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Section ll - Ac ronym s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Section lli - EAL Technical Basis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.0 Reactor Fuel 1.1 Co olant Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1.2 Containment High Radiation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 1.3 Irradiated Fuel or New Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 2.0 Reactor Pressure Vessel 2.1 Reactor Wate r Le vel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 2.2 R e acto r Powe r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 3.0 Primary Containment 3.1 Containment Pressure or Leakage . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . 43 4.0 Secondary Containment 4.1 Secondary Containment Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 4.2 Main Ste am Lin e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 5.0 Radioactivity Releaso 5.1 Effluent Monitors and Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 5.2 In. Plant Radiatio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 77 6.0 Loss of Power 6.1 Loss of AC or DC Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation . . . . . . . . . . . . . . . . 93 7.2 Loss of Decay Heat Removal Capabililty . . . . . . . . . . . . . . . . . . . . . . . . . 96 7.3 Loss of Assessment / Communications Capabililty . . . . . . . . . . . . . . . . . . . 99 8.0 External Events 8.1 S e cu rity E ve n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 8.2 Fire / Explosion and Toxic / Flammable Gases . . . . . . . . . . . . . . . . . . . . . . 112 8.3 Ma n. Mad e Ev e n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 8.4 N at u ral E v e n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 122 9.0 Other 9.1 General ................................................ 127
' LOS EAL Technical Basis Manual FIEv c. April 26,1995 Page 2 of 130 Section I- Introduction This manual contains the technical basis for the Emergency Action Levels as utilized in ERP-101, Classification of Emergencies. The format and use of this manualis as follows.
- 1. Heading and Sub-Heading There are nine major headings each containing one or more sub-headings. These are as follows:
1.0 Reactor Fuel 1.1 Coolant Activity 1.2 Containment High Radiation 1.3 Irradiated Fuel or New Fuel 2.0 Reactor Pressure Vessel '
2.1 Reactor Water Level '
2.2 Reactor Power 3.0 Primary Containment 3.1 Containment Pressure or Leakage 4.0 Secondary Containment 4.1 Secondary Containment Temperature 4.2 Main Steam Line 5.0 Radioactivity Release 5.1 Effluent Monitors and Dose 5.2 In-Plant Radiation 6.0 Loss of Power 6.1 Loss of AC or DC Power 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation 7.2 Loss of Decay Heat Removal Capabililty 7.3 ' Loss of Assessment / Communications Capabilitty 8.0 External Events 8.1 Security Events 8.2 Fire / Explosion and Toxic / Flammable Gases 8.3 Man Made Events 8.4 Natural Events 9.0 Other 9.1 General l
l 1
lgs EAL Technied Basis Mauil REv s. Apnl 26.1C95 Pega 3 et 130
- 2. Emergency Classification Level and Number identification The classifications range from Unusual Event through Alert, Site Area Emergency to General Emergency. For each sub-heading, there may not be an EAL in every classification level. Each EAL is individually and uniquely numbered. No two numbers are the same.
- 3. INITIATING CONDITION The Initiating Condition (as described in NUMARC NESP-007) is contained in this section.
ICs are a predetermined subset of conditions where either the potential exists for a radiological emerency or such an emerency has occurred. Additionally, ICs are the means by which EALs for different nuclear power plants are standardized.
- 4. EAL Each Emergency Action Level exactly as it is contained in ERP-101,
- 5. OPCON The operational condition (OPCON) that the EAL is applicable in is contained here. There are six OPCONs (1,2.3,4 and 5 and defueled) that are used. It should be noted that these OPCONs are entry level conditions. The EAL is applicable if the plant was in the OPCON at the start of the event. Subsequent positcns of the mode selector switch should be ignored for purposes of classification.
OPCON GRAPHIC OPCON
[i12141415101 1 I1]:13141s101 2 h-Itl s idi slol 3 QTfD141slol 4 1.112181416101 5 111:1a141s101 Defueled
- 6. BASIS The technical basis of each EAL is contained in this section. This includes any necessary ca'culations and also includes escalation references.
l
- 7. DEVIATION i Any deviations from the NUMARC NESP-007 methodology are contained in this section, if there are no deviations, NONE is used.
- 8. REFERENCES l All applicable references used in developing the technical basis for each EAL are contained in this section. l l
I
LoS EAL Technical Basis Manual REV c. Apnl 26,1995 Page 4 of 130 Section ll- Acronyms AC -
Alternating Current ADS -
Automatic Depressurization System APRM -
Average Power Range Monitor ARI -
Alternate Rod Insertion ARM -
Area Radiation Monitor ATWS - Anticipated Transient Without Scram BRP -
Bureau of Radiation Protection CDE -
Committed Dose Equivalent CFM -
Cubic Feet Per Minute CFR -
Code of Federal Regulations CRD -
Control Rod Drive CS -
Design Basis Accident DC -
Direct Current DEI -
Dose Equivalent lodine .
EAL -
Emergency Action Level ECCS -
Emergency Core Cooling Systems EDG -
Emergency Diesel Generator EPA - Environmental Protection Agency ERFDS -
Emergency Response Facility Data System ERP-C -
Emergency Response Procedure - Common FC -
Fuel Clad (Barrier)
FTS -
Federal Telephone System GPM -
Gallons Per Minute HCTL -
Heat Capacity Temperature Limit HPCI -
High Pressure Coolant injection IC -
Initiating Condition IRM -
Intermediate Range Monitor KV -
Kilovolt LCO -
Limiting Condition for Operation LGS -
Limerick Generating Station LOCA -
Loss of Coolant Accident LPCI -
Low Pressure Coolant injection MPH -
Miles Per Hour mR/hr -
Milli Roentgen Per Hour MSIV -
Main Steam Isolation Valve NFPB -
Normal Full Power Background NPSH -
Net Positive Suction Head NRC -
Nuclear Regulatory Commission NUMARC -
Nuclear Management and Resources Council ODCM -
Offsite Dose Calculation Manual OPCON -
Operating Condition PEMA -
Pennsylvania Emergency Management Agency PC -
Primary Containment (Barrier)
PSIG -
Pounds Square Inch Gauge RC -
Reactor Coolant (Barrier)
RCIC -
Reactor Core Isolation Cooling
LGS EAL Technical Basis Manual REV a, April 26,1995 Page 5 of 130 RCS - Reactor Coolant System RHR -
Residual Heat Removal RPS - Reactor Protection System RPV - Reactor Pressure Vessel RRCS - Redundant Reactivity Control System SBO -
Station Blackout .
SJAE - Steam Jet Air Ejector l SRM - Source Range Monitor SRV -
Top of Active Fuel TPARD -
Total Protective Action Recommendation Dose TRIPS - Transient Response implementation Plan Procedures Cl/cc Micro Curie Per Cubic Centimeter pCi/gm -
Micro Curie Per Gram !
UFSAR- Updated Final Safety Analysis Report l VDC -
Volts Direct Current t
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LGS EAL Technical Bas's Manual REV c. April 26,1995 Page 6 of 130 Section lli - EAL Technical Basis
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LCs EAL Technical Basis Manual REV a. April 26,1995 I Page 7 of 130 1.0 Reactor Fuel 1.1 Coolant Activity i
UNUSUAL EVENT - 1.1.1.a IC Fuel Clad Degradation EAL .
Reactor Coolant activity > 4 pC//gm Dose Equivalent lodine 131 OPCON lil2I814Islol BASIS Coolant activity in excess of Technical Specifications (> 4 pCl/gm) is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition. This level is chosen to be above any possible short duration spikes under normal conditior.s. An Unusual Event is only warranted when actual fuel clad damage is the cause of the elevated coolant sample (as determined by laboratory confirmation). However, fuel clad damage should be assumed to be the cause of elevated Reactor Coolant activity unless another cause is known, e.g., Reactor Coolant System chemical decontamination evolution (during shutdown) is ongoing with resulting high activity levels.
This event will be escalated to an Alert when Reactor Coolant activity exceeds 300 Cl/gm Dose Equivalent lodine 131 per EAL Section 1.1.2.
DEVIATION The OPCON applicability [1,2,3,4] is a deviation from NUMARC [all) in that the Technical Specifications only require only 1,2,3,4.
REFERENCES Technical Specification Section 3.4.5 NUMARC NESP-007, SU4.2 e
l LGS EAL Techncal Basis Manual REV c. April 26,1995 Page 8 of 130 1.0 Reactor Fuel 1.1 Coolant Activity UNUSUAL EVENT - 1.1.1.b IC Fuel Clad Degradation EAL SJAE Discharge > 2.1x10' mR/hr OPCON h i r l
- I4181ol BASIS The steam Jet air ejector radiation monitor in the Control Room would be one of the first indicators of a degrading core. The high-high alarm is set at the Technical Specification limit of 21,000 rnRcir. This instrument takes a sample after the recombiner. This indicator of elevated activity is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition.
This event will be escalated to an Alert when RCS activity exceeds 300 pCi/gm Dose Equivalent lodine 131 per EAL Section 1.1.2.
DEVIATION None REFERENCES Technical Specifications Section 3.3.7.12,3.11.2.6 NUMARC NESP-007, SU4.1 l
lgs EAL Tschnied Basis Manual REv a, Apnl 26,1995 Page 9 of 130 1.0 Reactor Fuel 1.1 Coolant Activity ALERT - 1.1.2 IC Loss of Fuel Clad EAL Reactor Coolant activity > 300 pC#gm Dose Equivalent lodine 131 OPCON lil2181dI5101 BASIS A reactor coolant sample activity of greater than > 300 Cl/gm was determined to indicate significant clad heating and is indicative of the loss of the fuel clad barrier. This concentration is well above that expected for lodine spikes and corresponds to 2.8% clad damage. 2.8% fuel clad damage is based upon NUREG-1228 core damage analysis.
Calculation of 300 Cl/cc equivalance to percent fuel clad damage is as follows (for purposes of this calculation, cc and gm are considered equivalent):
lodine isotope Dose Factors Cl/MWe Values (Time After Shutdown = 0)
(Rea Guide 1.109) (NUREG 1228) 1-131 4.39E-3 85000 1-132 5.23E-5 120000 1-133 1.04E-3 170000 1134 1.37E-5 190000 1-135 2.14E-4 150000 Time After Shutdown (T = 0) Ratios '
R,33 = 120000/85000(I-131) = 1.41(1-131)
R,33 = 170000/85000(I-131) = 2.00(1-131)
R,u = 190000/85000(1-131) = 2.24(1131)
R,33 = 150000/85000(1-131) = 1.76(1131)
Equation for Dose Equivalent lodine (Deli 33)
A i3,DF,33 +(Rin)A i3,DFin+(R s)A is DFi33 ist +(Riu)Ai33 DFiu+(Ri35)A i33 DFss i del,3, =
So;ve for A,3, assuming del,33 = 300 Cl/cc
- l t
l LGS EAL Technical Basis Manual REV a, Apnf 26,1995 Page to of 130 Assi4.39E-3+1.41 A,3,5.23E-5+2.00A,331.04E-3+2.24A33 ,1.37E-5+1.76Aisi2.14E-4 4.39E-3 300= i 4.39E-3 Therefore: A,3, = 189 Cl/cc l-131 Clad damage fraction (NUREG-1228, Table 4.1) = .02 Full Power = 1150 MWe Clad Activity 1-131 = (Cl/MWe) (MWe) (Clad Damage Fraction) i
= (85000Cl/MWe) (1150MWe) (.02) l
= 1.96E6 Cl Reactor Water Volume = 2.93E8 cc (ERP-C-1410)
Total Coolant Activity 1131 = (A,33) (Rx Water Volume) (Cl/pCl)
= (189 pCl/cc) (2.93E8ce) (1.0E-6Cl/ pCl)
= 5.54E4Cl Percent Clad Damage = Total Coolant Activity / Clad Activity 1-131
= (5.54E4) / (1.96E6)
= 2.8%
This event will be escalated to an Site Area Emergency when additional fission product barriers are lost per EAL Section 1.1.3.
DEVIATION None REFERENCES NUMARC NESP 007, FC EAL #1 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents, Table 2.2 l Reg. Guide 1.109, Table E-9 ERP -C-1410, Page 15
lgs EAL Technical Basis Manud REV a. Apnl 26.1995 Page 11 of 130 l
1.0 Reactor Fuel L 1.1 Coolant Activity SITE AREA EMERGENCY - 1.1.3 l
l lC Loss of Fuel Clad and Loss of Reactor Coolant System or Containment l
l EAL l Reactor Coolant activity > 300 yC//gm Dose Equivalent lodine 131 AND Identified breach of Primary Containment (Tech Specs Section 3.6.1.1) OR Drywell Pressure > 10 ps/g l
l l OPCON lil218Idl8101 1
BASIS This EAL indicates a loss of the fuel clad barrier and the loss of primary containment or loss of ,
the reactor coolant barrier. A reactor coolant sample activity of > 300 Ci/gm Dose Equivalent lodine 131 was determined to indicate significant clad heating and be indicative of the loss of the fuel clad barrier. Technical Specifications 3.6.1.1 define conditions which must be met to consider ;
I the primary containment barrier intact. Intentional venting of the containment is also included.
These conditions include:
- all primary containment penetrations required to be closed during accident conditions are either capable of being closed by an operable primary containment automatic isolation system or closed by at least one manual valve, flange or deactivated automatic valve secured in its closed position
- all primary containment equipment hatches are closed and sealed
- the primary containment airlock is in compliance with the requirements of Specification 3.6.1.3 l - primary containment leakage rates are within the limits of Specification 3.6.1.2
- the suppression chamber is in compliance with the requirements of Specification 3.6.2.1
- the sealing mechanism associated with each primary containment penetration is operable The continual increase in suppression pool pressure, or the sudden increase in pressure, to 10 psig indicates the presence of a large breach in the reactor coolant pressure boundary and subsequent release of high energy reactor coolant into the containment and thus a loss of the Reactor Coolant barrier.
This event will be escalated to a General Emergency through the loss or potential loss of the third fission product barrier per EAL Section 4.1.4,4.2.4 and 1.2.4.
._ . _-. _ . . . . . _ . . -, m - - -_. _ _ . _ _ .._. _
P lgs EAL Technical Basis Manual REV e, April 26,1995 Page 12 of 130 DEVIATION l An exception to the NUMARC Methodology was taken in that NUMARC states that the site- !
specific drywell pressure should be based on the drywell high pressure alarm setpoint which indicates a LOCA (50gpm leak). The high drywell pressure alarm is 1.68 psig and can be reached by a small primary system leak and/or loss of drywell cooling, which are addressed in EALs 3.1.1.a and 3.1.1.b, respectively. The value of 10 psig was selected in that it is larger than experience shows of blown packing and recirc seat leaks. The value of 10 psig is more representative of a LOCA condition and this suppression pool pressure is in the TRIPS for actions to protect the containment.
REFERENCES ;
NUMARC NESP-007, FC EAL #1, PC EAL #5 and RC EAL #2 Technical Specifications Section 3.6.1.1 T-102, Primary Containment Control, PC/P-5 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents to t
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i LGS EAL Technical Basis Manual REV a. Apnl 26.1995 Page 13 of 130 1.0 Reactor Fuel 1.2 Containment High Radiation ALERT - 1.2.2 IC Loss of Fuel Clad EAL Post LOCA Drywell Rad > 3.0xW R/hr OPCON 18I2181d15101
~
BASIS The reading of 300 R/hr (from shine) on the primary containment Post LOCA Drywell Radiation Monitors (RR 26191/291 A,B,C,D) has been calculated to correspond to 300 Cligm Dose Equivalent lodine 131 within the Reactor Coolant. This is based on a direct ratio of 9000 R/hr shine with 100% clad failure to 2.8% clad failure.
300 Ci/gm Dose Equivalent lodine 131 = 2.8% Clad Damage Since:
Drywell Rad of 9000 R/hr shine = 100% Clad Damage (2.8%)(9000 R/hr)/(100%) = approximately 250 R/hr Containment Dose Rate rounded to 300 R/hr for human factors for 2.8% Clad Damage OR 300 pCl/gm Dose Equivalent lodine 131 The value of 300 R/hr is used for human factors of the logarithmic meter and strip paper recorder.
This reading is significantly higher than readings expected during normal operations and also significantly lower than the readings expected with this coolant concentration released into containment. This concentration indicates significant clad heating and is indicative of the loss of the fuel clad barrier. This corresponds to approximately 2.8% clad damage based upon NUREG-1228 core damage analysis (detailed calculations are contained in the Basis for EAL 1.1.2).
This event will be escalated to a Site Area Emergency with the loss of an additional fission product barrier per EAL Section 1.2.3. ;
DEVIATION None l
f'
, LGS EAL Technical Basis Manuel REV e, Apnl 26,1996 Page 14 of 130 REFERENCES NUMARC NESP 007, FC EAL #4 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Flant Accidents GE Assumptions - Mitigation of Core Damage
., . . , . - . _ - _ _ _ _ _ - - - , . . , . _ _ . - . . _ _ _ ~ ~ _ . _ - _ -
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3 l-LGS EAL Trchnical Basis Manual REV a. April 26,1995 rW)e 15 of 130 1.0 Reactor Fuel 1.2 Containment High' Radiation SITE AREA EMERGENCY - 1.2.3.a IC Loss of Fuel Clad and Containment EAL Post LOCA Drywell Rad > 3.0x10' R/hr AND Identified breach of Primary Containment (Tech Specs Section 3.6.1.1)
OPCON 1812I 141s101 BASIS This EAL indicates the loss of the fuel clad barrier and the loss of the containment barrier. The reading of 300 R/hr (from shine) on a primary containment Post LOCA Radiation Monitors (RR-26-191/291 A,B,C,D) has been calculated to correspond to 300 pCl/gm Dose Equivalent lodine 131 within the RCS. This is based on a direct ratio of 9000 R/hr shine with 100% clad failure to 2.8% clad failure. This reading is significantly higher than readings expected during normal operations and also significantly lower than the readings expected with this coolant concentration released into containment. This concentration indicates significant clad heating and is indicative of the loss of the fuel clad barrier. This corresponds to approximately 2.8% clad damage based upon NUREG-1228 core damage analysis (detailed calculations are contained in the Basis for EAL 1.1.2).
Technical Specifications 3.6.1.1 define the conditions which must be met to consider the primary containment barrier intact. These conditions include:
- all primary containment penetrations required to be closed during accident conditions are either capable of being closed by an operable primary containment automatic isolation system or closed by at least one manual valve, flange or deactivated automatic valve secured in its closed position
- all primary containment equipment hatches are closed and sealed the primary containment airlock is in compliance with the requirements of Specification 3.6.1.3
- primary containment leakage rates are within the limits of Specification 3.6.1.2
+ the suppression chamber is in compliance with the requirements of Specification 3.6.2.1
- the sealing mechanism associated with each primary containment penetration is operabl.
This event will be escalated to a General Emergency when additional barriers are lost per EAL Sections 4.1.4, 4.2.4 and 1.2.4.
l lgs EAL Technical Basis Manuel REV a, April 26,1995 Page 16 of 130 DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4 and PC EAL #5.
Technical Specifications Section 3.6.1.1 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents GE Assumptions - Mitigation of Core Damage r
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l LGS EAL Tcchnical Basis Manual REV c. April 26,1995 Page 17 of 130 1.0 Reactor Fuel 1.2 Containment High Radiation SITE AREA EMERGENCY - 1.2.3.b IC Loss of Fuel Clad and Reactor Coolant System EAL ,
Post LOCA Drywell Rad > 4x10' R/hr OPCON lilaisidisipl BASIS The reading of 4x10' R/hr on primary containment Post LOCA Radiation Monitors (RR 191/291 A,B,C,D) indicates the loss of both the fuel and Reactor Coolant System (RCS) fission product barriers. The reading was calculated assuming an instantaneous release of the reactor coolant volume into the Primary Containment (direct reading not shine) at a coolant concentration of 300 pCl/gm Dose Equivalent lodine 131. This calculation is as follows:
Using attachment 5(B) figure 1, Curve 3 [1%) of ERP-C-1410 Time after Shutdown = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 15,000 R/hr Extrapolating to 2.8%
(15,000 R/hr/1%)(2.8) = 42,000 R/hr This is rounded conservatively to 40,000 R/hr for human factors considerations 2.8% clad damage is based upon NUREG 1228 core damage analysis, and by virtue of its release into containment, the loss of the RCS barrier (detailed calculations are contained in the Basis for EAL 1.1.2).
This event will be escalated to a General Emergency when the monitor reading exceeds a value corresponding to 20% fuel clad damage (3x105 R/hr) or the containment barrier is lost per EAL Section 1.2.4.
DEVIATION None
, 1 LOS EAL Technical Basis Manual
, REV a, April 26.1995 Page 18 of 130 REFERENCES NUMARC NESP-007, FC EAL #3 and RC EAL #3 ,.
NUREG 1228 - Source Term Estimation During incident Response to Nuclear Power Plant '
Accidents .
ERP-C-1410, pg 12, Attachment 5(B) i i
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f lgs EAL Technical Basis Manual REV c. April 26,1995 Page 19 of 130 1.0 Reactor Fuel 1.2 Containment High Radiation GENERAL EMERGENCY - 1.2.4 IC Loss of Fuel Clad and Reactor Coolant System and Potential Loss of Containment EAL Post LOCA Drywell Rad > 3x10' R/hr OPCON l il 2 I 81 d l 5 Iiil BASIS The primary containment Post LOCA Drywell Radiation Monitors (RR-26191/291 A,B,C,D) reading 3x10' R/hr indicates significant fuel damage, well in excess of that required for the loss of the Reactor Coolant System and Fuel Clad barriers. The reading was calculated assuming an instantaneous release of the Reactor Coolant volume into the Primary Containment (direct reading not shine) where the value corresponds to a release of approximately 20% of the gap region.
This calculation is as follows:
Using attachment 5(B) figure 1, Curve 3 [1%) of ERP-C-1410 Time after Shutdown = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (more conservative due to lower value for EAL) 1% fuel clad damage :
the dose rate = 15,000 R/hr Extrapolating to 20%
(15,000 R/hr/1%)(20%) = 300,000 R/hr Regardless of whether containment is challenged, this amount of activity in containment, if released, could have severe consequences and it is prudent to treat this as a potential loss of containment and declare a General Emergency. NUREG-1228, " Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents" states that releases of this magnitude are not possible if plant systems function as designed and any accident with a release of 20% or greater of the gap region must be considered severe.
DEVIATION None
\
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. LOS EAL Technical Basis Manual REV a April 26,1995
' Page 20 of 130 l REFERENCES-NUMARC NESP-007, FC EAL #4, RC EAL #3 and PC EAL #3 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents ERP-C- 1410, pg 12, Attachment 5(B)
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i LGS EAL Ttchnical Basis Manual I REV e. April 26,1995 f Page 21 of 130 l 1.0 Reactor Fuel 1.3 Irradiated Fuel or New Fuel l UNUSUAL EVENT - 1.3.1.a IC Potential increase in Refuel Floor Radiation due to Lowering Water Level Covering Fuel l EAL l Unexpected RPV level decrease to < 120"when reactor cavity is flooded up and fuel pool gates are in place OPCON I$1slaidislol BASIS )
During refueling operations, RPV level indication is read on Panel C602 (Instrument Ll-IR605).
This instrument has a range of 0 - 500" with normal level at 420".
An unexpected level decrease below 120", which is approximately 22 feet above the fuel l indicates the cavity is being drained. This is validated with visual observation of a decreasing ;
reactor cavity levelif the RPV head is removed. i The value of 120" is calculated from determining 22' above the fuel in the RPV. Technical l Specifications requires 22' above the fuel, which equals 264" by (22')(12"/1') = 264". Top of Active Fuel = -161", so 264" - 161" = 103" or conservatively 120" This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to public health and safety is very low. Classification as an Unusual l Event is warranted as a precursor to a more serious event.
This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor per EAL section 1.3.2.a. l DEVIATION
)
i None l l
REFERENCES NUMARC NESP-007, AU2.1 Technical Specifications l
f lgs EAL Technical Baris Manual nEv c. April 26,1995 Page 22 of 130 1.0 Reactor Fuel 1.3 Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.3.1.b IC Potential increase in Refuel Floor Radiation due to Lowering Water Level Covering Fuel EAL Fuel Pool Storage Lo Level alarm with visual observation of a water level decrease in the Spent Fuel Pool OPCON lil2181dislol BASIS A drop in the Spent Fuel Pool level or the RPV [when in refueling and flooded up with gates out] will result in a control room annunciator Fuel Pool Storage Lo Level Alarm. This Control Room alarm directs an operator to be dispatched to a local alarm panel to investigate the reason for the alarm. This alarm is validated with visual observation of a decreasing Spent Fuel Pool level. If the spent fuel pool level decreases below the inlet to the skimmer surge tank, without a planned event such as removing a large piece of equipment, there must be a leak in the spent fuel pool or the RPV. This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to public health and safety is very low. Classification as an Unusual Event is warranted as a precursor to a more serious event.
This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor per EAL section 1.3.2.
DEVIATION None REFERENCES NUMARC NESP-007, AU2.1, AU2.2 GP-6.1, Shutdown Operations - Refueling, Core Alteration and Core Off-loading GP-6.2, Shutdown Operations - Shutdown Condition Technical Specification Actions
l lgs EAL Technical Basis Manual REV a April 26,1995 Page 23 of 130 1.0 Reactor Fuel 1.3 Irradiated Fuel or New Fuel ALERT - 1.3.2.a ,
IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the '
Uncovering of irradiated Fuel Outside the Reactor Vessel EAL Unplanned general area radiation > S00 mR/hr on the refuel floor (Table 1-1) .
OPCON lil21*l415Iol BASIS Unexpected radiation levels which are at least 100 times higher than the normal background will generally indicate a fuel handling accident or loss of water covering the irradiated fuel. Readings may be from refuel floor Area Radiation Monitors or taken during a qualified radiological survey.
Table 1-1 monitors are as follows:
Table 1-1 Refuel Floor ARMS RIS29-M1-1(2)K600, Drywell Head Laydown RIS30-M1-1(2)K600, Dryer /Seperator Area RIS31-M1 1(2)K600, Spent Fuel Pool RIS32-M1-1(2)K600, New Fuel Storage Vault RIS33-M1-1(2)K600, Pool Plug Laydown Offsite doses during these accidents would be well below the EPA Protective Action Guidelines and the classification as an Alert is therefore appropriate. This radiation level could also be caused by an inadvertent criticality and is included even though the probability of this event occurring is low. Radiation increases above 500 mR/hr which were expected should not cause an Alert to be declared during a planned evolution. Additionally, surveys which identify " hot ;
spots" greater than 500 mR/hr should not cause an Alert to be declared.
This event will be escalated to a Site Area Emergency by way of offsite doses per EAL Sections 5.1.3.
DEVIATION None !
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! LGS EAL Technical Basis Manuel L REV c. April 26.1995 Page 24 of 130 REFERENCES
! NUMARC NESP-007, AA2.1, AA2.2, AA2.3 and AA2.4 NUREG 1228, Source Term Estimation During incident Response to Severe Nuclear Power Plant l
Accidents -
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lgs EAL Technical Basis Manual 4 REV a, April 26,1995 Page 25 of 130 1.0 Reactor Fuel !
1.3 Irradiated Fuel or New Fuel ALERT - 1.3.2.b IC Unexpected increase in Plant Radiation or Airbome Concentration EAL .
i A fuel handling accident causing a HiHi alarm from the Refueling Area Exhaust Duct monitor (2 mR/ht)
OPCON ITTibI4Ia101 BASIS
~
The Refueling Area Exhaust Duct monitor (RRSH-26-1(2)R606A,C and RRSH-26-1(2)R613B/D) alarm is indicative of a release rate which may exceed Technical Specifications from the refuel floor. Although the release rate does not in itself warrant the declaration of an Alert, this alarm could indicate that a fuel handling accident has taken place causing major damage to irradiated fuel. The event should be validated by a report from the scene of a fuel handling accident or a radiological survey. It is not intended to indicate planned or expected alarms.
This event will be escalated to a Site Area Emergency due to offsite doses per EAL Section 5.1.3.
DEVIATION None REFERENCES ,
NUMARC NESP-007, AA2.1 ;
EPA-26-24-5 t
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l t.GS EAL Technical Basis Manual l REV c. April 26.1995 Page 26 of 130 1
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lgs EAL Technicd Basis Manual REV a. April 26 1995 Page 27 of 130 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level ALERT - 2.1.2 IC Potential Loss of Reactor Coolant System EAL yp.
RPV level < -129 "
OPCON lilaislaisioi BASIS As reactor water level decreases from it's normal band, a multitude of automatic actions should have occurred. These automatic actions include a reactor scram, trip of tha recirculation pumps, isolation of selected systems, and automatic start of both HPCI and RCIC. Some of these activities are designed to limit inventory loss from the RPV and others are designed to provide inventory makeup to the RPV. Continued decrease in RPV level to -129" indicates that either these actions were unsuccessfulin limiting inventory loss from the RPV or that there is a breech in the primary systems which has caused the RPV to depressurize or is at a rate which can not be overcome by the aforementioned actuations. This RPV level (-129") value is also used for various other system isolations and actuations including: initiation set point for the low pressure Emergency Core Cooling Systems; start signal for the Emergency Diesel Generators; containment isolation signal; and, as a permissive to the Automatic Depressurization system.
Reactor water level decreasing to the low, low, low reactor water level (-129") set point is indicative of a major plant transient. Although actual core uncovery does not begin until RPV level has decreased to -161" (Top of Active Fuel), this value has been selected to characterize a loss of coolant event. When -129"is reached an automatic isolation of the M!ilVs will occur. Decay heat generated from the fuel after this isolation occurs will need to be removed and/or transferred to the containment. Core Submergence is the preferred method of heat removal from the nuclear fuel. Maintaining RPV level above 2/3 core height ensures that the Reactor Fuel Cladding integrity will remain intact. RPV level decreasing to -129" is an abnormal event and signifies the possibility that level may continue to degrade to the point where submergence no longer occurs, thus this event signifies a potential loss of reactor coolant barrier. If a LOCA has occurred, this event should be declared even if reflood to > -129" is successful.
Under an Anticipated Transient without Scram (ATWS) scenario, it is possible that actions will be taken to intentionally lower RPV level to below -129". Additionally, there are provisions under an ATWS scenario to permit the MSIVs to remain open even though RPV level has been reduced to below -129". These events will be classified under EAL Section 2.2, Reactor Power and should not be classified under this EAL.
. ._. _ - _. ~ . . . .._ .
I LGS EAL Technmal Bash Manuel REV c. April 26,1995 Page 28 of 130
- This event will be escalated to a Site Area Emergency based upon a loss or potentialloss of the Fuel Clad or Containment barriers per EAL Sections 1.1,1.2 and 2.1.
DEVIATION None REFERENCES ,
NUMARC NESP-007, RC EAL #5 T-101, RPV Control I
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f LGS EAL Technical easis Manual REV c. April 26.1995 Page 29 of 130 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level SITE AREA EMERGENCY - 2.1.3.a IC. Potential Loss of Fuel Clad and Loss of Reactor Coolant System EAL RPV level cannot be restored above -167 "
l OPCON lil21sl415lol BASIS Reactor Pressure Vessel Water level has decreased below and cannot be maintained Above the Top of Active Fuel (TAF)s -161 inches as indicated on RPV Fuel Zone Level Instruments. Core submergence ensures adequate core cooling. When RPV level decreases below the top of active fuel the ability to remove the decay heat generated from the nuclear fuel becomes suspect and the Fuel Clad Fission Product barriers can no longer be considered intact. Sustained partial or total core uncovery can result in the release of a significant amount of fission products to the reactor coolant. Sustained core uncovery can also result in a breach of the reactor coolant system.
This event signifies a potential loss of fuel cladding integrity as the ability to effectively remove heat from the nuclear fuel cannot be assured. The assumption can also be made that for level to decrease to below the top of active fuel, there is an inability to restore and maintain core submergence and a breach in Reactor Coolant system integrity. The combination of Emergency Core Cooling Systems (ECCS) and normal water sources are sufficient to restore level when the reactor coolant system is intact, thus there is a loss of Reactor Coolant system integrity when RPV level falls below and cannot be restored to above the top of active fuel.
Core uncovery is expected to occur under large break LOCA scenarios. This inventory loss and subsequent RPV prassure reduction provides a low pressure ECCS initiation signal and permits the injection of these low pressure water sources. When deciding on the ability to restore RPV level consideration should be given to the availability of low pressure sources (including non-ECCS sources), the injection status of these systems (i.e. awaiting start permissives, injecting as full flow, etc.), and the trend of indicated RPV level.
Prior to concluding that RPV level cannot be restored consideration must be given to injection system availability and status, reactor pressure and rate of depressurization, and trend of the rate at which RPV level is decreasing. Ample time should be allotted to analyze the ability of injection sources to restore levelin an expeditious manner (after completion of the blowdown stage of the postulated LOCA). Loss of core submergence by itself should not be utilized to classify this event .
If there is indication that injection systems will be successful in recovering RPV water level.
t1 l lgs EAL Technical Basis Manu11 ;
REV c. Apnl 26,1995 {'
Page 30 of 130
- Under an Anticipated Transient without Scram scenario is possible that actions will be taken to intentionally lower RPV level at or below -161 inches These events will be classified under EAL Section 2.2, Reactor Power and should not be classified under this EAL.
Escalation to a General Emergency would be based on the inability to restore RPV water level for a sustained period of time per EAL Section,2.1.4 DEVIATION I
None REFERENCES NUMARC NESP-007, FC EAL #2 and RC EAL #4 T-101, RPV Control, RC/L-7 T-111, Level Restoration / Steam Cooling T-117, Level / Power Control T-116, RPV Flooding l
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lgs EAL Technied Basis Manual REv a. April 26.1995 Page 31 of 130 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level
' SITE AREA EMERGENCY - 2.1.3.b IC Potential Loss of Fuel Clad and Loss of Reactor Coolant System EAL RPV level cannot be determined OPCON litrisl41stol B' ASIS i
inability to determine Reactor Pressure Vessel level may be due to reference leg boil-off, instrument power failure, or conflicting information on uncontrolled parameter oscillations. TRIP procedure guidance will require the flooding of the Reactor Pressure Vessel, thus ensuring core submergence. Based on difforences in calibration and design,all ranges of levelinstruments may not indicate exactly the same; this operational difference is expected and is not to be used for deciding that conflicting RPV level inoication exists. Multiple indications of level instruments pegged high is indication that the level is above the range and that it is known, also visual observation during refueling is indication of RPV water level, i If indeterminate Reactor Pressure Vessellevel is due to one of the reasons mentioned above, adequate core cooling would be rapidly assured using the guidance provided in the TRIP Procedures; however, if water level cannot be determined, it is conservative to assume that water level is actually below the top of active fuel and that both the Reactor Coolant System and Fuel Clad Fission Product Barriers are potentially lost.
Operator attention should be given to the possibility that under depressurized conditions, there is the possibility that gases may come out of solution and result in distorted RPV level indications.
Operators should be attentive to observe multiple level indications (particularly those which utilize separate reference legs) to ensure that actual RPV level is known and displayed. Unexplained and/or sudden changes in specific level indications may be a result of degassification of the coolant contained in the level instrumentation.
Escalation to a General Emergency would occur if minimum RPV flooding pressure cannot be established per EAL Section 2.1.4.b.
DEVIATION None
-. .. - - m _ _ ..
4 l
LOS EAL Techmcel Basis Manual ^
REV c, Apnf 26,1995 Page 32 of 130 REFERENCES NUMARC NESP-007, FC EAL #4 and RC EAL #5 l T-101, RPV Control, RC/L-1 T-112, Rapid Depressurization T-117, Level / Power Control ~
T 116, RPV Flooding ;
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i LGS EAL Technical Basis Manua! 4 REV a. Apnl 26,1995 !
Page 33 of 130 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level GENERAL EMERGENCY - 2.1.4.a l lC Loss of Fuel Clad, Loss of Reactor Coolant System and potential Loss of Containment 1
EAL RPV level cannot be restored above -204 "
OPCON lil21*ldl5Iol BASIS ,
Core submergence is the preferred method of core cooling and as such, the failure to reestablish RPV water level above the top of active fuel for an extended period of time could lead to significant fuel damage. This condition, - 204 ", could be indicative of a large break Loss Of Coolant Accident (LOCA) (where ECCS Systems are designed to maintain level at 2/3 core height) or a small LOCA with the inability of emergency core cooling systems to reflood the RPV.
-204" was chosen as it represents 2/3 core height.
The time basis for deciding whether or not vessel flooding can be accomplished is dependent on the rate of reactor depressurization, the availability of low pressure ECCS systems, and the rate of RPV inventory loss. Indications such as RPV level trend, ECCS injection flow rates, containment parameter trends, and low pressure ECCS system operability should be considered in rendering the decision as to the ability to reflood the RPV.
Calculations in the UFSAR indicate that reactor fuel should be recovered minutes after a design basis LOCA. The inability to reflood the reactor following a LOCA may indicate severo ECCS degradation and/or multiple failures, including the possibility that jet pumps have failed.
The failure of the fuel cladding (due to overheating) together with the loss of reactor pressure vessel integrity (which would occur during a LOCA) and the introduction of a large amount of energy into the primary containment indicates that all three fission product barriers are in Jeopardy or lost.
Ample time must be allotted for ECCS systems to reflood the RPV. For events starting from power operation, the failure to rapidly reflood could result in some core melting. Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time. Sustained operation with water level below the top of active fuel represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions. Partial core submergence may provide adequate cooling although it cannot be assumed.
I lgs EAL Technical Basis Manuti REV a. Apnl 26.1995 Page 34 of 130 The inability to restore level may indicate the presence of an break in one of the jet pumps. This failure would prevent core reflooding with even all injection systems intact when combined with a large loss of coolant accident.
Prior to concluding that RPV level cannot be restored consideration must be given to injection system availability and status, reactor pressure and rate of depressurization, and trend of the rate at which RPV level is decreasing. Ample time should be allotted to analyze the ability of injection sources to restore levelin an expeditious manner (after completion of the blowdown stage of the postulated LOCA). Loss of core submergence by itself should not be utilized to classify this event if there is indication that injection systems will be successful in recovering RPV water level.
DEVIATION -
None REFERENCES NUMARC NESP-007, FC EAL #2, RC EAL #4 and PC EAL #4 T-101, RPV Control T-111, Level Restoration / Steam Cooling, LR 11 T-112, Rapid Depressurization T-117, Level / Power Control ;
t T 116, RPV Flooding l
l lgs EAL Tschncal Basis Manual REV a. Apnl 26,1995 Page 35 of 130 2.0 Reactor Pressure Vessel i
2.1 Reactor Water Level GENERAL EMERGENCY - 2.1.4.b IC Potential Loss of Fuel Clad and Loss of Reactor Coolant System and Containment EAL RPV level cannot be determined AND RPV Flooding cannot be established per T-116 OPCON 111:18I416101 BASIS The decision to enter RPV Flooding is made when RPV water level cannot be determined. This judgement consists of evaluating all plant indications which can influence the ability to maintain adequate core cooling. Entry to RPV flooding requires rapid RPV depressurization. The minimum RPV Flooding Pressure is defined as the lowest differential pressure between the RPV and the Torus at which steam f!ow through the SRVs will be sufficient to remove all of the generated decay heat. Operation at the minimum reactor flooding pressure requires that a sufficient amount of water reach the core to carry away decay heat by boiling, which in tum requires that RPV water level increase. So RPV flooding not established requires containment flooding.
Inability to determine Reactor Pressure Vessel level may be due to reference leg boil-off, instrument power failure, or conflicting information on uncontrolled indication oscillations. TRIP procedure guidance will require the flooding of the Reactor Pressure Vessel, thus ensuring core submergence. Based on differences in calibration and design, all ranges of level instruments may not indicate exactly the same; this operational difference is expected and is not to be used for deciding that conflicting RPV levelindication exists. Levelindication pegged high is indication that the level is above the range and that it is known, also visual observation during refueling is indication of RPV water level.
If indeterminate Reactor Pressure Vessel level is due to one of the reasons mentioned above, adequate core cooling would be rapidly assured using the guidanca provided in the Emergency Operating Procedures; however, if water level cannot be determined, it is conservative to assume that water level is actually below the top of active fuel and that both the Reactor Coolant System and Fuel Clad Fission Product Barriers are potentially lost.
The minimum RPV flooding pressure will ensure that adequate core cooling exists independent of RPV level indication. Failure to establish the differential pressure between the RPV and the Torus in a timely manor can jeopardize the ability of the reactor coolant system to dissipate the decay heat generated.
6 lgs EAL Technical Basis Manual REV c. April 26.1995 Page 36 of 130 Eventual clad failure may occur due to overheating of the nuclear fuelif RPV flooding pressure cannot be established in a timely maner. The heat produced from the fuel can cause additional core damage. If the cause of the RPV level problem was caused by a LOCA, then both the Clad and the Reactor Coolant have been lost. This will occur with heat being added to the containment. Thus there is a loss of the Fuel Clad and Reactor Coolant barriers with a potential loss of the Containment barrier.
Ample time must be allotted for determining the failure of ECCS systems to pressurize the RPV.
Control room indications such as RPV level (used for trending), RPV Pressure, ECCS injection flow rates, Containment parameters, and injection system operability should all be used to gauge the effectiveness of the RPV Flood.
If the loss of level indication was caused by reference leg flashing, then level indicators can still be utilized to monitor the trend in RPV level. Actual RPV level will never be higher than indicated level.
In the event that the loss of level indication is only a result of degassification of the coolant contained in the level instrumentation piping, then it is anticipated that flooding pressure can be obtained.
RPV water level below the top of active fuel for a sustained period of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions. For events starting from power operation, some core melting can be expected.
Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time.
DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4, RC EAL #5 and PC EAL #5 T-101, RPV Control T-111, Level Restoration / Steam Cooling, LR 11 T-112, Rapid Depressurization T 117, Level / Power Control T 116, RPV Flooding
LGS EAL Techneal Basis Manuel REV a. Apnl 26,1995 Page 37 of 130 2.0 Reactor Pressure Vessel 2.2 Reactor Power ALERT - 2.2.2 IC Failure of Reactor Protection System lastrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful EAL Failure of Automatic RPS SCRAM to reduce reactor power < 4%
OPCON l il a laldis101 BASIS Entry into this EAL is based on a reactor parameter actually exceeding a RPS setpoint and the reactor is not brought to < 4% power conditions and maintained at that state with automatic RPS functions. The parameter must exceed the RPS setpoint by a significant margin eliminating minor setpoint drifts which are accounted for in the Technical Specification Margin of Safety.
Subsequent manual scram actions were successful in bringing the reactor to < 4% conditions.
Confirmation indications include control room annunciators, APRM power, IRM/SRM power level, SRM period, and Control rod position indication.
)
l A failure of the Reactor Protection System (RPS) to initiate and complete a reactor scram may Indicate that the design limits of the nuclear fuel has been compromised. RPS is designed to ,
automatically detect and generate a reactor scram signal when a limiting safety system setpoint !
is reached or exceeded. Control rod insertion following a scram signal is designed to be passive l (i.e. system deenergizes, control rod motive energy source is previously charged).
Assuming that shutdown (< 4% power) conditions cannot be established / maintained, an automatic scram signal failure followed by a successful manual scram would still constitute a scram failure and should be classified under this event.
Although the reactor may be brought initially < 4% power based on partial control rod insertion, there is a possibility that positive reactivity may be introduced by a number of factors. Xenon decay and factors associated with cooldown, lower fuel temperature (doppler), lower moderator temperature, and a lower presence of steam bubbles (volds) may all contribute to cause a power increase.
Suberitical conditions can be assured even with the most reactive control rod fully withdrawn from the core if the remaining 184 control rods fully insert. Any other control rod pattern resulting from partial control rod insertion must be carefully analywJ and/or monitored to detect the possibility of recriticality or local criticality.
LOS EAL Technx:al Basis MJtnual REv a. Apnl 26,1995 Page 38 of 130 Due to the buildup of Xenon in areas of the core that have previously been operating at high !
l power levels, attention should be applied to the possibility that control rods which previously had low worth (e.g. peripheral control rods) may now have significant control rod worth.
When the reactor is not shutdown as identified in the TRIPS, then entry into this EAL is warranted.
When partial control rod insertion occurs following a scram signal (either manual or automatic) judgement should be applied as to whether classification should occur. Multiple control rods failing to insert beyond notch position 02 may require actions to fully insert the control rods.
However, the reactor has been made subcritical, and for all intent the reactor will remain subcritical. TRIP guidance will govern the insertion of these control rods.
This EAL would be escalated with a failure of both manual and automatic scram signals with APRM power remaining above 4% per EAL Section 2.2.3.
DEVIATION None REFERENCES NUMARC NESP-007, SA2 T-100, Scram T 101, RPV Control, RC-1
lgs EAL Technical Basis Manual REV a. April 26,1995 Page 39 of 130 2.0 Reactor Pressure Vessel 2.2 Reactor Power SITE AREA EMERGENCY - 2.2.3 IC Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful EAL Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%
OPCON lil218141*lol BASIS A valid automatic or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication is greater than 4%. The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatic). The system is " fail safe", that is it deenergizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydraulic failure).
A failure of the Reactor Protection System to shut down the reactor (as indicated by reactor power remaining above 4%) is a degraded plant condition that together with suppression pool temperature approaching 110 F requires the injection of boron poison to shut down the reactor.
The RPV Control Trip Procedure establishes 4% power coincident with loss of scram capability as the initiating condition for various plant responses to ATWS events. With Reactor Power less than 4% the heat being generated in the core can be removed from the RPV and containment while actions are taken to bring the reactor suberitical.
A manual scram is defined as any set of actions by the reactor operator (s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor suberitical (i.e. mode switch to shutdown, manual scram push buttons, manual ARI initiation or manual Redundant Reactivity Control System push bottons). Taking the mode switch to shutdown ,
as part of the actions required by trip procedure is considered a manual scram action.
While the plant is being shutdown, significant heat is being generated in the core and the heat up rate of the suppression pool (due to heat rejection through SRVs) can increase which could approach the suppression pool temperature limit prior to shutting down. As the suppression pool heat increases towards the limiting temperature, the probability of causing a major over-pressure event increases substantially.
I LGS EAL Technical Basis Manual REVG Apnf 24.1995 Page 40 of 130 After an ATWS event, there is a potential that the Main Steam isolation Valves will remain open.
There is additional guidance in the Trip procedures to ensure that the MSIVs remain open even if RPV level is intentionally lowered to below the normal MSIV isolation level. This situation would allow the plant to remove heat and provide makeup through the normal steam / feed cycle. If this path is not available, or becomes unavailable during the transient, heat rejection will be to the suppression pool.
With Standby Liquid Control initiated and with partial or no control rods insertion there is a possibility that the neutron flux profile in the reactor core may become uneven or distorted.
Localized clad damage is possible if localized power levels increase significantly.
With reactor power remaining above 4% containment integrity is threatened as the ability of systems to remove all of the heat transferred to the containment may be exceeded. As the energy contained in the containment increases there may be a degradation in the ability to remove heat generated by the "at power" reactor core. There is therefore a potential loss of the containment or the fuel cladding (caused by overheating).
This event will be escalated based on suppression pool temperature exceeding 180*F per EAL Section 2.2.4.
DEVIATION None REFERENCES NUMARC NESP 007, SS2 :
T-100, Scram T-101, RPV Control, RC/L-2 T-117, Level / Power Control
)
LGS EAL Tzchrucal Basis Manual I REv a. Apnl 26,1995 l Page 41 of 130 l l
2.0 Reactor Pressure Vessel :
2.2 Reactor Power GENERAL EMERGENCY - 2.2.4 IC Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scrarn was NOT Successful and There is indication of an Extreme Challenge to the
. Ability to Cool the Core EAL Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%
AND Suppression Pool Temperature is > 180*F OPCON lilrlsteistol BASIS A valid automatic or manual scram signal is present as indicted by control room Indications and/or alarms and APRM indication is greater than 4%. In addition control room instrumentation indicates that Suppression Pool temperature is > 180 F.
Failure of all automatic and manual trip functions coincident with a high suppression pool temperature will place the plant in a condition where reactivity control capability is Jeopardized and heat removal capability is severely limited.
ECCS systems which may be used to cool the core, transfer heat from the reactor, and/or supply cooling water to the reactor all take a suction of the suppression pool. Operation with sustained high suppression pool temperatures may render these systems inoperable due to NPSH considerations.
The RPV Control Trip Procedure establishes 4% power coincident with loss of scram capability as the initiating condition for various plant responses to ATWS events. The timely initiation of Standby Liquid Control (prior to suppression poci temperature reaching 110 F would bring the reactor to below 4% power before suppression pool temperature approaches the heat capacity temperature limit curve limitations.
Under ATWS conditions, it is important to assure continuous, stable steam condensation capability. An elevated suppression pool temperature of 180 F would result in unstable steam condensation should rapid reactor depressurization occur (ADS activation). 180 F is the suppression pool heat capacity temperature limit. Maintaining the ability to condense steam will preclude the pressurization of the containment and prevent possible containment failure.
l
! lgs EAL Technical Basis Manual REV c, Apnl 26,1995 Page 42 of 130 Containment over pressurization, which would be an eventual result of sustained operation with heat being added to the containment and suppression pool temperature above 180 F would result in the loss of containment integrity and the inability to remove the heat generated from the fuel.
Fuel clad failure would result from the overheating of the fuel.
DEVIATIC" ,
None REFERENCES ,
NUMARC NESP-007, SG2 T-100, Scram T-101, RPV Control T-117, Level / Power Control, RC/L 2 1
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lgs EAL T:chnical Basis Manual REV c. Apnl 2G.1995 Page 43 of 130 3.0 Primary Containment 3.1 Containment Pressure or Leakage UNUSUAL EVENT - 3.1.1.a IC Reactor Coolant System Leakage EAL Drywell Pressure > f.68 ps/g AND Indication of a Leak into Containment OPCON lil*l3]Aj$lp]
BASIS Reaching 1.68 psig in the Drywell is likely indication of a primary system leak. Upon receipt of the 1,68 psig drywell pressure signal, system response includes a reactor scram, ECCS initiation (including HPCI), tripping of the drywell cooling fans and isolation of the cooling water to the drywell. These actuations may mask the trend in drywell pressure. For example, the scram will result in less heat being added to the containment and the cooling water isolation will result in no heat being removed. Trip procedure guidance will include establishment of cooling for the containment.
There will be control room annunciation indicating the presence of a high drywell pressure condition (1.0 psig) prior to the threshold value of 1.68 psig being reached. Actions initiated as a result of the indication of elevated drywell pressure conditions include stopping energy addition to the containment (e.g. Nitrogen make up), maximizing drywell cooling, and possibly lowering drywell pressure by venting (if permitted by procedures and plant conditions). Continued increase in drywell pressure or a sudden increase in drywell pressure may indicate a leak of high energy reactor coolant into the containment. l Indication of a Leak into Containment was added to qualify the pressure indication to avoid i declaring an emergency for situations where the pressure increase is clearly not due to a primary l system leak. For example, an emergency declaration is not appropriate if the high drywell j pressure is a result of a loss of Drywell Cooling. Indication of a leak should be determine by )
observing other cor:tainment indications such as sump level, ambient radiation, ambient temperature, and status of cooling systems.
This event will escalate to an Alert based on containment pressure reaching 10 psig per EAL Section 3.1.2 I
l l
LGS EAL Techmcal Basia Manuel REV c, April 26,1995 Page 44 of 130 DEVIATION .
4 None REFERENCES NUMARC NESP-007, SU5 T-100. Scram T-101, RPV Control T-102, Primary Containment Control ,
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LGS EAL Techncal Basis Manual REV a, April 26,1995 Page 45 of 130 3.0 Primary Containment 3.1 Containment Pressure or Leakage UNUSUAL EVENT - 3.1.1.b IC Reactor Coolant System Leakage EAL Unidentified Primary System Leakage > 10 ppm into the Drywell OR Identified Primary System Leakage > 25 ppm into the Drywell OPCON lil21sidl*lol BASIS Utiiizing the leak before break methodology, it is anticipated that there will be indicat;on(s) of minor reactor coolant system boundary integrity loss prior to this fault escalating to a major leak or rupture. Detection of low levels of leakage while pressurized is utilized to monitor for the potential of catastrophic failures.
This EAL is included as an Unusual Event because it may be a precursor of more serious conditions and, as a result, it is considered to be a potential degradation of the level of safety of the plant. The value of 10 gpm unidentified leakage is significantly higher than the expected pressurized leak rate from the reactor coolant system. The 10 gpm value for the unidentified pressure boundary leakage was selected as it is twice the Technical Specification value, indicating an increase beyond that assumed in Safety Analysis. It also is observable with normal control room indications. The EAL for identified leakage is set at a higher value (25 gpm) due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.
Technical Specification LCO required actions would necessitate a plant shutdown and subsequent depressurization, unless the source of the leak can be isolated, identified, and/or stopped.
Actions initiated by plant staff would include close monitoring of the calculated break size such that any sudden or gradualincrease in leak rate would be identified. A slow power reduction and ,
gradual depressurization would be necessitated due to the possibility that a sudden power and/or pressure surge could potentially worsen the break or cause a catastrophic failure.
The leak rate of 10 gpm may cause a high drywell pressure indication. Other indications of a leak of this magnitude would inc'ude an increase in drywell temperature or radiation.
This event will escalate to an Alert based upon high Drywell pressure per EAL Section 3.1.2.
- . . . _ . . _ m.- . - . _. _ _ -_ _ -
LOS EAL Technical Basis Manual -
REV c April 26.1995 Page 46 of 130 DEVIATION None REFERENCES !
NUMARC NESP-007, SUS )
Technical Specifications T-101, RPV Control T-102, Primary Containment Control r
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lgs EAL Technical Basis Manual REV s, April 26,1995 Page 47 of 130 ,
1 3.0 Primary Containment 3.1 Containment Pressure or Leakage l ALERT - 3.1.2 IC Loss of Reactor Coolant System EAL Drywell Pressure > 10 psig OPCON 1112181414I01 BASIG If drywell pressure exceeds 10 psig, there is a clear indication that a leak of sufficient magnitude exists that prevents drywell pressure stabilization.
Providing that RHR was not required to maintain reactor water level, these systems would normally be placed into the suppression pool cooling / spray mode following the receipt of the high pressure annunciators. This additional cooling, together with the reduction of heat being added to the containment (due to the reactor scram), should result in a reduction in primary containment pressure. The value of 10 psig is obtained from T-102, in that it is the pressure of the suppression pool where suppression pool sprays are initiated. For this EAL it will be used as the drywell pressure indicative of a loss of reactor coolant system. This value being used for the drywell is conservative in that it will be reached before the suppression pool. The continual increase in suppression pool pressure, or the sudden increase in pressure to 10 psig, indicates the presence of a large breach in the reactor coolant pressure boundary and subsequent release of high energy reactor coolant into the containment and thus a loss of the Reactor Coolant barrier.
This event will escalate to a Site Area Emergency based on a loss or potential loss of other barriers per EAL Sections 1.0,2.0 and 3.0.
DEVIATION An exception to the NUMARC Methodology was taken in that NUMARC states that the site-specific drywell pressure should be based e ine drywell high pressure alarm setpoint which indicates a LOCA (50gpm leak). The high dryuell pressure alarm is 1.68 psig and can be reached by a small primary system leak and/or '.oss of drywell cooling, which are addressed in EALs 3.1.1.a and 3.1.1.b, respectively. The ve',ue of 10 psig was selected in that it is larger than experience shows of blown packing and recirc seal leaks. The value of 10 psig is more representative of a LOCA condition and this suppression pool pressure is in the TRIPS for actions to protect the containment.
. . . _ . . .___. . _ _. .__ .~... ._. _ .._ _ . _ ..__.. .__ ._. _ _ _ _
i LOS EAL Techmcel Basis Manual 4 , REV e, April 26,1996 Page 48 of 130 REFERENCES.
. T-101, RPV Control T-102, Primary Containment Control, PC/P-5 i
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LGS EAL Technical Basis Manual REV c. April 26,1995 Page 49 of 130 3.0 Primary Containment 3.1 Containment Pressure or Leakage SITE AREA EMERGENCY - 3.1.3 IC Loss of Reactor Coolant System and Containment EAL Containment Failure indicated by a rapid, unexplained drop in Containment Pressure following initial pressure rise above 10 psig OPCON 11121*I418101 BASIS This EAL represents a condition where both the Reactor Coolant System and Containment Fission Product Barriers cannot be considered intact. Rapid, unexplained loss of pressure (i.e.,
not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity. Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, drywell pressure increasing to > 10 psig then dropping (without reason) indicates a loss of containment integrity.
Other indications such as Reactor Enclosure radiation levels, reactor enclosure area temperatures, stack radiation levels, and containment isolation status should be used to confirm the loss of containment integrity, if possible.
This Site Area Emergency declaration is based on the loss of Reactor Coolant System boundary as indicated by higher than expected containment pressurization followed by the loss of primary containment integrity as indicated by the rapid unexplained decrease in containment pressure.
This event would escalate to a General Emergency based on loss of the fuel clad per EAL Sections 1.0,2.0 and 3.0.
DEVIATION None REFERENCES NUMARC NESP 007, RC EAL #1 and PC EAL #1 T-101, RPV Control T-102, Primary Containment Control T-103, Secondary Containment Control
LGS EAL Technical Baso Manual I
REV a. Apnl 26,1995 Page 50 of 130 3.0 Primary Containment 3.1 Containment Pressure or Leakage GENERAL EMERGENCY - 3.1.4.a IC Loss of Reactor Coolant System and Containment wth potential loss of Fuel Clad EAL Containment Pressure > 62.5 ps/g OR Containment Venting via T-200 is required OPCON lil218Idlelol BASIS Should Containment Pressure exceed 62.5 psig, the Primary Containment Pressure Limit has been exceeded. This condition is reflective of failure of all Containment Pressure Control Systems and a loss of the pressure suppression mode. At this pressure the potential exists for uncontrolled and unpredictable breach of primary containment integrity and release of radioactivity to the environment.
This condition is also indicative of loss of the Reactor Coolant System fission product boundary since the only mechanism to pressurize containment to this pressure would be through the loss of the Reactor Coolant System.
This pressure is well above the maximum pressure expected to be present in primary containment during a design basis Loss of Coolant Accident. Before reaching this pressure procedural guidance is provided to vent the primary containment regardless of offsite dose consequences.
A controlled, monitored radiological release is preferred to an unisolable, uncontrolled, and potentially unmonitored release which could result from a containment failure. Entry into this EAL is also warranted when the primary containment is vented at pressures lower than 62.5 psig.
Containment failure could also result in subsequent loss of adequate core cooling due to loss of source water for ECCS, loss of ECCS pumps or piping integrity, or loss of Reactor Pressure Vessel support and integrity.
Thus this event reflects the loss of the Reactor Coolant and Containment Barriers and the potential loss of the Fuel Clad Barrier.
DEVIATION None
LGS EAL Technical Basis Manual REV a, April 26,1995 Page 51 of 130 ,
REFERENCES NUMARC NESP 007, FC EAL #4, RC EAL #5 and PC EAL #2 ,
T-104, Radioactivity Release Control
. T-112,~ Emergency Blowdown T 102, Primary Containment Control, PC/P-11 T-200, Primary Containment Venting L
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lgs EAL Techneal Basis Manual REV c. April 26.1995 Page 52 of 130 3.0 Primary Containment 3.1 Containment Pressure or Leakage GENERAL EMERGENCY - 3.1.4.b IC Loss of Fuel Clad and Reactor Coolant System and a potentialloss of Containment EAL Drywell Hydrogen 2 6 %
AND Drywell Oxygen 2 5 %
1' OPCON lil:18Idlelol BASIS This EAL is indicative of a potential loss of the Primary Containment fission product barrier, as well as the loss of the fuel clad and the RCS barriers.
Hydrogen gas concentrations exceeding 6% in the containment is representative of degraded fuel clad [approximately 15%), reactor coolant system leak and a challenge condition to containment.
This level of hydrogen is not expected to indicate an uncoolable core condition. If oxygen concentration exceeds 5%, an explosive mixture would exist and provide the potential for detonation. Hydrogen gas concentrations of this level are as a result of extensive zirc-water reaction, indicating loss of the fuel clad and loss of the reactor coolant system. It is unlikely that containment oxygen concentrations would exist due to the containment being inerted with nitrogen at power operations. The significant mitigative action is in the TRIPS is to require require Emergency Blowdown on the following conditions H,2 6% and O,2 5%.
A General Emergency is warranted because a LOCA and fuel clad damage have occurred with a potential for loss of containment due to an explosive mixture. This accident has produced significant levels of hydrogen, combined with high levels of oxygen would potentially produce an explosive mixture. With levels of hydrogen and oxygen this extreme there is a precursor to potential loss of containment and therefore warrants a General Emergency.
DEVIATION None l
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LGS EAL Technical Basis Manual REV a April 26,1995 ,
Page 53 of 130 REFERENCES
' NUMARC NESP-007, FC EAL #4, RC EAL #5, PC EAL #1 T-102, Primary Containment Control NUREG/BR-0150 Vol 1, Rev 3, Response Technical Manual (RTM-93) 4 L
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l lgs EAL Ttchne.al Basis Manual REV a. April 26.1995 Page 55 of 130 4.0 Secondary Containment 4.1 Secondary Containment Temperature ALERT - 4.1.2 IC Potentialloss of Reactor Coolant System and Potential loss of Containment EAL P
An Unisolable Primary System Leak is discharging into Secondary Containment AND A T 103 Temperature Max Safe Operating Value is exceeded in ONE area requiring a SCRAM OPCON lil218Isistol BASIS This EAL represents a challenge to both the Reactor Coolant and Containment barriers. The case of single area exceeding their Temperature Max Safe Operating Value indicates that there is a potential bypass of primary containment, as well as the potential loss of the reactor coolant pressure boundary by either a breech in high energy piping without isolation or interfacing systems LOCA. Increase in temperature in only one area indicates that the size of the leak is small enough to not cause a direct flow path to the environment. Temperature Max Safe Operating Value limits are located in T-103. '
TRIP guidance stipulates that when the Temperature Max Safe Operating Value limit has been exceeded for ONE area that the reactor be manually SCRAMMED.
There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e. primary leak into secondary and a fire within the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except those required for shutdown and cooling, at the Temperature Max Safe Operating Value Isolation levels. if the temperatures continue to rise to the Temperature Max Safe Operating Value it is indicative that an unisolable leak has occurred.
If an instrument line ruptures and cannot be isolated, these temperatures could reach the action levels and an ALERT is warranted.
This event will be escalated to a Site Area Emergency based upon verification of more than one area exceeding Temperature Max Safe Operating Value per EAL Section 4.1.3.
DEVIATION None
.. - . . . . . - _ - _ . _ _ . _ = . - . . _
LGS FAL Technical Ba:is Manual REV c, April 26,1995 Pa0e 56 of 130 REFERENCES NUMARC NESP-007, RC EAL #5, PC EAL #5 T-103, Secondary Containment Control, SSCfr-1 T-101, RPV Control T-112, Emergency Blowdown F
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lgs EAL Technical Basis Manual i
REV c. April 26,1995 Page 67 of 130 4.0 Secondary Containment 4.1 Secondary Containment Temperature SITE AREA EMERGENCY - 4.1.3 IC Loss of Reactor Coolant System and Containment EAL An Unisolable Primary System Leak is discharging into Secondary Containment AND T-103 Temperature Max Safe Operating Values are exceeded in TWO OR MORE areas requiring an Emergency Blowdown per T-112 OPCON lil21*Idisio1 BASIS This EAL represents a challenge to both the Reactor Coolant and Containment barriers. The case of multiple areas exceeding their Temperature Max Safe Operating Value indicates that there is a bypass of primary containment, as well as the loss of the reactor coolant pressure boundary by either a breech in high energy piping without isolation or interfacing systems LOCA.
Temperature Max Safe Operating Value limits are located in T 103.
TRIP guidance stipulates that when the Temperature Max Safe Operating Value limit has been exceeded for TWO OR MORE areas that the reactor be manually SCRAMMED and that an emergency blowdown be performed. The increase in Reactor Enclosure temperature in multiple areas may be an indication of a wide spread problem which may pose an immediate and direct threat to the Reactor Enclosure integrity, equipment and continued safe operation of the plant.
There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e. primary leak into secondary and a fire within the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except those required for shutdown and cooling, at the Temperature Max Safe Operating Value isolation levels. If the temperatures continua to rise to the Temperature May Safe Operating Value it is indicative that an unisolable leak has occurred.
This event will be escalated to a General Emergency based upon verification of Fuel Clad degradatiuon per EAL Section 4.1.4.
DEVIATION None
~_-_ . _ . . . - - _= -_ .. .. - ... -
LGS EAL Technical Basis Manuel REV ci, April 26,1995
-' Page 58 of 130 REFERENCES NUMARC NESP-007, RC EAL #5 and PC EAL #2
~ T-103, Secondary Containment Control, SSC/T-1
.T-101, RPV Control T-112, Emergency Blowdown 4
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LGS EAL Technical Ba;is Manual REV a. April 26.1993 !
Page 59 cf 130 4.0 Secondary Containment i
4.1 Secondary Containment Temperature GENERAL EMERGENCY - 4.1.4 Loss of Fuel Clad, Reactor Coolant System and Containment i BC EAL l I
An Unisolable Primary System Leak is discharging into Seccadary Containment AND T-103 Temperature Max Safe Operating Values are exceeded in TWO OR MORE areas requiring an Emergency Blowdown per T-112 AND T-103 Radiation Max Safe Operating Values are exceeded in the same TWO OR MORE areas OPCON lil Isidl*lol BASIS This EAL represents a challenge to the Reactor Coolant, Containment and Fuel Clad barriers. The case of multiple areas exceeding their Temperature Max Safe Operating Value indicates that there is a bypass of primary containment, as well as the loss of the reactor coolant pressure boundary by either a breech in high energy piping or an interfacing systems LOCA. T-103 identifies secondary containment areas with their Temperature Max Safe Operating Value and ,
their Radiation Max Safe Operating Values. For Radiation levels would not increase above these !
action levels unless fuel clad damage had occurred.
TRIP guidance stipulates that when the Temperature or Radiation Max Safe Operating Values have been exceeded for two or more areas that the reactor be manually scrammed and that emergency blowdown be performed. The increase in reactor enclosure temperature or radiation in multiple areas may be an indication of a wide spread problem which may pose an imn.ediate and direct threat to Reactor Building integrity, equipment and continued safe operation of the plant.
There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e. primary leak into secondary and a fire witilin the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except those required for shutdown and cooling, at the T-103 Temperature Max Safe Operating Value, i If the temperatures continue to rise to the Temperature Max Safe Opori..r g Value, it is indicative that an unisolable leak has occurred, and if the radiation levels rise above the Radiation Max Safe Operating Value it is an indication that fuel damage has occurred.
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- LGS EAL Technical Basis Manua!
- REV c. Apnl 26,1995 .
Page 60 of 130 .
DEVIATION
?
None ,
REFERENCES ,
NUMARC NESP-007, FC EAL #4, RC EAL #5 and PC EAL #2 T-103, Secondary Containment Control
< T-101, RPV Control T 112, Emergency Blowdown i
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I LGS EAL T:chnical Basis Manual REv c. April 26,1995 )
Page 61 of 130 4.0 Secondary Containment 4.2 Main Steam Line UNUSUAL EVENT - 4.2.1 IC Fuel C:ad Degradation EAL i
Main Steam Line HiHi Radiation (3xNFPB)
OPCON 1:1213 u1s1o1 BASIS Main Steam Line HiHi Radiation alarm (RE41/42N6,A,B,C,D) > 3 times normal full power background maybe indicative of minor fuel cladding degradation and warrants the declaration of an Unusual Event. This level is set to preclude most spurious events inc!uding resin intrusion.
The main steam line high-high radiation condition requires a manual Main Steam isolation Valve closure and a reactor scram. This transient may result in the introduction of fission product gases (previously contained in the gap area) to be suddenly released into the coolant due to the rapid down power transient and subsequent collapse of voids in the coolant.
This level of steam line activity is indicative of the release of gap activity to the coolant however, this level is not indication of a major failure of the fuel clad. The mechanics that caused main steam line radiation to increase to this level indicate there is a degradation of fuel clad integrity.
This event will escalate to an Alert based on the breach in the main steam line together with a failure of the MSIVs to isolate the main steam lines per EAL Section 4.2.2.
DEVIATION None REFERENCES NUMARC NESP-007, SU4.1 T-099, Post Scram Recovery T-100, Scram Technical Specifications NEDO-31400A
lgs EAL Technical Basis Manual REV a. April 26,1995 Page 62 of 130 4.0 Secondary Containment 4.2 Main Steam Line ALERT - 4.2.2 IC Potential Loss of Reactor Coolant System and Containment t EAL Failure of one or more Main Steam Lines to isolate on any MSIV Closure Signal OPCON h i r l s lai s101 BASIS The MSIV Closure Signal statement identifies that tne MSIVs should be closed. The reason for the isolation signalis not relative to the basis of this EAL. The fact that the MSIV's failed to isoiate the main steam line indicates the potential loss of the reactor coolant pressure boundary as well as a bypass potential of the primary containment.
Control room personriel should be attentive to indications as to if and where the steam is flowing.
Depending on the nature of the isolation signal. Radiation levels in the turbine enclosure and elsewhere may stay elevated. Actions should be taken to isolate the MSIVs as soon as possible.
10 CFR 100 release calculations assume the operability of the MSIVs and as such a steam line break may result in the release of radioactive materialin excess of the limits specified in 10 CFR 100.
This EAL is set below the threshold for deterrnining that the RCS and Containment Fission l
Product Barriers can no longer be considered intact; however, it represents a serious condition in that two of the three Fission Product Barriers are affected. A direct pathway from the reactor to the environment would result if there is a breach in the steam system.
Escalation of this event would be through identification of a significant leak downstream of the MSIVs indicative of actual or potentialloss of both RCS and Containment Fission Product Barriers per EAL Section 4.2.3.
DEVIATION l
None a
. i. .-_,.-_m._ . - . ._.. . _. _ .. _ - - _.. _
LGS EAL Technical Basis Manual .
~ REV a. Apnl 26,1995 Page 63 of 130
- REFERENCES ;
NUMARC NESP-007, RC EAL #5, PC EAL #2 10 CFR 100 i T-102, Primary Containment Control l T-103, Secondary Containment Control T-104, Radioactivity Release Control l
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V.
LOS EAL Technical Basis Manual REV c. Apnl 26,1995 Page 64 of 130 4.0 Secondary Containment 4.2 Main Steam Line SITE AREA EMERGENCY - 4.2.3 IC Loss of Reactor Coolant System and Containment EAL Main Steam Line Break discharging into the Turbine Enclosure AND North Stack > 1.08x10' pCUsec OPCON lil2laldl*191 B ASIS -
This event signifies that there is a direct path established for the transfer of main steam to inside i the Turbine Enclosure. Assumptions made in dose calculations regarding radioactive material transport (i.e. hold up, plate out, scrubbing, and retention) may be invalid. Additionally the transport time associated with a radiological release may be significantly shortened and there may be a higher percentage of short lived radioisotopes in any release. As both the reactor coolant pressure boundary and the primary containment are degraded; the extent of radioactive release is dependant on fuel clad integrity. Should a rapid reactor depressurization occur as a result of this event then there is a potential that a large amount of radiolodine may be released.
Steam Line Break discharging into the Turbine Enclosure is indicallve of lost Reactor Coolant System and lost Containment Fission Product Barriers. The term " discharging" indicates that the break cannot be isolated. The North Stack Wide Range Accident Monitor (RE26-076-4) high radiation alarm of > 1.08E5 pCl/sec provides confirmation of the Main Steam Line Break and indication that break is large enough to be significant.
This event will be escalated to a General Emergency based on the indication that the fuel clad integrity has also been threatened or lost per EAL Section 4.2.4.
DEVIATION None REFERENCES NUMARC NESP-007, RC EAL #5, PC EAL #2 T-103 Secondary Containment Control T-104, Radioactivity Release Control l
LGS EAL Techncal Basis Manual REv o, April 86,1995 Page 65 of 130 4.0 Secondary Containment 4.2 Main Steam Line GENERAL EMERGENCY - 4.2.4 IC Loss of Fuel Clad, Reactor Coolant System and Containment EAL Main Steam Une Break discharging into the Turbine Enclosure AND ,
North Stack > 1.08x10' pCIIsec OPCON lil21*Idl*1pl BASIS Control room indication and annunciators indicating that a main steam line break has occurred and that the steam line is discharging into the Turbine Enclosure. Additional indication is present that the main steam line has f ailed to isolate (i.e. steam line flow indication). There is indication that fuel cladding degradation has occurred.
This event signifies that there is a direct path established for the transfer of main steam to inside the Turbine Enclosure. Assumptions made in dose calculations regarding radioactive material transport (i.e. hold up, plate out, scrubbing, and retention) rnay be invalid. Additionally the transport time associated with a radiological release may be significantly shortened and there may ,
be a higher percentage of short lived radioisotopes in any release.
Both the reactor coolant pressure boundary and the primary containment are lost in the EAL and retention of radioactive material is dependent on fuel clad integrity. As there is indication of fuel clad integrity loss there is established a direct path for fission products to escape directly to the environs (bypassing primary and secondary containment). Thus all three of the fission product retention barriers have been compromised.
Due to the reactor scram and unisolable steam line leak there is a potential for a rapid release of fission product gases previously contained in the fuel pin gap. This may significantly increase the amount of gases present in the reactor coolant. The release of gases results from the rapid decrease in reactor power and the decrease in coolant void concentration.
If radiological release values exceed the limits established in TRIPS, then procedural guidance is provided to Emergency Depressurize the reactor. This will limit the amount of inventory and energy that would be available to escape through the breach, thus limiting the radiological release.
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LOS EAL Technical Basis Manual REV c. Apnl 26,1995 )
Page 66 of 130 j A Main Steam Line Break discharging into the Turbine Enclosure is indicative of loss of Reactor Coolant System and loss of Containment Fission Product Barriers. The term " discharging" indicates that the break cannot be isolated. ;
l The Main Steam Tunnel temperature higher than the isolation setpoint and 10 times the the North i I
Stack Wide Range Accident Monitor (WRAM) alarm of > 1.08E5pC//sec provides confirmation of the Main Steam Line Break and indication that break is large enough to be significant.
DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4, RC EAL #5 and PC EAL #2 T-103, Secondary Containment Control T-104, Radioactivity Release Control T-112, Emergency Blowdown
LGS EAL T*chnicti BIsis Minual REv a. April 26,1995 Page 67 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose UNUSUAL EVENT - 5.1.1.a IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer EAL North Stack or South Stack Rad monitor continuously in HiHi Alarm OR known Unmonitored Release continuously in progress for > 60 minutes AND Calculated maximum offsite dose rate exceeds 0.174 mrem /hr TPARD OR 0.342 mrem /hr child thyroid CDE based on a 60 minute average OPCON lilrlal41slol BASIS Releases in excess of 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE that continue for > 60 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The final integrated dose is very low and is not the primary concern. Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.
It is not intended that the releaso be averaged over 60 minutes, but exceed 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE limits for 60 minutes. For a monitored release, this EAL includes a 60 minute average for the dose projection with the release point radiation monitor above the HiHi alarm set point for the entire 60 minutes. For an unmonitored release [ie. blowout panel, inoperable radiation monitors), this EAL entry requires a known continuous unmonitored release lasting for greater than 60 minutes. Also, it is intended that the event be declared as soon as it is determined that the release will exceed 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE for greater than 60 minutes.
An indication or report is considered to be valid when it is verified by:
- 1. An instrument channel check
- 2. Indications on related or redundant instruments
- 3. By direct observation by plant personnel The response to the radiological monitor HiHi alarm condition in the control room is dose assessment entering their procedures to determine if dose projections are necessary [ NOTE: the HiHi alarm is conservatively set to be significantly below an Unusual Event value). If determined, dose projections utilizing the monitoring readings and actual meteorology will be performed. This projection will determine which classification, if any, is warranted. i I
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l lgs EAL Technical Baas Manual REv o. Apnl 26.1995 ;
Page 68 of 130 J l
The Total Protective Action Recommendation Dose (TPARD) is calculated by dividing the yearly ,
allowable Technical Specification limit (500 mrem /yr) by the number of hours per year (8760 I l
hr/yr), and then multiplying by a factor of 2 times Technical Specifications [ODCM).
TPARD = 2x(Tech Spec Limit)/(hours per year) ;
i
= 2(500 mrem /yr)/(8760 hr/yr)
= 0.114 mrem /hr The Committed Dose Equivalent (CDE) is calculated by dividing the yearly allowable Technical Specification limit (1500 mrem /yr) by the number of hours per year (8760 hr/yr), and then multiplying by a factor of 2 times Technical Specifications (ODCM].
CDE = 2x(Tech Spec Limit)/(hours per year)
= 2(1500 mrem /yr)/(8760 hr/yr)
= 0.342 mrem /hr This event will be escalated to an Alert when effluents increase per EAL Sections 5.1.2.a.
DEVIATION None REFERENCES NUMARC NESP-007, AU1.1, AU1.2 OffSite Dose Calculation Manual ST-6-104-880-0, Gaseous Effluent Dose Rate Determination
-. . = - - . . - -
l lgs EAL Technicd Bam Manud REv a. Apol 26,1995 l Page 69 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose UNUSUAL EVENT - 5.1.1.b IC Any Unplanned Release of Liquid Radioactivity to the Environment that Exceeds Two Times Radiological Technical Specifications for 60 Minutes or Longer EAL ,
Report indicates Liquid Release exceeds "IWO TIMES Offsite Dose Calculation Manual (ODCM 3.2.2 or 3.2.3) for > 60 minutes OPCON hirlaldislol BASIS Releases in excess of two times technical specifications that continue for > 60 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The final integrated dose is very low and is not the primary concern. Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.
It is not intended that the release be averaged over 60 minutes, but exceed two times technical specifications limits for 60 minutes. Further, it is intended that the event be declared as soon as it is determined that the release will exceed two times technical specifications for greater than 60 minutes.
An indication or report is considered to be valid when it is verified by:
- 1. An instrument channel check
- 2. Indications on related or redundant instruments
- 3. By direct observation by plant personnel The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm setpoints, etc).
This event will be escalated to an Alert when effluents increase per EAL Sections 5.1.2.b.
DEVIATION None l
-wa----sm m% w - -+ m
I LGS EAL Technied Basis Manual I REV O Apnl 26,1995 l Page 70 of 130 REFERENCES NUMARC NESP-007 AU1.2 Offsite Dose Calculation Manual T-104, Radioactivity Release Control
l l lgs EAL Tschnied Basis Manuti REv a, April 26,1905 I Page 71 of 130 5.0 Radioactivity Release l
5.1 Effluent R3l ease and Dose ALERT - 5.1.2.a IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer i EAL 4
North Stack or South Stack Rad monitor continuously in HiHi Alarm OR known Unmonitored Release continuously in progress for > 15 minutes AND Calculated maxium offsite dose rate exceeds 11.4 mrem /hr TPARD OR 34.2 mrem /hr child thyrold CDE based on a 15 minute average OPCON 1512181415101 BASIS Releases in excess of 11.4 mrem /hr TPARD or 34.2 mrem /hr CDE that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The primary concern is the final integrated dose [100 times greater than the Unusual Event] and the degradation in plant control implied by the fact that the release was not isolated within 15 minutes.
This EAL includes a 15 minute averate for the dose projection. For a monitored release, this EAL includes a 15 minute average for the dose projection with the release point radiation monitor above the HiHi alarm set point for the entire 15 minutes. For an unmonitored release [ie, blowout panel, inoperab!e radiation monitors], this EAL entry requires a known continuous unmonitored release lasting for greater than 15 minutes. Also, it is intended that the event be declared as ,
soon as it is determined that the release will exceed 11.4 mrem /hr TPARD or 34.2 mRom/hr CDE for greater than 15 minutes.
An indication or report is considered to be valid when it is verified by:
- 1. An instrument channel check
- 2. Indications on related or redundant instruments
- 3. By direct observation by plant personnel The response to the radiological monitor HiHi alarm condition in the control room is dose assessment entering their procedures to determine if dose projections are necessary [ NOTE: the HiHi alarm is conservatively set to be significantly below an Unusual Event value]. If determined, dose projections utilizing the monitoring readings and actual meteorology will be performed. This projection will determine which classification, if any, is warranted.
l lgs EAL Technical Basis Manual REV a. April 26.1995 Paga 72 of 130 The Total Protective Action Recommendation Dose (TPARD) is calculated by dividing the yearly allowable Technical Specification limit (500 mrem /yr) by the number of hours per year (8760 hr/yr), and then multiplying by a factor of 200 times Technical Specifications (ODCM).
l TPARD = 200x(Tech Spec Limit)/(hours per year) ;
= 200(500 mrem /yr)/(8760 hr/yr) l
= 11.4 mrem /hr The Committed Dose Equivalent (CDE) is calculated by dividing the yearly allowable Technical Specification limit (1500 mrem /yr) by the number of hours per year (8760 hr/yr), and then multiplying by a factor of 200 times Technical Specifications [ODCM).
CDE = 200x(Tech Spec Limit)/(hours per year)
= 200(1500 mrem /yr)/(8760 hr/yr)
= 34.2 mrem /hr This event will be escalated to a Site Area Emergency when actual or projected doses are determined to exceed 10CFR20 limits per EAL Sections 5.1.3.
DEVIATION None REFERENCES NUMARC NESP 007 AA1.1 AA1.2 Offsite Dose Calculation Manual
- - - , - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - w -cn - --e-n-~-- - - - -- ~
I lgs EAL Technied BIsis Mr.nual REv a. April 26,1995 Page 73 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose ALERT - 5.1.2.b IC Any Unplanned Release of Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EAL Report indicates Liquid Release exceeds TWO HUNDRED TIMES Offsite Dose Calculation Manual (ODCM 3.2.2 or 3.2.3) for > 15 mlnutes OPCON lil21*lal5lol BASIS Releases in excess of two hundred times technical specifications that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The primary concern is the final integrated dose [100 times greater than the Unusual Event) and the degradation in plant control implied by the fact that the release was not isolated within 15 minutes.
It is not intended that the release be averaged over 15 minutes, but exceed two hundred times technical specifications limits for 15 minutes. Further, it is intended that the event be declared as soon as it is determined that the release will exceed two hundred times technical specifications for greater than 15 minutes.
An indication or report is considered to be valid when it is verified by:
- 1. An instrument channel check
- 2. Indications on related or redundant instruments
- 3. By direct observation by plant personnel The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm setpoints, etc).
This event will be escalated to higher classifications based on plant conditions.
DEVIATION None
LGS EAL Technical Basis Manual REV et Apnl 26,1995 Page 74 of 130 REFERENCES NUMARC NESP-007 AA1.2 Offsite Dose Calculation Manual T 104, Radioactivity Release Control
l lgs EAL Technical Basis Manual REv e, April 26,1995 Pags 75 of 130 i
5.0 Radioactivity Release 5.1 Effluent Release and Dose SITE AREA EMERGENCY - 5.1.3 IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or Projected Duration of the Release EAL Projected offsite dose exceeds 100 mrem TPARD, OR Projected offsite dose exceeds 500 mrem child thyroid CDE, OR Actual offsite whole body dose rate exceeds 100 mrem /hr OPCON 11121814l5101 BASIS An actual or projected dose of 100 mrem Total Protective Action Recommendation Dose (TPARD) is based on the 10 CFR 20 annual average population exposure limit. TPARD is the sum of External Dose Equivalent + Committed Effective Dose Equivalent + 4-day Deposition Exposure.
This value also provides a desirable gradient (one order of magnitude) between the Site Area Emergency and General Emergency classifications. The 500 mrem integrated child thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TPARD and Child Thyroid Committed Dose Equivalent (CDE). Actual meteorology is used since it gives the most accurate dose projection.
Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with Pennsylvania Emergency Management Agency (PEMA) / Bureau of Radiation Protection (BRP).
This eventwill be escalated to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines per EAL Section 5.1.4.
DEVIATION None REFERENCES NUMARC NESP-007, AS1.1, AS1.3 and AS1.4 EPA 400
.._.,___m ,.w.._. . ,__. _ , ,_
I LOS EAL Technical Basis Manual REV a, April 26,1995 Page 76 of 130 5.0 Radioactivity Release 5.1 Effluent Release and Dose GENERAL EMERGENCY - 5.1.4 IC Boundary Dose Resulting from an Actual or imminent Release of Gaseous Radioactivity that Exceeds 1000 mR Whole Body or 5000 mR Child Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology EAL Projected offsite dose exceeds 1000 mrem TPARD, OR Projected offsite dose exceeds 5000 mrem child thyroid CDE, OR Actual offsite whole body dose rate exceeds 1000 mrem /hr OPCON 1112181415101 BASIS Actual or projected dose exceeding 1000 mrem Total Protective Action Recommendation Dose (TPARD) or 5000 mrom child thyroid CDE is greater than the EPA Protective Action Guidance.
TPARD is the sum of External Dose Equivalent + Committed Effective Dose Equivalent + 4-day Deposition Exposure. At these doses, public protective actions are required, consistent with the emergency class description of a General Emergency.
Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with Pennsylvania Emergency Management Agency (PEMA) / Bureau of Radiation Protection (BRP).
Actual meteorology is used since it gives the most accurate dose projection.
DEVIATION None REFERENCES NUMARC NESP-007, AGl.1, AG1.3 and AG1.4 EPA-400
LGS EAL Technied Basis Manual REV a Apnl 26,1995 Page 77 of 130 5.0 Radioactivity Release 5.2 in-Plant Radiation UNUSUAL EVENT - 5.2.1 BC Unexpected increase in Plant Radiation or Airborne Concentration EAL I
Inplant radiation level > 1x1f mR/hr requiring T 103 entry OPCON lil21alal5lol BASIS The value of 1x10' mR/hr indicates a radiation level of approximately 1000 times normal.
Unplanned increases in in-plant radiation levels represent a degradation in the control of radioactive material and represent a degradation in the level of safety of the plant. Planned evolutions which cause elevated radiation levels are not covered by this EAL.. T-103 identifies the systems that interface with the reactor coolant system. ,
An area monitor reading is considered to be valid when it is verified by:
- 1. an instrument channel check Indicating the monitor has not failed;
- 2. Indications on related or redundant instrumentation; or
- 3. direct observation by plant personnel This event will be escalated to an Alert when radiation levels increase in areas required for the safe shutdown of the plant resulting in impeded access per EAL Section 5.2.2.a.
DEVIATION None REFERENCES NUMARC NESP 007, AU2.4 T-103, Secondary Containment Control J l
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I lgs EAL Technied Basis Manus! j REv a. April 26,1995 l Page 78 of 130 l
5.0 Radioactivity Release 5.2 In-Plant Radiation ALERT - 5.2.2.a IC Release of Radioactive Material or increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Inplant radiation level > 5xM' mR/hr for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requiring T-103 entry i
OPCON lil2181dl5lol BASIS The EAL addresess radiation levels which would impede operation of systems required to maintain safe operations or to establish or maintain cold shutdown. Radiation levels could be indicated by ARM or radiological survey. T-103 lists the areas in the plant that locate systems interfacing with the reactor coolant system.
An area monitor reading is considered to be valid when it is verified by:
- 1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or,
- 3. Direct observation by plant personnel.
Unplanned increases inplant radiation levels represent a degradation in the control of radioactive materials and represent a degradation in the level of safety of the plant. Planned evolutions which cause elevated radiation levels are not covered by this EAL.
Access to the areas listed in T-103 may be necessary to perform manual actions to achieve or maintain cold shutdown.
This event will be escalated to a Site Area Emergency when loss of control of radioactive materials cause significant offsite doses per EAL Section 5.1.3.
DEVIATION None l
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LGS EAL Technical Basis Manual REV a. April 26,1995 Page 79 (f 130 I
REFERENCES NUMARC NESP-007, AA3.2 ,
T-103, Secondary Containment Control l m
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f lgs EAL Technical Basic Manual REVa April 26,1996 Page 80 of 130 5.0 Radioactivity Release 5.2 In-Plant Radiation ALERT - 5.2.2.b IC Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations EAL Control Room area radiation level > 15 mR/hr for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OPCON lil21*ldl5lol BASIS The EAL address radiation levels which would impede operation of systems required to maintain safe operations or to establish or maintain cold shutdown. Radiation levels could be indicated by ARM or radiological survey.
The Control Room general area level is set at 15 mR/hr and was chosen because continuous occupancy is required. This is consistent with General Design Criteria 19, which addresses continuous occupancy of the Control Room for 30 days after a design accident. Additionally, since the Control Room is shielded this radiation level represents a serious loss of control of radioactive material.
This event will be escalated to a Site Area Emergency when loss of control of radioactive materials cause significant offsite doses per EAL Section 5.1.3.
DEVIATION None REFERENCES NUMARC NESP-007 AA3.1
fi LGS EAL Tachnied Basis Manual REV a, April 26,1995 Page 81 of 130 I
6.0 Loss of Power !
6.1 Loss of AC or DC Power UNUSUAL EVENT - 6.1.1.a IC Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes ,
EAL 4
Loss of ALL Offsite Power for >15 minutes l
OPCON 151218141532J BASIS This EAL addresses the loss of off-site AC power supplying the station. Off-site power is fed ~
through transformers 10 and 20 to busses 101 and 201. Loss of off-site power will cause a reactor scram and a containment isolation. All eight (8) emergency Diesel Generators will be '
carrying the essentialloads for each unit (4 per unit). Balance of Plant systems that would assist in plant operations (i.e. condensate pumps, etc.) may be unavailable due to the loss of power.
Fifteen (15) minutes has been selected to exclude transient or momentary power losses. ;
However, an Unusual Event should be declared in less ban 15 minutes if it can be determined <
in less than 15 minutes that the power loss is not transient or momentary. Although no fission product release barriers are directly affected by the loss of offsite power, the plant is more vulnerable to a complete loss of AC Power with the dependancy on the emergency Diesel '
Generators to power emergency systems to remove heat and maintain the reactor core submerged and cooled. l Escal?. tion of this event to an Alert would be based on having a loss of all offsite AC power coincident with onsite AC power being reduced to a single power source in Modes 1,2, and 3 or having a loss of all offsite and onsite AC power in Modes 4 or 5 per EAL Section 6.1.2. l t
DEVIATION
, None REFERENCES NUMARC NESP-007, SU1 +
E-10/20, Loss of Offsite Power
l lgs EAL Technical Basis M:nual REv a. Apnl 26.1995 Page 82 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power UNUSUAL EVENT - 6.1.1.b IC Unplanned Loss of Required DC Power During Cold Shutdodwn or Refueling Mode for '
Greater than 15 Minutes EAL Loss of ALL safeguard DC Power indicated by < 105 VDC for > 15 minutes OPCON Mini *1disio1-BASIS:
The purpose of this EAL 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. The 125 volt DC Main Distribution Panel Busses are as follows:
1FA, Division I Safeguard 125/250 DC Bus 1FA 1FB, Division ll Safeguard 125/250 DC Bus 1FB 1FC, Division lli Safeguard 125 DC Bus 1FC 1FD, Division IV Safeguard 125 DC Bus 1FD 105 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is near the minimum voltage selected when battery sizing is performed.
This event will escalate to an Alert if the loss results in the inability to maintain cold shutdown, per EAL Section 7.2.2.
DEVIATION None 1
i: I LGS EAL Technical Basis Manual ,,
REV a, April 26.1995 Page 83 of 130 REFERENCES-NUMARC NESP-007, SU7 E 1FA, Loss of Division i Safeguard 125/250 DC Bus 1FA ,
~ E-1FB, Loss of Division 11 Safeguard 125/250 DC Bus 1FB i E 1FC, Loss of Division ill Safeguard 125 DC Bus 1FC E-1FD, Loss of Division IV Safeguard 125 DC Bus 1FD
I lgs EAL Technical Basis Manuil REV a. April 26.1995 Page 84 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power ALERT - 6.1.2.a IC AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout EAL Loss of ALL Offsite Power for > 15 minutes AND Only ONE 4 KV Safeguard Bus is available lil213I41sloj OPCON BASIS This EA. .s intended to provide an escalation from " Loss of offsite Power for greater than 15 minutes." This condition is a degradation of the offsite and onsite power systems such that any additional failure would result in a station blackout.
Depending on the 4 KV AC bus that remains energized there is a disparity in the systems that may be available. The ability to remove heat from the containment via suppression pool cooling may be lost due to the need to operate the 1 available RHR pump in other than suppression pool cooling (i.e. LPCI). As such there is a decrease in the systems available to remove heat transferred to the containment and there is an ongoing release of energy from the reactor to the containment (via SRVs, HPCI and/or RCIC operation). Fifteen (15) minutes has been selected to exclude transient or momentary power losses. However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power !oss is not transient or momentary. The ability to cool the nuclear fuel, remove decay heat, and control containment parameters is severely limited. Should equipment be unavailable prior to the loss of power functions necessary to maintain the plant in a cold shutdown condition may be threatened.
Escalation of this event would be based on the loss of the remaining Emergency Diesel Generator per EAL Section 6.1.3.
DEVIATION None REFERENCES NUMARC NESP-007, SAS E-1, Loss of All AC Power (Station Blackout)
t lgs EAL T:chnical Basis Manual REv a. Apn! 26,1995 Pagi 85 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power ALERT - 6.1.2.b IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode EAL Loss of ALL Offsite Power AND ALL 4 KV Safeguard Busses are unavailable for > 15 minutes OPCON 181818415101 BASIS Control room indications and annunciators indicating that all off-site and on-site AC power feeds have been lost. Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal, RHR Service Water, and Emergency Service Water. Although instrumentation (supplied through instrument invertors) and DC power loads would be available, their operability would be limited to the amount of stored ene,gy contained in their respective batteries. Instrumentation, communication equipment, and in-plant lighting and ventilation will be significantly hampered by the loss of all AC power.
Fifteen (15) minutes has been selected to exclude trans,ent or momentary power losses.
However, an Alert should be declared in less tnan 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary. i When in cold shutdown, refueling, or defueled mode this 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. It is assumed that the plant will be maintained in a cold shutdown condition and that if the plant is not able to be maintained in this mode then escalation to Site Area Emergency would be appropriate and be based on EAL Sections 2.1 or 5.1.
DEVIATION None I
I LGS EAL Technical Basis Manual REV 0, Apnl 26,1993 Pag 2 86 of 130 REFERENCES NUMARC NESP-007, SA1 E 1, Loss of All AC Power (Station Blackout)
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l lgs EAL TechnicrA Basis Manual REV e, April 26,1995 Page 87 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power SITE AREA EMERGENCY - 6.1.3.a IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EAL Loss of ALL Offsite Power AND ALL 4 KV Safeguard Busses are unavailable for > 15 minutes OPCON lil2laldlsiol BASIS Control room indications and annunciators indicating that all offsite and onsite AC power feeds have been lost. Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal, RHR .
Service Water, and Emergency Service Water. Although instrumentation (supplied through instrument invertors) and DC power loads would be available, their operability would be limited to the amount of ctored energy contained in their respective batteries. Instrumentation, communication equipment, and in plant lighting and ventilation will be significantly hampered by the loss of all AC power.
Fifteen (15) minutes has been selected to exclude transient or momentary power losses.
However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary.
Escalation of this event would be based on tne time that the Emergency Diesel Generator are unavailable per EAL Section 6.1.4.
DEVIATION None REFERENCES NUMARC NESP-007, SS1 f E-1, Loss of All AC Power (Station Binckout) l i
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l lgs EAL Technical Basis Manual ,
REV c. Apnl 26,199 Page 88 of 130 i
6.0 Loss of Power 6.1 Loss of AC or DC Power SITE AREA EMERGENCY - 6.1.3.b IC Loss of All Vital DC Power EAL Loss of ALL safeguard DC Power indicated by < 105 VDC for > 15 minutes OPCON Iti 18l*LsLJ BASIS:
A loss of all DC power compromises the ability to monitor and control plant functions.125 volt DC system provides control power to engineered safety features valve actuation, diesel generator auxiliaries, plant alarm and indication circuits as well as the control power for the associated load group. If 125 volt DC power is lost for an extended period of time (greater than 15 minutes) critical plant functions such as RPS Logic, Altemate Rod Insertion, Emergency Service Water Indication,4KV Breaker Controls, HPCI, RCIC and RHR pump controls required to maintain safe plant conditions may not operate and core uncovery with subsequent reactor coolant system and primary containment failure might occur. These busses are:
1FA, Division i Safeguard 125/250 DC Bus 1FA 1FB, Division il Safeguard 125/250 DC Bus 1FB 1FC, Division ill Safeguard 125 DC Bus 1FC 1FD, Division IV Safeguard 125 DC Bus 1FD Loss of all DC Power causes the loss of the following equipment:
- Alternate Rod insertion - ADS HPCI - RCIC Normal EDG Control - Normal Recirculation Pump Trip
- Containment Instrument Gas Compressors
- Other 4KV Circuit Breakers (RHR, CS, CRD)
Loss of ADS creates a loss of low pressure ECCS due to the inability to depressurize the reactor.
In addition, loss of these buses will eventually lead to MSIV closure and reactor trip due to the loss of the Containment Instrument Gas Compressor as a result of suction valve closure.
Subsequent to MSIV closure, much of the equipment noted above will be required for plant stabilization and shutdown.
A sustained loss of DC power will threaten the ability to remove heat from the reactor core.
resulting in eventual fuel clad damage. The loss of DC power will also result in the loss of the ability to remove heat from the containment. SRVs will remain operable in the relief mode and the
i lgs EAL Technical Basis Manuil REV a. Apri 26,1995 Page 89 of 130 heat addition to the containment could result in a loss of the primary containment as a fission product release barrier.
105 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety ,
related equipment. This voltage value incorporates a margin of at least 15 minutes of operation
, before the onset of inability to operate those loads. This voltage is near the minimum voltage selected when battery sizing is performed.
DEVIATION I
None REFERENCES NUMARC NESP-007, SS3 T-101, RPV Control T 102, Primary Containment Control E 1FA, Loss of Division l Safeguard 125/250 DC Bus 1FA E-1FB, Loss of Division ll Safeguard 125/250 DC Bus 1FB E 1FC, Loss of Division til Safeguard 125 DC Bus 1FC E 1FD, Loss of Division IV Safeguard 125 DC Bus 1FD
l LGS EAL Techncal Balis Manual REV a, Apnl 26,1995 Page 90 of 130 6.0 Loss of Power 6.1 Loss of AC or DC Power GENERAL EMERGENCY - 6.1.4 IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power EAL Loss of ALL Offsite Power AND ALL 4 KV Safeguard Busses are unavailable for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> OPCON 1112181d18101 BASIS 10 CFR 50.2 defines station blackout (SBO) as complete loss of AC power to essential and non-essent!al buses. SBO does not include loss of AC Power to busses fed by station batteries through inverters, nor does it assume a concurrent single failure or design basis accident.
Successful SBO coping maintains the following key parameters within given acceptable limits:
- 1. Reactor water level > -161" (TAF)
- 4. Containment pressure < 62.5 psig, design limit
- 5. Suppression pool temperature < 200 F, HPCl/RCIC lube oil temperature concem when suction aligned to suppression pool
- 6. Drywell temperature
<200 F indefinitely
<250 F 99 days
<320 F 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />
<340 F 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Successful extended SBO coping depends on ability to keep HPCl/RCIC available for injection, and ability to maintain RPV depressurized for low pressure injection should HPCI and RCIC become unavailable. Control power for HPCI, RCIC and SRVs is provided by 125V DC. The parameters listed above can be maintained as long as the batteries are intact. Two hours is the earliest the batteries would fail, and thus is the basis for the time limit in this EAL.
The significance of a station blackout r6%tive to the loss of fission product release barriers is that all three barriers will eventually be lost due to the inability to remove heat from the fuel and the containment. Although the RCS will be intact the longest, eventually SRVs will operate in the relief
n-f LGS EAL Technical Basis Manual REV e, Apnl 26,1995 Page 91 of 130 mode due to RPV overpressurization and if the containment has already failed then 'hera is a direct bypass of the RCS boundary.
DEVIATION None REFERENCES NUMARC NESP-007, SG1 E-1, Loss of All AC Power (Station Blackout)
T-101, RPV Control T-102, Primary Containment Control T 104, Radioactivity Release Control 10CFR50.2 l
l r n v v -u---- - _ _ _ . - - _ - _ _ _ _ _ - - -
l LOS EAL Technical Basis Manual REV 0, Aprd 26,1995 Page 92 of 130 This page intentionally left blank l
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LCS EAL Tcchnmal Basis Manu;J REV a. April 26,1995 Page 93 of 130 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation UNUSUAL EVENT - 7.1.1 lC Inability to Reach Required Shutdown Within Technical Specification Limits EAL Unable to bring the P! ant to the required OPCON within Tech Spec LCO action times OPCON lil218I*lslol
~
BASIS .
Limiting Conditions of Operation (LCOs) require the plant to be brought 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 under 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 Unusual Event is required when it is determined that the plant cannot be brought to the required operatirg mode within the allowable action statement time in the Technical Specifications. Declarabon of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, SU2 Technical Specifications
i LOS EAL Technical Basis Manual REU G April 26,1995 Page 94 of 130 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation ALERT - 7.1.2 IC Control Room Evacuation Has Been Initiated EAL Control Room evacuation procedures have been initiated OPCON I 12131415101 BASIS Control Room evacuation requires establishment of plant control from outside the control room (local control and remote shutdown panel) and support from the Technical Support Center and/or other emergency facilities as necessary. Control Room evacuation represents a serious plant situation since the level of control is not as complete as it would be without evacuation. The establishment of system control outside of the Control Room will bypass many protective trips and interlocks. In addition, much of the instrumentation and assessment tools available in the Control Room will not be available.
This event will be escalated to an Site Area Emergency if control cannot be established within fifteen minutes per EAL Section 7.1.3.
DEVIATION None REFERENCES NUMARC NESP-007, HA5 SE-1, Remote Shutdown SE-6, Attemate Remote Shutdown so-
LGS EAL Tochnmal Basis Manual REV o. Apnl 26,1995 Page 95 of 130 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation SITE AREA EMERGENCY - 7.1.3 IC Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EAL Failure to establish Remote Shutdown Control of the Plant within 15 minutes after evacuation of the Control Room OPCON ITTTalaisiol BASIS Transfer of safety system control has not been performed in an expeditious manner but it is unknown if any damage has occurred to the fission product barriers. The 15 minute time limit for transfer of control is based on a reasonable time period for personnel to leave the control room, arrive at the remote shutdown area, and reestablish plant control to preclude core uncovery and/or core damage. During this transitional period the function of monitoring and/or controlling parameters necessary for plant safety may not be occurring and as a result there may be a threat to plant safety.
This event will be escalated based upon system malfunctions or damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NEGB007, HS2 SE 1, Remote Shutdown SE-6, Alternate Remote Shutdown
l lgs EAL Technical Basis Manual REv o. Apnl 26,1995 Page 96 of 130 7.0 Internal Events 7.2 Loss of Heat Removal Capability UNUSUAL EVENT - 7.2.1 IC Inability to Maintain Plant in Cold Shutdown EAL Uncontrolled Re' actor Coolant temperature increase to > 200 *F OPCON hl 181415lpl BASIS This EAL addresses complete loss of normal functions required for core cooling during refueling and cold shutdown modes. Uncontrolled means that system temperature increase is not the result of planned actions by the plant staff.
This EAL is concemed with the ability to keep the reactor core cooled less than 200 F. The criteria of uncontrolled Reactor Coolant temperature increase > 200 F is met as soon as it becomes known that sufficient cooling cannot be restored in time to maintain the temperature <
200 *F regardless of the current temperature.
This event will be escalated to an Alert if alternate decay heat removal capability cannot be established per EAL section 7.2.2.
DEVIATION This EAL has been created to be a precursor to NUMARC NESP-007, SA3 which calls for an Alert to be declared for a loss of shutdown cooling function and reactor coolant temperature rises above 200 F. The reactor core can continue to be properly cooled as long as suppression pool cooling is available. While this event clearly warrants the declaration of an Unusual Event, LGS does not feelit meets the threshold for an Alert.
REFERENCES NUMARC NESP-007, SA3 ao
LGS EAL Technical Basis Manual '
REV a, Apnl 26.1995 Page 97 of 130 7.0 Internal Events 7.2 Loss of Heat Removal Capability ALERT - 7.2.2
.IC Inability to Maintain Plant in Cold Shutdown EAL Uncontrolled Reactor Coolant temperature increase to > 200 *F AND inability to establish alternate decay heat removal capability ,
OPCON HIsla141 slo _1 BASIS This EAL addresses complete loss of normal functions required for core cooling during refueling and cold shutdown modes. Uncontrolled means that system temperature increase is not the '
result of planned actions by the plant staff.
This EAL is concerned with the ability to keep the reactor core temperature less than 200 F. The criteria of uncontrolled Reactor Coolant temperature increase > 200 F is met as soon as it becomes known that sufficient cooling cannot be restored in time to maintain the temperature <
200 'F, regardless of the current temperature. The inability to establish Alternate methods of decay heat removal indicates that either alternate methods are unavailable to cool the core in the RPV or whan the steam is transferred to the Suppression Pool, Suppression Pool cooling is unavailable. Loss of Suppression Pool cooling will result in a continuing, uncontrolled increase in reactor coolant temperature.
This event will be escalated to a Site Area Emergency if boiling of the reactor coolant reduces level and produces high radiation and/or fuel damage as identified in EAL sections 1.2,1.3,2.1, and 5.1.
DEVIATION None REFERENCES NUMARC NESP-007, SA3
LGS EAL Technicd Basis Manual REV a. Apnl 26,1995 Page 98 of 130 7.0 Internal Events 7.2 Loss of Heat Removal Capability SITE AREA EMERGENCY - 7.2.3 eC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EAL Loss of Main Condenser as a heat sink AND Loss of Suppression Pool Heat Sink capabilities as evidenced by T-102 legs [SP/T, SP/L, PC/P, or DW/T] requiring an Emergency Blowdown OPCON lilrlaisisiol BASIS This EAL addresses complete loss of functions required to reach cold shutdown from Operational Conditions 1,2 or 3.
The normal method for rejecting heat is via the Main Condenser. If the Main Condenser is not available, heat may be rejected directly to the Suppression Pool utilizing SRVs. The number of SRVs required to reduce pressure will be dependent upon reactor pressure and power. A low Suppression Pool level would result in the SRV discharge spargers being exposed. This condition results in a loss of SRV functions as additional energy cannot be rejected to the Suppression Poolwithout increasing containment pressure. A high Suppression Pool temperature (180 F) would result in the Suppression Pool being at the high temperature limit whereby it can no longer fun: tion as a heat sink. With these conditions, reactor pressure cannot be reduced to i the shutdown cooling pressure interlock of 75 psig and shutdown cooling cannot be established. l Once the interlock is cleared, shutdown cooling can be utilized to reduce temperature to below 200 F.
l DEVIATION None REFERENCES NUMARC NESP-007, SS4 T-102, Primary Containment Control, SP/L-8 l
1
l LGS EAL Technical Basis Manual REv a, April 26,1995 Page 99 of 130 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability UNUSUAL EVENT - 7.3.1.a IC Unplanned Loss of Most or All Safety System Annunchtion or Indication in The Control Room for Greater Than 15 Minutes EAL Loss of All Annunciators in the Control Room for > 15 minutes OPCON lil21sidI616l BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Manager this loss of annunciators requires increased surveillance to safely operate the plant. Although loss of ALL annunciators is specified,if a large portion of annunciators or significant annunciators, as determined by the shift manager, are lost this EAL would then be appropriately entered. It is not intended that a detailed count of instrumentation be performed, but that only a rough approximation be used to determine the severity of the loss. The Plant Monitoring System is available to provide compensatory indication.
Fifteen minutes is used as a threshold to exclude transient or momentary power losses.
Unplanned loss of annunciators excludes scheduled maintenance and testing activities. Control Room panels with annunciators and direction for response are included in ON 122, Loss of Control Room Annunciators.
Reportability of Technical Specification imposed shutdowns, or the inability to comply with Technical Specification action statements is covered in EAL section 7.1, Technical Specifications.
This EAL is not applicable in cold shutdown or refueling modes due to the limited number of safety systems required for operation.
This event will be escalated to an Alert if a transient is in progress or if compensatory indications become unavailable per EAL Section 7.3.2.
DEVIATION None
LGS EAL Technical Beeis Manuel
' REV e, April 26.1996
- Page 100 of 130 -
REFERENCES NUMARC NESP 007, SU3 ON 122, Loss of Control Room Annunciators AIT A0004447, EP Self Assessment on Salem Loss of Annunciators b
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LGS EAL Technical Basis Manual REV a. Apnl 26,1995 Page 101 of 130 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability UNUSUAL EVENT - 7.3.1.b IC Unplanned Loss of All Onsite or Offsite Communications Capabilities EAL Loss of ALL Onsite communications (Table 7-1)
OR Loss of ALL Oftsite communications (Table 7-1)
OPCON 18I2131416101 BASIS This EAL recognizes a loss of communication ability that significantly degrades the plant operations staff's ability to perform tasks necessary for plant operations or the ability to communicate with offsite authorities. This EAL is separated into two groups of communications, Onsite and Offsite. A complete loss of either group is so severe, that the Unusual Event declaration is warranted. Table 7-1 is identified as follows:
Table 7-1 Communications Onsite Offsite Site Phones (Dimension 2000) X X PRELUDE System X X Plant Public Address X Station Radio X NRC (FTS-2000) X PA State Police Radio X County Police Radio X Load Dispatcher Radio X PECO Dial Network X There is no escalation to an Alert for loss of communications, although there is escalation to higher classifications if other communications for plant assessment is lost.
DEVIATION None l
1
LOS EAL Technical Basi 2 Manual REV c. April 26,1995 Page 102 of 130 REFERENCES NUMARC NESP-007, SUG Nuclear Emergency Plan l
1 l
l
)
___-____________-___________-________\
I lgs EAL Technic".1 Basis Manual REV a. Apnl 26,1905 Page 103 of 130 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability ALERT - 7.3.2 lC Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable EAL ,
Loss of All Annunciators in the Control Room for > 15 minutes AND Significant Plant Transient (Table 7 2) is in progress QR Plant Monitoring System (Plant Process Computer and ERFDS for Unit 1) is unavailable OPCON Ii1213I4161o1 BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Manager this loss of annunciators requires increased surveillance to safely operate the plant. Although loss of ALL annunciators is specified, if a large portion of annunciators or significant annunciators, as determined by the shift manager, are lost this EAL would then be appropriately entered. This EAL represents an increase in severity above 7.3.1.a in that the Plant Monitoring System (Plant Process Computer and ERFDS for Unit 1) can not provide compensatory indication, or that a significant transient is in progress. Table 7-2 significant plant transient includes response to automatic or manually initiated actions including:
Table 7-2 Plant Transients SCRAM Recirc runbacks > 25% thermal power Thermal power oscillations of 10% or greater Stuck Open Relief Valve (SORV)
ECCS injection Fifteen minutes is used as a threshold to exclude transient or momentary power loses. Control Room panels with annunciators and direction for restoration is included in ON-122, Loss of Control Room Annunciators.
Reportability of Technical Specification imposed shutdowns, or the inability to comply with Technical Specification action statements is covered in section 7.1, Technical Specifications.
I
I J
lgs EAL Techneal Basis Manual )
REv C. April 26,1995 Page 104 of 130 This EAL is not applicable in cold shutdown or refueling modes due to the limited number of safety systems required for operation.
This event will be escalated to a Site Area Emergency if a transient is in progress and the Plant Monitoring System (Plant Process Computer and ERFDS for Unit 1) are unavailable per EAL Section 7.3.3.
DEVIATION None REFERENCES NUMARC NESP-007, SA4 ON 122, Loss of Main Control Room Annunciators f
I l
r l
f lgs EAL Technical Basis Manual REV a. April 26,1993 Page 105 of 130 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability )
SITE AREA EMERGENCY - 7.3.3 IC Inability to Monitor a Significant Transient in Progress EAL
~ Loss of All Annunciators in the Control Room for > 15 m/nutes AND Significant Plant Transient (Table 7 2) is in progress AND Plant Monitoring System (Plant Process Computer and ERFDS for Unit 1) is unavailable OR Loss of good data on RM-11 as evidenced by all release point indications being magenta OPCON til21sidislol BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Manager this loss of annunciators requires increased surveillance to safely operate the plant. Although loss of ALL annunciators is specified, if a large portion of annunciators or significant annunciators, as determined by the shift manager, are lost this EAL would then be appropriately entered. This EAL represents an increase in severity above 7.3.2 in that the Plant Monitoring System (Plant Process Computer and ERFDS for Unit 1) can not provide compensatory indication. and that a significant transient is in progress. Table 7-2 significant plant transients include response to automatic or manually initiated actions including:
Table 7-2 Plant Transients SCRAM Recirc runbacks > 25% thermal power Thermal power oscillations of 10% or greater Stuck Open Relief Valve (SORV)
ECCS injection Planned maintenance or testing activities are included in this EAL due to the significance of this event. Control Room panels with annunciators and the restoration is included in ON 122, Loss of Control Room Annunciators.
The Control Room readouts from the RM-11 radiological monitoring Computer are included to ensure that potential releases or degraded core conditions can be monitored.
, .__. . -- -.m. . ..~.- . . .. ._ ~ . - - . - .. .- ._-- -. - .. . . ._ _ -. .. ~._
l
' LOS EAL Technical Basis Manual REV a Apnl 26 1995 Page 106 of 130 DEVIATION None REFERENCES NUMARC NESP-007, SS6 ON 122, Loss of Main Control Room Annunciators i
9 I
e l
m
lgs EAL Technical Basis Manual l
REv a. Apnl 26,1995 Page 107 of 130 8.0 External Events 8.1 Security Events UNUSUAL EVENT - 8.1.1 IC Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EAL 1
Confirmed security tnreat directed towards the station as evidenced by :
. Credible sabotage or bomb threat within the Protected Area, M
. Credible intrusion and attack threat to the Protected Area, E l
l
. Attempted intrusion and attack to the Protected Area, OR l - Attempted sabotage discovered within the Protected / Vital Area, OR
- Hostage / Extortion situation that threatens normal plant operations OPCON Dl2lal4lalol BASIS A security threat that is identified as being directed towards the station and represents a potential degradation in the level of safety of the plant. A security threat is satisfied if physical evidence '
supporting the threat exists, if information independent from the actual threat exists, or if a specific group claims responsibility for the threat. The Shift Manager will declare an Unusual Event subsequent to consulting with the Manager, Nuclear Security to determine the credibility of the security event.
Security threats which meet the threshold for declaration of an Unusual Event are:
- 1. Credible sabotage or bomb threat within the Protected Area
- 2. Credible intrusion and attack threat to the Protected Area
- 3. Attempted intrusion and attack to the Protected Area
- 4. Attempted sabotage discovered within the Protected / Vital Area
- 5. Hostage / Extortion situation that threatens normal plant operations Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or 10 CFR 50.72 and will not cause an Unusual Event to be declared.
This event will be escalated to an Alert based upon a hostile intrusion or act within the Protected Area per EAL Section 8.1.2.
LOS EAL Technical Basis Manual REVG April 26,1995 Page 108 of 130 DEVIATION None !
REFERENCES NUMARC NESP-007, HU4.1 and HU4.2
. Safeguards Contingency Plan Physical Security Plan l
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lgs EAL Technical Basis Manual REV c. April 26,1995 Page 109 of 130 8.0 External Events 8.1 Security Events ALERT - 8.1.2 IC Security Event in a Plant Protected Area EAL
'I Confirmed hostile intrusion or act within the Protected Area as evidenced by
- Actual attack and intrusion into the Protected Area, OR
- Suspected bomb, sabotage or sabotage device discovered in the Protected / Vital Area OPCON lilaisl4isiol BASIS This class of security event represents an escalated threat to the level of safety of the plant. This event is satisfied if physical evidence supporting the hostile intrusion or attack exists. The Shift Manager will declare an Alert subsequent to consulting with the Manager Nuclear Security to determine the validity of the entry conditions.
Security threats which meet the threshold for declaration of an Alert are:
- 1. Actual attack and intrusion into the Protected Area
- 2. Suspected bomb, sabotage or sabotage device discovered within the Protected Area This event will be escalated to a Site Area Emergency based upon a hostile intrusion or act in plant Vital Areas per EAL Section 8.1.3.
DEVIATION None REFERENCES NUMARC NESP-007, HA4.1 and HA4.2 Safeguards Contingency Plan Physical Security Plan
] I lgs EAL Technical Basis Manuel REV a. Apnl 26,1995 ;
Page 110 of 130 8.0 External Events i i
8.1 Security Events
- i SITE AREA EMERGENCY - 8.1.3 ,
IC Security Event in a Plant Vital Area EAL Confirmed hostile intrusion or act in plant Vital Areas as evidenced by :
- Actual attack and intrusion into a Vital Area, OR
- Confirmed bomb, sabotage or sabotage device discovered in a Protected / Vital Area l l
s I
OPCON lil*laldislol l
BASIS ;
i This class of security event represents an escalated threat to plant safety above that contained 4
in an Alert in that a hostile intrusion or attack has progressed from the Protected Area to a Vital ;
' Area. The Vital Areas are within the Protected Area and are generally controlled by key card i readers. These areas contain vital equipment which includes any equipment, system, device or material, the fe.iture, destruction or release of could directly or indirectly endanger the public health ,
and safety by exposure to radiation. Equipment or systems which would be required to function i to protect health and safety following such failure, destruction or release are also considered vital.
4 Security threats which meet the threshold for declaration of a Site Area Emergency are: j i
- 1. Actual attack and intrusion into a Vital Area '
- 2. Confirmed bomb, sabotage or sabotage device discovered in a Vital Area This event will be escalated to a General Emergency based upon the loss of physical control of l the Control Room or Remote Shutdown Capability DEVIATION ;
None REFERENCES NUMARC NESP 007, HS1.1 and HS1.2 Safeguards Contingency Plan Physical Security Plan i
i
lgs EAL TIchnical Basis Manual REV a. Apnl 26,1995 Page 111 of 130 8.0 External Events 8.1 Security Events GENERAL EMERGENCY - 8.1.4 IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold Shutdown EAL Security event resulting in the actualloss of physical control of the:
Control Room OR_ Remote Shutdown Capability OPCON lil21*l4151ol BASIS This class of security event represents conditions under which a hostile force has taken physical control of areas required to reach and maintain cold shutdown. Loss of Remote Shutdown Capability would occur if the control function of the Remote Shutdown Panels wac lost.
Security events which meet the threshold for declaration of a General Emergency are physical loss of the Control Room or the Remote and Alternate Shutdown Panels.
This situation leaves the plant in a very unstable condition with a high potential of multiple barrier failures.
DEVIATION None REFERENCES NUMARC NESP-007, HG1.1 and HG1.2 Safeguards Contingency Plan Physical Security Plan
LGS EAL TechnicaJ Basis Manual REV c. Apnl 26,1995 Page 112 of 130 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.a IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection EAL Fire within SE-8 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of verification of alarms OPCON 18121814l5101 BASIS This EAL addresses verified fires in Plant Vital Structures that house safety systems. These fires may be precursors to damage to safety systems contained in these structures. Verification that a fire exists is by operator actions to confirm that fire alarms received in the Control Room are not spurious or by any verbal notif;ation by plant personnel. Fifteen minutes has been established to allow plant staff to respond and control small fires or to verify that no fire exists.
Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network This event will be escalated to an Alert if the fire damages redundant trains of plant safety systems required for the current operating condition per EAL Section 8.2.2.a.
DEVIATION None REFERENCES NUMARC NESP-007, HU2 SE-8, Fire
lgs EAL Technicd Basis Manual REv a, Apnl 26.1995 Page 113 of 130 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.b lC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the Plant EAL Normal plant operation is impeded by:
Toxic or flammable gas concentrations confirmed within the Protected Area OR Control Room informed by Local, County or State Officials to evacuate personnel within the protected area due to an offsite gas release -
OPCON 1112131415I01 BASIS This EAL addresses toxic / flammable gas releases within the Protected Area in concentrations high enough to affect health of plant personnel or the safe operation of the plant. This includes releases that originate both on-site and off-site, A toxic / flammable gas is considered to be any substance that is dangerous to life or limb by reason of inhalation or skin contact. A gas release is considered to be impeding normal plant operations if concentrations are high enough to restrict normal operator movements. It also includes areas where access is only possible with respiratory equipment, as this equipment restricts normal visibility and mobility. It should not be construed to include confined spaces that must be ventilated prior to entry or situation involving the Fire Brigade who are using respiratory equipment during the performance of their duties unless it also affects personnel not involved with the Fire Brigade.
A combustible gas maintained at a concentration lower than the Lower Explosive Limit will not explode due to ignition.
An off-site event (such as a tanker truck accident or train derailment releasing toxic gases) may place the Protected Area within the evacuation area. This evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials.
DEVIATION None
LOS EAL Technical Basis Manual REV a Apnl 26,1995 Page 114 of 130 REFERENCES NUMARC NESP-007, HU3.1 and HU3.2 SE 2, Toxic Gas / Chlorine
l l
' LGS EAL Technical Basis Manual '
REV a. Apnl 26,1995 Page 115 of 130 8.0 External Events l 8.2 Fire / Explosion and Toxic / Flammable Gases i UNUSUAL EVENT - 8.2.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EA!
Report by plant personnel confirming the occurrence of an explosion in a Plant Vital Structure (Table 8-1)
OPCON fil21sI41slol BASf6 This EAL addresses a verified explosion in Plant Vital Structures that house safety systems and exhibit sufficient force to cause damage to Plant Vital Structures or safety systems. As used here, Explosion is a rapid, violent, unconfined combustion, or a catastrophic f ailure of pressurized equipment (Process storage tanks, gas cylinders, heat exchangers, etc.) that potentially imparts sufficient energy to damage nearby structures or materials. No attempt is made to assess tne
., r,1agnitude of the damage. The occurrence of the explosion with a report of damage (deformation / scorching) is sufficient for declaration. Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure ;
Turbine Enclosure Diesel Generator Enclosure -
Spray Pond Pump House / Spray Network Any security aspects of this event should be considered under EAL Section 8.1, Security Events.
- This event will be escalated to an Alert if the explosion damages one or more redundant trains
of plant safety systems required for the current operating condition per EAL Section 8.2.2.a.
DEVIATION ,
None
! REFERENCES NUMARC NESP-007, HU1.5
lgs EAL Technical Basis Manual REV a, Apnl 26,1995 '
Page 116 of 130 j 8.0 External Events 8.2 Fire / Explosicq and Toxic / Flammable Gases ALERT 8.2.2.a IC Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EAL Fire or explosion within Plant Vital Structures (Table 8-1) damages safe shutdown systems causing or Jeopardizing plant shutdown OPCON 111218I416101 BASIS This EAL recognizes that damage has occurred to safe shutdown systems required to achieve or maintain cold shutdown in that equipment required for a particular safety function is either inoperable or incapable of performing its function. This EAL is entered if the fire causes an ,
automatic shutdown of the plant or the fire damages safe shutdown equipment jeopardizing a plant shutdown. The safe shutdown systems are housed in the Plant Vital Structures.
Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network Degraded system performance or observation of damage that could degrade system performance is used as the indicator that the safe shutdown system was actually affected. A report of 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 damage. The occurrence of the fire or explosion with reports of damage (e.g., deformation, scorching) is sufficient for declaration.
This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.
DEVIATION None l
LGS EAL Techncal Basis Manual REV a Apnl 26.1995 Page 117 of 130 REFERENCES NUMARC NESP-007, HA2
lgs EAL Technmal Basis Manual REV a. Apnl 26,1995 Page 118 of 130 8.0 External Events 1 8.2 Fire / Explosion and Toxic / Flammable Gases ALERT - 8.2.2.b IC Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown l 1
EAL Plant personnel are unable to perform actions necessary for safe shutdown without appropriate personal protection equipment due to toxic or flammable gas concentrations confirmed within Plant Vital Structures (Table 8-1)
OPCOh . <l2181416101 BASIS This EAL recognizes that toxic / flammable gases have entered Plant Vital Structures and are affecting safe operation of the plant by impeding operator access to tho safety systems that must be operated manually in these structures. The cause and/or magnitude of the gas concentrations is not a concern, but rather that access is required to an area and is impeded. Plant Vital Structures that must be accessed are as follows:
Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, HA3.1 and HA3.2 l
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LCs EAL Technied Basis Manual REv a. April 26,1995 Page 119 of 130 1
8.0 External Events l 8.3 Man-Made Events UNUSUAL EVENT - 8.3.1.a l l
IC Natural and Destructive Phenomena Affecting the Protected Area EAL !
Confirmed report of any vehicle crash affecting Plant Vital Structures (Table 8-1)
OPCON lilrl3l41siol BASIS This EAL address crashes of vehicles, such as aircraft or large construction vehicles, whicn have the potential to cause damage to plant Vital Structures. Personal vehicles are not included in this ,
EAL since they don't have the potential to damage plant vital structures. Two vehicles involved in an accident within the Protected Area does not require classification. Although personal vehicles are excepted in this EAL, these events should be evalauated under EAL Section 8.1, Security Events. Table 81 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Netwcrk ,
This event will be escalated to an Alert if the crash causes damage to Plant Vital Structures par ,
EAL Section 8.3.2.
DEVIATION None REFERENCES NUMARC NESP-007, HU1.4 ,
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LG3 EAL Techncal Basis Manual REV a. April 26,1995 Page 120 of 130 8.0 External Events 8.3 Man-Made Events UNUSUAL EVENT - 8.3.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL Turbine failure resulting in casing penetration OR Damage to Generator releasing hydrogen requiring immediate plant shutdown OPCON 111:I81418101 BASIS Turbine failure of sufficient magnitue to cause observable damage to the turbine casing or seals of the turbine generator increases the potential for leakage of combustible fluids and gases (Hydrogen cooiing) to the Turbine Enclocure. The damage should be readily observabla and should not require equipment disassembly to locate.
This event will be escalated to an Alert based upon damage done by missi!ec generated by the failure or by radiological releases per EAL Sections 8.3.2 or 5.0.
DEVIATION An exception to the NUMARC Methodology was taken in that NUMARC states all operating mode applicability for this EAL. This EAL addresses main turbine rotating component failures, which do not occur during shutdown conditions. Therefore, only OPCONS 1,2 and 3 will be used for this EAL.
REFERENCES NUMARC NESP-007, HU1.6 i
r lgs EAL Tcchnic"t Basis Manual REV a, Apnl 26,1995 Page 121 of 130 8.0 External Events t
8.3 Man-Made Events ALERT - 8.3.2 IC Destructive Phenomena Affecting the Plant Vital Area EAL :
Confirmed report of damage to Plant Vital Structures (Table 8-1) from either a vehicle crash or missile impact l
OPCON hirislaisiol BASIS This EAL address crashes of vehicles or missile impacts that have caused damage to Plant Vital Structures, and thus damage may be assumed to have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to '
classification. The evidence of damage is sufficient for declaration. A vehicle crash includes aircraft and large motor vehicles, such as a crane. Missile impacts including flying objects from off site, on-site rotating equipment or turbine failure causing casing penetration. Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, HA1.5 and HA1.6
LGS EAL Technical Basis Manual REV c. Apr6l 26,1995 Page 122 of 130 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.a IC Natural and Destructive Phenomena Affecting the Protected Area EAL Earthquake >.005 g OPCON lil21*ldl5lol BASIS This EAL addresses a sensed earthquake. The magnitude of .005g (OC693) is the lowest detectable earthquake measured at LGS. An earthquake of this magnitude may be sufficient to cause minor damage to plant structures or equipment within the Protected Area. Damage is considered to be minor as it does not affect physical or structural integrity. This event is not expected to affect the capabilities of plant safety functions.
This event will be escalated to an Alert if the earthquake reaches an Operating Basis Earthquake per EAL Section 8.4.2.
DEVIATION None REFERENCES NUMARC NESP-007, HU1.1 SE 5, Earthquake LGS UFSAR,3.7.4.2.1
lgs EAL Technical Basis Manual REV a, Apnl 26,1995 Page 123 of 130 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report of a Tornado within the Site Boundary OR Wind speeds > 75 mph as indicated on site Metrological data for > 15 minutes OPCON 1112181415101 BASIS A tornado touching down within the Protected Area or wind speeds > 75 mph within the owner !
controlled Area are of sufficient velocity to have the potential to cause damage to Plant Vital Structures. The value of 75 was selected to maintain consistency with plant value and to coincide with the Beaufort Scale for Hurricane wind speed winds of 73136 mph. These conditions are indicative of unstable weather conditions and represent a potential degradation in the level of safety of the plant. Verification of a tornado will be by direct observation and reporting by station personnel. Verification of wind speeds > 75 mph will be via meteorological data in the control room. For purposes of this EAL, sustained is > 15 minutes.
This event will be escalated to an Alert if the tornado or high wind speeds cause damage to Plant Vital Structures per EAL Section 8.4.2. If it is determined that the tornado or high wind speeds have caused a loss of shutdown cooling, then escalation will be by EAL Section 7.3, Loss of Decay Heat Removal Capability.
DEVIATION None ;
REFERENCES NUMARC NESP-007, HU1.2 and HU1.7 SE-9, High Winds
lgs EAL T:chnical Basis Manual REV a. Apnl 26,1995 Page 124 of 130 8.0 External Events 8.4 Natural Events ALERT - 8.4.2.a IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Earthquake >.075 g OPCON lil21*ldl5lol BASIS This EAL addresses an earthquake that exceeds the Operating Basis Earthquake level of .075g (OOC 693) and is beyond design basis limits. An earthquake of this magnitude may be sufficient to cause damage to safety related systems and functions.
This event will be escalated to a higher emergency classification based. upon damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, HA1.1 SE-5, Earthquake LGS UFSAR,3.7 P
lgs EAL T:chnical Basis Manual REV s, April 26,1995 Page 125 of 130 8.0 External Events 8.4 Natural Events ALERT - 8.4.2.b IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Tornado or wind speeds > 75 mph causing damage to Plant Vital Structures (Table 8-1)
OPCON h l 2181 d l 5 lol BASIS This EAL addresses events that have caused damage to Plant Vital Structures, and thus damage may be assumed to have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification The evidence of damage is sufficient for declaration. Table 8-1 Plant Vital Structures are as follows:
Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, HA1.2 and HA1.3
LOS EAL Technical Basis Manual REVc Apnf 26,1995 Page 126 of 130 This page intentionally left blank
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LGS EAL TCchnical Basb Manual REV o, Apnl 26,1695 Page 127 of 130 9.0 Other 9.1 General UNUSUAL EVENT - 9.1.1 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of an Unusual Event EAL Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant OPCON lil218ldl5lol BASIS This EAL allows the Shift Manager to declare an Unusual Event upon the determination that the level of safety of the plant has degraded. Where the degradation is associated with equipment or system malfunctions, the decision that it is degraded should be made upon functionality, not operability. A system, subsystem, train, component or device, though degraded in equipment condition or configuration, should be considered functional if it is capable of maintaining respective system parameters within acceptable design limits. Examples include:
Intemal flooding in excess of sump handling capability affecting safety related areas of the plant.
Releases of radioactive materials requiring off-site response or monitoring are not expected to occur at this level unless further degradation of safety systems occurs. However, if one does occur, it will be classified under section 5.0, Radioactivity Releases.
DEVIATION None REFERENCES NUMARC NESP-007, HUS and HU1.3
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l lgs EAL Technied Basis Mmuil REv a, Apnl 26.1995 Page 128 of 130 9.0 Other 9.1 General ALERT - 9.1.2 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of an Alert EAL Events are in progress or have occurred which indicate an actual or potential substantial degradation of the level of safety of the plant OPCON lil2I*Idl$lol BASIS This EAL allows the Shift Manager to declare an Alert upon the determination that the level of safety of the plant has substantially degraded but is not explicitly addressed by other EALs.
This includes a determination by the Shift Manager that the TSC and OSC should be activated and command and control functions should be transferred for the event to be effectively mitigated.
Transfer of command and control functions is used as an initiator since an event significant to warrant transfer is a substantial reduction in the level of safety of the plant. Other examples are:
Internal flooding affects the operability of plant safety systems required to establish or maintain cold shutdown.
Releases that are expected will be limited to a small fraction of the EPA Protective Action Guidelines and will be classified under section 5.0, Radioactivity Releases.
DEVIATION None REFERENCES NUMARC NESP-007, HA6, FC EAL #5, RC EAL #6, PC EAL #6 w, w
Les EAL Txchnical Basis Manual REV a, April 26,1995 i Page 129 of 130 l l
9.0 Other. I l
9.1 General SITE AREA EMERGENCY - 9.1.3 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of Site Area Emergency l EAL I
Events are in progress or have occurred which indicate an actual or likely major failure of plant functions needed for protection of the public i
OPCON lil21sidislol l
BASIS This EAL allows the Shift Manager to declare a Site Area Emergency upon the determination of an actual or likely major failure of plant functions needed for protection of the public, but is not explicitly addressed by other EALs.
Releases are not expected to result in exposure levels which exceed the EPA Protective Action Guidelines except within the site boundary and will be classified under section 5.0, Radioactivity Releases.
DEVIATION None REFERENCES NUMARC NESP-007, HS3, FC EAL #5, RC EAL #6, PC EAL #6
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LGS EAL Technical Basis Manual REV e, Apnl 26,1995 Page 130 of 130 l
9.0 Other .
9.1 General i l
GENERAL EMERGENCY - 9.1.4 lC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant I Declaration of General Emergency i EAL i Events are in progress or have occurred which indicate an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity OPCON 1112131d16101 BASIS This EAL allows the Shift Manager to declare a General Emergency upon the determination of an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity, but is not explicitly addressed by other EALs.
I Releases may exceed the EPA Protective Action Guidelines for more than the immediate site area and will be classified under section 5.0, Radioactivity Releases.
DEVIATION None
, REFERENCES NUMARC NESP-007, HG2, FC EAL #5, RC EAL #G, PC EAL #6
. s lgs EAL Ttbla REV a, Apnl 26,1995 Page 1 of 20 LGS EAL Table Table of Contents 1.0 Reactor Fuel 1.1 Coolant Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 12
. Containment High Radiation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.3 Irradiated Fuel or New Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.0 Reactor Pressure Vessel 2.1 Re actor Wate r level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2 R e acto r P ow e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.0 Primary Containment 3.1 Containment Pressure or Leakage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.0 Secondary Containment 4.1 Secondary Containment Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 ,
4.2 Main Ste am Lin e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 5.0 Radioactivity Release 5.1 Effluent Release and Dose .................................. 10 5.2 I n - Pla nt R adi atio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 6.0 Loss of Power 6.1 Loss of AC or DC Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation . . . . . . . . . . . . . . . . 13 7.2 Loss of Decay Heat Removal Capabililty . . . . . . . . . . . . . . . . . . . . . . . . . . 14 7.3 Loss of Assessment / Communications Capabililty . . . . . . . . . . . . . . . . . . . 15 8.0 External Events 8.1 S ecu rity Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 8.2 Fire / Explosion and Toxic / Flammable Gases . . . . . . . . . . . . . . . . . . . . . . 17 8.3 Man- Made Eve nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 8.4 N atural Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 9.0 Other ,
9.1 General ................................................ 20 OPCON CONDITIOd liltl*laistol Power Operation
[11rlalalelol Startup I11slal41slol Hot Shutdown 111:1e141 101 Cold Shutdown til Isleistol Refueling hitielaist ol Defueled
lgs EAL Table REV a, April 26,1995 Page 2 of 20 1.0 Reactor Fuel 1.1 Coolant Activity CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 1.1.1.a i i l 21
- I d I4Ipl Reactor Coolant activity > 4 C//gm Dose Equivalent lodine 131 1.1.1.b 151213141:1o1 SJAE Discharge > 2.1x1# mR/hr ALERT 1.1.2 lil21slaistol Reactor Coolant activity > 300 C//gm Dose Equivalent lodine 131 SITE AREA 1.1.3 1112181'15101 EMERGENCY Reactor Coolant activity > 300 C//gm Dose Equivalent lodine 131 AND Identified breach of Primary Containment (Tech Specs Section 3.6.1.1) OR Drywell Pressure > 10 psig GENERAL None EMERGENCY
LGS EAL Tr. bis REV a Apnl 26,1995 Page 3 of 20 1.0 Reactor Fuel 1.2 Containment High Radiation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None ALERT 1.2.2 1:12181415I01 Post LOCA Drywell Rad > 3.0x10' R/hr SITE AREA 1.2.3.a lil218I4151pl EMERGENCY Post LOCA Drywell Rad > 3.0x1# R/hr AND Identified breach of Primary Containment (Tech Specs Section 3.6.1.1) 1.2.3.b 1812181416101 Post LOCA Drywell Rad > 4x10' R/hr GENERAL 1.2.4 lil218ldislol EMERGENCY Post LOCA Drywell Rad > 3x1# R/hr 1
c
lgs EAL Table REv a. Apnl 26,1995 Page 4 of 20 1.0 Reactor Fuel 1.3 Irradiated Fuel or New Fuel CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 1.3.1.a h l IsI41siol Unexpected RPV level decrease to < 120"when reactor cavity is flooded up and fuel pool gates are in place 1.3.1.b Iil21sidl5lol Fuel Pool Storage Lo Level alarm with visual observation of a water level decrease in the Spent Fuel Pool ALERT 1.3.2.a lil21sl41slol Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1) 1.3.2.b (1121:14I6101 A fuel handling accident causing a HiHi alarm from the Refueling Area Exhaust Duct monitor (2 mR/hr)
SITE AREA None EMERGENCY GENERAL None EMERGENCY Tebb 1-1 Refuel Floor ARMS RIS29 M1-1(2)K600, Drywell Head Laydown RIS30-M1-1(2)K600, Dryer /Seperator Area RIS31-M1-1(2)K600, Spent Fuel Pool RIS32-M1-1(2)K600, New Fuel Storage Vault RIS33-M1-1(2)K600, Pool Plug Laydown
LGS EAL Table REV a. Apnl 26,1995 Page 5 of 20 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None
\
ALERT 2.1.2 !il213Idl61ol
) RPV level < -129 "
l SITE AREA 2.1.3.a RTil:141slol EMERGENCY RPV level cannot be restored above -161 "
2.1.3.b lil2ialalalo1 -
RPV level cannot be determined GENERAL 2.1.4.a lil2ialaislol EMERGENCY RPV level cannot be restored above -204 "
2.1.4.b lil2181dl6Iol RPV level cannot be determined AND RPV Flooding cannot be established per T-116
LGS EAL Table REV a Apnl 26,1995 Page 6 of 20 2.0 Reactor Pressure Vessel 2.2 Reactor Power CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None ALERT 2.2.2 lil21aldlslol Failure of Automatic RPS SCRAM to reduce reactor power < 4%
SITE AREA 2.2.3 h l 21* I'Islol EMERGENCY Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%
GENERAL 2.2.4 l il 21*14 f$lFI EMERGENCY Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%
AND Suppression Pool Temperature is > 180'F
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LGO EAL Ttble REV a. April 26,1995 Page 7 of 20 3.0 Primary Containment 3.1 Containment Pressure or Leakage CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 3.1.1.a lil21sI41slol Drywell Pressure > 1.68 ps/g AND Indication of a Leak into Containment 3.1.1.b fil21sidislol Unidentified Primary System Leakage > 10 ppm into the Drywell M
Identified Primary System Leakage > 25 ppm into the Drywell ALERT 3.1.2 lil21sidl81cl Drywell Pressure > 10 ps/g SITE AREA 3.1.3 Iil213 41siol EMERGENCY Containment Failure indicated by a rapid, unexplained drop in Containment Pressure following initial pressure rise above 10 psig G'INERAL 3.1.4.a fil21slal*lol EMERGENCY Containment Pressure > 62.5 ps/g M
Containment Venting via T-200 is required 3.1.4.b 11121:141s1o1 Drywell Hydrogen 2 6 %
AND Drywell Oxygen 2 5 %
l LGS EAL Ttbis REV a. Apnl 26,1995 Page 8 of 20 4.0 Secondary Containment 4.1 Secondary Containment Temperature CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None ALERT 4.1.2 lil21*Idislol An Unisolable Primary System Leak is discharging into Secondary Containment AND A T 103 Temperature Max Safe Operating Value is exceeded in ONE area requiring a SCRAM SITE AREA 4.1.3 lil2181dislol EMERGENCY An Unisolable Primary System Leak is discharging into Secondary Containment AND T-103 Temperature Max Safe Operating Values are exceeded in TWO OR MORE areas requiring an Emergency Blowdown per T-112 GENERAL 4.1.4 lil21slaisiol EMERGENCY An Unisolable Primary System Leak is discharging into Secondary Containment AND T-103 Temperature Max Safe Operating Values are exceeded in 1WO OR ,
MORE areas requiring an Emergency Blowdown per T-112 AND T-103 Radiation Max Safe Operating Values are exceeded in the same TWO OR MORE areas
LGS EAL Ttble REV a. Apnl 26,1995 Page 9 of 20 4.0 Secondary Containment 4.2 Main Steam Line CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 4.2.1 lil21sl41slol Main Steam Line HiHi Radiation (3xNFPB)
ALERT 4.2.2 fil2ial*Islol Failure of one or more Main Steam Lines to isolate on any MSIV Closure Signal SITE AREA 4.2.3 lilrialalelol EMERGENCY Main Steam Line Break discharging into the Turbine Enclosure AND North Stack > 1.08x10' pCl/sec GENERAL 4.2.4 tilaisidistol EMERGENCY Main Steam Line Break discharging into the Turbine Enclosure AND North Stack > 1.08x10' pCl/sec
lgs EAL Table REV a. Apnl 26,1995 Page 10 of 20 5.0 Radioactivity Release 5.1 Effluent Release and Dose CLASSlFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 5.1.1.a lil2i>I41slol North Stack or South Stack Rad monitor continuously in HiHi Alarm OR known Unmonitored Release continuously in progress for > 60 minutes AND Calculated maximum offsite dose rate exceeds 0.114 mrem /hr TPARD OR 0.342 mrem /hr child thyroid CDE based on a 60 minute average 5.1.1.b lil21*ldl5lol Report indicates Liquid Release exceeds TWO TIMES Offsite Dose Calculation Manual (ODCM 3.2.2 or 3.2.3) for > 60 minutes ALERT 5.1.2.a lil:181dislol North Stack or South Stack Rad monitor continuously in HiHi Alarm OR known Unmonitored Release continuously in progress for > 15 minutes AND Calculated maxium offsite dose rate exceeds 11.4 mrem /hr TPARD OR 34.2 mrem /hr child thyroid CDE based on a 15 minute average 5.1.2.b 18121 1415101 Report indicates Liquid Release exceeds TWO HUNDRED TIMES Offsite Dose Calculation Manual (ODCM 3.2.2 or 3.2.3) for > 15 minutes SITE AREA 5.1.3 181213141s101 EMERGENCY Projected offsite dose exceeds 100 mrem TPARD, OR Projected offsite dose exceeds 500 mrem child thyroid CDE, OR Actual offsite whole body dose rate exceeds 100 mrem /hr GENERAL 5.1.4 lil2181415Iol EMERGENCY Projected offsite dose exceeds 1000 mrem TPARD, OR, Projected offsite dose exceeds 5000 mrem child thyroid CDE, OR <
Actual offsite whole body dose rate exceeds 1000 mrem /hr NOTE: CDE = Committed Dose Equivalent '
TPARD= Total Protective Action Recommendation Dose
I LGS EAL Table REV a. Apnl 26,1995 Page 11 of 20 5.0 Radioactivity Release 5.2 in-Plant Radiation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 5.2.1 lil218ldislol inplant rs..f: tion level > 1x1# mR/hr requiring T-103 entry ALERT 5.2.2.a 1512181415I01 Inplant radiation level > 5x1# mR/hr for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requiring T-103 entry 5.2.2.b Iil2181dislol Control Room area radiation level > 15 mR/hr for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SITE AREA None EMERGENCY GENERAL None EMERGENCY l
,.n_ - w,, , e ,-- -
LGS EAL Ts.bla REV a. Apnl 26,1995 Page 12 of 20 6.0 Loss of Power 6.1 Loss of AC or DC Power CLASSlFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 6.1.1.a 1 121:14I5101 Loss of ALL Offsite Power for >15 minutes 6.1.1.b U.18181818I#1 Lcas of ALL safeguard DC Power indicated by < 105 VDC for > 15 minutes ALERT 6.1.2.a 1i1218141e191 Loss of ALL Offsite Power for > 15 m/nutes AND Only ONE 4 KV Safeguard Bus is available 6.1.2.b 1112181415101 Loss of ALL Offsite Power AND ALL 4 KV Safeguard Busses are unavailable for > 15 minutes SITE AREA 6.1.3.a 1i121 14161*1 EMERGENCY Loss of ALL Offsite Power AND ALL 4 KV Safeguard Busses are unavailable for > 15 minutes 6.1.3.b I 121:141s101 Loss of ALL safeguard DC Power indicated by < 105 VDC for > 15 minutes GENERAL 6.1.4 Iil2l3lAISID] '
EMERGENCY loss of ALL Offsite Power AND ALL 4 KV Safeguard Busses are unavailable for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
LGS EAL Ttbla REV a. Apnl 26,1995 Page 13 of 20
- 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 7.1.1 fil2IsI41siol Unable to bring the Plant to the required OPCON within Tech Spec LCO action times ALERT 7.1.2 !il2ial41slol Control Room evacuation procedures have been initiated SITE AREA 7.1.3 Iil2laldislol EMERGENCY Failure to establish Remote Shutdown Control of the Plant within 15 minutes after evacuation of the Control Room GENEFlAL None EMERGENCY
I lgs EAL Ttble REV a, Apnl 26.1995 Page 14 of 20 7.0 Internal Events 7.2 Loss of Heat Removal Capability CLASSlFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 7.2.1 ht:Isidislol Uncontrolled Reactor Coolant temperature increase to > 200 *F ALERT 7.2.2 tilaisidislol Uncontrolled Reactor Coolant temperature increase to > 200 'F AND Inability to establish alternate decay heat removal capability GITE AREA 7.2.3 lil218I4181ol EMERGENCY loss of Main Condenser as a heat sink AND Loss of Suppression Pool Heat Sink capabilities as evidenced by T-102 legs
[SP/T 9P/L, PC/P, or DWrf] requiring an Emergency Blowdown GENERAL Nos..
EMERGENCY MI
l lgs EAL Ttble REv a, Apol 26.1993 Page 15 of 20 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 7.3.1.a lil21sI41siol Loss of All Annunciators in the Control Room for > 75 minutes 7.3.1.b 1112181418101 Loss of ALL Onsite communications (Table 7-1)
OR Loss of ALL Offsite communications (Table 71)
ALERT 7.3.2 15121:1415101 Loss of All Annunciators in the Control Room for > 15 minutes AND Significant Plant Transient (Table 7-2) is in progress OR Plant Monitoring System (Plant Process Computer and ERFDS for Unit 1) is unavailable SITE AREA 7.3.3 lil21sldlslol EMERGENCY Loss of All Annunciators in the Control Room for > 15 minutes AND Significant Plant Transient (Table 7-2) is in progress AND Plant Monitoring System (Plant Process Computer and ERFDS for Unit 1)is unavailable OR Loss of good data on RM 11 as evidenced by all release point indications being magenta GENERAL None EMERGENCY Table 7-1 Communications Tab;e 7-2 Plant Transients Onsite Offsite Site Phones (oimenmon 2000) X X SCRAM PRELUDE System X X Recire Runbacks > 25% thermal power Plant Public Address X Thermal power oscillations > 10%
Station Radio X Stuck open relief valve (s)
NRC (FTS-2000) X ECCS injection PA State Police Radio X County Police Radio X Load Dispatcher Radio X PECO Dial Network X s _a
lgs EAL Table REV a. Apnl 26,1995 Page 16 of 20 8.0 External Events 8.1 Security Events CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 8.1.1 lil21 141slol Confirmed security threat directed towards the station as evidenced by :
- Credible sabotage or bomb threat within the Protected Area, OR Credible intrusion and attack threat to the Protected Area, QR_
Attempted intrusion and attack to the Protected Area, OR Attempted sabotage discovered within the ProtectedNital Area, OR Hostage / Extortion situation that threatens normal plant operations ALERT 8.1.2 lil:Islaisiol Confirmed hostile intrusion or act within the Protected Area as evidenced by Actual attack and intrusion into the Protacted Area, QR
- Suspected bomb, sabotage or sabotage device discovered in the ProtectedNital Area SITE AREA 8.1.3 lil:181415Iol EMERGENCY Confirmed hostile intrusion or act in plant Vital Areas as evidenced by :
Actual attack and intrusion into a Vi'.al Area, QR_
Confirmed bomb, sabotage or sabotage device discovered in a ProtectedNital Area GENERAL 8.1.4 lil:Islaisiol EMERGENCY Security event resulting in the actual loss of physical control of the:
Control Room QR_ Remote Shutdown Capability l
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LGS EAL Tsbb REV a. Apnl 26,1995 Page 17 of 20 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases CLASSIF! CATION EMERGENCY ACTION LEVEL ,
UNUSUAL EVENT 8.2.1.a lil21sl4Islol Fire within SE-8 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of verification of alarms 8.2.1.b l'l218I415101 Normal plant operation is impeded by:
Toxic or flammable gas concentrations confirmed within the Protected Area OR Control Room informed by Local, County or State Officials to evacuate personnel within the protected area due to an offsite gas release 8.2.1.c lilalaldislol Report by plant personnel confirming the occurrence of an explosion in a Plant Vital Structure (Table 81)
ALERT 8.2.2.a 1812181415101 Fire or explosion within Plant Vital Structures (Table 81) damages safe shutdown systems causing or jeopardizing plant shutdown .
8.2.2.b 1112181415101 Plant personnel are unable to perform actions necessary for safe shutdown without appropriate personal protection equipment due to toxic or flammable gas concentrations confirmed within Plant Vital Structures (Table 8-1)
SITE AREA None EMERGENCY GENERAL None EMERGENCY Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network
I lgs EAL Ttbb REV a. Apnl 26,1995 Page 18 of 20 8.0 External Events 8.3 Man-Made Events CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 8.3.1.a lil2ial41*lol Confirmed report of any vehicle crash affecting Plant Vital Structures (Table 8-1) 8.3.1.b h I 21 a 141 *101 Turbine failure resulting in casing penetration OR Damage to Generator releasing hydrogen requiring immediate plant shutdown ALERT 8.3.2 hl21sI41*lol Confirmed report of damage to Plant Vital Structures (Table 8-1) from either a vehicle crash or missile impact SITE AREA None EMERGENCY GENERAL None EMERGENCY Tabl7 8-1 Plant Vital Structures Reactor Enclosu.e Control Enclosure Turbine Enclosure Dies:1 Generator Enclosure Spray Pond Pump House / Spray Network v- > + ' - - ,% - _
lgs EAL Ttbla REV a. Apnl 26,1995 Page 19 of 20 8.0 External Events 8.4 Natural Events CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 8.4.1.a lil2Isl4151ol Earthquake >.005 g 8.4.1.b I112181415101 Report of a Tornado within the Site Boundary OR Wind speeds > 75 mph as indicated on site Metrological data for > 15 minutes ALERT' 8.4.2.a lil21314Islol Earthquake >.075 g 8.4.2.b 111213i415101 Tornado or wind speeds > 75 mph causing damage to Plant Vital Structures (Table 8-1)
SITE AREA None EMERGENCY GENERAL None EMERGENCY Trble 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network
l LGS EAL Tabla REV a. Apnl 26,1995 Page 20 of 20 9.0 Other 9.1 General CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT 9.1.1 lil21aldislol Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant ALERT 9.1.2 1812181d15101 Events are in progress or have occurred which indicate an actual or potential substantial degradation of the level of safety of the plant SITE AREA 9.1.3 1112181418101 EMERGENCY Events are in progress or have occurred which indicate an actual or likely major failure of plant functions needed for protection of the public GENERAL 9.1.4 1812181415101 EMERGENCY Events are in progress or have occurred which indicate an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity
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LGS EAL NUMARC Companson REV a, Apnl 26,1995 Page 1 of 8 LGS EAL NUMARC Comparison Table of ContentsSection I - NUMARC EAL Versus LGS EAL Comparison Matrix . . . . . . . . . . . . . . . . . 2 Section 11 - Summary of Fission Product Barriers . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 F
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LGS EAL NUMARC Companson REV a. Apnl 26,1995 Page 2 of 8 Section 1 - NUMARC EAL Versus LGS EAL Comparison Matrix NUMARC LGS COMMENTS / EXCEPTIONS AU1.1 5.1.1.a AU1.2 5.1.1.a 5.1.1.b AU1.3 N/A LGS does not have telemetered perimeter monitors.
AU1.4 N/A LGS does not use automatic initiation of real time dose assessment.
AU2.1 1.3.1.a 1.3.1.b AU2.2 1.3.1.b AU2.3 N/A LGS does not have dry fuel storage capabilities.
AU2.4 5.2.1 AA1.1 5.1.2.a AA1.2 5.1.2.a 5.1.2.b AA1.3 N/A LGS does not have telemetered perimeter monitors.
AA1.4 N/A LGS does not use automatic initiation of real time dose assessment AA2.1 1.3.2.a 1.3.2.b AA2.2 1.3.2.a AA2.3 1.3.2.a This EAL is addressed by utilizing radiation levels which could be caused by uncovering the fuel.
AA2.4 1.3.2.a LGS does not have level indication on the Spent Fuel Pool.
This EAL is addressed by utilizing radiation levels which could be caused by uncovering the fuel.
AA3.1 5.2.2.b AA3.2 5.2.2.a AS1.1 5.1.3 LGS contains this within AS1.3 & 1.4, but does not specifically address.
LGS EAL NuMARc Companson REV a, Apnl 26.1995 Page 3 of 8 NUMARC LGS COMMENTS / EXCEPTIONS AS1.2 N/A LGS does not have telemetered perimeter monitors.
ASI .3 5.1. 3 AS1.4 5.'I.3 l
AG1.1 5.1.4 LGS contains this within AS1.3 & 1.4, but does not specifically address.
AG1.2 N/A LGS does not have telemetered perimeter monitors.
AG1.3 5.1.4 AG1.4 5.1.4 HU1.1 8.4.1.a HU1.2 8.4.1.b HU1.3 9.1.1 HU1.4 8.3.1.a HU1.5 8.2.1.c HU1.6 8.3.1.b This EAL addresses main turbine rotating component failures, which do not occur during shutdown conditions. Therefore, only OPCON 1,2 and 3 will be used for this EAL.
HU1.7 8.4.1.b LGS does not have any EAL addressing hurricanes or seiches which are not applicable. High winds are addressed.
HU2 8.2.1.a HU3.1 8.2.1.b HU3.2 8.2.1.b HU4.1 8.1.1 HU4.2 8.1.1 HUS 9.1.1 HA1.1 8.4.2.a ._
HA1.2 8.4.2.b HA1.3 8.4.2.b a
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l lgs EAL NUMARc Comparison REV a. Apnl 26.1995 Page 4 of 8 NUMARC LGS COMMENTS / EXCEPTIONS HA1.4 N/A LGS has no instrumentation other than meteorological which can be used to measure the impact of na'f al or destructive phenomena.
HA1.5 8.3.2 HA1.6 8.3.2 HA1.7 N/A There are no other site occurrences which have been identified.
HA2 8.2.2.a HA3.1 8.2.2.b HA3.2 8.2.2.b HA4.1 8.1.2 HA4.2 8.1.2 HA5 7.1.2 HA6 9.1.2 HS1.1 8.1.3 HS1.2 8.1.3 HS2 7.1.3 HS3 9.1.3 HG1.1 8.1.4 HG1.2 8.1.4 HG2 9.1.4 SU1 6.1.1.a SU2 7.1.1 SU3 7.3.1.a SU4.1 1.1.1.b 4.2.1 SU4.2 1.1.1.a LGS Technical Specifications only require only OPCONS 1,2,3,4.
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l LGS EAL NuMARC Companson l REv a Apnl 26,1995 Page 5 of 8 NUMARC LGS COMMENTS / EXCEPDONS SUS 3.1.1.a 3.1.1.b SU6 7.3.1.b SU7 6.1.1.b SA1 6.1.2.b SA2 2.2.2 SA3 7.2.1 EAL 7.2.1 was created to be a precursor to SA3 which calls 7.2.2 for an Alert to be declared for a loss of shutdown cooling function and reactor coolant temperature rises above 200 *F.
The reactor core can continue to be properly cooled as long as Torus cooling is available. While this event clearly warrants the declaration of an Unusual Event, LGS does not believe it rneets the threshold for an Alert.
SA4 7.3.2 SAS 6.1.2.a SS1 6.1.3.a SS2 2.2.3 SS3 6.1.3.b SS4 7.2.3 SS5 N/A LGS has not identified a specific EAL for this condition. Both conditions stated in this reference are addressed appropriately in EAL sections 7.2.2 and 2.1.2,2.1.3 and 2.1.4.
SS6 7.3.3 SG1 6.1.4 SG2.1 2.2.4 SG2.2 2.2.4 FC.1 1.1.2 1.1.3 FC.2 2.1.3.a 2.1.4.a FC.3 1.2.3.b
lgs EAL NUMARc Companson REV a. Apnl 26.1995 Page 6 of 8 NUMARC LGS COMMENTS / EXCEPTIONS FC.4 1.2.2 ,
1.2.3.a 1.2.4 2.1.3.b 2.1.4.b 3.1.4.a 3.1.4.b 4.1.4 4.2.4 FC.5 9.1.2 9.1.3 .
9.1.4 RC.1 N/A This barrier failure is not explicity addressed at LGS. There are several EALs, including Reactor Pressure Vessellevel and Primary Containment pressure which adequately address this situation.
RC.2 1.1.3 There is a deviation in that a high drywell pressure 10 psig 3.1.2 was selected duen to the fact that the alarm value of 1.68 3.1.3 psig can be reached by a small primary system leak and 10 psig is larger than experience shows of blown packing and recirc seal leaks. It is more representative of a LOCA condition and this torus pressure is in the TRIPS for actions to protect the containment.
RC.3 1.2.3 b 1.2.4 RC.4 2.1.3.a 2.1.4.a RC.5 2.1.2 2.1.3.b 2.1.4.b 3.1.4.a 3.1.4.b 4.1.2 4.1.3 4.1.4 4.2.3 4.2.4
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LGS EAL NUMARC Companson REV a. Apnl 26.1995 Page 7 of 8 NUMARC LGS COMMENTS / EXCEPTIONS RC.6 9.1.2 9.1.3 9.1.4 PC.1 3.1.3 3.1.4.b PC.2 3.1.4.a 4.1.3 4.1.4 4.2.2 4.2.3 4.2.4 PC.3 1.2.4 PC 4 2.1.4.a PC.5 1.1.3 1.2.3.a 2.1.4.b 4.1.2 PC.6 9.1.2 9.1.3 9.1.4 1
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LGS EAL NUMARC Companson l REV a. Apr8 26,1995 l Page 8 of 8 Section II- Summary of Fission Product Barriers The following summarizes the EALs whiCh resulted from the analysis performed of the fission product barrier methodology of NUMARC NESP-007 for Limerick Generating Station i UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 9.1.1 Emergency Director 1.1.2 RPV Coolant Actmty > 300 pCPgm DEI 1.1.3 RPV Coolant Activity > 300 pCi/gm DEI 1.2.4 Cont Rad > 3x10' Rhr (FC-4 bss and Judgement (FC-1 loss) (FC-1 bss) and cont integrity bss (PC- RC-3 loss and PC-3 potentialloss) 5 loss) or Dryweg > 10 psig (RC-2 bss) 1.2.2 Cont Rad > 3x10' Rhr (FC-4 loss) 2.1.4 a RPV <-204'(FC-2 loss ed RC-4 bss l 1.2.3.a Cont Rad > 3x10' Rhr (FC-4 bss) and and PC-4 potennat loss) j 2.1.2 RPV <-129'(RC-5 potentialloss) loss containment integnty (PC-5 loss) 2.1.4.b RPV level unknown and RPV flooding 3.1.2 Drywell > 10 psig (RC-2 bss) 1.2.3.b Cont Rad > 4x10' Rhr (FC-3 loss and not establiished (FC-4 potential loss, RC-3 bss) RC-5 bss and PC-5 loss) 4 1.2 RCS discharging into secondary cont (RC-5 potennat loss and PC-5 potential 2.1.3 a RPV <-161-(FC-2 potennalloss and 3.1.4 a Cont Press > 62.5 psig or vennng bss) RC-4 loss) required (FC-4 potennalloss and RC-5 loss and PC-2 less) 42.2 Failure of MSIVs (RC-5 potennat loss 2.1.3.b RPV level undetermined (FC-4 and PC-2 potentialloss) potential bss and RC-5 loss) 3.1.4 b Drywell H2 > 6% and O2 > 5% (FC-4 loss and RC-5 loss and PC-1 potennal 9.1.2 Emergency Director Judgement 3.1.3 Cont Press >10 pseg with rapid drop bss)
(RC-2 loss and PC-1 loss) 4.1.4 RCS dschargog into secondary cont 4.1.3 RCs dscharging into secondary cont and indicat:on of fuel damage (FC-4 (RC-5 bss and PC-2 loss) loss and RC-5 loss and PC-2 loss) 4.2.3 MS break into Turbine Building (RC-5 4.2.4 MS break into Turbine Buildiing with loss and PC-2 loss) indcation of itel damage (FC-4 loss and RC-5 loss and PC-2 bss) 9.1.2 Emergency Director Judgement 9.1.4 Emergency Director Judgement
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