ML20202G880
| ML20202G880 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom, Limerick  |
| Issue date: | 01/22/1999 |
| From: | Geoffrey Edwards PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9902050323 | |
| Download: ML20202G880 (74) | |
Text
I Station Support Department t:!!::
v PECO NUCLEAR ec m.,,, c-roo aro A Unit of PECO Energy $,5c
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January 22,1999 Docket Nos. 50-277 50-278 50-352 50-353 License Nos. DPR-44 DPR-56 NPF 39 NPF-85 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Peach Bottom Atomic Power Station, Units 2 and 3 Limerick Generating Station, Units 1 and 2 SupplementalInformation Conceming Revised Emergency Action Levels
Dear Sir / Madam:
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/
By letter dated April 16,1998 PECO Energy submitted revised Emergency Action Level (EAL) j guidelines for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. and Limerick Generating Station (LGS), Units 1 and 2. These revised EALs were developed based on the j' guidance provided in NUMARC/NESP-007, " Methodology for Development of Emergency Action /
Levels." 3\ {L\ s By letter dated November 16,1998. PECO Energy responded to an NRC request for additional information issued on September 24,1998, conceming the revised EALs. Subsequently, during a telephone conversation on December 23,1998, the NRC requested that PECO Energy provide supplemental information in order to clarify and update information previously submitted. This information is necessary in order for the NRC to continue its review of the revised EALs for PBAPS, Units 2 and 3, and LGS, Units 1 and 2.
Accordingly, the enciosures to this letter provide the requested information. Enclosure 1 contains information for PBAPS, Units 2 and 3, and Enclosure 2 contains similar information for LGS, Units 1 and 2. A desenption of what is contained in the enclosures is provided below.
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9902050323 990122 PDR ADDCK 05000277 ,
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January 22,1999 Page 2 l
Enclosure 1 (PBAPS IMe.ir,&tien)
Revised EAL 5.1, "Emuent Release and Dose" (Site Area Emergency - 5.1.3)
Revised EAL 5.1,"Emuent Release and Dose" (General Emergency 5.1.4)
Revised EAL Table page for EALs 5.1.3 and 5.1.4 Revised EAL 3.4, Primary Contamment Bamer" (PC.1, *Drywell Pressure") l Revised EAL 8.2,
- Fire /E=*=ian and Toxic / Flammable Gases" (Alert - 8.2.2.a)
Revised EAL Table page for EAL 8.2.2.a Revised EAL 7.2," Loss of Decay Heat Removat Capability" (Alert - 7.2.2) l Revised EAL 7.2," Loss of Decay Heat Removal Capability" (Site Area Emergency - 7.2.3)
Revised EAL Table page for EALs 7.2.2 and 7.2.3 PBAPS Procedure ON-125. " Loss of Shutdown Cooling - Procedure" Revised EAL 3.3,
- Reactor Coolant System Bamer" (RC.5, "Other indications")
Revised EAL 3.4, " Primary Contamment Bamer" (PC.5, "Other Indications")
Enclosure 2 (LGS Information)
Revised EAL 5.1, "Emuent Release and Dose" (Site Area Emergency - 5.1.3)
Revised EAL 5.1, "Emuent Release and Dose" (General Emergency 5.1.4)
Revised EAL Table page for EALs 5.1.3 and 5.1.4 Revised EAL 3.4, Primary Containment Bamer" (PC.1, "Drywell Pt essure")
' Revised EAL 8.2," Fire / Explosion and Toxic / Flammable Gases"(Alert - 8.2.2.a)
Revised EAL Table page for EAL 8.2.2.a Revised EAL 7.2," Loss of Decay Heat Removal Capability" (Alert - 7.2.2)
Revised EAL 7.2," Loss of Decay Heat Removal Capability" (Site Area Emergency - 7.2.3)
Revised EAL Table page for EALs 7.2.2 and 7.2.3 LGS Procedure GP-6.2,
- Shutdown Operations - Shutdown Condition Tech Spec Actions" Revised EAL 3.3," Reactor Coolant System Barrier" (RC.5, "Other Indications')
Revised EAL 3.4," Primary Containment Bamer" (PC.5,"Other indications")
If you have any questions or require additionalinformation, please do not hesitate to contact us.
Very truly yours, O
- r_ -_.
Garrett D. Edwards Director Licensing Enciosures cc: H. J. Miller, Administrator, USNRC. Region I (w/ enclosures)
A. C. McMurtray, USNRC Sernor Resident inspector, PBAPS (w/ enctosuresi A. L. Burritt, USNRC Senior Resicent inspector, LGS (w/ enclosures)
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ENCLOSURE 1 Peach Bottom Atomic Power Station Units 2 and 3 1
Supplemental information for i Revised Emergency Action Level Guidelines 1
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PBAPs EAL Tedncal Basis Manual ;
REV o1. January 15.1999 4
Page 67 of 128 5.0 Radioactivity Release 1 5.1 Emuent 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 -
4 EAL.
A valid reading on one or more of the fobowing radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes A. NQ Dose Projections are not available:
Main Stack 5.84 Cl/cc Vent Stack 2.08E-3 pCi/cc Torus Vent 203 cpm Note: If the dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.
98 Projected offsite dose using computer dose model exceeds 100 mrem TPARD QB '
500 mrem child thyroid CDE QR Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 100 ,
mrem /hrexpected to continue for more than one hour, QB Analysis of Field Survey results indicate child thyroid dose commitment of 500 mrem for one hour of inhalation OPCON " " * '
BASIS
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Valid means that a radiation monitor reading has been confirmed by the operators to be correct.
A 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.
Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE)is equal to the thyroid exposure due to iodine. The computerized dose model provides projected TPARD and CDE.
PSAPs EAL Techrucal Basia Manual ReV o1. January 15.1999 Pa9e 68 of 128 An actual or projected dose of 100 mrem Total Protective Action Recommendation Dose l (TPARD)is based on the 10 CFR 20 annual average population exposure limit. 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 considerston of the 1:5 ratio of the EPA Protectrve Action Guidelines for TPARD and Child Thyroid Committed Dose Equivalent (CDE). Actual meteorology is used, since it gives the most accurate dose projection.
Monitor indications are calculated using the computerized dose model with UFSAR source terms applicable to each monitored pathway in conjunction with annual average meteorology and a one hour release duration. The inputs are as follows:
Main Stack Vent Stack Torus Vent Stability Class E E E Wind Speed 11.4 mph 6.3 mph 6.3 mph Wind Direcbon 45' 22' 22' Accident LOCA LOCA LOCA Release Rate 5.84 Ci/cc 2.08E-3 Cl/cc 203 cpm 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 Protechve Acton Guidelines per EAL Sechon 5.1.4.
DEVIATION
-lNone REFERENCES NUMARC NESP-007, AS1.1, AS1.3 and AS1.4 EPA 400 i
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PBAPs EAL Techrucal Base Manusi
, REV c1. January 15,1999 l Page 69 of 128 '
5.0 Radioactivity Release 5.1 Effluent Release and Dose j
GENERAL EMERGENCY - 5.1.4 2
IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactmty a
that Exceeds 1000 mR Whole Body or 5000 mR Child Thyroid for the Actual or
' Projected Duration of the Release Using Actual Meteorology 4.
- EAL A valid reading on one or more of the following radiation monitors that exceeds or is l
expected to exceed the value shown for > 15 minutes AND Dose Projections are not ;
j available: l
' i Main Stack 58.4 Ci/cc
[ Vent Stack 2.08E-2 Cl/cc j
! Torus Vent 2000 cpm Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.
E Projected offsite dose using computer dose model exceeds 1000 mrem TPARD OR_ _
5000 mrem child thyroid CDE E
Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mrem /hrexpected to continue for more than one hour. OR Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mrem for one hour of inhalation OPCON unaau BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.
A monitor reading is considered to be valid when it is venfied by:
- 1. An instrument check indicating the monitor has not failed;
-2 Indications on related or redundant instrumentation; cr,
- 3. Direct observation by plant personnel.
Total Protective Action Recommendation Dose (TPARD)is equal to Total Effective Dose Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to the thyroid exposure due to iodine. The computerized dose model provides projected TPARD and CDE.
PBAPs EAL Techrace! Basis Manual REV D1, January 15.1999 Page 70 of 128 The 1000 mR TPARD and the 5000 mR child thyroid Integrated dose are based on the EPA 4 l protective action guidance. This is consistent wi'.h the emergency class description for a General l i Emergency. This level constitutes the upps.* level of the desirable gradient for the Site Area
( Emergency. Actual meteorology is specifically identified in the initiating condition since it gives the I most accurate' dose assessment.
Monitor indications are calculated using the computerized dose model with UFSAR source terms applicable to each monitored pathway in conjunction with annual average meteorology j cnd a one hour release duration. The inputs are as follows: '
Main Stack Vent Stack Torus Vent l Stability Class E E E Wind Speed 11.4 mph 6.3 mph 6.3 Wind Direction 45' 22* 22' Accident LOCA LOCA LOCA Release Rate 58.4 Ci/cc 2.08E-2 Cl/cc 2.026E+3 cpm Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with l Pennsylvar,ia Emergency Management Agency (PEMA) / Bureau of Radiation Protection (BRP).
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DEVIATION
] None REFERENCES NUMARC NESP-007, AG1.1, AG1.3 and AG1.4 EPA-400 l
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l PBAPs EAL Tcbie REV o1. January 10,1999 t ~I Page 11 of 25 SITE AREA IC Boundary Dose Resulting from an Actual or imminent Release of EMERGENCY Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or Projected Duration of the Release 5.1.3 " " " +
A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:
l Main Stack 5.84 Ci/cc Vent Stack 2.08E-3 Ci/cc Torus Vent 203 cpm Note: If the required dose projections cannot be completed within the 15 minute l period, then the declaration must be made based on the valid sustained monitor reading.
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!. Projected offsite dose using computer dose model exceeds l 100 mrem TPARD E 500 mrem child thyroid CDE E
Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 100 mrem /hr expected to continue for more than one hour, @
Analysis of Field Survey results indicate child thyroid dose commitment of 500 mrem for one hour of inhalation IC Boundary Dose Resulting from an Actual or Imminent Release of GENERAL EMERGENCY 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 5.1.4 ' ~ ~
A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:
Main Stack 58.4 Ci/cc Vent Stack 2.08E-2 Ci/cc Torus Vent 2000 cpm Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.
Projected offsite dose using computer dose model exceeds 1000 mrem TPARD % 5000 mrem child thyroid CDE M
Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mrem /hr expected to continue for more than one hour, @
Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mrem for one hour of inhalation NOTE: CDE = Committed Dose Equivalent TPARD = Total Protective Action Recommendation Dose
- P8APS EAL Technical Seas Manuel REV 09. January 1R 1999 i Page e of 128 3.0 Fission Product Barrier 3.4 Primary Containment Barrier PC.1 Drywell Pressure EAL LOSS Rapid, unexplained decrease in Drywell Pressure following initial increase
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Drywell pressure response not consistent with LOCA conditions POTENTIAL LOSS Drywell Pressure > 49 psig and increasing
.QB Drywell Hydrogen > 6% A,,N,,LQ N Drywell Oxygen > 5%
OPCON' -
BASIS '
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 relearse into containment from a-LOCA. Thus, drywell pressure not increasing under these corditions indicates a loss of containment integnty.- The 49 ps/g for potential loss of containment is based on the containment drywell design pressure and is equal to the peak presture expected from a DBA LOCA.
The specified valuefof 6% hydrogen concentration is the minimum which can support a deflagration. Likewise, the minimum concentration of ' oxygen required to support a deflagration is 5%. Combustion of hydrogen in the deflagration concentration range creates a traveling flame causing a rapid rise in primary containment pressure. A deflagration may result in a peak primary containment pressure high enough to rupture the primary containment or
- damage the drywell-to-torus boundary.
DEVIATION .
None REFERENCES NUMARC NESP-007, PC EAL #1 ON-110, Loss of Primary Containment -
T-101, RPV Control
- [ T-102, Primary Containmen' t Control w/ Bases
, T-103, Secondary Containment Cantrol-m ...__;. .- . _ . , _
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[ P8hs EAL Techneet Bass Manuel
' i' REV D1, January 15.1999 Page 110 of 138 8.0 ExternalEvents 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
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The following conditions exist u
Fire or explosion which potentially makes inoperable:
Two orMore subsystems of a Safe Shutdown System (Table 8-2) QR Two orMore
~ Safe Shutdown Systems QB Plant Vital Structures containing Safe Shutdown Equipment AND Safe Shutdown System or Plant Vital Structure is required for the present Operational Condition
.OPCON mm BASIS
, The primary concem of this EAL is the magnitude of the fire and the effects on Safe Shutdown Systems required for the present Operational Condition. A Safe Shutdown System is defined as any system required to maintain safe operation or to establish or maintain Cold Shutdown.
A system being " inoperable" means that it is incapable of performing the design function. For example, the LPCI System is intended to maintain adequate core cooling by covering the core
- to at least 2/3 core height following a DBA LOCA. In order for the system to be unable to l
maintain its intended function, multiple loops would need to be disabled by the fire. In addition to indication of degraded system performance, potentialinoperability may be determined by visual observation and other control room indications such as loss of indicating lights.
. Table 8 2 Safe Shutdown Systems Diesel Generators 4KV Safeguard Buses ADS HPCI- RCIC RHR (All Modes)
Core Spray HPSW ESW SBGTS ECW CAC/ CAD PCIS Control Room Ventilation l- Safe Shutdown Analysis is consulted to determine systems required for the applicable mode.
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PBAPS EAL Technmal Bases Manuel REV Dt. January 15,1999 Page 111 of 128 Two examples of applying this methodology are as follows:
Diesel Generators and 4 KV Safeguard Buses The fire disables multiple Diesel Generators or 4 KV Safeguard Buses so that the number of emergency power systems available would be decreased to below wnat would be required to mitigate an accident under the current operating conditions.
For 100% power, this could be conservatively interpreted as at least two Diesel Generators or 4 KV Buses disabled.
The fire disables multiple loops of LPCI so that adequate core submergence could not be assured following a DBA LOCA. For 100% power, this could also be conservatively interpreted as at least two loops disabled.
The EAL includes the condition that the fire must make "TWO OR MORE" subsystems or "TWO OR MORE" systems inoperable. In those cases where it is believed that the fire may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is ;
required under this EAL if the structure houses or otherwise supports Safety Systems required l for the present Operational Condition. '
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 or I made inoperable. 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.
Fire is defined as combustion characterized by the generation of heat and smoke. Sources of smoke such as overheated electncal equipment and slipping drive belts, for example, do not constitute fires. Observation of a flame is preferred, but is NOT required if large quantities of smoke and heat are observed.
This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.
DEVIATION '
None REFERENCES NUMARC NESP-007, HA2 PBAPS Safe Shutdown Analysis NUMARC Questions and Answers, June 1993," Hazards Questien #7"
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P8APs EAL Table REV D1. January 13,1999
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Page 20 of 25 8.0 External Events L - 8.2 Fire / Explosion and Toxic / Flammable Gases i
1 CLASSIFICATION . EMERGENCY ACTION LEVEL I UNUSUAL EVENT IC .- Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection 8.2.1.a '
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Fire within ON-114 Plant Vital Structures (Table 8-1) which is not extinguished l within 15 minutes of control room notification or verification of a control room alarm ]
l lC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe
.., Operation of the Plant 8.2.1.b -
l Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant QR !
Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event lC Natural and Destructive Phenomena Affecting the Protected Area l
I 8.2.1.c o-Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment ALERT ~ lC Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown i 8.2.2.a -
The following conditions exist:
Fire or explosion which potentially makes inoperable:
p Two orMore subsystems of a Safe Shutdown System (Table 8-2) QR
.Two cr More Safe Shutdown Systems QR Plant Vital Structures containing Safe Shutdown Equipment AND Safe Shutdown System or Plant Vital Structure is required for the present Operatenal Conoition
, F8APs EAL Tecmcei Bass Manuel REV of. January 13.1900 l _ Page 90 of 128 -
7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability ALERT ; 7.2.2 IC _ inability to Maintain Plant in Cold Shutdown EAL The following conditions exist Unplanned Loss of ey, Tech Spec required systems available to provide Decay Heat Removal functions -
8NQ Uncontrolled Temperature increase that either:
. Exceeds 212 *F (Excluding a <15 minute rise >212' F with a heat removal function restored)
Results in temperature rise approaching 212 'F l- (with NO heat removal function restored)
OPCON -
BASIS This EAL aodresses complete loss of functions required for core cooling during refueling and cold _ shutdown modes.' A loss of-Technical Specifications components is paired with exceeding :emperature limits to acknowledge additional plant capabilities to maintain plant cooling. Escalation to Site Area Emergency or General Emergency would be via Effluent Releasehn. Plant Radiation or Emergency Director Judgement ICs.
The statement " Unplanned loss of ALL Tech Spec required systems available to provide l
Decay Heat Removal functions"is intended to represent a complete loss of functions available, '
' or an inaceouate ability, to provide core cooling during the Cold Shutdown and Refueling Modes, including altamate decay heat removal methods. This EAL allows for actions taken in
~ ON-125,"Less of Shutdown Cooling - Procedure," to reestablish RHR in the Shutdown Cooling Mode or provide _for attemate methods of decay heat removal, with the intent of maintaining RCS temperature below 212* F. .
~ For loss of an in-service Decay Heat Removal system with other decay heat removat methods
= cvailable, a:::ons taken to provide for restoration of a decay heat removal function may require
. time to'im:::ement. If the event results in RCS temperature " momentarily" (for less than 15 minutes) ris:ng above 212'F_ with heat removal capability restored, Emergency C; rector / Shift Managemen: judgment will be required to determine whether heat removal systems are (dequate :: prevent boiling in the core and restoration of 'RCS temperature control.
l-1 Momentary not to exceed _15 minutes) unplanned excursions above 212* F, wnen alternate
- decay heat removal capabilities exist, should not be classified under this EAL.-
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PBAPS EAL Techncal Basis Manual REV D1, January 13,1999 l Page 91 of 128
" Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff.
The EAL guidance related to uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower than the cold shutdown temperature limit.
l This EAL is concemed with the ability to keep the reactor core temperature less than 212 'F, l The criteria of uncontrolled Reactor Coolant temperature increase > 212 *F is met as soon as it l becomes known that sufficient cooling cannot be restored in time to maintain the temperature
< 212 'F, regardless of the current temperature. The inability to establish altemate methods of decay heat removal indicates that either attemate 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.-
Escalation to the Site Area Emergency is by EAL IC, " Loss of Water Level in the Reactor Vessel that has or will uncover Fuel in the Reactor Vessel," or by Effluent Release /In-Plant Radiation ICs.
DEVIATION None REFERENCES NUMARC NESP-007, SA3
[ ON-125, Loss of Shutdown Cooling - Procedure Technical Specifications
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PBAPs EAL Technmal Base Manual
. REV o1, January 10,1999
. l Page 92 of 128 s 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 l Loss of TORUS heat sink capabilities as evidenced by T-102 legs requiring an Emergency Blowdown'
'OPCON r e ss BASIS:
This EAL addresses complete loss of functions, including ultimate heat sink, required for hot l shutdown with the reactor at pressure and temperature. Reactivity control is addressed in other. EALs. The loss of heat removal function is indicated by T-102 legs requiring an Emergency Blowdown which is directed when the Heat Capacity Temperature Limit (HCTL) curve is exceeded.
Under these conditions, there is an actual major failure of a system intended for protection of l the public. Thus, declaration of a Site Area Emergency is warranted. Escalation to General {
L Emergency would be via Effluent Release /In-Plant Radiation, Emergency Director Judgement, 1 or Fission Product Barrier Degradation ICs.
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f PBAPS EAL Tectncal Bases Manual
_ REV D1. January 10,1999
._ l Page 93 of 128 DEVIATION -
-None 2 REFERENCES l NEl 97-03, SSA T-102, Primary Containment Control, SP/L-8 f
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PBAPs EAL Table REV D1, January 15.1999 l Page 16 of 25 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None ALERT IC Inability to Maintain Plant in Cold Shutdown 7.2.2 EEEEEE The following conditions exist:
Unplanned Loss of ALL Tech Spec required sy w'ms available to provide Decay Heat Removal functions AND Uncontrolled Temperature increase that either.
. Exceeds 212 *F (Excluding a <15 minute rise >212' F with a heat removal function restored) 9.R,
. Results in temperature rise approaching 212 *F l (with N_Q heat removal function restored)
IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown SITE AREA EMERGENCY 7.2.3 cmmu
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Loss of TORUS heat sink capabilities as evidenced by T-102 legs requiring an Emergency Blowdown l
GENERAL. None EMERGENCY s
ON-125 PROCEDURE Rev. 2 Page 1 of 19 JPB:jb
.PECO Energy Company Peach Bottom Units 2 and 3 OM-125 LOSS OF SHUTDOWN COOLING - PROCEDURE (Revision 2 is a complete rewrite.)
1.O BYMPTOMS (An unexpected QE unexplained loss /inoperability of a required RHR Shutdown Cooling (SDC) subsystem.
- 2. O. OPERNEGE ACTIONS 2.1 IE unit conditions change such that a section of this
' procedure is no' longer applicable, IHEN exit the non-applicable section and enter the applicable section of this procedure.
2.2 IE the unit.is in MODE 3 with RPV pressure < 70 psig, IHEN perform section 2.6.
2.3 IE the unit is in MODE 4, IHEN perform section 2.7.
2.4 IE tNe unit is in MODE 5 with RPV water level < 458", IHEN perform section 2.8.
2.5- IE the unit is in MODE 5 with RPV water level = 458", IHEH perform section 2.9 2.6' Loss of RHR SDC in MODE 3 with RPV pressure < 70 psig 2.6.1 Monitor reactor coolant temperature and pressure in accordance with ST-O-080-500-2 (3) , " Recording and Monitoring Reactor Vessel Temperatures".
2.6.2 IE an RHR SDC isolation has occurred, THEN:
- 1. Initiate action to reset the isolation in accordance with GP-8.B.
.2. IE the isolation signal has cleared, IHEN reset the isolation in accordance with GP-8.B, "PCIS Isolation - Groups II &nd III".
QE
- 3. IE the isolation'is invalid AND can not be cleared, THEN bypass the isolation signal in accordance with Attachment 2 .
ON-125 PROCEDURE Rev. 2 Page 2 of 19 NOTE
' ' Reactor coolant forced circulation shall be established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from the discovery of no reactor coolant forced circulation. . Establishing forced circulation is required to satisfy Tech Spec 3.4.7 Required Action B.2.
2.6.3 If available, immediately initiate action to place an RHR SDC subsystem in operation in accordance with SO 10.1.B-2 (3 ) , "RHR System Shutdown Cooling Mode Manual Start".
2.6.4 IE no RER SDC subsystem in operation AHQ no Recirculation Pump in operation, IHEN place a-Recirculation Pump in operation in accordance with SO 2A.1. A-2 (3) , " Starting the First Recirculation Pump".
NOTE Some alternate decay. heat removal systems remove energy from the RPV by removing and adding reactor coolant. Use available systems at Shift Management discretion.
Use of condensate. systems is preferred over river or torus suctions when available.
CN-1 2.6.5 IE no RER SDC subsystem in operation, IHEN place available alternate decay. heat removal systems in operation. Refer to Attachment 1 for a listing of alternate decay heat removal systems.
NOTE ~
Natural circulation does NOT satisfy the
" requirements of Tech Spec 3.4.7 Required Action'B.2.
2.6.6 IE forced core circulation is not possible, THEN raise RPV water level to > +50" to promote natural circulation.
2.6.7 Immediately initiate action to restore the
, inoperable RHR SDC subsystem (s) to OPERABLE status.
2.6.8 Review CP Aid 95-04 (Unit 2) cr 95-05 (Unit 3) for expected decay heat load. (OP Aids are located on l the back of Panel 20 (30)C005. )
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I
.- . - _ _ . . . ~ . - . - - _.- _ _ .._ _ _ . . _ _ . _ _ . . _ _ _ _ _
l ON-125 PROCEDURE Rev. 2 Page 3 of 19 NOTE Ensure that the alternate decay heat removal system is available under current unit conditions. Multiple alternate decay heat removal systems may be used together to count as one alternate decay heat removal method.
2.6.9 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, verify an alternate method of decay heat removal is available for each required inoperable RHR SDC subsystem. Refer to the decay l heat load determined in the-previous step and to i
' Attachment 1 for the heat removal capabilities of l alternate decay heat removal systems.
2.6.10 Ensure. unit is in MODE 4 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.6.11 IE any of the Required Actions of Tech Spec 3.'4.7 are not met within the specified Completion Time, THEN enter LCO 3.0.3.
2.6.12L IE the unit can not be placed in MODE 4, THEN classify the event in accordance with ERP-101,
" Classification of Emergencies".
2.6.13 IE isolation signals have been bypassed in accordance with Attachment 2 AND it is no longer necessary to bypass the isolation signals, THEN restore the isolation logic in accordance with Attachment 3.
~
2.6.14 Refer to Tech Spec 3.4.7. Document non-compliance to LCO 3.4.7 and all Required Actions taken in the Tech Spec Action Log.
2.7 Loss of RHR SDC in MODE 4 2.7.1 Monitor reacter coolant temperature and pressure in accordance with ST-O-080-500-2 (3) , " Recording and Monitoring Reactor Vessel Temperatures"
, - - - - . -. . - .. - - . - - .. - . ~ ...-- - - - . - - _ ,~_ _ . - .. - .. - - _ ~
i J ON-125 PROCEDURE Rev. 2 Page 4 of 19 2.7.2 II an RHR SDC ' isolation has occurred, IHEE: ,
. , 1. Initiate action to reset the isolation in accordance with GP-8.B.
- 2. Reset the isolation in accordance with GP-8.B, I "PCIS. Isolation'- Groups II and III" IE the isolation ~ signal.has cleared.
RE
- 3. Bypass _the isolation signal.in accordance with Attachment.2 IE the isolation is invalid &HQ can not be cleared.
. NOTE ,
Reactor coolant forced circulation shall be established.within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from the
' discovery of no reactor coolant forced I
circulation. Establishing forced circulation is required to satisfy Tech ;
Spec 3.4.8 Required Action E.1.
2.7.3 -If available, immediately initiate action to place an RHR SDC subsystem in operation in accordance with
,SO 10.1. B-2 (3 ) , "RHR System Shutdown Cooling Mode Manual: Start".
2.7.4 IE no RHR SDC subsystem in operation AND no Recirculation Pump in operation, IHEN place a Recirculation Pump in operation in accordance with SO 2A.1. A-2 (3) , " Starting the First Recirculation Pump".
NOTE
- i. Some alternate decay heat removal systems remove energy from the RPV by removing and adding reactor coolant. Use available systems at Shift Management discretion.
Use of condensate systems is preferred over river or torus suctions when available.
CM-1 2.7.5 IE no RHR SDC subsystem in operation, IHEN place available alternate decay heat removal systems in
). operation. Refer to Attachment i for a listing of alternate decay heat removal systems.
l-l-
ON-125 PROCEDURE Rev. 2 j
Page 5 of 19 i r
NOTE Natural circulation does NOT satisfy the
. requirements of Tech Spec 3.4.8 Required l Action B.1.
i 2.7.6 IE forced core circulation is not possible, IHgg 1 raise RPV water level to > +50" to promote natural )
circulation. l i
2.7.7 Immediately initiate action-to restore the inoperable RHR SDC subsystem (s) to OPERABLE status.
2.7.8 Review OP Aid 95-04 (Unit 2) or 95-05 (Unit 3) for expected decay heat load.
l (OP Aids are located on the back of Panel 20(30)C005.)
NOTE Ensure that the alternate decay heat !
removal system is available under current unit conditions. Multiple alternate decay l heat removal systems may be used together l to count as one alternate decay heat removal method.
2.7.9 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, verify an alternate method of decay heat removal is available for each required inoperable RHR SDC subsystem. Refer to the decay heat load determined in the previous step and to Attachment 1 for the heat removal capabilities of alternate decay heat removal systems.
L 2.7.10 IE it is anticipated that the loss of RHR SDC will cause the unit to enter MODE 3, THEN the appropriate sections of GP-2 may be performed in order to make the Mode change in accordance with Technical Specifications.
2.7.11 IE the unit can not be maintained in MODE 4 AND the Tech Spec requirements specified in GP-2 for entry
! into MODE 3 can not be met, THEN classify the event
! in accordance with ERP-101, " Classification of Emergencies".
2.7.12 IE isolation signals ha.ve ceen bypassed in acccrdance with Attachment 2 AND it is no longer necessary to bypass the isolation signals, THEN restore the isolation logic in accordance with Attachment 3.
( 2.7.13 Refer to Tech Spec 3.4.8. Document non-compliance
!- to _
CO 3.4.8 and all Required Actions taken in the l Tech Spec Action Log.
i ON-125 PROCEDURE I Rev. 2 Page 6 of 19 2.8 Loss of RHR SDC in MODE 5 with RPV water level < 458"
, 2 8.1 Monitor reactor coolant temperature and pressure in accordance with ST-O-080-500-2 (3) , " Recording and Monitoring Reactor Vessel Temperatures".
l 2.8.2 IE an RHR SDC isolation has occurred, THEN: )
- 1. Initiate action to reset the isolation in accordance with GP-8.B.
- 2. Reset the isolation in accordance with GP-8.B, "PCIS Isolation - Groups II and III" IE the isolation signal has cleared.
- 3. Bypass the isolation signal in accordance with Attachment 2 IE the isolation is invalid AND can not be cleared.
2.8.3 If available, immediately initiate action to place an RER SDC subsystem in operation in accordance with SO 10.1.B-2 (3) , "RHR System Shutdown Cooling Mode Manual Start".
2.8.4 IE no RHR SDC subsystem in operation AND no Recirculation Pump in operation, THEH:
- 1. Place a Recirculation Pump in operation in accordance with SO 2A.1. A-2 (3) , " Starting the First Recirculation Pump" 98 NOTE Natural circulation may be used to satisfy the requirements of Tech Spec 3.9.8 Required Action C.1.
- 2. Raise and maintain RPV water level to > +50" to prcmote natural circulation.
ON-125 PROCEDURE Rev. 2 Page 7 of 19- l I
NOTE
- i L
Some alternate decay heat removal systems remove energy from the RPV by removing and adding reactor coolant. Use available systems at Shift Management discretion. ;
-Use of condensate systems is preferred over i river or torus suctions when available. '
CM-1 2.8.5 IE no RHR SDC subsystem in operation, 31DilE place I
.available alternate decay heat removal systems in operation. Refer to Attachment 1 for a listing of alternate decay heat removal systems. l l
2.8.6 Immediately initiate action to restore the '
l inoperable RHR SDC subsystem (s) to OPERABLE status.
l l 2.8.7 Review OP Aid 95-04 (Unit 2) or 95-05 (Unit 3) for expected decay-heat load. (OP Aids are located on the back of Panel 20 (30)C005.)
NOTE l Ensure that'the alternate decay heat i removal system is available under current unit conditions. Multiple alternate-decay heat removal systems may be used together to count as one alternate decay heat removal method.
2.8.8 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, verify an alternate method of decay heat removal is available for each required inoperable RHR SDC subsystem. Refer to the decay heat load determined in the previous step and to Attachment i for the heat removal capabilities of
(
alternate decay heat removal systems.
L
i ON-125 PROCEDURE Rev. 2 Page 8 of 19 1
2.8.9 IE an alternate decay heat removal method has HQI been verified to be available for each required inoperable RHR SDC subsystem within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, IHgH 3
l immediately initiate action to:
j
- 1. Restore Secondary Containment to OPERABLE status; AND
- 3. Restore isolation capability in each required I Secondary Containment penetration flow path not isolated.
2.8.10 IE isolation signals have been bypassed in I accordance with Attachment 2 AND it is no longer ;
necessary to bypass the isolation signals, THEN restore the isolation logic in accordance with Attachment 3.-
2.8.11 Document non-compliance.to LCO 3.9.8 and all -
Required Actions taken in the Tech Spec Action Log. ,
i 2.9 Loss of RHR SDC in MODE 5 with RPV water level a 458" 2.9.1 Monitor reactor coolant temperature and pressure in accordance with ST-O-080-500-2 (3) , " Recording and Monitoring Reactor Vessel Temperatures" .
2.9.2 IE an RHR SDC isolation has occurred, THEN:
- 1. Initiate action to reset the isolation in accordance with GP-8.B.
- 2. Reset the isolation in accordance with GP-8.B, "PCIS Isolation - Groups II and III" If the isolation signal has cleared.
- 3. Bypass the isolation signal in accordance with Attachment 2 IE the isolation is invalid AND can not be cleared.
2.9.3 If available, immediately initiate action to place an RHR SDC subsystem in operation in accordance with SO 10.1.B-2 (3 ) , "RER System Shutdown Cooling Mode Manual Start".
ON-125 PROCEDURE Rev. 2 Page 9 of 19
- 2.9.4 E no RHR SDC subsystem in operation AND no Recirculation Pump in operation, THEN:
n NOTE Natural circulation may be used to satisfy i the requirements of Tech Spec 3.9.7 Required Action C.1.
- 1. Maintain RPV water level > +458" to promote natural circulation.
QE
- 2. Place a Recirculation Pump in. operation in accordance with SO 2A.1. A-2 (3 ) , " Starting the First Recirculation Pump" J
NOTE Some alternate decay heat removal systems i remove energy from the RPV by removing and j adding reactor. coolant. Use available i systems at Shift Management discretion. .
Use of condensate systems is preferred over river or torus suctions when available. 1 CM-1 '
J 2.9.5 IE no RHR SDC subsystem in operation, IHEN place available alternate decay heat removal systems in I operation. Refer to Attachment 1 for a listing of alternate decay heat removal systems.
2.9.6 Immediately initiate action to restore the inoperable RHR SDC subsystem (s) to OPERABLE status.
2.9.7 Review OP Aid 95-04 (Unit 2) or 95-05 (Unit 3) for expected decay heat load. O
('P Aids are located on the back of Panel 20 (30)C005.)
o e
y
3 ON-125 PROCEDURE Rev. 2 Page 10 of 19 i
NOTE Ensure that the alternate decay heat removal system is available under current unit conditions. Multiple alternate decay heat removal systems may be used together to count as one alternate decay heat removal method.
2.9.8 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, verify an alternate method of decay heat removal is available for each required inoperable RHR SDC subsystem. Refer to the decay heat load determined in the previous step and to Attachment 1 for the heat removal capabilities of i alternate decay heat removal systems. .
2.9.9 IE an alternate decay heat removal method has NQI been verified to be available for each required i inoperable RHR SDC subsystem within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, THEN l immediately: 1
- 1. Suspend loading irradiated fuel assemblies into !
the RPV; M
- 2. Initiate action to restore Secondary Containment to OPERABLE status; M
- 4. Initiate action to restore isolation capability in each required Secondary containment penetration flow path not isolated.
2.9.10 IE istlation signals have been bypassed in accordance with Attachment 2 AND it is no longer necessary to bypass the isolation signals, THEN restore the isolation logic in accordance with j Attachment 3. 1 2.9.11' Document non-compliance to LCO 3.9.7 and all Required Actions taken in the Tech Spec Action Log.
ON-125 PROCEDURE l
Rev. 2 l Page 11 of 19
3.0 REFERENCES
3.1 QM-1, Letter to NRC dated . 5/21/75 (T03021) I 3.2 1 PBAPS TS 3.4.7, 3.4.8, 3.9.7, and 3.9.8 3 .' 3 INPO SERs 7-87, 2-87 3.4 GE SIL'No. 357, Control of Reactor Vessel Temperature / Pressure During Shutdown 3.5 IE Circular No. 81-11, Inadequate Decay Heat Removal During Reactor Shutdown 3.6 NEDO - 10625 3.7 INPO SER 23-86 J
- 1. 3.8 PORC 83-153 Calculation, Calc PM-940 and Calc PE-057 rev 7 3.9 Branch Technical Position ASB 9-2, Decay Heat Calc.
l
ON-125 PROCEDURE Rev. 2 Page 12 of 19 Attachment 1 ALTERNATE DECAY HEAT REMOVAL SYSTEMS System Hgjitt;, Limitations Removal Cacability RWCU 4.4 MW (One NRHX)
Fuel Pool Cooling 1.1 MW 1. Unit in MODE 5 (1 HX) 2. Reactor cavity flooded 2.2 MW 3. Fuel Pool Gates (2 HX) removed 3.3 MW (3 HX)
- Alternate Shutdown Cooling 20.5 MW Will inject low-in accordance with AO per RHR HX quality water into RPV 10 ~ 12-2 (3)
- Potential Make-uo Sourcesi
- 1. Feedwater
- 4. HPCI
- 5. Condensate
- 6. Core Spray
- 7. LPCI
- 9. Condensate Transfer 10,. Refuel Water Transfer
- 11. . Fire System 12 .SBLC k
4
ON-125 PROCEDURE '
Rev. 2 Page 13 of 19 Attachment 2 DEFEAT OF SHUTDOWN COOLING ISOLATION SIGNALS CAUTIONS *
- IE the automatic isolation capabilities of MO-2 (3) 017, *
- MO- 2 ( 3 ) 018 , MO-2 (3) 025A QB MO-2 (3) 025B are to be *
- defeated, THEN the following apply: *
- O Verify compliance with Tech Spec 3.3.6.1. *
- Cooling.
- NOTES
- 1. IE the unit is in MODE 4 QR 5, SED Shutdown Cooling System integrity is maintained, THEN for the Low RPV Level Shutdown Cooling Isolation, only one channel per trip system with an isolation signal available to either MO-2 (3 ) 17 QR MO-2 (3 ) 18 is required to be OPERABLE.
- 2. Steps 2 and 3 apply to Unit 2 only.
- 3. Steps 4 and 5 apply to Unit 3 only.
- 1. Perform an Independent Verification that the Shutdown Cooling isolation was due to an invalid signal:
PARAMETER (S) ACTUAL VALUE INITIALS IV INITIALS
. . .._ _. _ _ _ _ _ _ _ .. _ _. _ . . _ _ ... _ _ _ _ . _ _ . _ - . . . . . - . . _ _ . _ ~
7 ON-125 PROCEDURE Rev. 2 Page 14 of 19 1
'2. ' (UNIT.2 ONLY) IE MO-2-10-017, MO-2-10-018,.MO-2-10-025A QR MO I L 025B' isolate due to an invalid high drywell pressure signal QE i p
L due to an' invalid low reactor level signal, THEN perform the following-as applicable:
- a. For MO-2-10-017 : -1 I
17 . -1. Install'a jumper from terminal EE-17 to EE-18 in Panel !
p^ 20C015.
]
, I l l Performer Initials / Date DV. Initials / Date I L2.- Place switch 16A-S33, "PCIS RESET OUTBOARD", at Panel
[ 20C005A to the " GROUP II/III" position and allow it to -
[_ spring return to " NORMAL".
- 3. Operate MO-2-10-017 using the control switch as required.
- b. For MO-2-10-018:
- 1. Install a' jumper from terminal BD-10:to BB-11 in-Panel
'20C015.
Performer Initials / Date- DV-Initials / Date' )
- 2. - Place switch 16A-S32, "PCIS RESET INBOARD", at Panel-20C005A to the." GROUP II/III" position and allow it to spring return to " NORMAL" .
j
- 3. Operate MO-2-10-018 using the control switch as required.
- c. For.MO-2-10-025A:
- 1. : Lift lead from terminal CC-18 in Panel 20C041.
1 Performer Initials / Date DV Initials / Date
3.. Operate MO-2-10-025A using the control switch as required.
t
-3 4 4
-. = -, --.- - - .. . __ .- _ . _ . - . . ~ _- . _ . ~.
l ON-125 PROCEDURE Rev. 2
- Page 15 of 19
-d. For MO-2-10-025B:
1, Lift lead from terminal CC-13 in Panel 20C041.
Performer Initials / Date DV Initials / Date
- 3. Operate MO-2-10-025B using the control switch as required.
- 3. _ (UNIT 2 ONLY) IE valves MO-2-10-017 AND/OR MO-2-10-018 isolate due to an invalid high reactor vessel pressure signal, THEN perform the following as applicable:
- a. For MO-2-10-017:
- 1. Install a jumper from terminal AA-27 to CC-30 in Panel 20C042.
Performer Initials / Date DV Initials / Date
- 2. Place switch 16A-S33, "PCIS RESET OUTBOARD",.at Panel 20C005A to the " GROUP II/III" position and allow it to spring return to NORMAL.
- 3. Operate MO-2-10-017 using the control switch as required.
- b. For MO-2-10-018:
- 1. Install a jumper from terminal AA-27 to CC-32 in Panel 20C041.
Performer Initials / Date DV Initials / Date
- 2. Place switch 16A-S32, "PCIS RESET INBOARD", at Panel 20C005A to the " GROUP II/III" positicu and allcw it to spring return to NORMAL.
- 3. Operate MO-2-10-018 using the control switch as required.
l l
ON-125 PROCEDURE Rev. 2 Page 16 of 19
- 4. (UNIT 3 ONLY) IJE MO-3-10-017, MO-3-10-018, MO-3-10-025A OR MO 10-025B isolate due to an invalid high drywell pressure signal OR due t;o an invalid low reactor level signal, THEN perform the _
following as applicable:
- a. For MO-3-10-017:
- 1. Install a jumper from terminal EE-17 to EE-18 in Panel 30C015.
Performer Initials / Date DV Initials / Date
- 2. Place switch 16A-S33, "PCIS RESET OUTBOARD", at Panel 30C005A to the " GROUP II/III" position and allow it to spring return to " NORMAL".
- 3. Operate MO-3-10-017 using the control switch as required.
- b. For MO-3-10-018:
- 1. Install a jumper from terminal BB-10 to BB-11 in Panel 30C015.
Performer Initials / Date DV Initials / Date
- 2. Place switch 16A-S32, "PCIS RESET INBOARD", at Panel 3C005A to the " GROUP II/III" position and allow it to spring return to " NORMAL".
- 3. Operate MO-3-10-018 using the control switch as required.
- c. For MO-3-10-025A:
- 1. Lif t lead from terminal CC-19 in Panel 30C041.
i Performer Initials / Date DV Initials / Date
- 3. Operate MO-3-10-025A using the control switch as required.
- . . _- . -.= _ . _. .- - . _ ~ . . . . _ ~ _ , , _ - . .
Il ON-125 PROCEDURE j Rev. 2 1 Page 17 of 19
- d. For MO-3-10-025B:
1, Lift lead from terminal CC-13 in Panel 30C042. j Performer Initials / Date DV Initials / Date I
- 2. Push switch 10A-S32B, " SHUTDOWN CONTROL", for MO-3-10-025B l at Panel.30C003-02.
3.
Operate MO-3-10-025B using the control switch as required.
- 5. (UNIT 3-ONLY) IE valves Mo-3-10-017 AND/OR MO-3-10-018 isolate due to an invalid high reactor vessel pressure signal, THEN perform the following as applicable:
- a. For MO-3-10-017:
- 1. Install a jumper from terminal AA-27 to CC-30 in Panel
, 30C042.
1 Performer Initials / Date DV Initials / Date
- 2. Place switch 16A-S33, "PCIS RESET. OUTBOARD", at Panel 30C005A to the " GROUP II/III" position and allow it to spring return to NORMAL.
- 3. Operate MO-3-10-017 using the control switch as required.
! b. For MO-3-10-018:
f l ' 1. Install a jumper from terminal AA-27 to CC-32 in Panel 30C041.
Performer Initials / Date DV Initials / Date
- 2. Place switch 16A-S32, "PCIS RESET INBOARD", at Panel 30C005A to the " GROUP II/III" position and allow it to spring return to NORMAL.
- 3. Operate MO-3-10-018 using the control switch as required.
J 8
, - ON-125 PROCEDURE Rev. 2 i
Page 18 of 19
' Attachment 3 RESTORATION OF SHUTDOWN COOLING ISOLATION SIGNALS l
- 1.
WHEN'the defeat of the Shutdown Cooling isolation signals is no
. longer required, THEN continue in this attachment. I
! 2. : Remove Shutdown' Cooling from service in accordance with SO 10.2.B-2 (3) , " Residual Heat Removal Shutdown Cooling Mode Shutdown".
- 3. For Unit 2 ONLY: (Perform the applicable steps)
- a. Removt jumper from terminal EE-17 to EE-18 at Panel 20C015. 1 I
l i
Performer Initials / Date DV Initials / Date
- b. Remove jumper from terminal BB-10 to BB-11 at Panel 20C015.
I L
l Performer Initials / Date DV Initials / Date c.
Land lifted lead at. terminal CC-18 at Panel 20C041.
I Performer Initials / Date DV Initials / Date I
! # d. Land lif ted lead at terminal CC-13 at Panel 20C042.
l Performer Initials / Date DV Initials / Date
- e. Remove jumper from terminal AA-27 to CC-30 at Panel 20C042.
Performer Initials / Date DV Initials / Date
- f. Remove jumper from terminal AA-27 to CC-32 at Panel 20C041.
p Performer Initials / Date DV Initials / Date l
l' I
i p
l ON-125 PROCEDURE Rev. 2 Page 19 of 19
- 4. For Unit-3 ONLY: .(Perform the applicable steps)
- a. Rgmove jumper from terminal EE-17 to EE-18 at Panel 30C015.
Performer Initials / Date DV Initials /.Date-
- b. Remove jumper from terminal BB-10 to BB-11 at Panel 30C015.
~
l Performer Initials / Date DV Initials / Date
- c. . Land lifted lead at terminal CC-19 at Panel 30C041.
l- i i
Performer Initials / Date DV Initials / Date !
- d. Land lifted lead at terminal CC-1.: at Panel 30C042.
Performer Initials / Date DV Initials / Date J
- e. Remove jumper from terminal AA-27 to CC-30 at Panel 30C042.
Performer Initials / Date DV Initials / Date
- f. Remove jumper from terminal AA-27 to CC-32 at Panel 30C041.
Performer Initials / Date DV Initials / Date
- 5. Place Shutdown Cooling in' service in accordance with SO 10.1.B-2 (3 ) , " Residual Heat Removal System Shutdown Cooling Mode Manual Start".
l- 6. Inform Shift Management that all applicable steps of this attachment are complete.
Signature - Shift Management / Date j.
i l
4 Y
a a
6
, , , . .. _ . , . - . , , , _ _ _ . . _y_ ___., _ . , , _ . . , ,
.. . - . ~ . _ . - ~ . . ~ . _ . _ . - , - . _ . . _ _ . . _ ~ . . . . . . . . - . - - . ~ .
C8APs EAL Techracel Been Manuel CEV Df, January 15,1900 !
. l ' Page 44 of 128 3.0 Fission Product Barrier i
3.3 Reactor Coolant System Barrier RC.5 OtherIndications 4
EAL kQGE Not Applicable -
POTENTIAL LOSS RPV level cannot be determined OPCON. m um 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 cote submergence. Based on differences in calibration and design, all ranges of
" . level instruments may not indicate exactly the same; this operational difference is expected 4
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 RPV level is due to reference leg boil-off, it is an indicator of a potential loss of the Reactor Coolant System. Adequate core cooling would be rapidly assured using the guidance provided in the TRIP Procedures. If it can be determined that the cause is due to an instrument power or instrumentation failure, then it is not appropriate to classify the event as a potentialloss of the Reactor Coolant System.
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 i those which utilize separate reference legs) to ensure that actual RPV level is known and C
displayed. Unexplained and/or sudden changes in specific level indications may be a result of
' degassification of the coolant contained in the levelinstrumentation.
DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4 and RC EAL #5 T-101, RPV Control, RC/L-1 T-112, Rapid Depressurization
- T-117, Level / Power Control T-116,' RPV Flooding
m- . _ . . . _ . . . , . . . . _ _ _ . . _ . _ . _ . _ _._ , _ .m _- .. .- . . _ . _ _ _ - _ _ _ _ _ - _ _ . . . . _ _ _ . ...
4 j '
PBAPs EAL Tectwcal Basse Manual
_.l. REV o1, January 15,1990 i <
Page 53 of 126 3.0 Fission Product Barrier 3.4 Primary Containment Barrier i
3 PC,5 Other indications EAL 3 LQSE Not Applicable POTENTIAL LOSS' RPV level cannot be detemuned '
8NE RPV Flooding cannot be established per T-116 OPCON -
BASIS
'The decision to enter RPV Flooding is made when RPV waterlevel 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 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 waterlevelincrease. So RPV flooding not established requires containment flooding. This represents a potential loss of containment due to the potential need to vent containment in order to facilitate flooding. Additionally, it represents a potential inability to remove decay heat which may also lead to containment failure.
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. Level indication pegged high is indication that the levelis above the range and that it is known, also visual observation during refueling is indication of RPV water level.
if it can be determined that the loss of ability to monitor RPV level is due to an instrument
_ power or instrumentation failure, then .it is not appropriate to classify the event as a potential loss of the Primary Containment.
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.
. . .- -. . . . , - . ~ . - --- . .. .. . . . . . . . . .
l PBAPs EAL Techrucal Base Manuel l QEV Di. January 15,1999 '
l l Page 54 of 128 l l l 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, F.CCS l injection flow rates, Containment parameters, and injection system operability should all be i used to gauge the effectiveness of the RPV Flood.
L If the 1%s of level indication was caused by reference leg flashing, then level indicators can i
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 l contained in the level instrumentation piping, then it is anticipated that flooding pressure can i be obtained.
l 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.
l DEVIATION None I l
1 REFERENCES l NUMARC NESP-007, FC EAL #4, RCS EAL #5 and PC EAL #5 {
l T-101, RPV Control l
l T-111, Level Restoration / Steam Cooling, LR-11 l l T-112, Rapid Depressurization L T-117, Level / Power Control l T-116, RPV Flooding
]
l l l l
l I
l l
l i
t e
i .-
ENCLOSURE 2 4
Limerick Generating Station Units 1 and 2 -
SupplementalInformation for Revised Emergency Action Level Guidelines T
4
. . ~ _ . _ _ _ _ . _ _ . _ . - _ . _ _ _ . - ...._ ... _ .- _ . _ _ ~ _ _ _ . . _ _.- _ _.._ , .
LGS EAL Technical Bast Manual REV D1. January 19,1999 Page 67 of 126 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 Radioactmty ;
Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or Projected Duration of the Release EAL I 1
l A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes .A_MQ Dose Projections are not - '
available:
North Stack 4.16E+6 Ci/second South Stack 2.25E-3 Ci/cc l
l' Note: If the required dose projections cannot be completed within the 15 minute period, then 1 the declaration must be made based on the valid sustained monitor reading.
9E Projected offsite dose using computer dose model exceeds 100 mrem TPARD QR 500 mrem child thyroid CDE QB Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 100 mrem /hrexpected to continue for more than one hour, QE Analysis of Field Survey results l Indicate child thyroid dose commitment of 500 mrem for one hour of inhalation
! OPCON mu BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.
A 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.
Total Protective Action Recommendation Dose (TPARD)is equal to Total Effective Dose Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE)is equal to
' the thyroid exposure due to lodine. The computerized dose model provides projected TPARD and CDE.
An actual er projected dose of 100 mrem Total Protective Action Recommendation Dose
[ (TPARD) is cased on the 10 CFR 20 annual average population exposure limit. This value also V
- -. . . . ~ . .-.~ . .. .- - - . _ . . _ . . _ - - _ . _ - . . . . - .~ - - - . . - -
lgs EAL Teen 6ce Basie Manual i REV D1. January 15.1999 Page 68 of 126 provides a desirable gradient (c% order of magnitude) between the Site Area Emergency and
( General Emergency classifications. The 500 mrem integrated child thyroid dose was established l .in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TPARD and Child l'
[, Thyroid Committed Dose Equivalent (CDE). Actual meteorology is used, since it gives the most accurate dose i,rejedkri.
l Monitor indications are calculated using the computerized dose model with UFSAR source terms applicable to each monitored pathway in conjunction with annual average meteorology j
- and a one hour release duration. The inputs are as follows: '
North Stack South Stack !
L Stability Class E E Wind Speed 6.2 mph 6.2 mph
- Wind Direction 292* 292*
. Release Rate 4.16E+6 Ci/cc 2.25E-3 Ci/cc 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).
l This event will be escalated to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines per EAL Sechon 5.1.4.
I DEVIATION L
lNone REFERENCES l
o NUMARC NESP-007, AS1.1, AS1.3 and AS1.4 EPA 400 t
l 1
I i
i' l
. . - . -. . - - . ~ . . . . . . - _ . . .- . .-
i.
l LGS EAL Tschrncal Basis Manual REv c1 January 15,1999 Page 69 of 128 5.0 Radioactivity Release 5.1 Effluent Release and Dose CENERAL 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 1 EAL I
A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes A_HQ Dose Projections are not available:
North Stack 4.16E+7 pCi/second South Stack 2.25E-2 Ci/cc l
Note: If the required dose projections cannot be completed within the 15 minute period, then l the declaration must be made based on the valid sustained monitor reading.
! 98- l Projected offsite dose using computer dose model exceeds 1000 mrem TPARD OR 5000 mrem child thyroid CDE 98 Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 j l mrem /hr expected to continue for more than one hour, QB Analysis of Field Survey results I indicate child thyroid dose commitment of 5000 mrem for one hour of inhalation L
.OPCON " " ~
- BASIS 1
l.
Valid means that a radiation monitor reading has been confirmed by the operators to be correct. l A monitor reading is considered to be valid when it is verified by:
- 1. An instrument check indicating the monitor has not failed; l 2 Indications on related or redundant instrumentation; or, l
- 3. Direct observation by plant personnel.
Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose i Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to l the thyroid exposure due to iodine. The computerized dose model provides projected TPARD l and CDE. I I
The 1000 mR TPARD and the 5000 mR child thyroid integrated dose are based on the EPA
, protective action guidance. This is consistent with the emergency class desenction for a General l l Emergency. This level constitutes the upper level of the desirable gradient for the Site Area l
--- . . - . . - -. - . - . - ~ . . . . - . . . . _ . - . . .-
- -.. . . l lgs EAL Teerucal sasis Manual I,. - REV D1, January 15,1999
. .Ps9e 70 of 126
', Emergency. Actual metsorology is specifically identified in the initiating condition since it gives the most accurate dose assessment.
~
Monitor indica,tions are calculated using the computerized dose model with UFSAR source
. terms applicable to each monitored pathway in conjunction with annual average meteorology and a one hour release duration. The inputs are as follows:
North Stack South Stack Stability Class E E Wind Speed '
- 6.2 mph 6.2 mph Wind Direchon 292* 292*
Accident LOCA LOCA Release Rate ' 4.16E+7 Ci/cc - 2.25E-2 Cl/cc
! 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).
l DEVIATION
[None a REFERENCES
. 1 NUMARC NESP-007, AG1.1, AG1.3 and AG1.4 EPA-400
- r
lgs EAL Table REV 01 January 15.1999 l Page 11 of 24 l
SITE AREA IC Boundary Dose Resulting from an Actual or imminent Release of EMERGENCY Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or Projected Duration of the Release 5.1.3 """ m m A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:
North Stack 4.16E+6 Ci/second South Stack 2.25E-3 pCi/cc Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.
E Projected offsite dose using computer dose model exceeds 100 mrem TPARD @ 500 mrem child thyroid CDE i
M Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 100 mrem /hr expected to continue for more than one hour, M Analysis of Field Survey results indicate child thyroid dose commitment of 500 mrem for one hour of inhalation IC Boundary Dose Resulting from an Actual or Imminent Release of GENERAL EMERGENCY 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 5.1.4 -m A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:
North Stack 4.16E+7 Ci/second South Stack 2.25E 2 Ci/cc Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.
Projected offsite dose using computer dose model exceeds 1000 mrem TPARD @ 5000 mrem child thyroid CDE Analysis of Field Survey results indicate site boundary whole body oose rate exceeds 1000 mrem /hr expected to continue for more than one hour, OR Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mrem for one hour of inhalation NOTE: CDE = Committed Dose Equivalent TPARD =
Total Protective Action Recommendation Dose
LGS EAL Techneet Basis Monumi REV o1. January 13.1999 l Page 46 of 126 3.0 Fission Product Barrier 3.4 Primary Co'ntainment Barrier PC.1 DrywellPressure'
.EAL '
LQ33 Rapid, unexplained decrease in Drywell Pressure following initial increase
-QB D:ywell pressure response not consistent with LOCA conditions POTENTIAL LOSS Drywell Pressure > 44 psig and increasing Drywell Hydrogen > 6% AND Drywell Oxygen > 5%
OPCON- m"'
BASIS ,
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-pressurer thould increase as a result of mass and energy release into containment from a
-LOCA. Thus, drywell pressure not _ increasing under these' conditions indicates a loss of containment integrity. The 44 ps/g for potential loss of containment is based on the
- containment drywell design pressure and is equal to the peak pressure expected from a DBA LOCA.
The specified value of 6% hydrogen concentration is the minimum which can support a
' deflagration. Likewise, the minimum concentration of oxygen required to support a deflagration is 5%. Combustion of hydrogen in the deflagration concentration range creates a traveling flame causing'a rapid rise in primary containrfient pressure. A deflagration may result in.a peak primary containment pressure high enough to rupture the primary containment or damage the drywell-to-suppression pool boundary.
DEVIATION None REFERENCES NUMARC NESP-007, PC EAL #1 T-101, RPV Control l1T-102, Primary Containment Control w/ Bases T-103, Secondary Containment Control
l l
l; lgs EAL Techmcal Basis Manual REV o1, January 15.1999
- l - Page 110 of 126 -
8.0 Extemal Events L 8.2 Fire / Explosion and Toxic / Flammable Gases l
ALERT '8.2;2.a IC Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown l
EAL-The following conditions exist:
Fire or explosion which potentially makes inoperable:
Two or More subsystems of a Safe Shutdown System (Table 8-2) QB Two orMore Safe Shutdown Systems QR Plant Vital Structures containing Safe Shutdown
' Equipment AND l
Safe Shutdown System or Plant Vital Structure is required for the present.
Operational Condition OPCON ' ' ' ' '
BASIS The primary concern of this EAL is the megnitude of the fire and tile effects on Safe Shutdown ,
. Systems required for the present Operational Condition. A Safe Shutdown System is defined l as any system required to maintain safe operation or to establish or maintain Cold Shutdown. ;
A system being " inoperable" means that it is incapable of performing the design function. For ]
example, the LPCI System is intended to maintain adequate core cooling by covering the core i to at least 2/3 core height following a DBA' LOCA. In order for the system to be unable to maintain its intended function, multiple loops would need to be disabled by the fire. In addition to indication of degraded system performance, potentialinoperability may be determined by 4 visual observation and other control room indications such as loss of indicating lights.
- Table 8-2 Safe Shutdown Systems Diesel Generators 4KV Safeguard Buses ADS 1 HPCI RCIC RHR (All Modes)
Core Spray RHR Service Water ESW SGTS RERS CAC PCIS Control Room Ventilation
' Safe Shutdown Analysis is consu::ed to determine systems required for the applicable mode.
l I
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lgs EAL TCchncas Base Manuel REV ol. .:anuary 15.1999 l Page 111 of 126 Two examples of applying this methodology are as follows:
) -
Diesel Generators and 4 KV Safeguard Buses The fire disables multiple Diesel Generators or 4 KV Safeguard Buses so that the number of emergency power systems available would be decreased to below what <
would be required to mitigate an accident under the current operating conditions. l For 100% power, this could be conservatively interpreted as at least two Diesel l
Generators or 4 KV Buses disabled. i RHR - LPCI Mode The fire disables multiple loops of LPCI so that adequate core submergence could not be assured following a DBA LOCA. For 100% power, this could also be conservatively interpreted as at least two loops disabled.
The EAL includes the condition that the fire must make "TWO OR MORE" subsystems or "TWO OR MORE" systems inoperable In those cases where it is believed that the fire may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL if the structure houses or otherwise supports Safety Systems required for the present Operational Condition.
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 or made inoperable. 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 l actur.1 magnitude of damage. The occurrence of the fire or explosion with reports of damage (e.g., deformation, scorching) is sufficient for declaration.
Fire is defined as combustion characterized by the generation of heat and smoke. - Sources of
~
smoke such as overheated electrical equipment and slipping drive belts, for example, do not I constitute fires. Observation of a flame is preferred, but is NOT required if large quantities of smoke and heat are observed.
This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.
DEVIATION None REFERENCES NUMARC NESP-007, HA2 LGS Safe Shutdown Analysis NUMARC Cuestions and Answers, June 1993," Hazards Question #7"
lgs EAL Table 3 REV D1, January 19,1999 l' Page 20 of 24 8.0 External Events 8.2 Fire / Explosion and Toxic! Flammable Gases 1
CLASSIFICATION EMERGENCY ACTION LEVEL IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes !
UNUSUAL EVENT of Detection 8.2.1.a ""'""'
Fire within SE-8 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of control room notification or verification of a control room alarm IC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the Plant 8.2.1.b c"""'*
Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant
.QB Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event IC - Natural and Destructiva Phenomena Affecting the Protected Area 8.2.1.c '> -
Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equiprnent ALERT IC Fire or Exolosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown 8.2.2.a """mi The following conditions exist:
Fire or explosion which potentially makes inoperable:
Two orMore subsystems of a Safe Shutdown System (Table 8-2) QR Two or More Safe Shutdown Systems QR Plant Vital Structures containing Safe Shutdown Equipment AND Safe Shutdown System or Plant Vital Structure is required for the present Operational Condition W
l r
a
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lgs EAL Techrmcal Bass Manuel CEv ot, January 15,1999 l; Page 90 of 126 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability ALERT - 7.2.2
- .IC Inability to Maintain Plant in Cold Shutdown
'EAL The following conditions exist Unplanned Loss of gk Tech Spec required systems available to provide Decay Heat Removal functions MlE Uncontrolled Temperature increase that either:
. Exceeds 200 *F
- (Excluding a <15 minute rise >200* F with a heat removal function restored) 9.B e Results in temperature rise approaching 200 *F ;
l (with NO heat removal function restored) i OPCON ao BASIS This EAL addresses complete loss of functions required for core cooling during refueling and cold L shutdown modes.~ A loss of Technical Specifications components is paired with j exceeding temperature limits to acknowledge additional plant capabilities to maintain plant -
cooling. Escalation to Site Area Emergency or General Emergency would be via Effluent Release /in-Plant Radiation or Emergency Director Judgement ICs.
The statement " Unplanned Loss of A.16 Tech Spec required systems available to provide Decay Heat Removal functions"is intended to represent a complete loss of functions available,
.or an inadequate ability, to provide core cooling during the Cold Shutdown and Refueling Modes, including altemate decay heat removal methods. This EAL allows for actions taken in GP-6.2, " Shutdown Operations - Shutdown Condition Tech. Spec. Actions," to reestablish RHR in the Shutdown Cooling Mode or provide for altemate methods of decay heat removal, with
. the intent of maintaining RCS temperature below 200' F.
For loss of an in-service Decay Heat Removal system with other decay _ heat removal methods available, actions taken to provide for restoration of a decay heat removal function may require time to implement. If the event results in RCS temperature " momentarily" (for less than 15 minutes) rising above 200'F with heat removal capability restored, Emergency Director / Shift Management judgment will be required to determine whether heat removal systems are adequate to ' prevent boiling in the core and restoration of RCS temperature control.
Momentary (not to exceed 15 minutes) unplanned excursions above 200* F, when altemate
~ decay heat removal capabilities exist, should not be classified under this EAL.
i__i__m . _ __ m . . - - c-- , - =, _ , .., . . , , _- _ -,. . . , _
_ . . . _ , _ . . . ... _ _ . _ _ _ ~ _ ._ _ _ __. ._ _ _ _ _. __ _ _ . . _ . _ __
l l lgs EAL Techncal Basse Manual REU D1, January 15,1990 l Page 91 of126
' " Uncontrolled" means that system temperature increase is not the result of planned actions by l the plant staff. l The EAL guidance related to uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower !
than the cold shutdown temperature limit.
{
This EAL is concemed 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 attemate methods of decay heat removal indicates that either altemate methods are unavailable to cool the core in l the RPV or when the steam is transferred to the Suppression Pool, Suppression Pool cooling l
is unavailable. Loss of Suppression Pool cooling will result in a continuing, uncontrolled increase in reactor coolant temperature.
Escalation to the Site Area Emergency is by EAL IC, " Loss of Water Level in the Reactor i Vessel that has or will uncover Fuel in the Reactor Vessel," or by Effluent Release /In-Plant Radiation ICs.
DEVIATION None REFERENCES NUMARC NESP-007, SA3 GP-6.2, Shutdown Operations - Shutdown Conditions Tech. Spec. Actions Technical Specifications i
F l
l
1 lgs EAL Techneed Base Manuel REV D1, January 15,1999 l Page 92 of 126 7.0 Intemal 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 '
l Loss of SUPPRESSION POOL heat sink capabilities as evidenced by T-102 legs requiring an Emergency Blowdown OPCON rmma BASIS:
This EAL addresses complete loss of functions, including ultimate heat sink, required for hot shutdown with the reactor at pressure and temperature. Reactivity control is addressed in other EALs. The loss of heat removal function is indicated by T-102 legs requiring an Emergency Blowdown which is directed when the Heat Capacity Temperature Limit (HCTL) curve is exceeded.
Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted. Escalation to General Emergency would be via Effluent Release /In-Plant Radiation, Emergency Director Judgement, or Fission Product Barrier Degradation ICs.
l
LGS EAL Tectncal Base Manual j REV D1, January 15.1999 Page 93 of126 DEVIATION None REFERENCES.-
l ' NEl 97-03, SS4 ,
T-102, Primary Containment Control, SP/L-8 a
1 1
J j..
4 J
,.e.-_ .+~ d- - . , - . - . , a
-, , w - . _ . .
l lgs EAL Table l REV c1, January 15,1999 l
Page 16 of 24 l 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability l l
L CLASSIFICATION EMERGENCY ACTION LEVEL 1
l UNUSUAL EVENT None !
ALERT IC Inability to Maintain Plant in Cold Shutdown 7.2.2 -
! The following conditions exist:
Unplanned Loss of E Tech Spec required systems available to provide Decay Heat Removal functions AND Uncontrolled Temperature increase that either:
. Exceeds 200 *F (Excluding a <15 minute rise >200* F with a heat removal function restored)
.QE
. Results in temperature rise approaching 200 *F
\ (with NIQ, , heat removal function restored) i SITE AREA IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown '
EMERGENCY 7.2.3 =
l Loss of SUPPRESSION POOL heat sink capabilities as evidenced by T-102 legs requiring an Emergency B!awdown GENERAL None EMERGENCY P
l GP-6.2, RIv. 25 Page 1 of 17 HJB/AC:jml PECO Nuclear
,, LIMERICK GENERATING STATION GP-6.2 SHUTDOWN OPERATIONS - SHUTDOWN
- l. ' ,
CONDITION TECH SPEC ACTIONS 1.0 PURPOSE To provide necessary operating steps to be taken as directed by selected Technical Specification action statements.
The following actions are covered in this procedure:
Establishing Altemate Decay Heat Removal (in SECTION 3.1)
Establishing Alternate Reactor Coolant Circulation (in SECTION 3.2)
(Ref. 4.13)
Suspending Operations With A Potential for Draining the Vessel (in SECTION 3.3)
' Establishing Secondary Containment Integrity (in SECTION 3.4) 2.0 PREREQUISITES 2.1 ~ ' Vessel R9.1 defueled (Tech Spec definition 1.7 should be reviewed for
~ CORE ALTERATIONS) i
. - __ -_ -. . _ . - - - . _ . - - _ . _ - - _ _ _ _ ~-.
l GP 6.2, Rsv. 25 Page 2 of 17 HJB/AC:jml 1 3.0 PROCEDURE
,' lCAUTIONi ##
M,m?[whN.h w .
LReac Jf,c:
..m . .Ji (....,. She ' '
tonWater? Led (%jfj;. ..u.@musth grea . .. .,...,, . .... .. . , .ter thanL60 inches on) wea 4
m 1
IDE grentdMN7Cnches?ori LR 42U R6d81dbi6ffiserifrisidialinedlatioriIS itsNQf
'eventifTdigtnilidf ' ' **^ l ~ ~ . . , J " ' ' '
. M' l IDildisf5flfsdisi?Nid$I$tl$hMW N M Y$. M S- -
^
E'"'y gg y :1c; g g 'u>$wggg.y c:-- . , , . .mg a n,:gy,jgij.
c ~ , as.
, gj:y, ,
+
g.:3 egg ;,? - {
n g:-
l NOTE 1
Four RHR Shutdown Cooling Subsystems are available:
IE the associated RHR Heat Exchanger and RHR Pumps are OPERABLE AND are either aligned M able to be alianed, remotely M iocal manally, for SDC service THEN these SDC mode subsystems may be considered OPERABLE.
Reference SS1.8.B for SDC alignment using A
@ B RHR Pumps.
Reference S51.8.H for SDC alignment using C
@ D RHR Pumps.
l 3.1 ESTABLISHING ALTERNATE DECAY HEAT REMOVAL 3.1.1 SELECT AND VERIFY availability of an appropriate alternate decay heat removal system for each "less than fully operable RHR Shutdown Cooling Subsystem" using Decay Heat Removal Systems Capability Table based on time after shutdown. (Columns 1 & 3)
-- - - . - - . . . --- -. . - - - - . . - _ - . - - - . _ - . . ~ . . .
l l GP-6.2, Rsv. 25 Page 3 of 17 HJB/AC:jml 3.1.2 IE p_q systems from this list can adequately handle the decay heat load i
i based on time after shutdown estimate, THEN SELECT Atm. VERIFY availability of an appropriate alternate decay heat l removal system for each "less than fully operable RHR Shutdown l Cooling Subsystem", using Decay Heat Removal Systems Capability i
Table, based on calculation of decay heat provided by Reactor Engineering. (Columns 2 & 3)
NOTE Time estimates provided below are based on a core that has seen full power for 24 months. An alternate decay heat removal system can be selected from table below based on this decay heat load estimate. (Columns 2 & 3). For actual " Time Available l
' After Shutdown from full power" request Reactor Engineering estimation of current decay heat level.
i DECAY HEAT REMOVAL SYSTEMS CAPABILITY TABLE (1) (2) (3)
Operating System Step for Condition Time Available Max Heat Demonstration of f After Shutdown Removal Rate Decay Heat Removal from full power (MWth) Capability OPCON 4: 2 RHR HTX each (Reference with 1 RHR Pump 45 Min 58.6 3.1.5 qs Temp 180 *F) 3.1.6 oR__3.1.8 1 RHR HTX
! 1 RHR Pump 3 hrs 29.3 t 3.1.5 oR 2 RWCU Pumps /1 3.1.6 ORL3.1.8 NRHTX 58 days 2.9 3.1.6 OPCON5: 2 RHR HTX eacn (Reference with 1 RHR Pump 5 hrs 24.3 3.1.6 o_R Temp.125 'F) 3.1.8 i
1 RHR HTX 1 RHR Pump 42 hrs 12.3 3.1.6 oR_
3.1.8_
2 RWCU Pumps 1 NRHTX 193 days 1.35 3.1.6 Flooded Up: RHR Backup To Fuel Pool Cooling 60 hrs 10.7 3.1.7 3 Fuel Pool Cooling Htx & Pumps 8 days 7.0 3.1.7
' 2 Fuel Pool Cooling Htx & Pumps 20 days 4.7 3.1.7 2 RWCU Pumps 1 NRHTX 193 days 1.35 3.1.6 Flooded Up: 1 RHR Htx 1 RHR Pp 42 hrs 12.3 3.1.8 Flooded Up: U1 FPC Pps & Htxs U2 FPC Pps & Htxs 18 days 9.75 3.1.9 RWCU Pos & NRHTX i
l GP-6.2, Rsv. 25 Page 4 of 17 HJB/AC:jml 3.1.3 WHEN an alternate decay heat removal system has been verified as available, Itgg RECORD in Unified Narritive log ANQ NOTIFY shift supervision.
3.1.4 VERIFY that system selected for alternate decay heat removal is available at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.1.5 JE Alternate Shutdown Cooling using RHR, SRV's & Suppression Pool Cooling (NOT in OPCON 5) is required, ItgM PERFORM the following:
NOTE To ygtfy availability of this method for Tech Specs, only the following three steps must be performed:
- 1. VERIFY three SRV's will operate by ensuring ST-4-041-470
- is in surveillance.
- 2. VERIFY at least one loop of RHR is capable of being placed in Suppression Pool Cooling per S51.8.A.
NOTE An OPERABLE ECCS System may satisfy requirement of the following step.
- 3. VERIFY one loop of LPCI is lined up for operation per S51.1.A QB one loop of Core Spray is lined up for operation per S52.1.A. '
NOTE The following step will place alternate shutdown cooling method in service.
- 4. PERFORM S41.7.B.
J
l GP-6.2, Rav. 25 Page 5 of 17 HJB/AC:Jml 3.1.6 E RWCU System is required, THEN PERFORM the following:
' NOTE
- 1. RWCU may be verified as an available decay heat removal method by calculation QB demonstration. RT-3-044-310
- may be performed to verify availability by calculation. A method of core circulation must also be verified as available (i.e. RHR SDC or RPV Recire Pump).
4 2. For each plant shutdown period, steps 3.1.6.1. through 3.1.6.3. are only required for the initial decay heat removal demonstration. ,
I
- 1. ESTABLISH Reactor coolant circulation oniv by placing one RHR l Shutdown Cooling subsystem in service per SS1.8.B with the heat exchanger in full flow bypass ;
QB one Recirc Pump in service per S43.1.A.
- 2. PLACE RWCU in service per S44.7.B.
- 3. PERFORM RT-6-044-310
- to verify RWCU can adequately remove the decay heat load.
i NOTE l
The following step requires at least the same number of RWCU pumps AN_D_ non-regenerative heat exchangers available for service as utilized during the performance of step 3.1.6.3.. l
- 4. E subsequent demonstrations are performed, THEN VERIFY RWCU system is in service AND is capable of being placed in the decay heat removal mode per S44.7.B ANQ is capable of being supplied with cooling water from the same source as used in step 3.1.6.3..
4
_ . _. _ _ _ _ _ _ _ u-w--- .,n-
l GP-6.2, R:,v. 25 Page 6 of 17 j HJB/AC:Jml
! 3.1.7 E Fuel Pool Cooling System (With Reactor Cavity Flooded \
l N Spent Fuel Pool Gates Removed) is required, AL.D. !
Ttgg PERFORM the following:
. i 1
4 NOTE l
l For each plant shutdown period, steps 3.1.7.1. through 3.1.7.3. are only required for the initial decay heat remove demonstration. l
- 1. PERFORM the following to establish Reactor coolant l i
circulation:
- a. PLACE one RHR shutdown cooling subsystem in service per S51.8.8 i 3
QB S51.8.H with the heat exchanger in full flow
' bypass QB PLACE one Recire Pump in service per S43.1.A.
- b. IMPLEMENT Natural circulation per S53.7.A.
(Ref. 4.13)
. 2. E not already done, THEN PLACE Fuel Pool Cooling in service per S53.1.A, QB using RHR as a back-up per S51.8.G QB SS1.8.J.
- 3. ENSURE RT-1-053-310
- has been performed AND Shift Update Notice is issued for Fuel Pool Cooling decay heat removal.
3.1.8 E Shutdown Cooling is required for decay heat removal,
, THEN PERFORM the following:
- 1. PLACE one RHR Shutdown Cooling Subsystem in service per S51.8.B QB S51.8.H.
3.1.9 E Reactor Coolant Natural Circulation AND Alternate Decay Heat Removalis required (flooded up only)
THEN PERFORM SP-187, Reactor Coolant Natural Circulation.
l GP-6.2, Rev. 25 Page 7 of 17 HJB/AC:jml 3.2 ESTABLISHING ALTERNATE REACTOR COOLANT CIRCULATION NOTE ECCS Tech bpec 3.5.2 must be complied with while removing LPCI loops from dedicated LPCI service.
j 3.2.1 PERFORM one of the following to establish alternate reactor coolant circulation:
l
- 1. PLACE one RHR shutdown cooling subsystem in service per S51.8.B l QB S51.8.H, QR PLACE one Recirc Pump in service per S43.1.A.
OB
- 2. IMPLEMENT Natural circulation per S53.7.A. (Ref. 4.11) i 3.2.2 W_!iENn2 RHR Shutdown Cooling Pump is in operation as required by Tech Specs, THEN PERFORM ST-6-107-641
- to monitor Reactor coolant pressure AND temperature.
3.3 SUSPENDING OPERATIONS WITH A POTENTIAL FOR DRAINING THE VESSEL 3.3.1 POST a sign at Reactor console *0C603 to remind operators that operations with potential for draining the Vessel are suspended.
l l
l-
l GP-6.2, R v. 25 Page 8 of 17 HJB/AC:jml 3.3.2 SUSPEND the following work activities:
CRD maintenance including drive removal / replacement.
(HCU insert or withdraw line maintenance is allowed if performed in accordance with recommendations in GE SIL 419.)
NOTE Testing includes troubleshooting, logic testing, routine / surveillance tests QB any other tests which could cause valve / system operations.
Work / testing / operations which have potential to remove Vessel inventory other than normal methods of makeup / letdown, on any line QB system which connects to a Reactor Vessel penetration below Vessel level required to be maintained for current plant conditions.
The following are exceptions and this work is allowed:
- a. Work, testing, or other operations outboard of a closed valve or blind flange that has been blocked specifically to prevent draining the vessel is allowed.
- b. Work, testing, or other operations on instrument lines outboard of excess flow check valves is allowed.
- c. Work, testing or other operations on small pipe (2 inch diameter or smaller) outboard of check valves is allowed.
- d. Controlled filling and venting, in accordance with PORC approved procedures, of systems connected to the reactor vessel whose normal operation is not considered a potential for draining the vessel provided that two primary containment isolation valves in each penetration can be closed and guidance for valve closure is specified in the procedure.
- e. Filling, in accordance with PORC approved procedures, of closed piping systems from the reactor vessel to closed primary containment isolation valves.
- f. . Nuclear instrumentation replacements when a
, qualified seal tube and seal tube isolation valve is utilized. (NOTE: The seal tube and valve are the RPV boundary during the period that the NI
, is removed.)
l
.! GP-6.2, Rav. 25 Page 9 of 17 HJB/AC:jml 3.4 ESTABLISHING SECONDARY CONTAINMENT INTEGRITY NOTE
- 1. jf in OPCON 4 with Reactor Cavity Shield Plugs installed M Tech Spec action statements require establishing Secondary Containment integrity, M the Reactor Enclosure Secondary Containment Zone is extended to include the common refuel floor, THEN both the Reactor Enclosure .
M the Common Refuel Floor Secondary Containment Integrity needs to be established to satisfy Tech. Spec. actions, Otherwise. only Reactor Enclosure Secondary Containment needs to be established to satisfy Tech. Spec actions. l
- 2. Steps 3.4.1 through 3.4.11 establish Reactor Enclosure Secondary Containment <
Integrity. 1 3.4.1 ENSURE the following surveillance tests are still in surveillance:
- ST-4-076-322 * )
- 1 ST-2-026-436 * '
- ST-2-042-454 * ;
ST-2-042-455 * !
- ST-2-042-460 * ,
- l 3.4.2 VERIFY ST 6-076-200
- is in surveillance.
-l GP-6.2, R:v. 25 Page 10 of 17
, HJB/AC:jml 3.4.3 VERIFY the following monthly surveillance tests are still in surveillance:
--' ST-2-026-621
- ST-2-042-656 * '
-1 :ST-2-042-658 *
- ST-2-042-659 *-
- ST-2-076-600 * '
- 3.4.4 REVIEW outstanding Clearances /TPA's to determine whether there is any impact on Reactor Enclosure Secondary Containment Integrity
.i Ann CORRECT any problems. '
3.4.5 ~ PERFORM ST-6-076-360 *,
s.
-3.4.6 MAINTAIN Reactor Enclosure differential pressure greater than 0.25 inches water negative by placing Reactor Enclosure HVAC in service per
. S76.1.B.
l 3.4.7 IE Reactor Enclosure HVAC is unavailable, L THEN INITIATE SGTS per S76.8.B.
p
- i. NOTE
- .. The following step is performed prior to
- ANQ during the time Secondary Containment Integrity is needed per Tech. Specs. action statement.
4
- 3.4.8 CHANGE ST-6-107-591
- with a temporary change per A-3, to include
.the additional daily ANQ shiftly channel checks for Secondary Containment Integrity as
<4 , prescribed in Appendix 1.
! -3.4.9 UNIT 1 ONLY
)
VERIFY SGD-76-206-1, " Reactor Enclosure to SGTS," is locked open.
3.4.10 UNIT 2'ONLY
= VERIFY SGD-76-506-2, " Reactor Enclosure to SGTS," is locked open.
- - 3,4.11
- IE Reluel Floor Secondary Containment is p_g1 established, e THEN VERIFY SGD 76-206-3, " Refuel Floor to SGTS," is closed.
Y
1 I
l GP-6.2, Rsv. 25 l l Page 11 of 17 H' JB/AC:jml NOTE
- 1. WHEN in OPCON 5 with Reactor Well Cavity filled '
93 in OPCON 4 without Reactor Cavity Shield Plugs installed, THEN Reactor Enclosure M Refuel Floor Secondary Containment Integrity will need to be established to satisfy any applicable required Tech Spec actions.
- 2. IE Secondary Containment Integrity must be established due to loss of all ECCS required per Tech Spec 3.5.2 M the Reactor Wel! Cavity is filled, removing the Fuel Pool Gates reduces the ECCS requirements to zero. This may eliminate the need to establish Secondary Containment integrity.
- 3. Steps 3.4.12 through 3.4.20 establish Refuel Floor Secondary Containment integrity.
1 l
! ~ 3.4.12 ENSURE the following 18 month surveillance tests are still in surveillance:
ST-2-072-106-0 ST-2-072-107-0 l
ST-6-076-310-0 l -
- l -
- I -
- l -
ST-2-076-402-0 ST-2-076-403-0 l
3.4.13 ENSURE the following quarterly surveillance test is still in surveillance:
ST-6-076-201-0 l 3.4.14 ENSURE the following monthly surveillance tests are still in surveillance:
- ST-2-076-602 0 ST-2-076-603-0 3.4.15 ENSURE one of the following monthly surveillance test stillin surveillance:
ST-6-076-250-1 OR ST-6-076-250-2 l 3.4.16 VERIFY SGD-76-206-3, " Refuel Floor to SGTS," is locked open.
l t
t I l
- l. GP-6.2, Rsv. 25 l Page 12 of 17
! HJB/AC:jml l
3.4.17 REVIEW outstanding Clearances /TPA's to determine whether there is any I impact on Refuel Floor Secondary Containment Integrity ANQ correct any problems.
3.4.18 PERFORM ST-6-076-360 0.
3.4.19 MAINTAIN Refuel Floor differential pressure greater than 0.25 inches water negative by placing Refuel Floor HVAC in service per S76.1.A.
3.4.20 IE Refuel Floor HVAC is unavailable, i
THEN INITIATE SGTS per S76.8.B.
l 4
+
I GP-6.2, R;v. 25 Page 13 of 17 l HJB/AC:jml
4.0 REFERENCES
4.i BLP-22707, RHR Shutdown Cooling I
4.2 UFS R 9.1.3.1.d, Fuel Pool Cooling 4.3 UFSAR Fig. 5.4-14, RHR Process Diagram 4.4 UFSAR Fig. 5.4-17, RWCU Process Diagram 4.5 NEDO-24810B, Volume 1, Fig. 2-4, Decay Heat 4.6 GE SIL #401, BWR Shutdown Cooling - AlD 68 4.7 IE Bulletin No. 80-12, Decay Heat Removal System Operability 4.8 INPO SOER 82-2, Inadvertent Reactor Pressure Vessel Pressurization 4.9 Safety Evaluation for Mod 5107 (File: RES 5-3) 4.10 IE Circular 81-11 4.11 GE-NE-A0005873-12, LGS Alternate Shutdown Cooling with Natural Circulation 5.0 TECHNICAL SPECIFICATIONS 5.1 3.5.2 6.0 INTERFACING PROCEDURES 6.1 S41.7.B, Use of SRV's and Suppression Pool Cooling as an Alternate Shutdown Cooling Method
. 6.2 S43.1.A, Startup of Recirculation System 6.3 S44.7.0, Using Reactor Water Cleanup as an Alternate Method of Decay Heat Removal 6.4 S51.1.A, Set Up of RHR System for Automatic Operation in LPCI Mode 6.5 S51.8.A, Suppression Pool Cooling Operation (Startup and Shutdown) and Level Control 6.6 S51.8.8, Shutdown Cooling / Reactor Coolant Circulation Operation (Startup and Shutdown) 6.7 ' S51.8.H, Use of Dedicated LPCI Pumps for Shutdown Cooling / Reactor Coolant Circulation Operation (Startup and Shutdown) 6.8 S51.8.G, RHR System Back Up to Fuel Pool Cooling (Startup and Shutdown) 6.9 S51.8.J, A or B RHR Full Flow Test and Suppression Pool Cooling 6.10 S52.1.A, Core Spray Setup for Service Operation J
\ GP 6.2, Rsv. 25 i
Page 14 of 17 HJB/AC:jml 6.11 S53.1.A, Startup of Fuel Pool Cooling System l 6.12 S53.7.A, Alignment of Fuel Pool Cooling System For Natural Recirculation and Establishing / Restoring Natural Recirculation 6.13 S76.8.B, initiation of Reactor Enclosure of Refueling Floor Secondary Containment isolation 6.14 ST-6-076-360 *, Reactor Enclosure Integrity Check 6.15 ST-6-076-360-0, Refuel Floor Integrity Check 6.16 ST-6-107-591 *, Daily Surveillance Log /Opcons 1,2,3 6.17 ST-6-107-641 *, Rx Vessel Temperature and Pressure Monitoring 6.18 RT-1-053-310 *, Demonstration of Fuel Pool Cooling Capability as an Alternate Decay Heat Removal Method 6.19 RT-3-044-310 *, Demonstration of Reactor Water Cleanup Capability as an Alternate Decay Heat Removal Method
'6.20 RT-4-041-470 *,
6.21 RT-6-044-310 *, Demonstration of Reactor Water Cleanup as an Alternate Decay Heat Removal Method l
l l
l l
f
~
l-GP-6.2, Rev. 25 Page 15 of 17 HJB/AC:jml-ATTACHMENT 1 ,
Page 1 of 3 DAILY SURVEILLANCE ADDITIONS DAILY SURVEILLANCE LOG - (DAY and NIGHT SHIFT) - CONTP.OL ROOM t OPCON __/._.
D N Date __/__/._
DAY-SHIFT NIGHT-SHIFT TECH SPEC NOTES
. Pcnol *00681' RDG. INIT. RDG. INIT.
Rx. Enxi./ Outside PDI-76 *98A N/A N/A 4.6.5.1.1.a Must be > 0.25 in.
1 Press. water negative *
(in.wtr.vac.) 4.6.5.1.2.a :
PDI-76 *98B N/A N/A ,
4
4 l GP-6.2, Rsv.' 25
' Page 16 of 17 HJB/AC:jml
- ATTACHMENT 1
- Page 2 of 3
) , DAILY SURVEILLANCE ADDITIONS DAILY SURVEILLANCE LOG - (DAY-SHIFT) - AUX. EQUIPMENT ROOM t
- OPCON _.
Date __/__/._
- 0C609 *0C611 1 PARAMETER RDG. INIT. RDG. INIT.
4 D/W Press PIS-42 (PSIG)
. C71 *N650A C71 *N650D 1
L C71 *N650C . C71 *N650B t
. RX Vessel i Wtr Level
, LIS-42 (IN)
B21*N681A B21*N681D B21*N681C B21*N681B
- greq% s 4 4 pp .~//~' * *0C633 *00606
' ~ '~
~
PARAMETER Mihy;j ^
^ -
4 O - %' 'o RDG. INIT. RDG. INIT. ' ,, '+
RX Enxl Vent i
Rad RISH 26
!.' (MR/HR)
"B" Vent *K609B *K609A "A" Vent 4
4 "D" Vent *K609D *K609C "C" Vent 4
E i-
.g-- , - -
> . ~ - _ _ . _ . . _ _ . . . - . _ . - - . - _ ._ _ . _ _ . _ __ . . . _ __ _ - _ _ . _ _ .
I GP-6.2, Rnv. 25 Page 17 of 17 HJB/AC:jml ATTACHMENT 1 Page 3 of 3
, DAILY SURVEILLANCE ADDITIONS DAILY SURVEILLANCE LOG - (NIGHT-SHIFT) - AUX. EQUIPMENT ROOM OPCON _
Date __/__./_.
l
- 0C609
' *0C611 l PARAMETER I RDG. INIT. RDS. INIT.
l D/W Press i PIS-42 (PSIG)
C71 *N650A C71 *N650D C71 *N650C C71 *N650B RX Vessel' Wtr Level LIS-42 (lN)
B21*N681A B21 *N681 D B21*N681C B21*N681B I stgj ' + , < g; .
~
- '/ '
- 00633 *0C606 ,-> s PARAMETER >
- W -
,,/ '
RDG. INIT. RDG. INIT.
RX Enxi Vent Rad RISH-26 (MR/HR)
"B" Vent - *K609B *K609A "A" Vent l l l
'D" Vent *K609D *K609C l "C" Vent l
- ,. . , . . - . -. - - . _ - . . _ - . ~ . , . . ---.-- . .-. ~ - ~.- ~ .. -
- lgs EAL Tecnrucal 8aese Manual REV Ot. January 15, W l
] Page M of 126 -
L l
L 3.0 Fission Product Barrier
, 3.3 ' Reactor Coolant System Barrier L
!. . RC.5 l OtherIndications l .EAL , ,
i; M
i- Not Applicable i'
^
, POTENTIAL LOSS
- RPVlevel cannot be determined L
< OPCON mm
' BASIS -
e ,
' 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 toL be used for deciding that conflicting RPV level _ indication exists. Multiple indications of level instruments pegged high is indication 91st the level is above the range and 'l
' that it is known, also visual observation during refueling is indication of RPV water level.
-l
- - If indeterminate RPV level is due to reference leg boil-off, it is an indicator of a potential loss of the Reactor Coolant System. Adequate' core cooling would be rapidly assured using the guidance provided in the TRIP Procedures. If it can be determined that the cause is due to an instrument power or instrumentation failure, then it is not appropriate to classify the event as a
- potentialloss of the Reactor Coolant System. 1
- 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 Lindications.10perators should be attentive to observe multiple level indications (particularly lthose which utilize separate reference legs) to ensure that actual RPV level is known and displayeo. Unexplained and/or sudden changes in specific level indications may be a result of degassification of the coolant contained in the levelinstrumentation.
DEVIATION None:
L L REFERENCES-L NUMARC NESP 007, FC EAL #4 and RC EAL #5
- T-101; RPV Control, RC/L-1
- T-112. Rapid Depressurization i T-117 4Level / Power Control LT-1.16. RPV Flooding
-m's ' ..,4-d
~ . . --,s.-+_~ .
. - . . - - - - - - - - . ~ . - - - - . ~ - . . . - . . . - ... .- ._..
1 LGS EAL Techncal Bases Manual REV D1. January 13.1999
[
Page 53 of 126 L ,
1 1 3.0 Fission Product Barrier
~
3,4 Primary Containment Barrier
? ' PC.5 Other Indications EAL kQS.E Not Applicable -
POTENTIAL LOSS j: RPV level cannot be determined g
RPV Flooding cannot be established per T-116
'l-L -OPCON. -
1 BASIS The_ decision to enter RPV Flooding is made when RPV water level cannot be deiermined. 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 Suppression. Pool at which steam flow through the SRVs will be. sufficient to l 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. This represents a potential loss of containment due to the potential need to vent containment in order to facilitate flooding. Additionally, it represents a potential inability to remove decay heat which may also lead to containment failure.
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. Level indication pegged high is indication that the levelis above the range and that it is known, also visual observation during refueling is indication of RPV water level.
If it can be determined that the loss of ability to monitor RPV fevel is due to an instrument )
power or instrumentation failure, then it is not appropnate to classify the event as a potential i loss of the Primary Containment.
The' minimum RPV flooding pressure will ensure that_ acequate core cooling exists independent of RPV level indication. Failure to establisn the differential pressure between the
^
RPV and the Suporession Pool in a timely manor can jeoparcize the ability of the reactor l coolant system to cissipate the decay heat generated.
s a
>Ut -p re-ev- ~ 4v.--rr w**-r ve^ - - *- -----r' *- _ * -_ _ _ --'
~ lgs EAL Technscal Bases Manual REV o1. January 15.1999 l Page 54 of 126 l
1 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 causad 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 I indicated level. I 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.
1 RPV water 'evel 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 I expected. Even under these conditions vessel failure and containment failure with resultant f
release to the public would not be expected for some time.
DEVIATION None REFERENCES i
i NUMARC NESP-007, FC EAL #4, RCS 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 i